Miscellaneous Waste-Form FEPs Rev 01, ICN 00

ANL-WIS-MD-000009

April 2004

1. PURPOSE 
This scientific analysis report focuses on Waste-Form and Waste-Form Colloid features, events, 
and processes (FEPs) to be considered in the TSPA model for the License Application. 
The U.S. Department of Energy (DOE) must provide a reasonable assurance that the 
performance objectives for the Office of Repository Development (ORD) radioactive waste 
repository described in 10 CFR 63.113 [DIRS 156605] can be achieved for a 10,000-year 
postclosure period. The requirements for the postclosure performance assessment to provide this 
assurance are given in 10 CFR 63.114 [DIRS 156605]. The performance assessment is an 
analysis that (10 CFR 63.2 [DIRS 156605]): 
1. Identifies the features, events, processes (except human intrusion), and sequences of 
events and processes (except human intrusion) that might affect the Yucca Mountain 
disposal system and their probabilities of occurring during 10,000 years after disposal 
2. Examines the effects of those features, events, and processes and sequences of events 
and processes upon the performance of the Yucca Mountain disposal system 
3. Estimates the dose incurred by the reasonably maximally exposed individual, 
including the associated uncertainties, as a result of releases caused by all significant 
features, events, processes, and sequences of events and processes, weighted by their 
probability of occurrence. 
To demonstrate that regulatory-specified performance objectives of 10 CFR 63.113 
[DIRS 156605] can be achieved for a 10,000-year postclosure period, the ORD is implementing 
a FEP analysis and scenario-development methodology based on the work of Cranwell et al. 
(1990 [DIRS 101234]). The methodology, incorporated into a total system performance 
assessment (TSPA), provides a systematic approach for considering, as completely as 
practicable, the possible future states of a repository system. The TSPA seeks to include all 
possible future states using a finite set of scenario classes. A scenario is a well-defined, 
connected sequence of FEPs that outlines a possible future condition of the proposed repository 
system. A scenario class is a set of related scenarios sharing sufficient similarities that can 
usefully be aggregated for the purposes of screening or analysis. The objective of FEP analysis 
and scenario development is to define a limited set of scenario classes and scenarios that can 
reasonably be analyzed quantitatively while still maintaining comprehensive coverage of the 
range of possible future states of the disposal system. 
FEPs are a fundamental aspect of a performance assessment, where (NRC 2003 [DIRS 163274], 
Section 3): 
1. A feature is an object, structure, or condition with a potential to affect disposal system 
performance. 
2. An event is a natural or human-caused phenomenon with a potential to affect disposal 
system performance and that occurs during an interval that is short compared to the 
period of performance. 
April 2004 1 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
3. A process is a natural or human-caused phenomenon with a potential to affect disposal 
system performance and that operates during all or a significant part of the period of 
performance. 
The identification of FEPs that are relevant to the functioning of the disposal system 
conceptually produces the initial domain or parameter space of the disposal system model. A 
screening process omits those portions of the domain that are not pertinent. The formal and 
defensible selection of the pertinent domain of FEPs when developing the conceptual model of 
the disposal system is one aspect that sets performance assessment apart from typical scientific 
or engineering analyses. Because of the nature of FEP screening and model development, 
several iterations of the performance assessment process are potentially necessary to eliminate 
those FEPs of negligible influence (excluded FEPs) and to improve the modeling of those 
retained FEPs (included FEPs). 
1.1 SCOPE 
This scientific analysis report is governed by Technical Work Plan for Waste Form Degradation 
Modeling, Testing, and Analyses in Support of LA (BSC 2004 [DIRS 167796]) and Work 
Package AWFM01, WF Modeling and Analysis for LA. 
This report addresses the waste form (WF) and WF colloid FEPs, as extracted and updated from 
the LA FEP list (DTN: MO0307SEPFEPS4.000 [DIRS 164527]). The scope of this report is 
two-fold: 
1. It summarizes the TSPA-LA disposition (i.e., how the FEP is implemented) for the 
WF and WF colloid FEPs that are included in the TSPA-LA analysis model, and 
relates the FEPs to the scientific analysis reports or model reports in which these 
dispositions are developed and documented. 
2. It documents the screening argument (i.e., technical basis and rationale) for the WF 
and WF colloid FEPs that are excluded from the TSPA-LA analysis based on criteria 
identified in Section 4.2 of this report. 
1.2 FEPS IDENTIFICATION AND ANALYSIS - BACKGROUND 
1.2.1 FEP Identification 
The development of a comprehensive list of FEPs potentially relevant to post-closure 
performance of the Yucca Mountain repository is an ongoing, iterative process based on sitespecific 
information, design, and regulations. The first step of the FEP analysis process is the 
identification of FEPs potentially relevant to the performance of the potential Yucca Mountain 
repository. An initial list of FEPs relevant to Yucca Mountain was developed from a 
comprehensive list of FEPs from radioactive waste disposal programs in other countries 
(BSC 2001 [DIRS 154365], Section 2.1) and was supplemented with additional YMP-specific 
FEPs from project literature, technical workshops, and reviews (BSC 2001 [DIRS 154365], 
Sections 2.2 through 2.4). 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). Although Total System Performance Assessment for the Site Recommendation (CRWMS 
April 2004 2 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
M&O 2000 [DIRS 153246]) was cancelled along with other TSPA-SR process-level reports, it 
adequately explained the initial FEPs listing. 
For TSPA-LA, a reorganization and re-evaluation of the FEP list was undertaken in accordance 
with The Enhanced Plan for Features, Events, and Processes (FEPs) at Yucca Mountain 
(BSC 2002 [DIRS 158966], Section 3.2) and KTI Letter Report, Response to Additional 
Information Needs on TSPAI 2.05 and TSPAI 2.06 (Freeze 2003 [DIRS 165394]). The FEPs 
were then assigned to one or more FEP analysis reports, each of which addresses a specific 
technical area. In some cases, where a FEP covered multiple technical areas, it was shared by 
multiple FEP analysis reports. In these cases, a single FEP analysis report may provide only a 
partial technical basis for the screening of the FEP. The full technical basis for these shared 
FEPs is addressed collectively by all of the sharing FEP analysis reports. 
The reorganization and reevaluation resulted in a preliminary list of 47 WF and WF colloid FEPs 
as extracted from the LA FEP list (DTN: MO0307SEPFEPS4.000 [DIRS 164527]. Subsequent 
evaluation of the scope of these WF FEPs resulted in 7 FEPs being reassigned from this report to 
other FEP analysis reports and 1 FEP being reassigned from other FEP analysis reports to this 
report. The resulting 41 WF and WF colloid FEPs that are addressed in this report are listed in 
Table 1.2-1. Additional WF FEPs (clad related) are addressed in Clad Degradation  FEPs 
Screening Arguments (BSC 2003 [DIRS 165057]). 
The second step of the FEP analysis process is the screening of potentially relevant FEPs based 
on the criteria identified in Section 4.2.1. A schematic of the FEP identification and screening 
process is given in Figure 1. Although shown as sequential activities, iteration occurs as new 
information becomes available that might result in new FEPs and/or changes to the technical 
bases for screening. 
Table 1.2-1 provides the FEP number, name, description, and screening decision for each of the 
41 WF FEPs. Note that the screening decisions are based on information documented in Section 
6.2 of this report. Note also that FEP names and descriptions have been revised from the LA 
FEP list (DTN: MO0307SEPFEPS4.000 [DIRS 164527]) for FEPs 2.1.09.06.0A and 
2.1.09.07.0A to accommodate a splitting of the FEPs and that FEP descriptions have been 
modified for FEPs 1.2.04.04.0A, 2.1.01.01.0A, 2.1.01.03.0A, 2.1.02.04.0A, 2.1.02.06.0A, 
2.1.02.07.0A, 2.1.02.29.0A, 2.1.09.02.0A, 2.1.09.16.0A, 2.1.09.17.0A, and 2.1.13.02.0A to 
correct minor errors. Table 1.2-1 also identifies shared FEPs and, for included FEPs, it identifies 
the supporting model and analysis reports that further describe the implementation of the FEP. 
April 2004 3 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Figure 1. Illustration of the FEP Screening Process 
Source: BSC 2003 [DIRS 166296], Figure 3.2-2 
1.2.2 Integration of WF and WF Colloid FEP Screening into a FEP Database 
Under a separate scope of work, the Office of Repository Development FEP team is constructing 
a FEP database (BSC 2004 [DIRS 168024]) and the associated LA FEP list 
(DTN: MO0307SEPFEPS4.000 [DIRS 164527]) to assist the Yucca Mountain Project during 
the license review process. Fields within each entry provide a unique identification number, a 
description of the FEP, its origin, and mapping to related FEPs and to the assigned scientific 
analysis reports or model reports. Fields also provide summaries of the screening arguments 
with references to supporting documentation, and, for all included FEPs, statements of 
disposition of the FEP within the TSPA. The FEP discussions in this scientific analysis report 
provide summaries and dispositions for the waste-form FEPs to be included in the FEP database. 
April 2004 4 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 1.2-1. Waste-Form and Waste-Form Colloid FEPs for TSPA-LA 
LA FEP 
Number 
LA Screening 
Decision 
Shared 
FEP 
Supporting Reports for 
Included FEPs LA FEP Description 
An igneous intrusion in the form of a dike occurs through the 
repository, intersecting the repository drifts. Magma, 
pyroclastics, and volcanic gases enter the drift and interact with 
the EBS components including the drip shields, the waste 
packages, pallet, and invert. This leads to accelerated drip 
shield and waste package failure (e.g. attack by magmatic 
volatiles, damage by flowing or fragmented magma, thermal 
effects) and dissolution or volatilization of waste. 
FEP Name 
Included DE BSC 2004 [DIRS 167654] 1.2.04.04.0A Igneous 
intrusion 
interacts 
with EBS 
components 
The waste inventory includes all potential sources of radiotoxicity 
and chemical toxicity. It consists of the radionuclide inventory 
(typically in units of curies), by specific isotope, of anticipated 
radionuclides in the waste, and the non-radionuclide inventory 
(typically in units of density or concentration) that consists of 
both physical, e.g., CSNF, DNSF, and HLW, and chemical waste 
constituents. The radionuclide composition of the waste will vary 
due to initial enrichment, burnup, the number of fuel assemblies 
per container, and the decay time subsequent to discharge of the 
fuel from the reactor. 
Also consider that the fuel types, matrices, radionuclide mixes, 
and non-radionuclide inventories in DSNF and HLW may differ 
from CSNF. Additional waste types should be considered if they 
are proposed for disposal at Yucca Mountain. 
2.1.01.01.0A Waste Included N/A BSC 2003 [DIRS 161961] 
inventory 
2.1.01.02.0A Interactions Colocation refers to the disposal of CSNF, DSNF, HLW, and 
possibly other wastes in close proximity within the repository. 
Colocation might affect thermal outputs, chemical interactions, or 
radionuclide mobilization. 
Ex N/A N/A cluded 
between 
colocated 
waste 
(low consequence) 
2.1.01.02.0B Interactions Codisposal refers to the disposal of different waste types within 
the same waste package. Codisposal might affect chemical 
interactions or radionuclide mobilization. At Yucca Mountain, the 
DSNF will be combined with HLW canisters within a waste 
package. This codisposal with HLW within a waste package is 
unique to the DSNF and does not apply to the CSNF placement 
within waste packages. 
The DSNF will be contained within canisters that will be placed 
within the waste packages. The use of canisters within the 
waste package is not typical of the CSNF placement within 
waste packages. Also, some DSNF waste packages may 
contain only DSNF canisters, while others may contain both 
DSNF and HLW canisters. 
N/A Included BSC 2003 [DIRS 161962] 
between 
codisposed 
waste 
BSC 2003 [DIRS 163693] 
BSC 2004 [DIRS 167619] 
April 2004 5 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 1.2-1. Waste-Form and Waste-Form Colloid FEPs for TSPA-LA (Continued) 
LA FEP 
Number 
LA Screening 
Decision 
Shared 
FEP 
Supporting Reports for 
Included FEPs LA FEP Description FEP Name 
2.1.01.03.0A Heterogeneit Commercial spent nuclear fuel (CSNF), DOE-owned spent 
nuclear fuel (DSNF), and high-level waste (HLW) shipped to the 
repository may 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 may vary due to initial uranium 
enrichment, possible plutonium enrichment, and burnup of the 
fuel, among other factors. The physical state within the waste 
form may also vary. For example, damaged fuel pellets or 
extremely high-burnup fuels may have greater surface area 
exposed to any water penetrating a waste package than 
undamaged, low-burnup spent fuel. Given these potential 
differences in isotopic composition and physical condition, the 
mass of radionuclides available for transport may vary 
significantly among waste packages. 
N/A Included BSC 2004 [DIRS 167619] 
y of waste 
inventory 
BSC 2003 [DIRS 161961] 
The different physical (structure, geometry), chemical, and 
radiological properties of the many forms of CSNF, DSNF, and 
HLW could result in differences in the corrosion and alteration 
rates of the waste package composition. This could affect 
repository chemistry, breach times, dissolution rates, and 
availability of radionuclides for transport. 
2.1.02.01.0A DSNF DSNF to be disposed of in Yucca Mountain has a variety of fuel 
types that include metallic uranium fuels; oxide and MOX fuels; 
Three Mile Island rubble; and heterogeneous fuels such as UAlx, 
U-ZrHx, and graphite fuels. In general, the composition and 
structure of these spent fuels are significantly different from the 
commercial spent nuclear fuel (CSNF), and degradation, 
alteration, and dissolution may be different from the CSNF 
degradation. 
Included N/A BSC 2003 [DIRS 163693] 
degradation 
(alteration, 
dissolution, 
and 
radionuclide 
release) 
Processes to be considered in this FEP include alteration and 
dissolution of the various DSNF waste forms, phase separation, 
oxidation of spent fuels, selective leaching, and the effects of the 
high-integrity can (HIC) on DSNF degradation. 
April 2004 6 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 1.2-1. Waste-Form and Waste-Form Colloid FEPs for TSPA-LA (Continued) 
LA FEP 
Number 
LA Screening 
Decision 
Shared 
FEP 
Supporting Reports for 
Included FEPs LA FEP Description FEP Name 
2.1.02.02.0A CSNF Alteration of the original mineralogy of the commercial spent 
nuclear fuel (CSNF) (under wet or dry conditions) and dissolution 
of the uranium-oxide matrix can influence the mobilization of 
radionuclides. The degradation of UO2 could be affected by a 
number of variables, such as surface area, burnup, temperature, 
overall solution electrochemical potential (Eh), pH, and 
especially solutions containing significant concentrations of 
calcium, sodium, carbonate and silicate ions, as well as 
availability of organic complexing materials. In turn, these water 
properties are affected by the alteration of the cladding and 
matrix. 
Included N/A BSC 2004 [DIRS 167321] 
degradation 
(alteration, 
dissolution, 
and 
radionuclide 
release) 
2.1.02.03.0A HLW glass Glass waste forms are thermodynamically unstable over long Included N/A BSC 2004 [DIRS 167619] 
degradation 
(alteration, 
dissolution, 
time periods, and will alter on contact with water. Radionuclides 
can be mobilized from the glass waste by a variety of processes, 
including degradation and alteration of the glass, phase 
separation, congruent dissolution, precipitation of silicates, 
coprecipitation of other minerals including iron corrosion 
products, and selective leaching. 
and 
radionuclide 
release) 
Ex N/A N/A cluded 2.1.02.04.0A Alpha recoil During decay of certain radionuclides, alpha particles may be 
enhances emitted with sufficiently high energies that the daughter nuclide 
recoils appreciably to conserve system momentum. A potential 
result of recoil is that certain radionuclides, such as 234U, exhibit 
substantially greater dissolution rates (with the same solubility 
limits) and can be transported preferentially. 
(low consequence) 
dissolution 
Included N/A BSC 2004 [DIRS 167619] 2.1.02.05.0A HLW glass Cracking of the HLW glass on cooling and during handling 
cracking means that the surface area of the glass is greater than the 
surface area of a monolithic block. The increase in the surface 
area could affect the rate of glass alteration and radionuclide 
dissolution. 
2.1.02.06.0A HLW glass HLW glass recrystallization could occur and would lead to a less Included N/A BSC 2004 [DIRS 167619] 
recrystallizati corrosion-resistant waste form. Recrystallization is a slow 
process and typically occurs only if a high glass temperature is 
maintained over a prolonged period. 
on 
April 2004 7 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 1.2-1. Waste-Form and Waste-Form Colloid FEPs for TSPA-LA (Continued) 
LA FEP 
Number 
LA Screening 
Decision 
Shared 
FEP 
Supporting Reports for 
Included FEPs LA FEP Description FEP Name 
2.1.02.07.0A Radionuclide While in the reactor at the high temperatures, radionuclides such 
release from 
Included N/A BSC 2004 [DIRS 167321] 
as I and Cs may migrate and preferentially accumulate in cracks 
in the fuel matrix, grain boundaries of the UO2, and in the gap 
between the fuel and cladding. If the waste package fails and 
the cladding perforates, the release rate of this fraction of the 
radionuclides could be rapid. In addition, reactions between 
materials within the spent fuel assembly or waste package infill 
such as I and Pb may affect this process. 
gap and 
grain 
boundaries 
2.1.02.08.0A Pyrophoricity DSNF can contain pyrophoric material. Pyrophoric material 
from DSNF 
Ex N/A N/A cluded 
could ignite and produce an adverse effect on repository 
performance. Pyrophoric events could affect the thermal 
behavior of the system and could contribute to degradation of the 
waste package, waste form, and cladding. 
(low consequence) 
Included N/A BSC 2003 [DIRS 161962] 2.1.02.09.0A Chemical If waste packages and/or DSNF canisters are not completely 
effects of filled, then the unfilled inert-gas or air-filled volume could 
influence water-chemistry calculations. void space 
in waste 
package 
Ex N/A N/A cluded 2.1.02.10.0A Organic / Degradation of cellulose in the waste could affect the long-term 
cellulosic performance of the disposal system. (low consequence) 
materials in 
waste 
2.1.02.25.0A DSNF DSNF to be disposed of in Yucca Mountain has a variety of fuel 
types that may not be similar to the CSNF to be disposed. Some 
of the fuel types may have initial cladding-degradation 
characteristics that are different from those for the CSNF. 
Therefore, the effectiveness of DSNF cladding as a barrier to 
radionuclide mobilization might be different from CSNF. This 
FEP addresses all types of DSNF cladding except Naval SNF 
cladding. 
Ex N/A N/A cluded 
cladding (low consequence) 
2.1.02.28.0A Grouping of Several hundred distinct types of DOE-owned SNF (DSNF) may 
DSNF waste 
Included N/A BSC 2003 [DIRS 163693] 
potentially be stored at YMP. These represent many more types 
than can viably be examined for their individual effect on the 
repository. A limited number of representative or bounding 
degradation models must be selected and/or abstracted. As a 
result, the effects on repository performance of the details 
related to the many distinct types of DSNF can not be evaluated. 
types into 
categories 
April 2004 8 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 1.2-1. Waste-Form and Waste-Form Colloid FEPs for TSPA-LA (Continued) 
LA FEP 
Number 
LA Screening 
Decision 
Shared 
FEP 
Supporting Reports for 
Included FEPs LA FEP Description FEP Name 
2.1.02.29.0A Flammable DOE-owned Spent Nuclear Fuel (DSNF) to be disposed of in Ex N/A N/A cluded 
gas 
generation 
from DSNF 
Yucca Mountain will have a small percent of carbide fuel. When 
carbide is exposed to water, flammable gases such as methane 
and its minor constituents ethane, ethylene, and acetylene 
(referred as ethyne by the International Union of Pure and 
Applied Chemistry) are generated. If these gases ignite, 
localized increases in temperature can occur, which might affect 
fuel degradation. The area around the ignition point may be 
mechanically and/or thermally perturbed, which could affect 
waste container or host-rock properties in the area of the EBS. 
(low consequence) 
2.1.03.06.0A Internal Aggressive chemical conditions within the waste package could 
contribute to corrosion from the inside out. Effects of different 
waste forms, including CSNF and DSNF, are considered in this 
FEP. 
N/A WP Exclud ed 
corrosion of (low consequence) 
waste 
packages 
prior to 
breach 
Included N/A BSC 2003 [DIRS 161962] Chemical characteristics of the water in the waste packages (pH 
and dissolved species) may be affected by interactions with 
materials or steel used in the waste packages or waste forms 
and by the inflowing water from the drifts and near-field host 
rock. 
2.1.09.01.0B Chemical 
characteristi 
cs of water 
in waste 
package 
The in-package chemistry, in turn may influence dissolution and 
transport as contaminants move through the waste, EBS, and 
down into the unsaturated zone. 
EBS Included 2.1.09.02.0A Chemical Corrosion products produced during degradation of the waste 
interaction 
BSC 2003 [DIRS 161962] 
form, metallic portions of the waste package, and metals in the 
drift (rock bolts, steel in invert, gantry rails) may affect the 
mobilization and transport of radionuclides. Corrosion products 
may facilitate sorption/desorption and coprecipitation/dissolution 
processes. Corrosion products may form a rind around the fuel 
that could (1) restrict the availability of water for dissolution of 
radionuclides or (2) inhibit advective or diffusive transport of 
water and radionuclides from the waste form to the EBS. 
Corrosion products also have the potential to retard the transport 
of radionuclides to the EBS. Finally, corrosion products may 
alter the local chemistry, possibly enhancing dissolution rates for 
specific waste forms, or altering radionuclide solubility. 
BSC 2004 [DIRS 167619] 
with 
corrosion 
products 
BSC 2003 [DIRS 166845] 
April 2004 9 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 1.2-1. Waste-Form and Waste-Form Colloid FEPs for TSPA-LA (Continued) 
LA FEP 
Number 
LA Screening 
Decision 
Shared 
FEP 
Supporting Reports for 
Included FEPs LA FEP Description FEP Name 
2.1.09.04.0A Radionuclide Degradation of the waste form will mobilize radionuclides in the Included N/A BSC 2003 [DIRS 163152] 
solubility, 
solubility 
limits, and 
speciation in 
the waste 
form and 
EBS 
aqueous phase. Factors to be considered in this FEP include the 
initial radionuclide inventory, justification of the limited inventory 
included in evaluations of aqueous concentrations, and the 
solubility limits for those radionuclides. 
2.1.09.06.0A Reduction- Included N/A BSC 2003 [DIRS 161962] The redox potential in the waste package influences the 
oxidation of waste-form materials and the in-package solubility of 
radionuclide species. Local variations in the in-package redox 
potential can occur. 
oxidation 
potential in 
waste 
package 
2.1.09.07.0A Reaction Chemical reactions, such as radionuclide dissolution/ Included N/A BSC 2003 [DIRS 161962] 
kinetics in 
waste 
package 
precipitation reactions and reactions controlling the reductionoxidation 
state, may not be at equilibrium within the waste 
package. 
2.1.09.10.0A Secondary Inclusion of radionuclides in secondary uranium mineral phases, Ex N/A N/A cluded 
phase 
effects on 
dissolved 
radionuclide 
concentratio 
ns 
such as neptunium in schoepite and uranium silicates, could 
affect radionuclide concentrations in water in contact with the 
waste form. During radionuclide alteration, the radionuclides 
could be chemically bound to immobile compounds and result in 
a reduction of available radionuclides for mobilization. 
(low consequence) 
2.1.09.13.0A Complexatio The presence of organic complexants in water in the EBS could 
augment radionuclide transport by providing a transport 
mechanism in addition to simple diffusion and advection of 
dissolved material. Organic complexants may include materials 
found in natural groundwater such as humates and fulvates, or 
materials introduced with the waste or engineered materials. 
Ex N/A N/A cluded 
n in EBS (low consequence) 
2.1.09.15.0A Formation T of rue colloids are colloidal-sized assemblages (between Ex N/A N/A cluded 
true 
(intrinsic) 
colloids in 
EBS 
approximately 1 nanometer and 1 micrometer in diameter) 
consisting of hydrolyzed and polymerized radionuclides. They 
may form in the waste package and EBS during waste -form 
degradation and radionuclide transport. True colloids are also 
called primary colloids, real colloids, Type I colloids, 
Eigenkolloide, and intrinsic colloids (or actinide intrinsic colloids, 
for those including actinide elements). 
(low consequence) 
April 2004 10 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 1.2-1. Waste-Form and Waste-Form Colloid FEPs for TSPA-LA (Continued) 
LA FEP 
Number 
LA Screening 
Decision 
Shared 
FEP 
Supporting Reports for 
Included FEPs LA FEP Description FEP Name 
2.1.09.16.0A Formation of Pseudo-colloids are colloidal-sized assemblages (between Included N/A BSC 2003 [DIRS 166845] 
approximately 1 nanometer and 1 micrometer in diameter) of 
nonradioactive material that have radionuclides bound or sorbed 
to them. Natural pseudo-colloids include microbial colloids, 
mineral fragments (clay, silica, iron oxyhydroxides), and humic 
and fulvic acids. This FEP addresses radionuclide-bearing 
pseudo-colloids formed from host-rock materials and all 
interactions of the waste and EBS with the host rock 
environment except corrosion. 
pseudocolloids 
(natural) in 
EBS 
Included N/A BSC 2003 [DIRS 166845] 2.1.09.17.0A Formation of Pseudo-colloids are colloidal-sized assemblages (between 
approximately 1 nanometer and 1 micrometer in diameter) of 
nonradioactive material that have radionuclides bound or sorbed 
to them. Corrosion product pseudo-colloids include iron 
oxyhydroxides from corrosion and degradation of the metals in 
the EBS and silica from degradation of cementitious materials. 
pseudocolloids 
(corrosion 
product) in 
EBS 
2.1.09.18.0A Formation T of his FEP addresses the formation and transport of microbial 
colloids in the waste and EBS. 
Ex N/A N/A cluded 
microbial 
colloids in 
EBS 
(low consequence) 
2.1.09.23.0A Stability of For radionuclide-bearing colloids to affect repository Included N/A BSC 2003 [DIRS 166845] 
colloids in 
EBS 
performance, the colloids in the dispersion must remain 
suspended, that is, be stable, for time scales that are long 
relative to time required for groundwater travel. Further, they 
must carry significant concentrations of radionuclides. The 
stability of smectite colloids, which is applicable for YMP 
groundwater colloids and waste-form colloids, is determined 
primarily by ionic strength but also to an extent by pH. The 
stability of iron oxyhydroxide colloids, which is applicable to 
corrosion-product colloids, is determined by both ionic strength 
and pH. 
2.1.09.25.0A Formation of Dissolved radionuclides and other ions may coprecipitate to form Included N/A BSC 2003 [DIRS 166845] 
colloids 
(waste form) 
by 
coprecipitati 
on in the 
EBS 
colloids. Coprecipitates may consist of radionuclides bound in 
the crystal lattice of a dominating mineral phase or may consist 
of radionuclides engulfed by a dominating mineral phase. 
April 2004 11 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 1.2-1. Waste-Form and Waste-Form Colloid FEPs for TSPA-LA (Continued) 
LA FEP 
Number 
LA Screening 
Decision 
Shared 
FEP 
Supporting Reports for 
Included FEPs LA FEP Description FEP Name 
2.1.11.03.0A Exothermic Exothermic reactions liberate heat and will alter the temperature Ex N/A EBS cluded 
of the disposal system and affect the properties of the repository 
and surrounding materials. Oxidation of uranium metal fuels 
such as represented by N-Reactor fuels is one example of a 
possible exothermic reaction. Hydration of concrete used in the 
underground environment is another example of a possible 
exothermic reaction in the EBS. 
reactions in 
the EBS 
(low consequence) 
2.1.11.08.0A Thermal BSC 2003 [DIRS 161962] EBS Included Temperature changes may affect chemical and microbial 
processes in the waste and EBS. effects on 
chemistry 
and 
microbial 
activity in the 
EBS 
Ex N/A EBS cluded 2.1.11.09.0B Thermally- Temperature differentials may result in convective flow in the 
EBS. Convective flow within the waste and EBS could influence 
in-package chemistry. 
driven flow 
(convection) 
in the EBS 
(low consequence) 
Microbes are known to produce inorganic acids, methane, 
organic byproducts, carbon dioxide, and other chemical species 
that could change the longevity of materials in the repository and 
the transport of radionuclides from the near field. The rate of 
microbial gas production will depend on the nature of the 
microbial populations established, the prevailing conditions 
(temperature, pressure, geochemical conditions), and the 
organic or inorganic substrates present. Initial analysis indicates 
the most important source of nutrient in the YMP repository will 
be metals. Other possible nutrients include cellulosic material, 
plastics, and synthetic materials. Minimal amounts of organics 
are mandated by regulation. 
Ex N/A EBS cluded 2.1.12.04.0A Gas 
generation 
(CO2, CH4, 
H2S) from 
microbial 
degradation 
(low consequence) 
Ex N/A EBS cluded 2.1.12.07.0A Effects of Radioactive gases may exist or be produced in the repository. 
radioactive 
gases in 
EBS 
These gases may subsequently escape from the repository. 
Typical radioactive gases include 14C (in 14CO2 and 14CH4) 
produced during microbial degradation, tritium, fission gases (Ar, 
Xe, Kr), and radon. 
(low consequence) 
April 2004 12 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 1.2-1. Waste-Form and Waste-Form Colloid FEPs for TSPA-LA (Continued) 
LA FEP 
Number 
LA Screening 
Decision 
Shared 
FEP 
Supporting Reports for 
Included FEPs LA FEP Description FEP Name 
2.1.13.01.0A Radiolysis Alpha, beta, gamma, and neutron irradiation of water can cause Excluded N/A WP 
dissociation of molecules, leading to gas production and 
changes in chemical conditions (Eh, pH, concentration of 
reactive radicals). 
(low consequence) EBS 
2.1.13.02.0A Radiation Radiolysis due to the alpha, beta, gamma-ray, and neutron 
damage in 
Excluded N/A WP 
irradiation of water could result in the enhancement for the 
movement of the radionuclides from the surface of a degraded 
waste form into groundwater flow. When radionuclides decay, 
the emitted high-energy particle could result in the production of 
radicals in the water or air surrounding the spent nuclear fuel. If 
these radicals migrate (diffuse) to the surface of the fuel, they 
may then enhance the degradation/corrosion rate of the fuel 
(UO2). This effect would increase the dissolution rate for 
radionuclides from the fuel material (fuel meat) into the 
groundwater flow. Strong radiation fields could lead to radiation 
damage to the waste forms and containers (CSNF, DSNF, 
DHLW), drip shield, seals and surrounding rock. 
(low consequence) EBS 
EBS 
N/A Included BSC 2003 [DIRS 161962] 2.2.08.12.0B Chemistry of Inflowing water chemistry may be used in analysis or modeling 
water that requires initial water chemistry in the waste package. (FEP 2.2.08.12.0B is not 
in the included FEP table 
in the currently controlled 
flowing into 
the waste 
package REV 02 of this model 
report. It will be added to 
the next ICN/REV of the 
model report. 
Included 3.1.01.01.0A Radioactive Radioactivity is the spontaneous disintegration of an unstable 
atomic nucleus that results in the emission of subatomic 
particles. Radioactive isotopes are known as radionuclides. 
Radioactive decay of the fuel in the repository changes the 
radionuclide content in the fuel with time and generates heat. 
Radionuclide quantities in the system at any time are the result 
of the radioactive decay and the growth of daughter products as 
a consequence of that decay (i.e., ingrowth). Over a 10,000- 
year performance period, these processes will produce daughter 
products that need to be considered in order to adequately 
evaluate the release and transport of radionuclides to the 
accessible environment. 
SZ BSC 2003 [DIRS 161961] 
decay and 
in-growth 
UZ 
BIO 
April 2004 13 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
ANL-WIS-MD-000009 REV 01 
INTENTIONALLY LEFT BLANK 
April 2004 14 Miscellaneous Waste-Form FEPs 
2. QUALITY ASSURANCE 
The scientific analysis documented in this report was performed in accordance with AP-SIII.9Q, 
Scientific Analyses and evaluated in accordance with AP-2.27Q, Planning for Science Activities, 
and was determined to be quality-affecting and subject to Quality Assurance Requirements and 
Description (DOE 2004 [DIRS 168669]). These activities are subject to the QARD because 
they are associated with the characterization of the waste form in support of performance 
assessment (BSC 2004 [DIRS 167796], Section 8). No special safety management controls, 
beyond the applicable procedures, apply to this work. The process control evaluations are 
performed in accordance with AP-SV.1Q, Control of the Electronic Management of Information. 
The controls on the electronic management of data are implemented in the governing procedures 
per Section 8 of Technical Work Plan for Waste Form Degradation Modeling, Testing, and 
Analyses in Support of LA (BSC 2004 [DIRS 167796]). This waste-form FEPs analysis report 
was developed in accordance with the Office of Repository Development quality assurance 
program using approved procedures identified in Technical Work Plan for Waste Form 
Degradation Modeling, Testing, and Analysis in Support of LA (BSC 2004 [DIRS 167796]) and 
Work Package AWFM01, WF Modeling and Analysis for LA. Additional guidance has been 
provided in the Scientific Processes Guidelines Manual (BSC 2002 [DIRS 160313]). 
Based upon the Q-List, the waste form/engineered barrier is considered to be not important to 
safety but important to waste isolation, and thus has Safety Category SC (BSC 2003 
[DIRS 165179], Table A-2). 
April 2004 15 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
ANL-WIS-MD-000009 REV 01 
INTENTIONALLY LEFT BLANK 
April 2004 16 Miscellaneous Waste-Form FEPs 
3. USE OF SOFTWARE 
The analyses and arguments presented herein are based on regulatory requirements, results of 
analyses presented, and documentation in other scientific analysis reports, model reports, or 
technical literature and, thus, required no computational software. This report was developed 
using only commercially available software, used solely for visual display or graphical 
representation of data, and considered to be controlled under the Software Configuration Control 
system (LP-SI.11Q-BSC). SigmaPlot, Scientific Graphic Software, Version 3.06, Jandel 
Corporation, and Microsoft Excel 97 SR-2 are commercial software packages used in this 
analysis to display the data visually using only standard built-in mathematical functions. 
SigmaPlot is also used to plot data from the analysis. No calculations are performed with these 
software packages. No routines or macros were developed using this commercial software. The 
values used and displayed are approximate and are used only to identify a range of expected 
values. The analyses in this report provide the basis for the decision to either include or exclude 
the FEP. If a FEP is to be included in a model, any values or data are developed according to 
appropriate quality assurance procedures in the applicable subject summary or model abstraction 
report. 
April 2004 17 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
ANL-WIS-MD-000009 REV 01 
INTENTIONALLY LEFT BLANK 
April 2004 18 Miscellaneous Waste-Form FEPs 
4. INPUTS 
Section 4.1 and Table 4-1 identify all input data, parameters, and product output used in this 
analysis. Data and product output have been obtained from controlled source documents and 
other appropriate sources in accordance with the controlling procedures AP-SIII.9Q and 
AP-3.15Q. Section 4.2 addresses the FEPs screening criteria and relevant definitions detailed in 
10 CFR Part 63 [DIRS 156605] as identified in Project Requirements Document (Canori and 
Leitner 2003 [DIRS 166275]). 
4.1 DIRECT INPUTS 
The information supporting the screening of those LA FEPs dispositioned as Included is 
extracted from WF and WF colloid analysis/model reports identified in this FEPs report and 
summarized in Table 1.2-1. These reports are direct inputs that explicitly address the disposition 
of the included FEPs in the TSPA-LA. When cited in this report to support screening decisions 
to Exclude FEPs, the analysis or model reports and other inputs may be used as direct inputs or 
in a supportive or corroborative manner. Table 4-1 summarized the direct inputs (and Table 6-1 
summarizes the indirect inputs) for both included and excluded FEPs. Attachment V contains 
the qualification of references (per AP-SIII.2Q and AP-SIII.9Q) in Table 4-1 that are not 
qualified for general use, but are qualified for direct input for use in this analysis report. The LA 
FEPs often use information developed in those reports as the basis for screening decisions, 
screening arguments, and TSPA dispositions. Uncertainties in the cited data are considered in 
the screening of the FEPs in the TSPA-LA. Uncertainties concerning included FEPs are 
developed in the associated reports. 
April 2004 19 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 4-1. Direct Inputs to Waste-Form FEPs Screening Decision to Include 
Description Input Category Direct Use In Source Input 
BSC (Bechtel SAIC Company) 2002. Gamma and 
Neutron Radiolysis in the 21-PWR Waste Package 
from Ten to One Million Years. 000-00C-DSU0- 
00700-000-00A. Las Vegas, Nevada: Bechtel SAIC 
Company. ACC: MOL.20021004.0002 [DIRS 
161912] 
entire Data 6.2. Section 38 Justification in screening out radiolysis due to 
gamma linear energy transfer (LET). . Recent 
calculations by Tang have shown that neutron 
irradiation is negligible and gamma dose has been 
significantly reduced. 
BSC (Bechtel SAIC Company) 2002. Radionuclide 
Screening. ANL-WIS-MD-000006 REV 01. Las 
Vegas, Nevada: Bechtel SAIC Company. ACC: 
MOL. 20020923.0017 [DIRS 160059] 
Sections 6.2.2, The radionuclides of importance to dose calculations 
and 6.2.41 
Data entire 
were assessed.The assessment considered nominal, 
human intrusion and igneous events, as well as 
groundwater and atmospheric exposure scenarios 
and times. 
BSC (Bechtel SAIC Company) 2003. Risk 
Information to Support Prioritization of Performance and 3.5. 
Assessment Models. TDR-WIS-PA-000009 REV 01 
ICN 01, with errata. Las Vegas, Nevada: Bechtel 
SAIC Company. ACC: DOC.20031014.0003; 
MOL.20021017.0045. [DIRS 168796] 
TSPA parameters and model components that affect 
dose. Dose is insensitive to DSNF and HLW 
degradation rates, resulting in low risk prioritization 
for these components. 
Section 6.2.3 Sections 3.3.6 Data 
BSC (Bechtel SAIC Company) 2003. Dissolved 
Concentration Limits of Radioactive Elements. ANLWIS-
MD-000010 REV 02. Las Vegas, Nevada: 
Bechtel SAIC Company. ACC: 
DOC.20030624.0003 [DIRS 163152] 
Section 6.2.1 Section 6.5.1.2 Data Because waste forms are conservatively considered 
chemically unchanged, the dissolved concentration 
of radioelements in water in the basalt would have 
the same dependency on water chemistry as does 
waste dissolving in regions not influenced by an 
igneous intrusion. 
5.2.22 Data Sections Sections 6.5 Abstracted solubility models for 14 elements 
through 6.18 and 6.2.22 (plutonium, neptunium, uranium, thorium, americium, 
actinium, protactinium, radium, technetium, iodine, 
strontium, carbon, lead, and cesium) may be used to 
constrain the maximum radionuclide concentrations. 
Section 6.2.25 Analyses of the relative release rates of neptunium Data entire 
and uranium in the fuel degradation experiments in 
support of a close correlation between neptunium 
and uranium concentrations in the leachates 
Section 6.2.25 Alternative model (not utilized in the base-case) of Section 6.6.4.5 Data 
incorporation of neptunium into uranyl minerals 
captures the experimental results very well, and 
eliminates the built-in conservatism in the 
conventional pure phase solubility approach and thus 
gives more realistic source terms. 
April 2004 20 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 4-1. Direct Inputs to Waste-Form FEPs Screening Decision to Include (Continued) 
Description Input Category Direct Use In Source Input 
Section 6.2.36 Parameterizes dissolved concentrations of Data Section 6 
radionuclides as a function of in-package and EBS 
solution chemistry over a range of carbon dioxide 
partial pressures of 10-5 to 10-1.5  in effect to two 
orders of magnitude greater than atmospheric. 
BSC 2003. DSNF and Other Waste Form 
Degradation Abstraction. ANL-WIS-MD-000004 REV 
02. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: DOC.20030711.0002 [DIRS 163693] 
Section 6.2.3 Data entire DSNF degradation model development. 
Data Attachment I N-Reacto Sections 6.4 r SNF represents the overwhelming 
and 8 majority of the DSNF (85%) but less than ~5 % of the 
total waste to be stored at Yucca Mountain 
Section 6.2.17 Degradation rates for N-Reactor SNF are adopted as Data Section 6.3 
a surrogate for all DSNF. Even when upper bound 
assumptions for the rate of degradation are used for 
the entire inventory of DSNF, the contribution of the 
DSNF to the postclosure site dose, due to 
degradation and or pyrophoric behavior is negligible. 
Sections 6.2.6 
and 6.2.9 
Data Section 6.3 Development of upper-limit, conservative, and best 
estimate degradation models for each of 11 DSNF 
waste groups for TSPA-LA. 
Data Section 6.2.6 Descripti Sections 6.3, on of immobilized ceramic plutonium waste 
6.3.12 form. 
Section 6.2.19 Rapid corrosion rate of the metallic uranium that is Data Table 9 
the matrix of N-Reactor SSNF compared with other 
SNF. 
The DSNF release rate model has no explicit 
dependence on water chemistry effects that could 
result from chemical interactions between or within 
waste packages. 
Section 6.2.4 Data Section 6.3 
Section 6.2.6 Data Section 7 DSNF degradation model abstraction validation 
BSC (Bechtel SAIC Company) 2003. EBS 
Radionuclide Transport Abstraction. ANL-WIS-PA- 
000001 REV 01. Las Vegas, Nevada: Bechtel SAIC 
Company. [DIRS 162728] 
Table 20 and 
Section 7.5.1 
6.2.34 Data Sections Consider ation of dryout near the waste packages, as 
and 6.2.35 well as evaporation/condensation within the drift and 
underneath the drip shield. 
April 2004 21 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 4-1. Direct Inputs to Waste-Form FEPs Screening Decision to Include (Continued) 
Description Input Category Direct Use In Source Input 
BSC (Bechtel SAIC Company) 2004. Igneous 
Intrusion Impacts on Waste Packages and Waste 
Forms. MDL-EBS-GS-000002 REV 01D. Las 
Section w 6.2.1 Igneous intrusi o on re uld nder the drip shields, Section 6.5.1.1 Data 
waste packages and cladding in the Zone 1 
emplacement drifts ineffective so they would provide 
no further protection to the waste. These evaluations Vegas, Nevada: Bechtel SAIC Company. 
ACC: MOL.20040414.0095. TBV-5788 provide further corroboration to the TSPA 
assumption that the waste packages in Zone 1 
provide no further protection to waste forms after an 
intrusive event. 
BSC (Bechtel SAIC Company) 2003. In-Package 
Chemistry Abstraction. ANL-EBS-MD-000037 REV 
02. Las Vegas, Nevada: Bechtel SAIC Company. 
entire Data 6.2.14 Sections General , reference describing the basis of the 
6.2.15, 6.2.20, conceptual model for in-package chemistry of water. 
6.2.23, 6.2.24, 
6.2.35, 6.2.38, 
and 6.2.40 
ACC: DOC.20030723.0003 [DIRS 161962] 
Section 6.2.4 Section 6.6.2 Data Support for the conclusion that DHLWG tends to 
generate high pH solutions while DOE SNF 
generates neutral pH solutions, and together 
combined with the waste package materials the high 
pH values are buffered to lower values. Also, the 
DSNF codisposal and high-level radioactive waste 
glass waste packages are expected to be generally 
cooler than commercial SNF packages at the time of 
disposal. 
Table 10 and 
Sections 6.3.2, 
6.3.3, 6.6.1.1, 
Section 6.2.21 Support for the conclusion thatcorrosion products Data 
produced during degradation of the waste form 
metallic portions of the waste package, and metals in 
the drift (rock bolts, steel in invert, gantry rails) may 
affect the mobilization and transport of 
radionuclides.) 
6.6.1.2, and 
6.6.2.2 
6.2.15 Data Sections Supp Sections 8.5 ort for the conclusion that it is reasonable to 
and 8.11 and 6.2.36 expect little impact of carbon dioxide production on 
in-package pH and ionic strength, and that only 
minor amounts of organic matetrials might appear in 
the WP. 
4.4.4, Data Section 6.2.34 T Sections hermodynamic data for EQ6 runs in support of FEP 
8.2, 8.4, 8.5, 2.1.11.08.0A. 
and 8.6 
April 2004 22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 4-1. Direct Inputs to Waste-Form FEPs Screening Decision to Include (Continued) 
Description Input Category Direct Use In Source Input 
BSC (Bechtel SAIC Company) 2003. Initial 
Radionuclide Inventories. ANL-WIS-MD-000020 
REV 00. Las Vegas, Nevada: Bechtel SAIC 
Company. [DIRS 161961] 
Section 6.6 and Provid Data Sections 5.1, es the inventories of radionuclides for three 
Table 16 6.2.2, 6.2.3, major categories of waste forms, currently planned 
6.2.5, 6.2.18 and for disposal at Yucca Mountain: CSNF, DSNF, and 
6.2.41 DHLW Glass. 
BSC (Bechtel SAIC Company) 2003. Waste Form 
and In-Drift Colloids-Associated Radionuclide 
Concentrations: Abstraction and Summary. MDLEBS-
PA-000004 REV 00 ICN 01. Las Vegas, 
Nevada: Bechtel SAIC Company. ACC: 
DOC.20031222.0012 [DIRS 166845] 
BSC (Bechtel SAIC Company) 2004. FEPs 
Screening of Processes and Issues in Drip Shield 
and Waste Package Degradation. ANL-EBS-PA- 
000002 REV 02. Las Vegas, Nevada: Bechtel SAIC 
Company. ACC: DOC.20040408.0001. [DIRS 
162200] 
BSC (Bechtel SAIC Company) 2004. CSNF Waste 
Form Degradation: Summary Abstraction. ANL-EBS- 6.2.2.3, 6.4.1.5 
MD-000015 REV 01, with errata. Las Vegas, 
Nevada: Bechtel SAIC Company. ACC: 
DOC.20030708.0004; DOC.20031224.0001; 
DOC.20040202.0002 [DIRS 167321] 
BSC (Bechtel SAIC Company) 2004. Multiscale 
Thermohydrologic Model. ANL-EBS-MD-000049 
REV 01. Las Vegas, Nevada: Bechtel SAIC 
Company. ACC: DOC.20040301.0004. [DIRS 
163056] 
Sections 5.9, 
6.3.1, 6.3.2, 
6.3.3, and 6.3.4 
Sections 5.2.30, Effect of waste form colloids-associated Data 
6.2.21, 6.2.27, 
6.2.28, 6.2.29, 
6.2.30, 6.2.31, 
concentrations on chemical interactions. 
and 6.2.32 
Maximum allowable waste package interior 
pressures due to the release of radioactive gases to 
the interior of the waste package will be too low to 
damage the waste package. 
Section 
6.2.13.5 
6.2.37 Data Section 
Characteristics of CSNF waste degradation 
processes and CSNF degradation model. 
Sections 6.2.6 
6.2.7, and 6.2.12 
Sections 1, 6.2, Data 
Thermal effects on liquid or two-phase flow within the 
waste and EBS are explicitly accounted for in the 
Multiscale Thermohydrologic Model. Explicit 
accounting for thermal effects on liquid or two-phase 
flow within the waste and EBS, including 
consideration of dryout near the waste packages, as 
well as evaporation/condensation within the drift and 
underneath the drip shield. 
The effect of waste glass compositions on the 
calculated degradation rate is taken into account 
through the range of values of the model parameter 
kE. Ranges for the values of kE in acidic and alkaline 
solutions are selected based on the results of 
laboratory tests with glasses that provide a wide 
range of compositions that bounds the range of 
entire Data 6.2.34 Section 
BSC (Bechtel SAIC Company) 2004. Defense HLW Sections 6.5.2 
Glass Degradation Model. ANL-EBS-MD-000016 
REV 01 ICN 01. Las Vegas, Nevada: Bechtel SAIC 
Company. ACC: DOC.20040223.0006 [DIRS 
167619] 
6.2. Data Section 3 
and 6.2.5 and 6.5.4.2 
concentrations of key glass components in DHLW 
glasses, such as aluminum. The glass degradation 
model accounts for the heterogeneity of the waste 
April 2004 23 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 4-1. Direct Inputs to Waste-Form FEPs Screening Decision to Include (Continued) 
Description 
inventory. 
The glass degradation model addresses degradation 
of glass exposed to humid air and dripping water, 
and glass immersed in water. The glass degradation 
rate is calculated as a function of pH and 
temperature. 
The surface area used to calculate the radionuclide 
release rate bounds the effects of thermal cracking 
during manufacture and impact cracking during 
subsequent handling. The exposure factor 
(fexposure) is used to model the combined effects of 
the added surface area due to cracking, the fraction 
of the surface that is accessible to water, and the 
reactivity of glass in tight cracks relative to glass at a 
free surface. 
The glass degradation model is insensitive to the 
presence of crystalline phases that are either formed 
by devitrification of the glass melt during manufacture 
or intentionally added to the waste form. 
Input Category Direct Use In Source Input 
Section 6.2.3 
and 6.2.8 
Section 8.2.2 Data 
Sections 6.2.4 
and 6.2.10 
Section 8.2.1 Data 
Section 7.5.1 
and Table 7.5-1 
6.2.11 Data Section 
The effects of dissolved iron on the glass dissolution 
rates are addressed by using the results of tests in 
which iron products were added to the test solution 
when determining model parameter values for pH 
dependence () in acidic and alkaline solutions. 
The effects of glass alteration phases on the glass 
dissolution rate are addressed by including the 
results of tests in which alteration phases form when 
determining model parameter values for kE. 
HLW glass degradsation rate in water would not be 
significantly affected by the types of crystalline 
inclusion phases that could form during the slow 
cooling of the HLW glass exposed to Zone 2 
conditions. 
Sections 
6.5.4.1 and 
6.5.4.3 
6.2.21 Data Section 
Section 6.2.1 Data Section 7.5 
CRWMS M&O 1999. BWR Source Term Generation Attachment XV Data 
and Evaluation. BBAC00000-01717-0210-00006 
REV 01. Las Vegas, Nevada: CRWMS M&O. ACC: 
MOL.20000113.0334 [DIRS 136428] 
Attachment IV BWR radionuclide source term. 
April 2004 24 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 4-1. Direct Inputs to Waste-Form FEPs Screening Decision to Include (Continued) 
Description Input Category Direct Use In Source Input 
CRWMS M&O 1999. PWR Source Term Generation Attachment XV Data 
and Evaluation. BBAC00000-01717-0210-00010 
REV 01. Las Vegas, Nevada: CRWMS M&O. acc: 
MOL.20000113.0333 [DIRS 136429] 
Attachment IV PWR radionuclide source term. 
CRWMS M&O 2000. In-Drift Microbial 
Communities. ANL-EBS-MD-000038 REV 00 ICN 
01. Las Vegas, Nevada: CRWMS M&O. ACC: 
MOL.20001213.0066 [DIRS 151561] 
6.2.26 Data Sections Estimates of microbe massses growing in the Sections 6.3 
repository system suggest that the effects on the bulk and 6.2.34 and 7 
in-drift geochemistry should be small. Bounding 
biomass estimates are ~100grams (dry) per meter of 
drift; insufficient to produce high chelate levels. 
DOE (U.S. Department of Energy) 2002. DOE 
Spent Nuclear Fuel Information in Support of TSPASR. 
DOE/SNF/REP-047, Rev. 2. Idaho Falls, 
Idaho: U.S. Department of Energy, Idaho Operations 
Office. ACC: DOC.20030905.0020 [DIRS 158405] 
Uranium-metal spent fuel more chemically reactive Attachment I Data Section 6.2 
than other spent fuel. 
The total inventory of the Group 3 (Pu/U carbide) Section 6.2.6, 
6.2.17, and 
Data Appendix D 
SNF is less than .001 % of DSNF waste. 
6.2.18 
Current plans call for five glass logs and one DSNF 
canister to be placed in each codisposal package. 
Section 6.2.3 
and 6.2.4 
Data Section 9 
DOE (U.S. Department of Energy) 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. 
[DIRS 155943] 
Section 
1.4.2.2.2 
Established Fact Section 6.2.18 Description of repository site bedrock geology. 
Rechard, R.P., ed. 1995. Methodology and 
Results. Volume 2 of Performance Assessment of 
the Direct Disposal in Unsaturated Tuff of Spent 
Nuclear Fuel and High-Level Waste Owned by U.S. 
Department of Energy. SAND94-2563/2. 
Albuquerque, New Mexico: Sandia National 
Laboratories. TIC: 237102 [DIRS 101084] 
Sections 6.2.6 
per Attachment V) and 6.2.16 
Significant amounts of DSNF cladding are likely to be Section 11.3.1, Data (qualified 
damaged pp. 11 to 24 
Data (qualified 
per Attachment V) 
Section 6.2.38 Support for the conclusion that the effects of alpha pp. 67 to 74 Sunder, S.; Shoesmith, D.W.; and Miller, N.H. 1997. 
"Oxidation and Dissolution of Nuclear Fuel (UO2) by 
the Products of the Alpha Radiolysis of Water." 
Journal of Nuclear Materials, 244, 66-74. 
Amsterdam, The Netherlands: Elsevier. TIC: 
246914. [DIRS 143860] 
radiolysis on spent fuel corrosion (dissolution) will be 
transitory and will become minor as alpha dose rates 
decrease 
April 2004 25 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 4-1. Direct Inputs to Waste-Form FEPs Screening Decision to Include (Continued) 
Description Input Category Direct Use In Source Input 
Thornton, T.A. 1998. "HPPP Issue 1; Preliminary 
TSPA for a Pyrophoric Event Involving N-Reactor 
SNF Waste Packages." Interoffice correspondence 
from T.A. Thornton (CRWMS M&O) to J.S. Clouet, 
S.S. Sareen, and D. Stahl, September 21, 1998, 
LV.WP.TAT.09/98-179, with attachment. 
MOL.19981019.0001 [DIRS 125082] 
(qualifi entire Data ed Section 6.2.13 Examples of the potential for pyrophoric behavior in 
uranium metal and uranium metal-based spent per Attachment V) 
nuclear fuel. 
Temperatures of waste forms as they are disributed 
through the magma as the result of an igneous 
Section 6.2.1 Data entire DTN: MO0402SPAHWCIG.002. Heat and Water 
Chemistry Output From Igneous Intrusion. Submittal 
date: 02/23/2004. [DIRS 168762] event. 
April 2004 26 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
4.2 CRITERIA 
The NRC provides guidance in Yucca Mountain Review Plan, Final Report (NRC 2003 
[DIRS 163274]) in accordance with 10 CFR 63.114(d), (e), and (f) [DIRS 156605] on the 
screening process to exclude FEPs. The guidance states criteria that will be used as follows 
(Section 2.2.1.2.1.3): 
Acceptance Criterion 1The Identification of a List of Features, Events, and Processes Is 
Adequate 
 The Safety Analysis Report contains a complete list of features, events, and processes, 
related to the geologic setting or the degradation, deterioration, or alteration of 
engineered barriers (including those processes that would affect the performance of 
natural barriers), that have the potential to influence repository performance. The list is 
consistent with the site characterization data. Moreover, the comprehensive features, 
events, and processes list includes, but is not limited to, potentially disruptive events 
related to igneous activity (extrusive and intrusive); seismic shaking (high-frequencylow 
magnitude, and rare large-magnitude events); tectonic evolution (slip on existing 
faults and formation of new faults); climatic change (change to pluvial conditions); and 
criticality. 
How Addressed 
FEPs related to the WF and WF colloid processes evaluated in this report are listed in 
Table 1.2-1. Those subject matter analysis model reports that provide supporting technical 
discussions relevant to a specific FEP are identified in the table. Documentation of the evolution 
of the YMP FEP list is provided in Section 1.2.1. 
Acceptance Criterion 2Screening of the List of Features, Events, and Processes Is 
Appropriate 
 The U.S. Department of Energy has identified all features, events, and processes related 
to either the geologic setting or to the degradation, deterioration, or alteration of 
engineered barriers (including those processes that would affect the performance of 
natural barriers) that have been excluded; 
 The U.S. Department of Energy has provided justification for those features, events, and 
processes that have been excluded. An acceptable justification for excluding features, 
events, and processes is that either the feature, event, and process is specifically 
excluded by regulation; probability of the feature, event, and process (generally an 
event) falls below the regulatory criterion; or omission of the feature, event, and process 
does not significantly change the magnitude and time of the resulting radiological 
exposures to the reasonably maximally exposed individual, or radionuclide releases to 
the accessible environment; and 
 The U.S. Department of Energy has provided an adequate technical basis for each 
feature, event, and process, excluded from the performance assessment, to support the 
April 2004 27 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
conclusion that either the feature, event, or process is specifically excluded by 
regulation; the probability of the feature, event, and process falls below the regulatory 
criterion; or omission of the feature, event, and process does not significantly change the 
magnitude and time of the resulting radiological exposures to the reasonably maximally 
exposed individual, or radionuclide releases to the accessible environment. 
How Addressed 
The above criterion permits an exclusion argument for FEPs that are not consistent with the 
regulations because of waste characteristics, repository design, or site characteristics, provided 
NRC staff finds sufficient justification. Additionally, a FEP may be excluded from TSPA-LA if 
it is demonstrated that the likelihood of a specific occurrence is below the quantitative 
probability of one in 10,000 of occurring over a period of 10,000 years. The final criterion 
permits exclusion of FEPs that do not significantly change the calculated expected annual dose 
provided the NRC staff finds sufficient support in accompanying discussions or calculations, 
including the use of either bounding or representative estimates. The sequence implemented in 
excluding a WF FEP is provided in Section 6.1. Generally, a regulatory-type screening criterion 
was examined first, followed by a screening rationale based on either a low probability or a low 
consequence criterion. In Section 6.2 are brief discussions, labeled as Screening Arguments, 
providing the rationale for excluding a FEP. A summary of screening arguments and the 
exclusion basis is provided in Section 6. 
4.2.1 Regulatory Basis for NRC Guidance 
This scientific analysis report complies with the NRCs criteria for FEP screening, given in 
10 CFR Part 63 [DIRS 156605]. The criteria that can be used to exclude a FEP from TSPA-LA 
are given in the following subsections. 
Low Probability Criterion 4.2.1.1 
The low probability criterion is explicitly stated in 10 CFR 63.114(d) [DIRS 156605]: 
Consider only events that have at least one chance in 10,000 of occurring over 
10,000 years. 
and supported by 10 CFR 63.342 [DIRS 156605]: 
DOEs performance assessments shall not include consideration of very unlikely 
features, events, or processes, i.e., those that are estimated to have less than one 
chance in 10,000 of occurring within 10,000 years of disposal. 
The low-probability criterion (i.e., very unlikely FEPs) is stated as less than one chance in 
10,000 of occurring in 10,000 years (i.e., 10-4/104 yr, or a probability of 10-8). 
Furthermore, it is stated in 10 CFR 63.342 [DIRS 156605] that: 
DOEs assessments for the human-intrusion and ground-water protection 
standards shall not include consideration of unlikely features, events, or 
April 2004 28 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
processes, or sequences of events and processes, i.e., those that are estimated to 
have less than one chance in 10 and at least one chance in 10,000 of occurring 
within 10,000 years of disposal. 
This criterion for unlikely FEPs corresponds to an annualized probability of less than 10-5, but 
greater than or equal to 10-8, which is the upper boundary for very unlikely FEPs. 
4.2.1.2 Low-Consequence Criteria 
Criteria for low-consequence screening arguments are provided in 10 CFR 63.114(e) and (f) 
[DIRS 156605], which indicates that performance assessments shall: 
(e) Provide the technical basis for either inclusion or exclusion of specific 
features, events, and processes in the performance assessment. Specific 
features, events, and processes must be evaluated in detail if the magnitude 
and time of the resulting expected radiological exposures to the reasonably 
maximally exposed individual, or radionuclide releases to the accessible 
environment, would be significantly changed by their omission. 
(f) Provide the technical basis for either inclusion or exclusion of degradation, 
deterioration, or alteration processes of engineered barriers in the performance 
assessment, including those processes that would adversely affect the 
performance of natural barriers. Degradation, deterioration, or alteration 
processes of engineered barriers must be evaluated in detail if the magnitude 
and time of the resulting radiological exposures to the reasonably maximally 
exposed individual, or radionuclide releases to the accessible environment, 
would be significantly changed by their omission. 
Furthermore, it is stated in 10 CFR 63.342 [DIRS 156605] that 
DOEs performance assessments need not evaluate the impacts resulting from any 
features, events, and processes or sequences of events and processes with a higher 
chance of occurrence if the results of the performance assessments would not be 
changed significantly. 
Because the relevant performance measures differ for different FEPs (e.g., effects on 
performance can be measured in terms of changes in concentrations, flow rates, transport times, 
or other measures as well as overall expected annual dose), there is no single quantitative test of 
significance. 
Some FEPs have a beneficial effect on the TSPA, as opposed to an adverse effect. As identified 
in 10 CFR 63.102(j) [DIRS 156605], the concept of a performance assessment includes that: 
The features, events, and processes considered in the performance assessment 
should represent a wide range of both beneficial and potentially adverse effects on 
performance (e.g., beneficial effects of radionuclide sorption; potentially adverse 
effects of fracture flow or a criticality event). Those features, events, and 
processes expected to materially affect compliance with [10 CFR] 63.113(b) or be 
April 2004 29 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
potentially adverse to performance are included, while events (event classes or 
scenario classes) that are very unlikely (less than one chance in 10,000 over 
10,000 years) can be excluded from the analysis.  
Yucca Mountain Review Plan, Final Report (NRC 2003 [DIRS 163274], Section 2.2.1), states: 
In many regulatory applications, a conservative approach can be used to decrease 
the need to collect additional information or to justify a simplified modeling 
approach. Conservative estimates for the dose to the reasonably maximally 
exposed individual may be used to demonstrate that the proposed repository 
meets U.S. Nuclear Regulatory Commission regulations and provides adequate 
protection of public health and safety. The total system performance 
assessment is a complex analysis with many parameters, and the U.S. Department 
of Energy may use conservative assumptions to simplify its approaches and data 
collection needs. However, a technical basismust be provided. 
On the basis of these statements, those FEPs that are demonstrated to have only beneficial effects 
on the radiological exposures to the reasonably maximally exposed individual, or radionuclide 
releases to the accessible environment, can be excluded on the basis of low consequence because 
they have no adverse effects on performance. 
4.2.1.3 By-Regulation Criteria 
Regulations which specify characteristics, concepts, and definitions may serve as the basis for 
exclusion of FEPs by regulation, as allowed by Yucca Mountain Review Plan, Final Report 
(NRC 2003 [DIRS 163274]). These include the characteristics, concepts and definitions 
pertaining to the reference biosphere and geologic setting and the RMEI. Also pertinent are 
characteristics, concepts, and definitions that must be considered during the FEP screening, such 
as the areal extent of the accessible environment and of the controlled area, and the spatial 
relationship to the distance from the repository to the RMEI. 
4.3 CODES AND STANDARDS 
There are no formally established codes or standards directly applicable to this analysis. 
Regulations used by the DOE to set conformance criteria are in: 
 10 CFR 63. Energy: Disposal of High-Level Radioactive Wastes in a Geologic 
Repository at Yucca Mountain, Nevada. [DIRS 156605] 
April 2004 30 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
5. ASSUMPTIONS 
5.1 GENERAL ASSUMPTIONS 
There are two general assumptions used indirectly throughout this analysis model report in 
screening WF FEPs. These assumptions are not specific to the disposition of any single FEP but 
rather are used throughout the FEP analyses contained in Section 6.2 of this report. These 
general assumptions do not need further confirmation. 
5.1.1 Consistency with Design 
Assumption: The TSPA is based on an assumption that the repository will be constructed, 
operated, and closed according to the design used as the basis for the FEP screening. 
Rationale: Unless a FEP can be excluded because of a low probability of the phenomenon ever 
occurring, the FEP screening decision is based, at least in part, on the waste package design in 
Waste Package Design Methodology Report (McKenzie 2002 [DIRS 158010]). This assumption 
is justified because a only a major change in the waste package or EBS design would require a 
reevaluation of the screening decision for FEPs dependent on design requirements. 
Confirmation Status: Confirmation is not required for this general assumption. 
Use in the Analysis: This general assumption is used throughout the report. 
5.1.2 Waste Type and Inventory 
Assumption: Only the wastes described in Initial Radionuclide Inventories (BSC 2003 [DIRS 
161961]) were considered in these analyses. 
Rationale: This assumption is based on the fact that Initial Radionuclide Inventories (BSC 2003 
[DIRS 161961]) represents the latest compilation of waste forms. 
Confirmation Status: Confirmation is not required for this general assumption. 
Use in the Analysis: This general assumption is used throughout the report. 
5.2 FEPS-SPECIFIC ASSUMPTIONS 
5.2.1 Igneous Intrusion Interacts with EBS Components (1.2.04.04.0A) 
It is conservatively assumed in addressing this FEP that the igneous intrusion would render the 
drip shields, waste packages, cladding, and waste forms in Zone 1 emplacement drifts ineffective 
as barriers, providing no further protection to the waste and that contact with magma causes the 
waste form to be distributed throughout the cooled magma leaving the waste forms chemically 
unchanged, and thus that the dissolved concentration of radioelements in water has the same 
dependency on water chemistry as for waste not contacted by magma. This assumption is used 
in Section 6.2.1 to support the screening decision to include this FEP. The assumption, since it is 
conservative with respect to the calculation of dose, does not require confirmation. 
April 2004 31 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
5.2.2 Waste Inventory (2.1.01.01.0A) 
No assumptions were used in addressing this FEP. 
5.2.3 Interactions between Colocated Waste (2.1.01.02.0A) 
No assumptions were used in addressing this FEP. 
5.2.4 Interactions Between Codisposed Waste (2.1.01.02.0B) 
No assumptions were used in addressing this FEP. 
5.2.5 Heterogeneity of Waste Inventory (2.1.01.03.0A) 
No assumptions were used in addressing this FEP. 
5.2.6 DSNF Degradation (Alteration, Dissolution, and Radionuclide Release) 
(2.1.02.01.0A) 
It is conservatively assumed in addressing this FEP that the cladding of the DSNF provides no 
retardation of the degradation of the SNF upon contact with water. This assumption is primarily 
based on observations of the extent of cladding damage to N-Reactor SNF. This assumption is 
used in Section 6.2.6 to support the screening decision to include this FEP in the TSPA-LA. The 
assumption requires no further confirmation because it represents the worst possible cladding 
condition for the DSNF. 
5.2.7 CSNF Degradation (Alteration, Dissolution, and Radionuclide Release) 
(2.1.02.02.0A) 
No assumptions were used in addressing this FEP. 
5.2.8 HLW Glass Degradation (Alteration, Dissolution, and Radionuclide Release) 
(2.1.02.03.0A) 
No assumptions were used in addressing this FEP. 
5.2.9 Alpha Recoil Enhances Dissolution (2.1.02.04.0A) 
No assumptions were used in addressing this FEP. 
5.2.10 HLW Glass Cracking (2.1.02.05.0A) 
No assumptions were used in addressing this FEP. 
5.2.11 HLW Glass Recrystallization (2.1.02.06.0A) 
No assumptions were used in addressing this FEP. 
32 ANL-WIS-MD-000009 REV 01 April 2004 Miscellaneous Waste-Form FEPs 
5.2.12 Radionuclide Release From Gap and Grain Boundaries (2.1.02.07.0A) 
No assumptions were used in addressing this FEP. 
5.2.13 Pyrophoricity from DSNF (2.1.02.08.0A) 
Any actual breach of the waste package and container holding the potentially pyrophoric 
N-Reactor SNF would probably be small enough such that exposure of the SNF to air or water 
vapor would occur slowly and thereby limit the rate of reaction, since the waste package and 
container are initially sealed with an inert gas. However, for the purpose of estimating the 
potential effect of pyrophoricity on release, it is conservatively (but non-mechanistically) 
assumed that all N-Reactor SNF-containing waste packages that breach during the regulatory 
period undergo a pyrophoric event within one TSPA time step. It is also assumed that (1) breach 
of an N-Reactor SNF-containing waste package results in a pyrophoric condition that causes the 
two adjacent waste packages to fail, (2) the various types of waste packages are evenly 
distributed throughout the repository, and (3) the probability of failure of a waste package 
containing pyrophoric (i.e., N-Reactor) SNF is the same as that of any other waste package. This 
assumption is used in Section 6.2.13 and Attachment I to provide partial support to the screening 
argument decision to exclude this FEP from the TSPA-LA. This assumption requires no 
confirmation because it conservatively represents the worst possible effect of the pyrophoric 
event on adjacent waste packages. 
5.2.14 Chemical Effects of Void Space in the Waste Package (2.1.02.09.0A) 
No assumptions were used in addressing this FEP. 
5.2.15 Organic/Cellulosic Materials in Waste (2.1.02.10.0A) 
No assumptions were used in addressing this FEP. 
5.2.16 DSNF Cladding (2.1.02.25.0A) 
It is assumed in this FEP that the DSNF cladding does not provide a retardation barrier to the 
release of radionuclides. This is based on the observation of significant amounts of damage to 
the N-Reactor SNF (over 80 percent by weight of all DSNF) cladding (Abrefah et al. 1995 
[DIRS 151125]), and the fact that there has been insufficient characterization of the various other 
types of DSNF cladding to establish it as intact for the purpose of the TSPA-LA. This 
assumption is used in Section 6.2.16 to support the designation of this FEP as excluded. This 
assumption requires no confirmation because it conservatively recommends the worst possible 
cladding condition for use in the analysis. 
5.2.17 Grouping of DSNF Waste Types into Categories (2.1.02.28.0A) 
In addressing this FEP, it is assumed that the entire DSNF inventory (with the exception of naval 
SNF) can be conservatively represented by N-Reactor SNF. The basis for this assumption is that 
the rate of degradation of the N-Reactor SNF can be shown to be higher than all but a miniscule 
fraction of the rest of the DSNF. This assumption is used in Section 6.2.17 to support the 
screening decision to include this FEP in the TSPA-LA. This assumption does not require 
April 2004 33 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
further confirmation because it represents a bounding condition for use in the analysis of all the 
DSNF types. 
5.2.18 Flammable Gas Generation from DSNF (2.1.02.29.0A) 
No assumptions were used in addressing this FEP. 
5.2.19 Internal Corrosion of Waste Packages Prior to Breach (2.1.03.06.0A) 
In addressing this FEP it is assumed that only a limited amount of water will be allowed by waste 
acceptance criteria inside a waste package. This assumption is based on the planned inert gas 
backfill of waste packages and containers. This assumption is used in Section 6.2.19 to support 
the screening decision to exclude this FEP from the TSPA-LA. It does not require confirmation 
because waste acceptance criteria do not allow for damaged waste packages and limit the amount 
of water in undamaged waste packages. 
5.2.20 Chemical Characteristics of Water in Waste Package (2.1.09.01.0B) 
No assumptions were used in addressing this FEP. 
5.2.21 Chemical Interaction with Corrosion Products (2.1.09.02.0A) 
No assumptions were used in addressing this FEP. 
5.2.22 Radionuclide Solubility, Solubility Limits, and Speciation in the Waste Form and 
EBS (2.1.09.04.0A) 
In addressing this FEP it is conservatively assumed that under repository condition no strontium 
solubility-controlling solid exists and its release is considered to be controlled by the dissolution 
rate of waste forms and the waste inventory (BSC 2003 [DIRS 163152], Section 6.18). The 
basis for this assumption is that strontium is quite soluble and it does not contribute to dose 
significantly. This assumption is used in Section 6.2.22 to support the screening argument to 
include this FEP in the TSPA. This assumption is conservative and is supported by Dissolved 
Concentration Limits of Radioactive Elements (BSC 2003 [DIRS 163152]) and so requires no 
further confirmation for its usage in this document. 
5.2.23 Reduction-Oxidation Potential in Waste Package (2.1.09.06.0A) 
No assumptions were used in addressing this FEP. 
5.2.24 Reaction Kinetics in Waste Package (2.1.09.07.0A) 
No assumptions were used in addressing this FEP. 
5.2.25 Secondary Phase Effects on Dissolved Radionuclide Concentrations (2.1.09.10.0A) 
In addressing this FEP it is assumed that the constant ratio of neptunium to uranium in the 
leachate is the result of the formation of solid solutions in the secondary uranium phases. The 
basis for this assumption is the experimental observations that the neptunium and uranium 
April 2004 34 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
behave coherently both in the release from the waste form and in the process of secondary phase 
formation, and this results in lower neptunium concentrations observed in the leachate than 
would be expected from solubility limits. This assumption is used in Section 6.2.25 to 
demonstrate that the use of the mechanistic model for the dissolution/precipitation reactions are 
conservative in terms of predicting neptunium transport, and therefore support the screening 
argument to exclude this FEP from TSPA. This assumption is supported by Dissolved 
Concentration Limits of Radioactive Elements (BSC 2003 [DIRS 163152]) and so requires no 
further confirmation for its usage in this document. 
5.2.26 Complexation in EBS (2.1.09.13.0A) 
No assumptions were used in addressing this FEP. 
5.2.27 Formation of True (Intrinsic) Colloids in EBS (2.1.09.15.0A) 
No assumptions were used in addressing this FEP. 
5.2.28 Formation of Pseudo-Colloids (Natural) in EBS (2.1.09.16.0A) 
No assumptions were used in addressing this FEP. 
5.2.29 Formation of Pseudo-Colloids (Corrosion Product) in EBS (2.1.09.17.0A) 
No assumptions were used in addressing this FEP. 
5.2.30 Formation of Microbial Colloids in EBS (2.1.09.18.0A) 
The potential influence of organic complexants from microbial metabolism on radionuclide 
transport is assumed not to affect the stability of inorganic (mineral) colloids (BSC 2003 
[DIRS 166845], Sections 5.9 and 6.3.4) (i.e., inorganic colloid stability is determined by fluid 
chemistry [ionic strength and pH]). Therefore this assumption is used in section 6.2.30 to 
support the decision to exclude this FEP on the basis of low consequence. This assumption is 
supported by the analyses in Waste Form and In-Drift Colloids-Associated Radionuclide 
Concentrations: Abstraction and Summary (BSC 2003 [DIRS 166845]) and so does not require 
further confirmation for its usage in this report. 
5.2.31 Stability of Colloids in EBS (2.1.09.23.0A) 
No assumptions were used in addressing this FEP. 
5.2.32 Formation of Colloids (Waste-Form) by Coprecipitation in EBS (2.1.09.25.0A) 
No assumptions were used in addressing this FEP. 
5.2.33 Exothermic Reactions in the EBS (2.1.11.03.0A) 
No assumptions were used in addressing this FEP. 
April 2004 35 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
5.2.34 Thermal Effects on Chemistry and Microbial Activity in the EBS (2.1.11.08.0A) 
No assumptions were used in addressing this FEP. 
5.2.35 Thermally Driven Flow (Convection) in the EBS (2.1.11.09.0B) 
It is assumed in addressing this FEP that a single mixing cell in the waste package may be used 
for determining transport out of the waste package. The basis for this assumption is the 
determination that thermal convection inside the waste package would tend to erase 
concentration gradients and distribute radionuclides uniformly throughout the package. This 
assumption is used in Section 6.2.35 to support the screening argument to exclude this FEP from 
the TSPA-LA. This assumption does not require validation because having the concentration of 
dissolved materials dispersed uniformly would slightly over-predict the transport of 
radionuclides out of the waste package, and is thus conservative. 
5.2.36 Gas Generation (CO2, CH4, and H2S) from Microbial Degradation (2.1.12.04.0A) 
In addressing this FEP it is assumed that the potential impacts of high carbon dioxide levels from 
microbial degradation on radioactive element solubilities will be small. The basis for this 
assumption is the analysis of dissolved concentrations of radioactive elements by Finch and 
Fortner (2002 [DIRS 162004]), which encompassed carbon dioxide partial pressures as high as 
1.0 to 1.5 atmospheres. This analysis showed this effect to be small. This assumption is used in 
Section 6.2.36 in support of the screening decision to exclude this FEP from the TSPA-LA based 
on low consequence. 
5.2.37 Effects of Radioactive Gases in EBS (2.1.12.07.0A) 
No assumptions were used in addressing this FEP. 
5.2.38 Radiolysis (2.1.13.01.0A) 
No assumptions were used in addressing this FEP. 
5.2.39 Radiation Damage in EBS (2.1.13.02.0A) 
No assumptions were used in addressing this FEP. 
5.2.40 Chemistry of Water Flowing into Waste Package (2.2.08.12.0B) 
No assumptions were used in addressing this FEP. 
5.2.41 Radioactive Decay and In-Growth (3.1.01.01.0A) 
No assumptions were used in addressing this FEP. 
April 2004 36 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
6. SCIENTIFIC ANALYSIS DISCUSSION 
This analysis model report addresses the 41 FEPs that have been identified as waste-form FEPs. 
These FEPs generally represent waste form and colloid formation processes that are potentially 
relevant to repository performance. FEPs representing cladding processes are addressed 
elsewhere (BSC 2003 [DIRS 165057]). 
For FEPs that are excluded based on 10 CFR Part 63 [DIRS 156605], the screening argument 
identifies the exclusion criterion (low probability [Section 4.2.1.1] or low consequence 
[Section 4.2.1.2]) and provides a short summary of the technical basis. No waste-form FEPs 
have an excluded screening decision based solely on regulatory requirements or regulatoryspecified 
conditions (Section 4.2.1.3). 
For FEPs that are included in the TSPA, the disposition describes how the FEP has been 
incorporated in the TSPA model, either directly or through an abstraction. Supporting analysis 
or model reports that provide a more detailed discussion of the implementation of included FEPs 
are also identified. 
Corroborating and supporting information used in this scientific analysis activity is listed and 
statused in Table 6-1. This report identifies all sources of corroborating or supporting 
information used in this model report. No explicit calculations using product output or technical 
information were performed as a part of this activity, so the analysis of uncertainties is not 
involved. 
April 2004 37 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 6-1. Supporting Information for Waste-Form FEPs Disposition 
Accession 
or TIC 
number 
Type of 
Reference UsedFrom Informa tion Description 
DIRS 
Number 
corroborating/ Evidence that a significant fraction of the N- 243197 151125 Abrefah, J.; Gray, W.J.; Ketner, G.L.; Marschman, Section 3.1 and 
S.C.; Pyecha, T.D.; and Thornton, T.A. 1995. K- Figures 3.1 to 3.5 supporting 
Basin Spent Nuclear Fuel Characterization Data 
Report. PNL-10778. Richland, Washington: 
Pacific Northwest National Laboratory. 
Reactor spent fuel is currently damaged and so 
will be damaged at the time of emplacement in 
the repository. 
entire corroborating T / here is a small but finite chance of spontaneous 
supporting 
248558 151226 Abrefah, J.; Huang, F.H.; Gerry, W.M.; Gray, 
W.J.; Marschman, S.C.; and Thornton, T.A. 
1999. Analysis of Ignition Testing on K-West 
Basin Fuel. PNNL-11816. Richland, 
Washington: Pacific Northwest National 
Laboratory. 
ignition of N-Reactor SNF since uranium hydride 
inclusions tend to be concentrated near the 
exposed uranium metal fuel surface of damaged 
U-metal SNF. Because the fraction of N-Reactor 
SNF with cladding damaged enough to expose 
the metallic uranium core is quantitatively 
unknown but potentially significant, essentially all 
the SNF elements can be regarded as damaged 
for the purposes of this analysis. 
corroborating/ The presence of uranium hydride inclusions are 
supporting 
Tables 4.2, 4.3 
believed to be responsible for the observed 
decreased ignition temperature observed during 
ignition testing of damaged/corroded N-Reactor 
SNF samples compared to 
unirradiated/undamaged samples. 
entire corroborating/ Draft Input - Validation of Dissolved 
supporting 
163864 MOL.20030 Adkins, H.E. 2003. "Transmittal of Draft Version 
619.0100 Concentrations, Argonne National Laboratory. of Validation of Dissolved Concentrations, 
Argonne National Laboratory." Interoffice 
memorandum from H.E. Adkins (BSC) to Y. 
Chen, June 19, 2003, 0617037708, with 
attachment. 
ASTM C 1454-00. 2000. Standard Guide for 
Pyrophoricity/Combustibility Testing in Support of 
Pyrophoricity Analyses of Metallic Uranium Spent 
Nuclear Fuel. West Conshohocken, 
Pennsylvania: American Society for Testing and 
Materials. 
248977 152779 entire corroborating AST / M Standard Guide for evaluating the 
supporting potential for pyrophoric behavior of uranium 
metal-based spent fuel. 
April 2004 38 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 6-1. Supporting Information for Waste-Form FEPs Disposition (Continued) 
Accession 
or TIC Type of 
number Reference Used Informa From tion Description 
DIRS 
Number 
240716 101565 Bruno, J.; Cera, E.; Observatio Eklund, U.-B.; Eriksen, T.E.; entire corroborating/ n that a coherent relationship exists 
and Grive, M. 1998. Some Preliminary Modeling between neptunium and uranium in spent fuel 
Results of Long Term Radiolytic Mass Balance dissolution experiments in the Spanish Nuclear 
Experiments. Presentation at the Spent Fuel Waste Program in support of the conclusion that 
Workshop 1998, Las Vegas, Nevada, May 18-20, the ratio of neptunium to uranium in the leachate 
1998. [Barcelona, Spain]: Quantisci. is equal to the ratio in the spent fuel used in 
experiments. 
supporting 
BSC (Bechtel SAIC Company) 2001. In-Drift 
Microbial Communities Model Validation 
Calculations. CAL-EBS-EV-000001 REV 00 ICN 
02. Las Vegas, Nevada: Bechtel SAIC 
Company. 
entire corroborating Calculated / microbial populations will not be 
sufficient to generate significant concentrations 
162733 MOL.20020 
supporting 107.0156 
of radionuclide-chelating organics 
BSC (Bechtel SAIC Company) 2001. Internal Section 6, Table 6- corroborating/ Support for the conclusion that partial pressures 
Pressurization Due to Fuel Rod Rupture in Waste 1 from radioactive gas will not be great enough to 
Packages. CAL-EBS-ME-000005 REV 01. Las 
Vegas, Nevada: Bechtel SAIC Company. 
153755 MOL.20010 
supporting 629.0076 
compromise waste package integrity that would 
result in a waste package breach. 
BSC (Bechtel SAIC Company) 2003. In-Package Attachment II 
Chemistry for Waste Forms. ANL-EBS-MD- 
000056 REV 00, with errata. Las Vegas, 
Nevada: Bechtel SAIC Company. 
168795 DOC.20031 corroborating/ Supports screening out gamma radiation as 
affecting radiation changes in waste 014.0005 supporting 
BSC (Bechtel SAIC Company) 2003. Clad 
Degradation  FEPs Screening Arguments. ANLWIS-
MD-000008 REV 01. Las Vegas, Nevada: 
Bechtel SAIC Company. 
165057 DOC.20040 entire corroborating Cladding F / EPs screening arguments 
322.0001 supporting 
Appendix A, Table corroborating/ WP temperature grasdients will be low at 1000 
supporting years. 
A-3 
BSC (Bechtel SAIC Company) 2004. Defense 
HLW Glass Degradation Model. ANL-EBS-MD- 
000016 REV 01 ICN 01. Las Vegas, Nevada: 
Bechtel SAIC Company. 
167619 DOC.20040 entire corroborating Summary ta / ble indicating FEPs which Model 
Report helps resolve. 223.0006 supporting 
BSC (Bechtel SAIC Company) 2003. Dissolved 
Concentration Limits of Radioactive Elements. 
ANL-WIS-MD-000010 REV 02. Las Vegas, 
Nevada: Bechtel SAIC Company. 
163152 DOC.20030 Section 6.2, Table corroborating/ Description of solubility FEP 
supporting 624.0003 6.2-1 
BSC (Bechtel SAIC Company) 2003. WF DSNF and entire corroborating/ FEPs-related model reports 
Other Waste Form Degradation Abstraction. 
163693 DOC.20030 
supporting 711.0002 
April 2004 39 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 6-1. Supporting Information for Waste-Form FEPs Disposition (Continued) 
Accession 
or TIC Type of 
number Reference Used Informa From tion Description 
DIRS 
Number 
ANL-WIS-MD-000004 REV 02. Las Vegas, 
Nevada: Bechtel SAIC Company. 
BSC (Bechtel SAIC Company) 2003. W In-Package entire corroborating F / reports related to FEPs 
Chemistry Abstraction. ANL-EBS-MD-000037 
REV 02. Las Vegas, Nevada: Bechtel SAIC 
Company. 
161962 DOC.20030 
723.0003 supporting 
BSC (Bechtel SAIC Company) 2003. Waste 
Form and In-Drift Colloids-Associated 
Radionuclide Concentrations: Abstraction and 
Summary. MDL-EBS-PA-000004 REV 00 ICN 
01. Las Vegas, Nevada: Bechtel SAIC 
Company. 
entire corroborating Identificati / on of responsible input waste form 
model reports 
166845 DOC.20031 
supporting 222.0012 
BSC (Bechtel SAIC Company) 2003. Total 
System Performance Assessment-License 
Application Methods and Approach. TDR-WISPA-
000006 REV 00 ICN 01. Las Vegas, Nevada: 
Bechtel SAIC Company. 
BSC (Bechtel SAIC Company) 2004. FEPs 
Screening of Processes and Issues in Drip Shield 
and Waste Package Degradation. ANL-EBS-PA- 
000002 REV 02. Las Vegas, Nevada: Bechtel 
SAIC Company. 
BSC (Bechtel SAIC Company) 2004. Features, 
Events, and Processes in UZ Flow and Transport. 
ANL-NBS-MD-000001 REV 02. Las Vegas, 
Nevada: Bechtel SAIC Company. 
corroborating/ Logic schematic for FEPs identification Figure 3.2-2 166296 DOC.20031 
supporting 215.0001 
Analysis of clogging of waste package breached FEP 2.1.03.10.0A corroborating/ 162200 DOC.20040 
zones by corrosion products. supporting 408.0001 
UZ FEPs document; atmospheric transport of 168782 DOC.20040 corroborating/ 
supporting 
Section 6.3.9 
gases FEP discussion. 331.0002 
BSC (Bechtel SAIC Company) 2004. Engineered Section 6.10 
Barrier System: Physical and Chemical 
Environment Model. ANL-EBS-MD-000033 REV 
02. Las Vegas, Nevada: Bechtel SAIC Company. 
167461 DOC.20040 corroborating/ Repository dust on the top of the waste 
packages is composed of non-corrosive 212.0004 supporting 
minerals. 
BSC 2003. Dike/Drift Interactions. MDL-MGR-GS- entire corroborating/ Due to reactions and decreasing pressures, 
000005 REV 00. Las Vegas, Nevada: Bechtel 
SAIC Company. 
165923 DOC.20031 
H2O, SO2 and CO2 released from the intrusion 
event will not make it to zone 2 in sufficient 
quantities to effect waste packages and waste 
022.0001. supporting 
forms in Zone 2 emplacement drifts. 
BSC 2004. Igneous Intrusion Impacts on Waste 
Packages and Waste Forms. MDL-EBS-GScorroborating/ 
Corrosive gases released during dike intrusion 
will not enter Zone 2 emplacement drifts in 
Sections 5.4 and 167654 MOL.20040 
supporting 6.5.2.2 414.0095 
April 2004 40 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 6-1. Supporting Information for Waste-Form FEPs Disposition (Continued) 
Accession 
or TIC Type of 
number Reference Used Informa From tion Description 
sufficient quantities to effect waste packages and 
DIRS 
Number 
000002 REV 01D. Las Vegas, Nevada: Bechtel 
SAIC Company. waste forms. 
entire corroborating Np is inco / rporated into uranium-based alteration 
phases in approximately the same 
168411 MOL.20040 Buck, E.C.; McNamara, B.K.; Douglas, M.; and 
202.0085. Hanson, B.D. 2003. Possible Incorporation of 
Neptunium in Uranyl (VI) Alteration Phases. 
PNNL-14277. Richland, Washington: Pacific 
Northwest National Laboratory. 
supporting 
concentrations as presented in the mother 
solutions. 
entire corroborating Examin / ation of corrosion products from spent 
fuel drip tests by electron energy loss 
100388 240702 Buck, Fi E.C.; nch, R.J.; Finn, P.A.; and Bates, 
supporting J.K. 1998. "Retention of Neptunium in Uranyl 
Alteration Phases Formed During Spent Fuel 
Corrosion." Scientific Basis for Nuclear Waste 
Management XXI, Symposium held September 
28-October 3, 1997, Davos, Switzerland. 
McKinley, I.G. and McCombie, C., eds. 506, 87- 
94. Warrendale, Pennsylvania: Materials 
Research Society. 
spectroscopy (EELS) analyses in a transmission 
electron microscope. 
entire corroborating Predicti / on that the substitutions Pu6+ in U6+ and 
(Np5+, Pu5+)in U6+ are likely to occur in most 
235501 100389 Burns, P.C.; Ewing, R.C.; and Miller, M.L. 1997. 
"Incorporation Mechanisms of Actinide Elements 
into the Structures of U6+ Phases Formed During 
the Oxidation of Spent Nuclear Fuel." Journal of 
Nuclear Materials, 245, ([1]), 1-9. Amsterdam, 
The Netherlands: North-Holland. 
supporting 
U6+ structures based on an analysis of the 
crystal-chemical properties of the U-O bond, Np- 
O bond, and Pu-O bond. 
Canori, G.F. and Leitner, M.M. 2003. Project 
Requirements Document. TER-MGR-MD-000001 
REV 02. Las Vegas, Nevada: Bechtel SAIC 
Company. 
166275 DOC.20031 entire corroborating Project requ / irements document 
222.0006 supporting 
entire corroborating Analys / es of the relative release rates of 247873 161997 Chen, Y. 2001. "An Empirical Np Solubility 
Model Based on Spent Fuel Dissolution 
Experiments." "Back to the Future - Managing 
the Back End of the Nuclear Fuel Cycle to Create 
a More Secure Energy Future," Proceedings of 
the 9th International High-Level Radioactive 
Waste Management Conference (IHLRWM), Las 
Vegas, Nevada, April 29-May 3, 2001. La 
Grange Park, Illinois: American Nuclear Society. 
neptunium and uranium in the fuel degradation 
experiments in support of a close correlation 
between neptunium and uranium concentrations 
supporting 
in the leachates. 
corroborating/ Analyses of the relative release rates of pp. 385-397 254363 162709 Chen, Y. 2003. "Using Reactive Transport 
Modeling to Evaluate the Source Term at Yucca neptunium and uranium in the fuel degradation supporting 
April 2004 41 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 6-1. Supporting Information for Waste-Form FEPs Disposition (Continued) 
Accession 
or TIC Type of 
number Reference Used Informa From tion Description 
experiments in support of a close correlation 
DIRS 
Number 
Mountain." Computers & Geosciences, 29, ([3]), 
385-397. [New York, New York]: Pergamon. between neptunium and uranium concentrations 
in the leachates 
248663 161996 Chen, Y.; Loch, A.R.; Wolery, T.J.; Steinborn, 
T.L.; Brady, P.V.; and Stockman, C.T. 2002. 
"Solubility Evaluation for Yucca Mountain TSPASR." 
Scientific Basis for Nuclear Waste 
Management XXV, Symposium held November 
26-29, 2001, Boston, Massachusetts. McGrail, 
B.P. and Cragnolino, G.A., eds. 713, 775-782. 
Warrendale, Pennsylvania: Materials Research 
Society. 
pp.775-782 corroborating Anal / yses of the relative release rates of 
neptunium and uranium in the fuel degradation 
experiments in support of a close correlation 
between neptunium and uranium concentrations 
supporting 
in the leachates 
corroborating/ Analyses of the relative release rates of pp. 471-488 246426 145441 Chen, Y.; Siegmann, E.; Mattie, P.; NcNeish, J.; 
Sevougian, S.D.; and Andrews, R. 1999. "A 
Mechanistic Model of Spent Fuel Dissolution, 
Secondary Mineral Precipitation, and Np 
Release." Scientific Basis for Nuclear Waste 
Management XXII, Symposium held November 
30-December 4, 1998, Boston, Massachusetts, 
U.S.A. Wronkiewicz, D.J. and Lee, J.H., eds. 
556, 471-478. Warrendale, Pennsylvania: 
Materials Research Society. 
neptunium and uranium in the fuel degradation 
experiments in support of a close correlation 
between neptunium and uranium concentrations 
supporting 
in the leachates 
142613 225504 Croff, A.G. 1980. A User's Manual for the entire corroborating Radionucli / de inventory calculational code 
ORIGEN2 Computer Code. ORNL/TM-7175. 
Oak Ridge, Tennessee: Oak Ridge National 
Laboratory. 
supporting 
Croff, A.G. 1980. ORIGEN2 - A Revised and 
Updated Version of the Oak Ridge Isotope 
Generation and Depletion Code. ORNL-5621. 
Oak Ridge, Tennessee: Oak Ridge National 
Laboratory. 
entire corroborating Radionucli / de inventory values 101554 NNA.19910 
supporting 326.0094 
CRWMS M&O 1997. Waste Package Design 
Basis Events. BBA000000-01717-0200-00037 
REV 00. Las Vegas, Nevada: CRWMS M&O. 
section 7.2.2.7, 
Table 7.2.2.7-1 
corroborating/ Release of radioactive gases to the interior of the 
waste package will be too low to compromise 
100264 MOL.19971 
supporting 006.0075 
waste package integrity. 
100362 MOL.19981 CRWMS M&O 1998. "Waste Form Degradation, Table 5-1 and 
008.0006 
corroborating/ EDA-II design factors used in thermal analysis 
Radionuclide Mobilization, and Transport Through Figure 5-1 
the Engineered Barrier System." Chapter 6 of 
supporting 
April 2004 42 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 6-1. Supporting Information for Waste-Form FEPs Disposition (Continued) 
Accession 
or TIC Type of 
number Reference Used Informa From tion Description 
DIRS 
Number 
Total System Performance Assessment-Viability 
Assessment (TSPA-VA) Analyses Technical 
Basis Document. B00000000-01717-4301-00006 
REV 01. Las Vegas, Nevada: CRWMS M&O. 
Page A6-29 Figure corroborating/ Dissolution Rates for High-Level Waste, Metallic 
supporting Carbide, and Ceramic Spent Nuclear Fuel 
A6-1 
entire corroborating Supp / ort for mechanistic neptunium reactivitytransport 
model 
supporting 
Table 5-1 and 
Figure 5-1 
corroborating/ Support for judgement that preferential 
condensation will not be important. 
supporting 
CRWMS M&O 1998. Immobilized in Ceramic. 
Volume II of Report on Intact and Degraded 
Criticality for Selected Plutonium Waste Forms in 
a Geologic Repository. BBA000000-01717-5705- 
00020 REV 01. Las Vegas, Nevada: CRWMS 
M&O. 
105649 MOL.19981 entire corroborating Supp / ort for the conclusion that DSNF codisposal 
and high-level radioactive waste glass waste 217.0112 supporting 
packages are expected to be generally cooler 
than commercial SNF packages at the time of 
disposal for an approximate waste package 
center-to-center separation distance of ~5 m. 
CRWMS M&O 1998. Waste Form Degradation 
and Radionuclide Mobilization Expert Elicitation 
Project. Las Vegas, Nevada: CRWMS M&O. 
entire corroborating Radioactivit / y values in support of the conclusion 
that four radionuclides will dominate the total 
100374 MOL.19980 
supporting 804.0099 
repository mass inventory (238U, 235U, 239Pu, 
and 236U). 
CRWMS M&O 1999. License Application Design Section 4.4 
Selection Report. B00000000-01717-4600-00123 
REV 01. Las Vegas, Nevada: CRWMS M&O. 
corroborating/ Support for the conclusion that preferential 
condensation may not be important with the line 
103955 MOL.19990 
supporting 528.0303 
loading of the EDA-II design 
CRWMS M&O 2000. Design Analysis for UCF 
Waste Packages. ANL-UDC-MD-000001 REV 
00. Las Vegas, Nevada: CRWMS M&O. 
entire corroborating Investigati / on of the sensitivity of internal waste 
package pressures presuming fuel rods rupture, 
144128 MOL.20000 
supporting 526.0336 
causing release of gases from the fuel matrix to 
the waste package interior 
CRWMS M&O 2000. Evaluation of Codisposal 
Viability for HEU Oxide (Shippingport PWR) DOE- Summary 
Owned Fuel. TDR-EDC-NU-000003 REV 00. 
Las Vegas, Nevada: CRWMS M&O. 
corroborating/ Characteristics of DOE spent fuel; HEU SNF Executive 147651 MOL.20000 
supporting 227.0240 
CRWMS M&O 2000. Evaluation of Codisposal 
Viability for UZrH (TRIGA) DOE-Owned Fuel. 
TDR-EDC-NU-000001 REV 00. Las Vegas, 
Nevada: CRWMS M&O. 
147650 MOL.20000 Executive 
Summary 
corroborating/ characteristics of DOE spent fuels; UZrH 
207.0689 supporting 
April 2004 43 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 6-1. Supporting Information for Waste-Form FEPs Disposition (Continued) 
Accession 
or TIC Type of 
number Reference Used Informa From tion Description 
DIRS 
Number 
CRWMS M&O 2000. In-Drift Microbial 
Communities. ANL-EBS-MD-000038 REV 00 
ICN 01. Las Vegas, Nevada: CRWMS M&O. 
entire corroborating Microbi / al respiration is potentially able to 151561 MOL.20001 
increase in-drift CO2 production by two orders of supporting 213.0066 
magnitude. 
Section 6.3 and 7.2 corroborating/ Potential sources for the release of gases from 
the fuel matrix to the waste package interior. 
supporting 
CRWMS M&O 2000. Initial Cladding Condition. Section 6.3.3 
ANL-EBS-MD-000048 REV 00 ICN 01. Las 
Vegas, Nevada: CRWMS M&O. 
corroborating/ Fission and radioactive-gas diffusion rates 151659 MOL.20001 
supporting 002.0145 
CRWMS M&O 2000. Relative Contribution of 
Individual Radionuclides to Inhalation and 
Ingestion Dose. CAL-WIS-MD-000002 REV 00. 
Las Vegas, Nevada: CRWMS M&O. 
153595 MOL.20010 entire corroborating Supp / ort for the conclusion that the DSNF 
codisposal and high-level radioactive waste 103.0211 supporting 
glass waste packages are expected to be 
generally cooler 
entire corroborating Supp / ort for the contention that omission of 
DSNF cladding degradation as a system-wide 
supporting 
feature from the performance assessment 
models will not significantly change the 
interpretation of expected dose 
CRWMS M&O 2000. Unsaturated Zone Flow and Sections 3.3, 3.6 corroborating/ Support for the conclusion that any gas 
Transport Model Process Model Report. TDR- produced in the repository will tend to flow 
NBS-HS-000002 REV 00 ICN 02. Las Vegas, through the fracture complex with some gas 
Nevada: CRWMS M&O. diverted to the matrix at unit interfaces or where 
fracture networks pinch out. These flow paths 
151940 MOL.20000 
supporting 831.0280 
will disperse flammable gas concentrations to 
levels below the ignition point. 
DOE (U.S. Department of Energy) 2000. DOE 
Spent Nuclear Fuel Grouping in Support of 
Criticality, DBE, TSPA-LA. DOE/SNF/REP-0046, 
Rev. 0. Idaho Falls, Idaho: U.S. Department of 
Energy, Idaho Operations Office. 
248046 118968 entire corroborating/ Classification of DSNF types 
supporting 
247664 149432 DOE (U.S. Department of Energy) 1997. 
EM/RW Repository Task Team Report: Grouping 
Method to Minimize Testing for Repository 
Emplacement of DOE SNF. DOE/SNF/REP-008, 
Rev. 0. Idaho Falls, Idaho: U.S. Department of 
Energy. 
corroborating/ Potential for pyrophoric reactions: oxide fuels, 
metal fuels, and carbide fuels. 
pp. 20-21 
supporting 
April 2004 44 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 6-1. Supporting Information for Waste-Form FEPs Disposition (Continued) 
Accession 
or TIC Type of 
number Reference Used Informa From tion Description 
DIRS 
Number 
page 21 corroborating Urani / um-metal spent fuel is more chemically 
reactive than other spent fuel. 
supporting 
DOE (U.S. Department of Energy) 1998. Total Sections 3.5.3 and corroborating/ HLW glass is not a major contributor to total 
System Performance Assessment. Volume 3 of 5.5.3 activity. 
Viability Assessment of a Repository at Yucca 
Mountain. DOE/RW-0508. Washington, D.C.: 
U.S. Department of Energy, Office of Civilian 
Radioactive Waste Management. 
100550 MOL.19981 
007.0030 supporting 
DOE (U.S. Department of Energy) 1998. Update Appendix A, 
to Assessment of Product Output Disposal in 
Unsaturated Tuff of Spent Nuclear Fuel and High- 
Level Waste Owned by U.S. Department of 
Energy. DOE/SNF/REP-015. Washington, D.C.: 
U.S. Department of Energy. 
corroborating/ Broken Peach-Bottom Core-1 pellets have an 
estimated cladding failure rate of 35% 
243859 122980 
Section 6.9, p. 41 supporting 
Table 1-1, p.1-8 
and Appendix B, p. supporting 
corroborating/ 1.663 MTHM of Peach Bottom Core 1 pellets 
contained in 103 waste packages, will be 
disposed in the repository. 14 
Appendix B, p. 14 corroborating/ 24.667 MTHM of high-integrity fuel (contained in 
supporting 545 waste packages) from the Fort St. Vrain and 
and Table 1-1, 
Peach-Bottom Core 2 reactors p.1-8 
entire corroborating HLW / glass will constitute a large volume fraction 
of the total volume of waste in the repository 
supporting 
DOE 2002. Waste Acceptance System 
Requirements Document. DOE/RW-0351, REV 
04. Washington D.C.: U.S. Department of Energy, and 4.3.17 
Office of Civilian Radioactive Waste 
Management. 
Section 4.2.11, 
4.3.1.B, 4.3.7.B, 
corroborating/ Restriction on organic and chemically reactive 
materails in the waste canisters emplaced in the 
158873 MOL.20020 
supporting 326.0056 
repository 
DOE (U.S. Department of Energy) 2002 DOE 
Spent Nuclear Fuel Information in Support of 
TSPA-SR. DOE/SNF/REP-0047, Rev. 2. 
[Washington, D.C.; U.S. Department of Energy, 
Office of Environmental Management. 
Section 1 and 
Appendix D 
corroborating/ The only fuel waste capable of producing 252089 158405 
flammable gasses is U/Th carbide DSNF waste. 
Only a small fraction(~1%) of the DSNF fuel 
contains U/Th carbide. Since the carbide spent 
supporting 
fuels will amount to less than 0.04% of the waste 
packages,and all DNSF waste will be widely 
dispersed throughout the repository, any 
flammable gas produced will be widely 
dispersed. 
April 2004 45 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 6-1. Supporting Information for Waste-Form FEPs Disposition (Continued) 
Accession 
or TIC Type of 
number Reference Used Informa From tion Description 
DIRS 
Number 
entire corroborating Char / acteristics of DSNF waste to be disposed in 
the repository. 
supporting 
Finch, R.J. and Fortner, J.A. 2002. Dissolved 
Concentration Limits: Neptunium in Alteration 
Phases, A Progress Report. Argonne, Illinois: 
Argonne National Laboratory. 
162004 MOL.20031 entire corroborating Analys / is of dissolved concentrations of 
119.0180. radioactive elements which includes high carbon supporting 
dioxide partial pressures. 
corroborating/ Experimental results on neptunium incorporation 
into dehydrated schoepite and estimated that the 
Section 6 248663 161979 Finch, R.J.; Fortner, J.A.; Buck, E.C.; and Wolf, 
S.F. 2002. "Neptunium Incorporation into 
Uranium(VI) Compounds Formed During 
Aqueous Corrosion of Neptunium-Bearing 
Uranium Oxides." Scientific Basis for Nuclear 
Waste Management XXV, Symposium held 
November 26-29, 2001, Boston, Massachusetts. 
McGrail, B.P. and Cragnolino, G.A., eds. 713, 
647-654. Warrendale, Pennsylvania: Materials 
Research Society. 
supporting 
amount of neptunium incorporated in dehydrated 
schoepite can be as high as 2 percent of the 
host solid based on EELS measurement. 
Foster, A.R. and Wright, R.L., Jr. 1973. Basic 
Nuclear Engineering. 2nd Edition. Boston, 
Massachusetts: Allyn and Bacon. 
corroborating/ Basic Nuclear Engineering textbook - it requires 
a minimum displacement energy, Ed, of 
p. 296-297 247586 144061 
supporting 
approximately 25 ev to eject the struck atom 
from its lattice site. displacement of atoms 
corresponds to a threshold event, P(T) is 
modeled as a Heaviside step function. 
163911 MOL.20030 Friese, J.I. 2003. "Transmittal of Draft Version of entire corroborating/ Draft Input - Existing Evidence for the Fate of 
617.0128 supporting Neptunium in the Yucca Mountain Repository. Existing Evidence for the Fate of Neptunium in 
the Yucca Mountain Repository, Pacific 
Northwest National Laboratory." Interoffice 
memorandum from J.I. Friese (BSC) to Y. Chen, 
June 17, 2003, 0613037658, with attachment. 
162634 MOL.20040 Friese, J.I.; Buck, E.C.; McNamara, B.K.; Hanson, Entire corroborating/ Distribution of neptunium in the repository. 
303.0023 supporting B.D.; and Marschman, S.C. 2003. Existing 
Evidence for the Fate of Neptunium in the Yucca 
Mountain Repository. PNNL-14307. Richland, 
Washington: Pacific Northwest National 
Laboratory. 
Gray, W.J. 1988. Effect of Surface Oxidation, 
Alpha Radiolysis, and Salt Brine Composition on 
entire corroborating Laboratory / experiments show little efffect of 
alpha radiolysis on the leaching behavior of 
168471 HQX.19881 
supporting 021.0025 
April 2004 46 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 6-1. Supporting Information for Waste-Form FEPs Disposition (Continued) 
Accession 
or TIC Type of 
number Reference Used Informa From tion Description 
spent fuel and UO2 samples. 
DIRS 
Number 
Spent Fuel and UO2 Leaching Performance. 
PNL/SRP-6689. Richland, Washington: Pacific 
Northwest Laboratory. 
Gray, W.J. and Einziger, R.E. 1998. Initial 
Results from Dissolution Rate Testing of NReactor 
Spent Fuel Over a Range of Potential 
Geologic Repository Aqueous Conditions. 
DOE/SNF/REP-022, Rev. 0. Washington, D.C.: 
U.S. Department of Energy. 
corroborating/ Rapid corrosion rate of the metallic uranium 
uranium that is the matrix of N-Reactor SSNF 
Section 4 245380 109691 
supporting 
compared with other SNF 
Kaplan, I. 1963. Nuclear Physics. 2nd edition. p. 307 
Reading, Massachusetts: Addison-Wesley 
Publishing. 
corroborating/ Commercial fuel cladding will remain intact and 
should substantially reduce alpha dose rates to 
247872 149367 
supporting 
groundwater 
248909 152482 Lam, P.S.; Sindelar, R.L.; and Peacock, H.B. Jr. Section 8 
1997. Vapor Corrosion of Aluminum Cladding 
Alloys and Aluminum-Uranium Fuel Materials in 
Storage Environments (U). WSRC-TR-97-0120. 
Aiken, South Carolina: Westinghouse Savannah 
River Company. 
corroborating/ Aluminum-clad uranium metal DSNF does not 
have the amount of damage/corrosion (and supporting 
consequent extent of hydriding and potential for 
pyrophoricity) nor the propensity to form uranium 
hydrides as a result of corrosion, shown by the 
N-Reactor SNF 
Lawrence, L.A. 1999. K Basin Spent Nuclear Fuel Section 5.1 and 
Characterization. HNF-3534, Rev. 0. Richland, 
Washington: Duke Engineering & Services 
Hanford, Inc. 
corroborating/ Estimate of approximately 33% (by means of 
visual survey) for the fraction of failed N-Reactor 
168475 MOL.20040 
supporting Table 1 325.0053 
spent fuel assemblies for both the K-East and KWest 
basins. 
243883 142133 Lederer, C.M. and Shirley, V.S., eds. 1978. 
Table of Isotopes. 7th Edition. New York, New 
York: John Wiley & Sons. 
corroborating/ Table of isotopes p. 1464 
supporting 
corroborating/ Uranium hydride inclusions tend to be 248059 149429 Marschman, S.C.; Pyecha, T.D.; and Abrefah, J. Section 3.4.2 
1997. Metallographic Examination of Damaged N 
Reactor Spent Nuclear Fuel Element 
SFEC5,4378. PNNL-11438. Richland, 
Washington: Pacific Northwest National 
Laboratory. 
concentrated near the exposed uranium metal supporting 
fuel surface of damaged N-Reactor SNF 
McKenzie, D.G., IV. 2002. Waste Package 
Design Methodology Report. TDR-MGR-MD- 
000006 REV 02. Las Vegas, Nevada: Bechtel 
SAIC Company. 
158010 MOL.20020 entire corroborating/ Waste package and EBS design description. 
404.0085 supporting 
April 2004 47 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 6-1. Supporting Information for Waste-Form FEPs Disposition (Continued) 
Accession 
or TIC Type of 
number Reference Used Informa From tion Description 
DIRS 
Number 
164527 MO0307SEPFEPS4.000. LA FEP List. Submittal LA FEP list entire corroborating/ 
date: 07/31/2003. supporting 
corroborating/ Yucca Mountain Review Plan for screening 
FEPs 
Section 2.2.1.2 254568 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. 
supporting 
Many of the he p.8 corroborating/ avy radionuclides emit alpha 
particles with energies greater than 4.0 MeV 
233705 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. 
supporting 
entire corroborating/ significant portion of the radionuclides in the 
expected YMP inventory are alpha emitters. 
supporting 
Propp, W.A. 1998. Graphite Oxidation 
Thermodynamics/Reactions. DOE/SNF/REP- 
018, Rev. 0. Idaho Falls, Idaho: U.S. 
Department of Energy. 
enntire corroborating that hy / drocarbon production is not a problem 
with carbide fuels 
247663 149395 
supporting 
252663 161925 Quinones, J.; Grambow, B.; Loida, A.; and 
Geckeis, H. 1996. "Coprecipitation Phenomena 
During Spent Fuel Dissolution. Part 1: 
Experimental Procedure and Initial Results on 
Trivalent Ion Behaviour." Journal of Nuclear 
Materials, 238, ([1]), 38-43. Amsterdam, The 
Netherlands: Elsevier. 
Support for the conclusion that the large 
discrepancy in Np solubility is due to the 
incorporation of actinides into secondary 
corroborating/ 
supporting 
pp. 38-43 
uranium phases, that was not accounted for in 
conventional solubility calculations. 
Rechard, R.P., ed. 1995. Methodology and 
Results. Volume 2 of Performance Assessment 11-24) 
of the Direct Disposal in Unsaturated Tuff of 
Spent Nuclear Fuel and High-Level Waste Owned 
by U.S. Department of Energy. SAND94-2563/2. 
Albuquerque, New Mexico: Sandia National 
Laboratories. 
237102 101084 Section 11.3.1, p. corroborating/ Estimate of the amount of damaged cladding in 
DSNF. For approximately 80% of the DSNF that supporting 
is N-Reactor SNF, the cladding will be 
significantly damaged at the time of 
emplacement in their canisters. Up to 50 percent 
may be already perforated 
Reilly, M.A. 1998. Spent Nuclear Fuel Project 
Technical Databook. HNF-SD-SNF-TI-015, Rev. 
5. Richland, Washington: Fluor Daniel Hanford. 
p.30 corroborating/ It is possible that additional U-hydride will form in 
N-Reactor SNF during interim storage. 
247814 149433 
supporting 
April 2004 48 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 6-1. Supporting Information for Waste-Form FEPs Disposition (Continued) 
Accession 
or TIC Type of 
number Reference Used Informa From tion Description 
DIRS 
Number 
247643 149368 Sanchez, L.C.; Rath, J.S.; Aguilar, R.; Trellue, 
H.R.; Cochrane, K.; Taylor, L.L.; and Wilson, J.R. 
1998. Nuclear Dynamics Consequence Analysis 
(NDCA) for the Disposal of Spent Nuclear Fuel in 
an Underground Geologic Repository. 
DOE/SNF/REP-0033, REV 0. Three Volumes. 
Washington, D.C.: U.S. Department of Energy. 
entire corroborating Radionucli / de inventories for final waste forms 
supporting 
Schulz, W.W. 1972. Descripti Shear-Leach Processing of entire corroborating/ on and analysis of the pyrophoric 
N-Reactor Fuel--Cladding Fires. ARH-2351. 
Richland, Washington: Atlantic Richfield Hanford. operations. 
243079 159406 
behavior of N-Reactor SNF during reprocessing supporting 
entire corroborating Supp / ort for the conclusion that water will not 
intrude until gamma and beta emitters have 
246431 113368 Shoesmith, D.W. and Sunder, S. 1992. "The 
Prediction of Nuclear Fuel (UO2) Dissolution 
Rates Under Waste Disposal Conditions." 
Journal of Nuclear Materials, 190, 20-35. 
Amsterdam, The Netherlands: Elsevier. 
supporting 
decayed to very low concentrations 
Simmons, A.M. 2004. Yucca Mountain Site 
Description. TDR-CRW-GS-000001 REV 02. Two 
volumes. Las Vegas, Nevada: Bechtel SAIC 
Company. 
corroborating/ Description of gas flow in the unsaturated zone. Section 7.9.1.8 166960 DOC.20040 
supporting 120.0004 
corroborating/ Description of two pyrophoric events concerning 
uranium metal-based nuclear fuel. 
pp 89-93 165881 103316 Solbrig, C.W.; Krsul, J.R.; and Olsen, D.N. 1994. 
supporting "Pyrophoricity of Uranium in Long-Term Storage 
Environments." Proceedings of the Topical 
Meeting on DOE Spent Nuclear FuelChallenges 
and Initiatives, Salt Lake City, Utah, December 
13-16, 1994. Pages 89-93. La Grange Park, 
Illinois: American Nuclear Society. 
246920 143815 Sunder, S. and Shoesmith, D.W. 1991. 
Chemistry of UO2 Fuel Dissolution in Relation to 
the Disposal of Used Nuclear Fuel. AECL- 
10395. Pinawa, Manitoba, Canada: Whiteshell 
Laboratories, Atomic Energy of Canada Limited. 
entire corroborating Supp / ort for the conclusion that water will not 
intrude (i.e., the waste container will not fail) until supporting 
gamma and beta emitters have decayed to very 
low concentrations, and that strong gamma and 
beta fields associated with the used fuel will 
decrease by a factor >103 in the first few 
hundred years after disposal. 
Radiation fields decrease dramatically over the 
first 1,000 years after the fuel is out-of-reactor 
corroborating/ 
supporting 
p. 67 246914 143860 Sunder, S.; Shoesmith, D.W.; and Miller, N.H. 
1997. "Oxidation and Dissolution of Nuclear Fuel 
(UO2) by the Products of the Alpha Radiolysis of 
April 2004 49 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 6-1. Supporting Information for Waste-Form FEPs Disposition (Continued) 
Accession 
or TIC Type of 
number Reference Used Informa From tion Description 
DIRS 
Number 
Water." Journal of Nuclear Materials, 244, 66- 
74. Amsterdam, The Netherlands: Elsevier. 
Swanson, J.L.; Bray, L.A Kjarmo, H.E.; Ryan, 
J.L.; Matsuzaki, C.L.; Pitman, S.G.; Haberman, 
J.H. Laboratory Studies of Shear/Leach 
Processing of Zircaloy Clad Metallic Uranium 
Reactor Fuel. PNL-5708, December 1985. 
corroborating/ Analysis of the cause of the ignition and burning 
of sheared metallic uranium spent fuel elements 
Sections 7. and 168468 MOL.20040 
supporting 10. 325.0049 
in the dissolvers at West Valley. The most 
probable cause for the ignition was concluded to 
be the sensitization of fuel element weld beads 
by the very acidic dissolver solutions 
125082 MOL.19981 Thornton, T.A. 1998. "HPPP Issue 1; 
019.0001 
entire corroborating Supp / ort for the conclusion that overall repository 
performance is very insensitive to the DSNF Preliminary TSPA for a Pyrophoric Event 
Involving N-Reactor SNF Waste Packages." 
Interoffice correspondence from T.A. Thornton 
(CRWMS M&O) to J.S. Clouet, S.S. Sareen, and 
D. Stahl, September 21, 1998, LV.WP.TAT.09/98- 
179, with attachment. 
supporting 
degradation rate 
entire corroborating DSNF / contribution to postclosure dose is 
negligible 
107796 MOL.19980 Thornton, T.A. 1998. "HPPP Issue 1; 
730.0142 supporting Preliminary TSPA for N-Reactor SNF Waste 
Packages." Interoffice correspondence from T.A. 
Thornton (CRWMS M&O) to J.S. Clouet, S.S. 
Sareen, and D. Stahl, June 1, 1998, 
LV.WP.TAT.05/98-111, with attachment. 
corroborating/ Support for the conclusion that the large 
discrepancy in Np solubility is due to the 
pp. 194-200 243085 113466 Werme, L.O. and Spahiu, K. 1998. "Direct 
Disposal of Spent Nuclear Fuel: Comparison 
Between Experimental and Modelled Actinide 
Solubiluties in Natural Waters." Journal of Alloys 
and Compounds, 271-273, 194-200. Lausanne, 
Switzerland: Elsevier. 
supporting 
incorporation of actinides into secondary 
uranium phases, that was not accounted for in 
conventional solubility calculations. 
Wick, O.J., ed. 1980. Plutonium Handbook: A 
Guide to the Technology. Volumes I & II. La 
Grange Park, Illinois: American Nuclear Society. 
corroborating/ Density of pure plutonium metal Section 7.1 248002 143651 
supporting 
100191 NNA.19940 Radi Wilson, M.L.; Gauthier, J.H.; Barnard, R.W.; Barr, entire corroborating onuclid / e inventory values 
112.0123 supporting G.E.; Dockery, H.A.; Dunn, E.; Eaton, R.R.; 
Guerin, D.C.; Lu, N.; Martinez, M.J.; Nilson, R.; 
Rautman, C.A.; Robey, T.H.; Ross, B.; Ryder, 
E.E.; Schenker, A.R.; Shannon, S.A.; Skinner, 
L.H.; Halsey, W.G.; Gansemer, J.D.; Lewis, L.C.; 
Lamont, A.D.; Triay, I.R.; Meijer, A.; and Morris, 
April 2004 50 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table 6-1. Supporting Information for Waste-Form FEPs Disposition (Continued) 
Accession 
or TIC Type of 
number Reference Used Informa From tion Description 
DIRS 
Number 
D.E. 1994. Total-System Performance 
Assessment for Yucca Mountain  SNL Second 
Iteration (TSPA-1993). SAND93-2675. 
Executive Summary and two volumes. 
Albuquerque, New Mexico: Sandia National 
Laboratories. 
corroborating/ Description of a prophoric event concerning 
metallic uranium fuel elements. 
255359 165884 Wood, D.; Snowden, S.; Howe Jr, H.; Thomas, L.; pp. 113-115 
Moon, D.; Gregg, H.; Miller, P. 1994. "Regarding 
the chemistry of metallic uranium stored in steel 
drums." Journal of Nuclear Materials, 209, 113- 
115. New York, New York: Elsevier. 
supporting 
156713 MOL.20010 YMP (Yucca Mountain Site Characterization 
322.049, 
Table 1.3-1 and p. corroborating/ Relative amounts of CSNF,HLW glass, and 
supporting DSNF to be disposed in the repository. Project) 2001. Yucca Mountain Site 
Characterization Project Requirements Document 
(YMP-RD). YMP/CM-0025, Rev. 4, DCN 02. Las 
Vegas, Nevada: Yucca Mountain Site 
Characterization Office. 
1.3-6 
MOL.20011 
107.0002 
April 2004 51 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
6.1 APPROACHES 
The NRC requires consideration and evaluation of FEPs as part of the performance assessment 
activities. More specifically, NRC regulations allow the exclusion of FEPs from the TSPA if 
they can be shown to be of low probability or of low consequence. The specified criteria can be 
summarized in the form of two FEP screening statements as follows. 
1. The event has at least one chance in 10,000 of occurring over 10,000 years (see 10 
CFR 63.114(d) [DIRS 156605]). 
2. The magnitude and time of the resulting radiological exposure to the RMEI, or 
radionuclide release to the accessible environment, would be significantly changed by 
its omission (see 10 CFR 63.114 (e and f) [DIRS 156605]). 
Additionally, the acceptance criteria in Yucca Mountain Review Plan, Final Report (NRC 2003 
[DIRS 163274], Section 2.2.1.2.1.3) calls for evaluating the FEPs based on the regulations. This 
criterion can be summarized in the form of a third FEP screening statement. 
3. The FEP is not excluded by regulation. 
If there are affirmative conditions for all three screening criteria, the FEP is included in the 
TSPA-LA model. If there is a negating condition in any of the three screening criteria, the FEP 
is excluded from the TSPA-LA model. 
Evaluation of the FEPs against these screening statements may be done in any order. In practice, 
by-regulation criteria were examined first, and then either low probability or low consequence 
criteria were examined. FEPs that were retained on one criterion (e.g., regulatory guidance) 
were also considered against the other criteria (probability and consequence). Consequently, the 
application of the analysts judgment regarding the order in which to apply the criteria does not 
affect the final decision. Allowing the analyst to choose the most appropriate order to apply the 
criteria prevents needless work, such as developing quantitative probability arguments for lowconsequence 
events or complex consequence models for low-probability events. 
Regardless of the specific approach chosen to perform the screening, the screening process is in 
essence a comparison of the FEP against the criteria specified in Section 4.2. The outcome of 
the screening is independent of the particular methodology selected to perform the screening. 
Alternative interpretations of data as they pertain directly to the FEPs screening are provided in 
the sections for each FEP, as discussed below. The FEPs screening decisions may also rely on 
the results of analyses performed and documented as separate activities. Alternative approaches 
related to separate activities and analyses are addressed in the specific analysis model reports for 
those analyses and are not discussed in this report. 
6.2 WASTE-FORM FEPS SCREENING AND ANALYSES 
This section addresses the 41 waste-form FEPs identified for TSPA-LA (see Section 1.2.1 and 
Table 1.2-1). Each FEP is screened for inclusion or exclusion in TSPA-LA, using the criteria 
outlined in Section 4.2.1 based on requirements defined by the NRC (10 CFR 63.114(d)(e)(f) 
[DIRS 156605]). 
April 2004 52 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
The screening decisions, screening arguments, and TSPA dispositions for the waste-form FEPs 
in some cases reference and utilize model reports or analysis reports currently in process or for 
which there are planned ICNs. If the final version of these model reports and analysis reports are 
inconsistent with the analyses in this report, this report will be modified per the AP-SIII.9Q 
checking and review process and per the requirements of the controlled version of the technical 
work plan at the time of the revisions. 
6.2.1 Igneous Intrusion Interacts with EBS Components 
FEP Number: 
1.2.04.04.0A 
FEP Description: An igneous intrusion in the form of a dike occurs through the repository, 
intersecting the repository drifts. Magma, pyroclastics, and volcanic gases enter the drift and 
interact with the EBS components including drip shields, waste packages, pallet, and invert. 
This leads to accelerated drip shield and waste package failure (e.g. attack by magmatic volatiles, 
damage by flowing or fragmented magma, or thermal effects) and dissolution or volatilization of 
waste. 
Descriptor Phrases: 
Igneous intrusion (drip shield damage) 
Igneous intrusion (waste package damage) 
Igneous intrusion (cladding damage) 
Chemical effects of igneous intrusion in EBS (solubility) 
Thermal effects of igneous intrusion in EBS 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: The effects of an igneous intrusion into the repository are implemented in 
the TSPA through the delineation of two zones. Zone 1 encompasses portions of the drifts where 
EBS components are directly contacted by magma. Zone 2 encompasses the remainder of the 
drifts, where EBS components are not contacted by magma, but are exposed to high temperature 
and magmatic gases. Details of how an igneous intrusion interacts with EBS components is 
described in Igneous Intrusion Impacts on Waste Packages and Waste Forms (BSC 2004 [DIRS 
167654]. A summary of the effects of igneous intrusion is presented here. 
Damage and Thermal Effects in Zone 1 
Igneous Intrusion Impacts on Waste Packages and Waste Forms (BSC 2004 [DIRS 167654], 
Section 6.5.1) indicates the igneous intrusion would render the drip shields, waste packages, and 
cladding in the Zone 1 emplacement drifts ineffective so they would provide no further 
protection to the waste. The exposed waste form would then be distributed throughout the cooled 
magma, though exactly how is unknown (Section 5.2.1 of this analysis report). Spent fuel is a 
refractory material and considered to be chemically unaffected by exposure to the intruding 
April 2004 53 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
magma. Likewise, waste glass degradation behavior is considered likely to be minimally 
affected by any consequent devitrification (BSC 2004 [DIRS 167619], Section 7.5.4). As the 
waste forms will be distributed throughout the magma body, the temperature of those waste 
forms can be considered the same as the temperature of the cooling magma. These values are 
presented in DTN: MO0402SPAHWCIG.002 [DIRS 168762]. 
Damage and Thermal Effects in Zone 2 
The impacts of magmatic heat conduction from Zone 1 to Zone 2 emplacement drifts were 
modeled by numerical simulations of nonsteady state heat conduction with radial flow in Igneous 
Intrusion Impacts on Waste Packages and Waste Forms (BSC 2004 [DIRS 167654], Section 
6.5.2.1 and Attachment I). The simulation modeling of heat conduction away from the intruded 
drifts demonstrates that the initial 1,150C temperature reduces to about 30C at the center of the 
drift in about 30 years, and the maximum temperature rise expected in the Zone 2 emplacement 
drifts is less than 1C. The drift-rock in Zone 2 provides an effective thermal insulation barrier 
to the impacts of high temperature of intruded magma and there would not be any impact of 
igneous heat on the waste packages, drip shield, or cladding in Zone 2 emplacement drifts. 
An analysis of gas flow from Zone 1 to Zone 2 emplacement drifts is detailed in Dike/Drift 
Interactions (BSC 2003 [DIRS 165923]). This report shows that due to reactions and decreasing 
pressures, H2O, SO2 and CO2 released from the intrusion event will not make it to Zone 2 in 
sufficient quantities to affect waste packages and waste forms in Zone 2 emplacement drifts. 
Effects of other corrosive gases is discussed in Sections 5.4 and 6.5.2.2 of Igneous Intrusion 
Impacts on Waste Packages and Waste Forms (BSC 2004 [DIRS 167654]). These sections also 
conclude that corrosive gases released during dike intrusion will not enter Zone 2 emplacement 
drifts in sufficient quantities to affect waste packages and waste forms. 
Chemical Effects 
After postintrusive magma cooling and reversion to normal in-drift environmental conditions, 
seepage water flowing through the intruded basalt will be chemically altered by reaction with 
basaltic minerals. As demonstrated in Igneous Intrusion Impacts on Waste Packages and Waste 
Forms (BSC 2004 [DIRS 167654]), this alteration involves significant changes to the pH and 
ionic strength of the solution. The pH and fCO2 values obtained through reaction of seepage 
water with basalt should be used in conjunction with Dissolved Concentration Limits of 
Radioactive Elements (BSC 2003 [DIRS 163152]) to describe the solubility of the waste forms. 
Because waste forms are conservatively considered chemically unchanged (BSC 2004 
[DIRS 167654], Section 6.5.1.2), the dissolved concentration of radioelements in water would 
have the same dependency on water chemistry in regions influenced by an igneous intrusion as 
those regions that are not. Thus, the solubility tables given in Dissolved Concentration Limits of 
Radioactive Elements (BSC 2003 [DIRS 163152]) should be used to determine concentrations of 
radionuclides in basalt-filled drifts the same way they are used in magma-free drifts. 
Supporting Documents: 
BSC 2004 [DIRS 167654] 
April 2004 54 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Related FEPs: 
N/A 
6.2.2 Waste Inventory 
FEP Number: 
2.1.01.01.0A 
FEP Description: The waste inventory includes all potential sources of radiotoxicity and 
chemical toxicity. It consists of the radionuclide inventory (typically in units of curies), by 
specific isotope, of anticipated radionuclides in the waste, and the nonradionuclide inventory 
(typically in units of density or concentration) that consists of both physical, e.g., CSNF, DNSF, 
and HLW, and chemical waste constituents. The radionuclide composition of the waste will vary 
due to initial enrichment, burnup, the number of fuel assemblies per container, and the decay 
time subsequent to discharge of the fuel from the reactor. 
Also consider that the fuel types, matrices, radionuclide mixes, and nonradionuclide inventories 
in DSNF and HLW may differ from CSNF. Additional waste types should be considered if they 
are proposed for disposal at Yucca Mountain. 
Descriptor Phrases: 
Radionuclide inventory 
Nonradionuclide inventory 
CSNF inventory 
DSNF inventory 
HLW glass inventory 
Gap inventory 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: Modeling the waste inventory in TSPA-LA can be divided into two tasks. 
The first is to select those radionuclides important to dose calculations. The second is to 
determine which radionuclides are present in each type of waste and in what quantity. 
The radionuclides of importance to dose calculations were assessed in Radionuclide Screening 
(BSC 2002 [DIRS 160059]). This information was provided to Initial Radionuclide Inventories 
(BSC 2003 [DIRS 161961]) and is reproduced in Table 13 of that report. The table presents 
those radionuclides relevant to the nominal, igneous disruption, and human intrusion scenario 
classes and further breaks down the list into the radionuclides of importance for the periods from 
100 to 10,000 years (for TSPA-LA calculations) and from 10,000 to 1,000,000 years (for EIS 
calculations). Nonradioactive waste is not modeled. 20 radionuclides of interest were identified 
from the TSPA-LA screening. The separate groundwater protection standard required that 228Ra 
be added to the list. The precursors of the screened-in radionuclides were also required, which 
added another seven radionuclides. As a result, 28 isotopes were modeled in the TSPA: 227Ac, 
ANL-WIS-MD-000009 REV 01 April 2004 55 Miscellaneous Waste-Form FEPs 
241 
228
Am, 243Am, 14C, 245Cm, 135Cs, 137Cs, 129I, 237Np, 231Pa, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 226Ra, 
Ra, 90Sr, 99Tc, 229Th, 230Th, 232Th, 232U, 233U, 234U, 235U, 236U, and 238U (BSC 2003 [DIRS 
161961], Table 16). Four additional radionuclides were added to the inventory to retain the 
capability to do 1-million year EIS calculations: 36Cl, 210Pb, 79Se, and 126Sn (BSC 2003 [DIRS 
161961], Table 16). 
Nominal average inventories of radionuclides for each type of waste are documented in Initial 
Radionuclide Inventories (BSC 2003 [DIRS 161961]). In the analysis, three types of uncertainty 
are applied to the nominal inventories: computational methods and nuclear data, completeness 
of records, and future decisions. Application of these uncertainties resulted in weighted average 
grams per package for the 32 radionuclides and 3 waste types (CSNF, DSNF, and HLW). 
Table 21 of Initial Radionuclide Inventories (BSC 2003 [DIRS 161961]) lists these values. The 
CSNF inventory for 14C includes minor contributions (18 percent) from activation of CSNF 
assembly hardware outside the cladding. Additional contributions from activated mineral 
deposits (crud) were found to be insignificant and were not considered as part of the inventory 
(BSC 2003 [DIRS 161961]). 
The weighted average waste inventory values (grams per package for each radionuclide for each 
waste type) are input to GoldSim for use in the TSPA-LA model. The Initial Radionuclide 
Inventories (BSC 2003 [DIRS 161961]) also provides GoldSim with radionuclide decay rates. 
As the drip shield, waste package, cladding and waste form are subjected to degradation 
processes or disruptive events, the radionuclides can be mobilized and transported. 
Supporting Documents: 
BSC 2003 [DIRS 161961] 
Related FEPs: 
3.1.01.01.0A, Radioactive decay and ingrowth 
2.1.01.03.0A, Heterogeneity of waste inventory 
6.2.3 Interactions between Colocated Waste 
FEP Number: 
2.1.01.02.0A 
FEP Description: Colocation refers to the disposal of CSNF, DSNF, HLW, and possibly other 
wastes in close proximity within the repository. Colocation might affect thermal outputs, 
chemical interactions, or radionuclide mobilization. 
Descriptor Phrases: 
Colocated waste (chemical interactions between waste packages) 
Colocated waste (thermal interactions, preferential condensation) 
Other waste types (intermediate-level radioactive waste, low-level radioactive waste, and 
nonradioactive waste) 
Screening Decision: 
Excluded (low consequence) 
April 2004 56 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Screening Argument: The following waste forms will be part of the initial license application 
submittal (BSC 2003 [DIRS 161961]): 
 CSNF from boiling-water and pressurized-water reactors 
 DSNF (including naval SNF), approximately 80 percent by weight of which is the 
N-Reactor SNF currently stored at Hanford 
 HLW in the form of glass logs in stainless steel canisters. 
Regulatory restrictions will prevent colocation of these waste forms with low- and intermediatelevel 
nuclear waste, toxic chemical wastes, or wastes disposed by future human activity. These 
waste forms will be placed in specially designed waste packages with all of the HLW and some 
of the DSNF form designated for codisposal. Present plans call for five glass logs and one 
DSNF canister to be placed in each codisposal package (DOE 2002 [DIRS 158405], Section 9). 
The codisposal packages will be colocated randomly within an array comprised predominantly of 
CSNF waste packages, and will contain a relatively small fraction of the total waste. 
The codisposal waste packages are expected to be generally cooler (CRWMS M&0 2000 
[DIRS 147650]; CRWMS M&0 2000 [DIRS 147651]) than CSNF waste packages at the time of 
disposal. Preferential condensation due to this colocated waste could possibly occur under the 
dripshield just above the relatively cooler codisposal (DSNF and HLW) waste packages. 
Condensate droplets that would eventually fall, preferentially onto the codisposal waste 
packages, could increase glass degradation rates, DSNF clad degradation, and dissolution, thus 
releasing HLW-DSNF radionuclides. Preferential condensation may be less important with the 
line loading of the EDA-II design (CRWMS M&O 1999 [DIRS 103955], Table 5-3; 
CRWMS M&O 1998 [DIRS 100362], Table 5-1 and Figure 5-1), which is designed to keep the 
waste packages at nearly equal temperatures. Moreover, it was determined in Section 6.10 of 
Engineered Barrier System: Physical and Chemical Environment Model (BSC 2004 
[DIRS 167461], Tables 6.10-7 and 6.10-8) that brines formed by deliquescence in the repository 
dust on the top of the waste packages have pH and water chemistry (particularly pH) 
characteristics that are not problematic for the codisposal waste forms. This is because the HLW 
glass (BSC 2004 [DIRS 167619], Section 8.1) degradation model has only a slight dependence 
on pH in the range of the deliquescence brines, and the DSNF degradation model (BSC 2003 
[DIRS 163693], Section 8) has no pH dependence. Nevertheless, since no credit is taken for 
DSNF cladding, and sensitivity calculations indicate dose is not sensitive to degradation of HLW 
glasses (BSC 2003 [DIRS 168796], Sections 3.3.6 and 3.5) or release of DSNF (BSC 2003 
[DIRS 163693], Section 6.4), preferential condensation due to colocation of waste is excluded 
based on low consequence. 
The nature of flow through fractured, unsaturated host rock and a small, but sufficient, separation 
of neighboring waste packages mean that any chemical interactions between colocated waste 
packages would be unlikely and of negligible consequence. Unsaturated flow in the drift is 
largely vertical and colocated packages in the same disposal elevation would not interact 
geochemically. Hence, significant changes in the rates of waste package degradation and 
radionuclide mobilization due to chemical reactions between colocated waste packages are not 
April 2004 57 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
expected and chemical interactivity between colocated waste packages can therefore be excluded 
based on low consequence. 
The possible effects of colocation with low- and intermediate-level radioactive waste, or toxic 
chemical wastes are moot. These waste types are not part of the license application design. In 
10 CFR 63.73 [DIRS 156605], the NRC requires prompt notification if there are significant 
deviations from the design criteria and design bases stated in the license application. Any 
significant deviations with the potential to affect repository performance will be evaluated, and 
corrected, as needed. This ensures a low consequence (it is unlikely that there will be significant 
effects from non-design waste) in the event that the design is not followed. Therefore, the 
effects of colocation with other waste types are excluded based on low consequence. 
TSPA Disposition: 
N/A 
Supporting Documents: 
BSC 2003 [DIRS 163693] 
Related FEPs: 
2.1.01.01.0A, Waste inventory 
2.1.01.02.0B, Interactions between codisposed waste 
2.1.01.03.0A, Heterogeneity of waste inventory 
2.1.02.01.0A, DSNF degradation (alteration, dissolution, and radionuclide release) 
2.1.02.03.0A, HLW glass degradation (alteration, dissolution, and radionuclide release) 
6.2.4 Interactions Between Codisposed Waste 
FEP Number: 
2.1.01.02.0B 
FEP Description: Codisposal refers to the disposal of different waste types within the same 
waste package. Codisposal might affect chemical interactions or radionuclide mobilization. At 
Yucca Mountain, the DSNF will be combined with HLW canisters within a waste package. This 
codisposal with HLW within a waste package is unique to the DSNF and does not apply to the 
CSNF placement within waste packages. 
The DSNF will be contained within canisters that will be placed within the waste packages. The 
use of canisters within the waste package is not typical of the CSNF placement within waste 
packages. Also, some DSNF waste packages may contain only DSNF canisters, while others 
may contain both DSNF and HLW canisters. 
Descriptor Phrases: 
Codisposed waste (chemical interactions between DSNF and HLW) 
Effects of HLW pour canisters 
Effects of DSNF high integrity canisters 
Screening Decision: 
Included 
April 2004 58 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Screening Argument: 
N/A 
TSPA Disposition: Codisposal issues include chemical interactions between the two degrading 
waste forms and the corroding HLW glass and DSNF canisters that occur through the solution 
chemistry within a breached waste package. The TSPA-LA will not take credit for the HLW 
glass pour canisters or the DSNF canisters or cladding as barriers to radionuclide release, but will 
account for the effects of container corrosion on the in-package chemistry. Different codisposal 
waste packages will have different amounts of DSNF and HLW glass, and thus different relative 
surface areas accessible to water contact. The HLW glass release calculation depends on the 
surface area of glass contacted by water (BSC 2004 [DIRS 167619], Section 8.2.1), but the SNF 
degradation model assumes complete dissolution in one TSPA time step and is thus not sensitive 
to exposed surface area (CRWMS M&O 1998 [DIRS 100362], Tables 6-29 and A6-12). The 
geometric surface area used to represent all HLW glass in the HLW glass model (BSC 2004 
[DIRS 167619], Section 8.2.1) is based on a weighted average of canister sizes. Neither the 
HLW glass degradation model nor the radionuclide inventory model distinguishes the contents of 
codisposal packages. 
Codisposal waste packages are specially designed waste packages that will contain the HLW 
glass and DSNF waste forms. These packages are expected to be generally cooler 
(CRWMS M&O 1998 [DIRS 105649]; CRWMS M&O 2000 [DIRS 153595]) than CSNF 
packages. Current plans call for most codisposal waste packages to contain five glass logs and 
one DSNF canister (DOE 2002 [DIRS 158405], Section 9), but some codisposal waste packages 
will contain two HLW glass canisters, and one potential configuration will contain five 
(Hanford) HLW glass canisters with no DSNF (BSC 2003 [DIRS 161961], Table 17). The 
codisposal packages will be colocated randomly within an array comprised predominantly of 
CSNF waste packages, and will contain a relatively small fraction of the total waste. Interactions 
of condensed water or seepage with DSNF and HLW glass in a breached codisposal waste 
package were simulated in In-Package Chemistry Abstraction (BSC 2003 [DIRS 161962]), 
which uses degradation models for DSNF (BSC 2003 [DIRS 163693] and HLW glass (BSC 
2004 [DIRS 167619]) with degradation models for other waste package components. Condensed 
and seepage water compositions are used to calculate the range of water chemistries (e.g., pH) 
that could occur in a breached codisposal waste package. Degradation of HLW glass tends to 
generate high pH solutions (BSC 2003 [DIRS 161962], Section 6.6.2), while DSNF degradation 
generates neutral pH solutions (BSC 2003 [DIRS 161962], Section 6.6.2). Those calculations 
indicate that the combined effects of degrading pour canisters, DSNF, and other waste package 
materials will moderate the pH values attained in condensed water and seepage water (BSC 2003 
[DIRS 161962]). 
The resulting in-package water chemistry is provided to TSPA-LA as part of In-Package 
Chemistry Abstraction (BSC 2003 [DIRS 161962]). The abstraction provides TSPA-LA 
parameters for the pH, Eh, ionic strength, and the total carbonate, chloride, and fluoride 
concentrations under condensed water and seepage conditions in a codisposal package. The 
DSNF release rate model has no explicit dependence on water chemistry effects and will not be 
affected by changes in the solution chemistry that occur due to chemical interactions between 
codisposal packages or between waste forms within a package (BSC 2003 [DIRS 163693]). 
April 2004 59 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
The glass degradation rate model does contain an explicit pH dependence and will be affected by 
changes in the solution pH within a codisposal package (BSC 2003 [DIRS 161962]. The glass 
model does not depend on other properties of the solution. 
Thermal effects and preferential condensation due to codisposed waste are discussed as part of 
FEP 2.1.01.02.0A. 
Supporting Documents: 
BSC 2003 [DIRS 161962] 
BSC 2003 [DIRS 163693] 
CRWMS M&O 1998 [DIRS 105649] 
CRWMS M&O 2000 [DIRS 153595] 
BSC 2004 [DIRS 167619] 
Related FEPs: 
2.1.01.02.0A 
6.2.5 Heterogeneity of Waste Inventory 
FEP Number: 
2.1.01.03.0A 
FEP Description: Commercial spent nuclear fuel (CSNF), DOE-owned spent nuclear fuel 
(DSNF), and high-level waste (HLW) shipped to the repository may 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 may vary due to 
initial uranium enrichment, possible plutonium enrichment, and burnup of the fuel, among other 
factors. The physical state within the waste form may also vary. For example, damaged fuel 
pellets or extremely high-burnup fuels may have greater surface area exposed to any water 
penetrating a waste package than undamaged, low burnup spent fuel. Given these potential 
differences in isotopic composition and physical condition, the mass of radionuclides available 
for transport may vary significantly among waste packages. 
The different physical (structure, geometry), chemical, and radiological properties of the many 
forms of CSNF, DSNF, and HLW could result in differences in the corrosion and alteration rates 
of the waste-package composition. This could affect repository chemistry, breach times, 
dissolution rates, and availability of radionuclides for transport. 
Descriptor Phrases: 
Variation in radionuclide activity per waste package 
CSNF waste package configurations 
DSNF and HLW codisposed waste package configurations 
Spatial heterogeneity in-package (waste form, cladding, waste type) 
Screening Decision: 
Included 
April 2004 60 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Screening Argument: 
N/A 
TSPA Disposition: As discussed in Initial Radionuclide Inventories (DOE 2003 
[DIRS 161961]) the repository waste types are quite heterogeneous in type (SNF versus glass) 
and in inventory per package. CSNF, DSNF, and HLW glass shipped to the repository will 
contain quantities of radionuclides that will vary from waste package to waste package, canister 
to canister, and fuel assembly to fuel assembly. The different physical, chemical, and 
radiological properties of the various CSNF, DSNF, and HLW glass waste forms could result in 
differences in their corrosion rates. This heterogeneity is represented in TSPA-LA by sampling 
from distributions for radionuclide inventory (for release calculations). However, for postclosure 
TSPA, the only simulations that approach the regulatory dose limit in 10,000 years are those 
where many packages breach. With many packages breached, the heterogeneity of the 
inventory, while included, is of minor importance and is characterized with the uncertainty 
parameters for the average CSNF, DSNF, and HLW glass radionuclide inventory in average 
CSNF and codisposal packages. 
As discussed in Defense HLW Glass Degradation Model (BSC 2004 [DIRS 167619]), one effect 
of the heterogeneity of the waste inventory is variations in the compositions of waste glasses 
made to immobilize specific wastes at different DOE sites. The effect of waste glass 
compositions on the calculated degradation rate is taken into account through the range of values 
of the model parameter kE. Ranges for the values of kE in acidic and alkaline solutions are 
selected based on the results of laboratory tests with glasses that provide a wide range of 
compositions that bounds the range of concentrations of key glass components in HLW glasses, 
such as aluminum (BSC 2004 [DIRS 167619], Sections 6.5.4.2 and 6.5.2). The glass 
degradation model accounts for the heterogeneity of the waste inventory through the range of 
parameter values. The range of glass degradation rates calculated using the glass degradation 
model developed in Defense HLW Glass Degradation Model (BSC 2004 [DIRS 167619]) can be 
used with the average radionuclide concentrations for the entire HLW inventory. 
Supporting Documents: 
BSC 2004 [DIRS 167619] 
BSC 2003 [DIRS 161961] 
Related FEPs: 
2.1.01.01.0A, Waste Inventory 
6.2.6 DSNF Degradation (Alteration, Dissolution, and Radionuclide Release) 
FEP Number: 
2.1.02.01.0A 
FEP Description: DSNF to be disposed of in Yucca Mountain has a variety of fuel types that 
include metallic uranium fuels; oxide and mixed oxide fuels; Three Mile Island rubble; and 
heterogeneous fuels such as UAlx, U-ZrHx, and graphite fuels. In general, the composition and 
structure of these spent nuclear fuels are significantly different from the commercial spent 
nuclear fuel (CSNF), and degradation, alteration, and dissolution may be different from the 
April 2004 61 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
CSNF degradation. Processes to be considered in this FEP include alteration and dissolution of 
the various DSNF waste forms, phase separation, oxidation of spent fuels, selective leaching, and 
the effects of the high-integrity can (HIC) on DSNF degradation. 
Descriptor Phrases: 
DSNF inventory characteristics (surface area, fuel type) 
Time-dependent THC characteristics in-package (temperature, pH, Eh, water 
concentration) 
Degradation (corrosion/oxidation) of DSNF fuel matrix 
Alteration of DSNF fuel matrix 
Release from DSNF fuel matrix (congruent dissolution) 
Release from DSNF fuel matrix (phase separation) 
Release from DSNF fuel matrix (selective leaching) 
Degradation (corrosion) of DSNF non-fuel metals 
Release from DSNF (non-fuel metals) 
N-Reactor uranium metal fuel 
Pu ceramic waste 
Naval fuel 
Spatial heterogeneity of DSNF degradation 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: DSNF to be disposed of in Yucca Mountain is composed of a variety of fuel 
types that are not similar to the CSNF disposed waste. The various DSNF types have been 
classified into eleven groups for the purpose of TSPA-LA analyses (DOE 2000 [DIRS 118968]). 
Section 6.3 of DSNF and Other Waste Form Degradation Abstraction (BSC 2003 
[DIRS 163693]) describes the development of upper-limit, conservative, and best estimate 
degradation models for each of 11 DSNF waste groups for TSPA-LA. These degradation 
models are summarized in Table 6 of DSNF and Other Waste Form Degradation Abstraction 
(BSC 2003 [DIRS 163693]). 
The largest single DSNF type is the N-Reactor SNF, which comprises approximately 84 percent 
by weight of the total DSNF. When group-specific degradation models applicable to individual 
waste packages are not required, the N-Reactor SNF uranium metal (Group 7) models are 
adopted as a surrogate for all DSNF groups, except for the naval SNF (Group 1) (BSC 2003 
[DIRS 163693]). The CSNF degradation model (BSC 2004 [DIRS 167321]) is used as the 
surrogate for naval SNF. For all groups (except naval DSNF), the upper-limit model produces 
complete dissolution of the waste form during a single code time step upon exposure of the waste 
form to groundwater (BSC 2003 [DIRS 163693], Section 7). 
In addition to being the largest DSNF inventory type by weight, the N-Reactor SNF degradation 
models can be applied to the other DSNF groups (with the exception of Naval fuel) because a 
significant fraction of the N-Reactor SNF will be damaged at the time of emplacement in their 
62 ANL-WIS-MD-000009 REV 01 April 2004 Miscellaneous Waste-Form FEPs 
canisters and because the N-Reactor SNF degradation modelpredicts dissolution rates greater 
than most other groups. The two groups with potentially greater dissolution rates than N-Reactor 
SNF are mixed-carbide-fissile fuel waste particles in a nongraphite matrix (Group 3) and the 
immobilized ceramic Pu disposal waste form. Group 3 waste would not serve as an appropriate 
DSNF degradation surrogate because the total inventory of the Group 3 SNF is less than 0.001 
percent of DSNF waste (DOE 2002 [DIRS 158405], Appendix D). For the immobilized ceramic 
Pu disposal waste form, the degradation model is either the instantaneous release of all the Pu 
from the waste package upon postclosure contact with groundwater, a ceramic dissolution rate, 
or ten times the ceramic dissolution rate. The models do not take credit for the glass-matrix 
coverage of surface areas on the Pu ceramic disks. Although the large majority of the excess 
defense plutonium was originally meant to be disposed in the repository as this can-in-canister 
form, the DOE Office of Environmental Management has recently decided to use the majority of 
the excess defense plutonium as MOX fuel in commercial reactors as the preferred disposition 
path rather than immobilization in a ceramic waste form (67 FR 19432 [DIRS 162618]). 
Up to 50 percent of the DSNF cladding may be already perforated (Rechard 1995 
[DIRS 101084], Section 11.3.1, p. 11-24). Because the cladding integrity of most DSNF will not 
be extensively characterized, the TSPA takes no credit for delayed degradation of the DSNF due 
to DSNF cladding or delayed release of radionuclide by the DSNF cladding (Section 5.2.6 of this 
analysis report). Also, the TSPA takes no credit for DSNF canister integrity (i.e., once the waste 
package has leaked or failed, it is conservatively assumed that the DSNF is directly exposed to 
the water or air of the repository environment). 
An upper-limit degradation model is used for DSNF in TSPA-LA because preliminary analyses 
(Thornton 1998 [DIRS 125082]); Thornton 1998 [DIRS 107796]) have shown that overall 
repository performance is not sensitive to the DSNF degradation rate. This model yields 
complete dissolution of DSNF waste forms during a single code time-step upon exposure of the 
waste form to groundwater. The DSNF waste is a small percent of the total waste to be disposed 
and the waste stream plume is dominated by CSNF degradation even when upper-limit DSNF 
degradation models are used. The formulation of the conservative and best-estimate models is 
described in DSNF and Other Waste Form Degradation Abstraction (BSC 2003 [DIRS 163693], 
Section 6.2.7). 
Supporting Documents: 
BSC 2003 [DIRS 163693] 
Related FEPs: 
2.1.01.02.0A, Interactions between colocated waste 
2.1.01.02.0B, Interactions between codisposed waste 
6.2.7 CSNF Degradation (Alteration, Dissolution, and Radionuclide Release) 
FEP Number: 
2.1.02.02.0A 
FEP Description: Alteration of the original mineralogy of the commercial spent nuclear fuel 
(CSNF) (under wet or dry conditions) and dissolution of the uranium-oxide matrix can influence 
April 2004 63 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
the mobilization of radionuclides. The degradation of UO2 could be affected by a number of 
variables, such as surface area, burnup, temperature, overall solution electrochemical potential 
(Eh), pH, and especially solutions containing significant concentrations of calcium, sodium, 
carbonate and silicate ions, as well as availability of organic complexing materials. In turn, these 
water properties are affected by the alteration of the cladding and matrix. 
Descriptor Phrases: 
CSNF inventory characteristics (surface area, burnup) 
Time-dependent THC characteristics in-package (temperature, pH, Eh, water 
concentration) 
Degradation of CSNF uranium oxide fuel matrix 
Alteration of CSNF uranium oxide fuel matrix 
Release from CSNF fuel matrix (congruent dissolution) 
Release from CSNF fuel matrix (phase change) 
Release from CSNF fuel matrix (selective leaching) 
Degradation (corrosion) of CSNF nonfuel metals 
Release from CSNF (nonfuel metals) 
Spatial heterogeneity of CSNF degradation 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: As described in CSNF Waste Form Degradation: Summary Abstraction 
(BSC 2004 [DIRS 167321]), the base case CSNF model is designed to provide the fractional 
release rate of radionuclides (Fi) when the CSNF matrix is dissolved or otherwise altered upon 
exposure to water or humid air. The CSNF alteration and dissolution processes that are 
considered are described in Section 6.2 of CSNF Waste Form Degradation: Summary 
Abstraction (BSC 2004 [DIRS 167321]). The mathematical form of the model is described in 
Section 1 of CSNF Waste Form Degradation: Summary Abstraction (BSC 2004 [DIRS 
167321]). It has six model parameters (A, a0, a1, a2, a3, and a4) and four independent variables 
(absolute temperature, pCO3, pO2, and pH). The effects of surface area are included through the 
effective specific surface area parameter (A) as described in CSNF Waste Form Degradation: 
Summary Abstraction (BSC 2004 [DIRS 167321], Section 6.4.1.5). Effects of other factors, 
including burnup, secondary phase formation, and selective or congruent release, are discussed 
in CSNF Waste Form Degradation: Summary Abstraction (BSC 2004 [DIRS 167321], Section 
6.2.2.3). The determination of the model parameter values over the independent variables factor 
space of interest is described in CSNF Waste Form Degradation: Summary Abstraction (BSC 
2004 [DIRS 167321], Section 6.4.1). 
In TSPA-LA, the CSNF degradation model parameters are used along with the in-package 
chemistry to calculate a CSNF degradation rate at each model time step. The degradation model 
parameters are determined by sampling from parameter distributions that incorporate 
uncertainty, while the in-package chemistry variables are calculated as part of the TSPA 
simulation. 
April 2004 64 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Supporting Documents: 
BSC 2004 [DIRS 167321] 
Related FEPs: 
2.1.01.02.0A, Interactions between colocated waste 
2.1.01.02.0B, Interactions between codisposed waste 
6.2.8 HLW Glass Degradation (Alteration, Dissolution, and Radionuclide Release) 
FEP Number: 
2.1.02.03.0A 
FEP Description: Glass waste forms are thermodynamically unstable over long time periods 
and will alter on contact with water. Radionuclides can be mobilized from the glass waste by a 
variety of processes, including degradation and alteration of the glass, phase separation, 
congruent dissolution, precipitation of silicates, coprecipitation of other minerals including iron 
corrosion products, and selective leaching. 
Descriptor Phrases: 
HLW glass inventory characteristics (surface area, composition) 
Time-dependent THC characteristics in-package (temperature, pH, water concentration) 
Degradation (corrosion) of HLW glass 
Alteration of HLW glass 
Release from HLW glass (congruent dissolution) 
Release from HLW glass (phase separation) 
Release from HLW glass (selective leaching) 
Release from HLW glass (precipitation of silicates and other minerals) 
Spatial heterogeneity of HLW glass degradation 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: The rate expression for glass degradation provided in Defense HLW Glass 
Degradation Model (BSC 2004 [DIRS 167619]) was developed to address this FEP. The model 
addresses degradation of glass exposed to humid air or dripping water, and glass immersed in 
water. The glass degradation rate is calculated as a function of pH and temperature by using the 
following equation: rate G = kE 10(pH) exp(-Ea/RT) (BSC 2004 [DIRS 167619], Section 6.5.2 and 
Equation 9). This rate accounts for the combined effects of water diffusion, ion exchange, and 
hydrolysis processes that lead to glass degradation when contacted by water. Explicit 
dependencies are given for variables tracked in TSPA calculations, including pH, temperature, 
and relative humidity (rate G is defined as zero for relative humidities less than 44 percent). The 
effects of environmental processes specific to the disposal conditions that affect glass 
degradation, such as water condensation and dripping, are captured in the model by the range of 
values of the kE term. The results of tests in which those processes occur are used to determine 
bounding model parameter values. The range of kE also accounts for the range of glass 
April 2004 65 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
durabilities (compositions) and the evolution of the solution chemistry contacting the glass. 
Although the rate equation does not depend on time, variables used in the modelpH, 
temperature, and relative humiditydepend on time; therefore, those values must be obtained 
from other models. The release of radionuclides from HLW glass is modeled to be congruent 
with other glass matrix components. The release rates of boron measured in laboratory tests are 
used to determine the matrix degradation rate. The radionuclide release rate is calculated as the 
product of the glass degradation rate, the exposed surface area, and the radionuclide inventory. 
The model report (BSC 2004 [DIRS 167619]) provides equations for calculating the glass 
degradation rate and the surface area; the radionuclide inventory is provided by another model. 
Surrogate glasses having glass matrix compositions representative of HLW glasses were used in 
laboratory tests. Except for the absence of radionuclides, these glasses provided levels of glass 
and glass phase separation and devitrification phases representative of HLW glasses. The 
maximum rates calculated by the model were selected to bound the rates measured 
experimentally after the precipitation of silicate, iron silicate, and mineral phases. In this way, 
processes that affect the glass alteration rate, congruent dissolution, selective leaching, and 
precipitation of alteration phases, have been implicitly incorporated in the model through model 
parameters, even though only the temperature and pH dependencies are calculated explicitly. 
The range of values of the rate coefficient kE provides rates consistent with those measured under 
various test conditions in which glasses were reacted in humid air, dripping water, and various 
immersion conditions. A triangular distribution of values for kE is skewed to low values to 
reflect the greater likelihood that glass in the disposal system will be contacted by water vapor 
(which gives the lowest measured rates) rather than immersion (which gives the highest 
measured rates). A minimum relative humidity of 44 percent is required for glass degradation to 
occur (BSC 2004 [DIRS 167619], Section 6.5.5.3.2). 
The range of kE also reflects the effects of glass composition and, to a large extent, the 
radionuclide inventory on glass degradation, since the waste glasses are formulated based on the 
waste compositions. Calculation of the radionuclide release rate accounts for the spatial 
heterogeneity of HLW glass (i.e., the distribution of various HLW glass) by using a 
representative glass log to represent all HLW glasses. The mass and surface area of glass that is 
available for corrosion, and the average radionuclide inventory, are calculated based on the 
characteristics and predicted numbers of glass logs to be received from different production 
facilities. 
In TSPA-LA, the HLW degradation model parameters and factors are used along with the inpackage 
chemistry to calculate a HLW degradation rate at each model time step. The 
degradation model parameters and factors are determined by sampling from parameter 
distributions that incorporate uncertainty, while the in-package chemistry parameters are 
calculated as part of the TSPA simulation. The composition of the representative glass is not 
defined in the HLW glass degradation model or used to calculate the degradation rate. A 
representative composition is specified in the in-package chemistry model for calculating the 
effects of glass dissolution on the chemistry. 
Supporting Documents: 
BSC 2004 [DIRS 167619] 
April 2004 66 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Related FEPs: 
2.1.01.02.0A, Interactions Between Colocated Waste 
2.1.01.02.0B, Interactions Between Codisposed Waste 
6.2.9 Alpha Recoil Enhances Dissolution 
FEP Number: 
2.1.02.04.0A 
FEP Description: During decay of certain radionuclides, alpha particles may be emitted with 
sufficiently high energies that the daughter nuclide recoils appreciably to conserve system 
momentum. A potential result of recoil is that certain radionuclides, such as 234U, exhibit 
substantially greater dissolution rates (with the same solubility limits) and can be transported 
preferentially. 
Descriptor Phrases: 
Alpha decay recoil enhances waste dissolution 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: The effects of alpha recoil can be excluded based on low consequence. 
Omission of the effects of alpha recoil will not significantly change radionuclide releases to the 
accessible environment, because the alpha recoil enhanced dissolution rates have been shown to 
be much less than the dissolution rate due to chemical processes. Even when conservatively 
assuming that all radioactive decay results in alpha recoil, it does not cause significant increases 
in the dissolution rate of the different waste forms. The processes investigated in this argument 
only involve direct nuclear effects, and not indirect nonnuclear effects such as thermal-enhanced 
dissolution due to heat generation from the radioactive decay of SNF. The following is the 
detailed discussion. 
Of the various radioactive decay modes (i.e., alpha (), beta (), gamma (), spontaneous fission 
[SF], isomeric transition, etc.) the most important for fissile materials is the alpha decay 
mechanism. This mechanism is the dominant decay mode for heavy radionuclides. All the 
heavy nuclides above 209Bi are radioactive and tend to decay into more stable nucleus 
configurations (i.e., atomic masses of less than 200 atomic mass units). Because these heavy 
radionuclides need to lose significant quantities of mass in order to become more stable, in 
general they will decay by the mode that results in the largest mass loss. Thus, the most likely 
decay mode will be alpha decay, which has the largest mass and associated kinetic energy. 
Although many of the heavy radionuclides emit alpha particles with energies greater than 
4.0 MeV (Parrington et al. 1996 [DIRS 103896], p. 48), there is no gamma decay or beta decay 
with energy greater than 4.0 MeV and only a few with energy greater than 1.0 MeV. Thus, alpha 
recoil mechanisms will bound the effects due to beta and gamma recoil. Other special decay 
modes such as isomeric transition and spontaneous fission decay have probabilities of occurrence 
that are orders of magnitude less than that of alpha decay. Information in Attachment II of this 
analysis report indicates that isomeric transition occurs for 108mAg, 242mAm, 93mNb, and 121mSn; 
none of which are significant in terms of mass contribution. Also, information from Lederer and 
April 2004 67 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Shirley (1978 [DIRS 142133], p. 1464) indicates that the spontaneous fission half-lives are 
several orders of magnitude longer than that for other decay modes and are, thus, insignificant. 
The number of atom displacements per alpha decay can be calculated by noting that when a 
recoil nucleus strikes an atom, it requires a minimum displacement energy, Ed, of approximately 
25 eV to eject the struck atom from its lattice site (Foster and Wright 1973 [DIRS 144061], 
p. 296). The total number of displacements caused by a single alpha decay event is given by 
Equation 1 (Foster and Wright 1973 [DIRS 144061], p. 296). Equation 1 gives displacement 
units: 
 Tm 
d 
where 
 
To simplify the analysis, the probability for kinetic energy transfer is treated as being a uniform 
distribution over the applicable energy range: 
T E . m d 
The total number of displacements which originate from a primary knock-on is described by the 
Kinchin-Pease Model (Foster and Wright 1973 [DIRS 144061], p. 297): 
0 
 
P(T ) T = .E = 
1 K( E,T ) =
i 
d d 
T E T . 
2 2 E E 
Disp(E) 
0 E 
 = =    P(T )K(E,T ) (T )dT P(T )K(E,T ) (T )dT (Eq. 1) 
P(T) = probability that an atom (primary knock-on), struck by either the emitted 
alpha or the alpha recoil atom, receiving energy T is displaced. 
K(E,T) = probability for the transfer of kinetic energy T to the primary knock-on atom 
of energy E. 
v(T) = total number of displacements in a cascade originating from a primary knockon 
atom whose energy is T. 
The expression is integrated over the energy range starting at the displacement threshold energy, 
Ed, and ending at the maximum energy that can be transferred to an atom, Tm. Since the 
displacement of atoms corresponds to a threshold event, P(T) is modeled as a Heaviside step 
function (Foster and Wright 1973 [DIRS 144061], p. 297): 
0 d (Eq. 2) d 
0 < if T E 
if T . Ed 
1 
(Eq. 3) 
d i (Eq. 4) T / . 2E if T E 
E / E if T E 2 > i  i d 
April 2004 
 
 . (T )= + i T E = 
where 
Ei = energy required for ionization. 
The 2Ed in the denominator accounts for the displacement of the knock-on atom and the 
additional Ed for the striking atom to also leave the displacement site. The model also 
68 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
reasonably concludes there is an ionization threshold (EI  1000A) below which displacements 
take place and above which only ionization takes place. 
The total number of displacements is given by Equation 5, which includes ionization 
interactions. 
Tm T 
Ed 
dT 
2 2. E T E 
..
. d m 
2d
= Disp( t ) m 
2 2
d 
 
E T i m T E 
d i E E2 2 2 2. . E T E E T E i 
d d m d d m 
= 
.
.... 
..
..
i d E E 
. E (T E ) 4 2 . . E (T E ) 
238 
+ >   dT dT if T E 
m d T E +
Ed 4 
2 2 2 ET E im i + 
. . 
d m d d m d 
69 
if T  E m i 
(Eq. 5) 
if T > E m i 
April 2004 
. 
For the example involving 238U decaying to 234Th, there are two cases to examine. The first is 
for the alpha recoil atom, and the second is for the emitted alpha particles. From the analysis in 
Attachment III of this analysis report, it can be seen that the alpha recoil atom (234Th) has a value 
0.072 MeV for Tm, which is much less than the ionization energy (typically approximately 1000A 
equal to 0.234 MeV (million electron volts). This results in Dispmax() equal to 720 
displacements per alpha decay. (This does not include replacement collisions along with 
focusing and channeling effects, which will significantly lower the estimate for displacements.) 
If each of the secondary displacements follows a bifurcation process (i.e., 2N equals 720), this 
would correspond to 9.49 bifurcation levels. This means that the maximum number of atom 
monolayers of the SNF fuel meat through which recoil nuclei (due to alpha decay) could pass 
through and enter bounding groundwater is approximately 10. This is only for recoil atoms 
traveling in the direction of the groundwater. Attachment III of this analysis report indicates that 
only half of the recoil could be in the proper direction. This also does not mean that all the 
(secondary) displaced atoms within ten monolayers of the material surface would enter the 
groundwater; it only indicates that it is possible. 
The second case, the emitted alpha particles, involves a much larger number of possible 
displacements. This is due to its larger kinetic energy. From Lederer and Shirley (1978 
[DIRS 142133]), it can be identified that the maximum energy for alpha particles emitted from 
U is 4.196 MeV. Since this energy is above the ionization value, the second version of 
Equation 5 that contains two terms must be used to take into account the ionization by the alpha 
particles. Using the maximum alpha energy results in a value of 4,550 for Disp( ). When added 
to the recoil atom value, the net displacements are 5,270, which corresponds to approximately 
12.4 bifurcation levels. 
The enhancement to the dissolution rate due to alpha decay can now be upper bounded. This rate 
will correspond to the number of alpha decays per unit time within the first thirteen halfmonolayers 
of the material surface that result in nuclei recoiling toward the groundwater. The 
worst case density for thorium (or even uranium) will be approximately that of pure plutonium 
metal, with a maximum density of 19.84 g/cm3 (Wick 1980 [DIRS 143651], Table 7.1) and a 
ANL-WIS-MD-000009 REV 01 
if T  E m i 
Miscellaneous Waste-Form FEPs 
monolayer thickness of approximately 3.0 A (3.0x10-10m). The combination of these two 
parameters results in 0.039 grams of SNF material within the first thirteen monolayers of a 
surface area of 1.0 m2 that are in the direction of the material surface. When this surface density 
(0.039 g/m2) is multiplied by the fractional rate at which the SNF material experiences 
radioactive -decay, it can be expressed in the units used in the CSNF (BSC 2004 [DIRS 
167321]; DSNF (BSC 2003 [DIRS 163693]) degradation analysis reports for dissolution of both 
CSNF and DSNF material due to chemical processes. The fractional rate was determined from 
radiological half-lives along with some data from Attachment II of this analysis report, which 
contains inventory information for the Yucca Mountain repository. These data identify the major 
constituents on a mass basis of the final waste forms (SNF and HLW): 238U, 235U, 239Pu, 236U, 
and 240Pu. Key radionuclides from Attachment II of this analysis report are analyzed in Table 
6.2-1, which identifies their maximum alpha recoil enhancement for dissolution rates. Data in 
column 5 of this table are computed by noting that the fractional decay rate for the radionuclides 
is given by their decay constant ( = ln(2)/). To put these values from Table 6.2-1 into proper 
perspective, a comparison to Figure 2 must be made. Figure 2 identifies the dissolution rates of 
different waste forms due to chemical (nonnuclear) processes. As can be identified, the alpha 
recoil dissolution rates are much less than the values due to chemical processes. 
In summary, the radioactive decay processes that directly increase fuel-meat dissolution are 
bounded by alpha recoil rates. Even when conservatively assuming that all radioactive decays 
result in an alpha recoil, they will not cause significant increases to the dissolution rate of the 
different waste forms. The processes investigated in this argument are only for direct nuclear 
effects and not indirect nonnuclear effects, such as thermally enhanced dissolution due to heat 
generation from SNF radioactive decay. Thus, the overall dissolution due to direct nuclear decay 
processes has been eliminated from performance-assessment calculations on the basis of low 
consequence. 
Table 6.2-1. Alpha Recoil Enhanced (From  and -Recoil Atom) Dissolution Rates Due to the Major 
Mass-Based Constituents of SNF and HLW to be Disposed of in the Yucca Mountain Repository 
Fraction Decay Rate (b) 
(1/yr) 
1.55 x 10 .10 
9.85 x 10-10 
2.88 x 10-5 
2.96 x 10-8 
70 
NOTES: a Data obtained from Attachment II of this analysis report. The original source of 
information, in radioactivity units, was DOE (1998 [DIRS 122980]). Mass unit values were 
derived by Sanchez et al. (1998 [DIRS 149368]). 
b The fraction decay rate, also known as the decay constant, is given by =ln(2)/, where 
 is the radionuclide half-life given by values in column 3. 
c Each monolayer thickness is 3.0 A (3.0x10-10m), and the density is upper bounded at 19.86 
g/cm3 (theoretical density of pure plutonium metal ([Wick 1980 [DIRS 143651]). 
.Decay Rate in 
13 Mono-Layers(c) 
(g/m2.yr) 
6.05 x 10-12 
3.84 x 10-11 
1.12 x 10-6 
1.15 x 10-9 
April 2004 
Nuclide Half.Life (a) 
ID Decay Mode (years) 
, , SF 238U 4.47 x 109 
235U 7.04 x 108 , , SF 
2.410 x 104 239Pu , , SF 
236U 2.342 x 107 , , SF 
ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Figure 2. Comparison of Dissolution Rates for High-Level Radioactive Waste, Metallic Carbide, and 
Ceramic Spent Nuclear Fuel 
Source: CRWMS M&O 1998 [DIRS 100362], p. A6-29, Figure A6-1 
TSPA Disposition: 
N/A 
Supporting Documents: 
None. 
Related FEPs: 
None. 
6.2.10 HLW Glass Cracking 
FEP Number: 
2.1.02.05.0A 
FEP Description: Cracking of the HLW glass on cooling and during handling means that the 
surface area of the glass is greater than the surface area of a monolithic block. The increase in 
the surface area could affect the rate of glass alteration and radionuclide dissolution. 
Descriptor Phrases: 
Increased dissolution surface area from HLW glass cracking 
April 2004 71 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Screening Decision: 
Included 
Screening Argument: 
N/A 
exposure) is 
TSPA Disposition: The expression for exposed glass surface area developed to address this FEP 
is provided in Defense HLW Glass Degradation Model (BSC 2004 [DIRS 167619]). The surface 
area used to calculate the radionuclide release rate bounds the effects of thermal cracking during 
manufacture and impact cracking during subsequent handling. The exposure factor (f 
used to model the combined effects of the added surface area due to cracking, the fraction of the 
surface that is accessible to water, and the reactivity of glass in tight cracks relative to glass at a 
free surface. The value of fexposure is selected for each realization from a triangular distribution 
with a minimum and most expected value of 4 and a maximum value of 17. The initial exposed 
surface area is calculated by using the following equation (BSC 2004 [DIRS 167619], Section 
6.5.6, Equation 37): 
S0 = fexposure ~ (2ro
2 + 2ro ~ Lo) (Eq. 6) 
where ro is the internal radius of the glass canister, and Lo is the length of the glass canister. 
The change in the surface area as the glass degrades is calculated as the sum of the product of the 
specific surface area of a glass log and the mass of glass available at the beginning of the time 
step using the following equation (BSC 2004 [DIRS 167619], Section 6.5.6, Equation 39): 
(Eq. 7) S = fexposure ~ 2.70 ~ 10-3 m2/kg ~ (2,710 kg - Mt kg) 
where M gives the mass loss during all previous time steps (and resulting loss of surface area). 
The value of M is calculated as the product of the glass degradation rate used for that time step 
and duration. The mass loss is used to calculate the loss in surface area. 
Supporting Documents: 
BSC 2004 [DIRS 167619] 
Related FEPs: 
None 
6.2.11 HLW Glass Recrystallization 
FEP Number: 
2.1.02.06.0A 
FEP Description: HLW glass recrystallization could occur and would lead to a less corrosionresistant 
waste form. Recrystallization is a slow process and typically occurs only if a high glass 
temperature is maintained over a prolonged period. 
April 2004 72 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Descriptor Phrases: 
Increased corrosion from HLW glass crystallization 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: The effect of crystallization on the glass corrosion rate is implicitly 
accounted for in Defense HLW Glass Degradation Model (BSC 2004 [DIRS 167619]) by the 
range of model parameter values used to represent uncertainty. The ranges of model parameter 
values were determined from experimentally measured dissolution rates of glasses having 
compositions similar to HLW glasses. 
Although the effects of crystallization are included in the TSPA-LA model, in situ devitrification 
of HLW glass is not expected to occur at disposal temperatures (i.e., at temperatures less than 
400C). As part of model validation, the glass degradation rate was shown to be insensitive to 
the presence of crystalline phases that are either formed by devitrification of the glass melt 
during manufacture or intentionally added to the waste form. This is based on literature results 
that show the effects of devitrification on glass dissolution rate are small and near the uncertainty 
in the measurement of the rate. 
In summary, even though the effects recrystallization on the glass degradation rate are expected 
to be insignificant, the effects are captured in TSPA-LA through the range of parameter values 
used in the degradation model used to represent uncertainty. 
Supporting Documents: 
BSC 2004 [DIRS 167619] 
Related FEPs: 
None. 
6.2.12 Radionuclide Release from Gap and Grain Boundaries 
FEP Number: 
2.1.02.07.0A 
FEP Description: While in the reactor at the high temperatures, radionuclides such as I and Cs 
may migrate and preferentially accumulate in cracks in the fuel matrix, grain boundaries of the 
UO2, and in the gap between the fuel and cladding. After the waste package fails and the 
cladding perforates, the release rate of this fraction of the radionuclides could be rapid. In 
addition, reactions between materials within the spent nuclear fuel assembly or waste package 
infill such as I and Pb may affect this process. 
April 2004 73 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Descriptor Phrases: 
Rapid release from fuel gap and grain boundaries 
Chemical binding of gap inventory 
Migration to gap and grain boundaries after emplacement 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: Gap-and-grain boundary release applies mainly to commercial spent nuclear 
fuel and similar fuel designs (such as MOX) based on UO2-based fuel pellets. In such SNF a 
portion of some of the volatile or soluble fission product radionuclides can migrate to the fuel 
pellet grain boundaries and open gap areas of the fuel rods under the influence of the high 
temperature gradients operative under reactor operation. Further migration under the 
significantly cooler environment of repository disposal is not expected to occur. The gap-andgrain-
boundary inventory is assumed in the TSPA to be accessible for dissolution into any water 
that penetrates the fuel cladding. 
As described in CSNF Waste Form Degradation: Summary Abstraction (BSC 2004 [DIRS 
167321]), the release of the gap and grain boundary radionuclide inventory is modeled as an 
instantaneous release fraction (fi) where the subscript i refers to 137Cs, 129I, 90Sr, and 99Tc. 
Available experimental data are used to estimate the mean values, the ranges and the distribution 
functions for fi (BSC 2004 [DIRS 167321], Section 6.3). The gap and grain boundary release is 
modeled as an instantaneous release in one GoldSim time step. These distributions are sampled 
at the beginning of each TSPA GoldSim realization to determine the values of fi that are used for 
the instantaneous release fractions when fuel cladding fails during the course of the realization. 
At the high temperatures fuel pellets experience during reactor operation, a significant migration 
of gaseous or soluble radionulides to the gap and grain boundaries occurs. At the much lower 
temperatures expected after waste package emplacement, the thermally driven migration of 
radionuclides to the gap and grain boundaries is expected to be low enough that it can be 
considered to be included in the uncertainties in the instantaneous release fraction of the CSNF 
(BSC 2004 [DIRS 167321], Section 6.2.1.2). 
Supporting Documents: 
BSC 2004 [DIRS 167321] 
Related FEPs: 
2.1.02.02.0A, CSNF Degradation (Alteration, Dissolution, and Radionuclide Release) 
6.2.13 Pyrophoricity from DSNF 
FEP Number: 
2.1.02.08.0A 
FEP Description: DSNF can contain pyrophoric material. Pyrophoric material could ignite and 
produce an adverse effect on repository performance. Pyrophoric events could affect the thermal 
April 2004 74 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
behavior of the system and could contribute to degradation of the waste package, waste form, 
and cladding. 
Descriptor Phrases: 
Effects of pyrophoricity (heat generation) 
Effects of pyrophoricity (waste package degradation) 
Effects of pyrophoricity (waste form degradation) 
Effects of pyrophoricity (cladding degradation) 
Screening Decision: 
Excluded low consequence 
Screening Argument: DSNF may potentially include pyrophoric material, particularly the 
N-Reactor SNF currently being removed from the K-basin and placed in dry storage at Hanford. 
Because N-Reactor SNF is composed primarily of zirconium-clad metallic uranium, it must be 
assessed for its potential for pyrophoric behavior. However, a significant amount of the cladding 
has been breached exposing metallic uranium, which has shown pyrophoric behavior in the past 
(Thornton 1998 [DIRS 125082]). Pyrophoric materials can spontaneously ignite and produce an 
adverse effect on repository performance by producing heat and dispersible particulates and 
gases. For the purposes of this analysis, pyrophoricity is understood to be the capability to ignite 
spontaneously if the temperature, chemical, or physical or mechanical conditions of the storage 
or transportation environments are oxidizing (ASTM C 1454-00 [DIRS 152779], Section 3.2), 
which could occur as a result of waste package degradation, mechanical impact, or overheating 
during handling or transportation. A pyrophoric event is defined as ignition, followed by rapid 
chemical oxidation or self-sustained burning. A pyrophoric event has the potential for increasing 
the release and dispersal rate of radionuclides, which could impact repository performance. 
Uranium hydride inclusions in the metallic uranium matrix of damaged N-Reactor SNF (i.e., fuel 
elements that have damaged cladding) could provide an ignition source. The open literature 
contains descriptions of pyrophoric behavior in uranium metal and uranium metal-based SNF 
(such as Magnox SNF). Although N-Reactor SNF has significant damaged cladding and is 
known to contain such uranium hydride inclusions, there have not been any pyrophoric reactions 
during normal handling and testing of N-Reactor SNF samples at the K-basins or Pacific 
Northwest National Laboratory in support of the N-Reactor fuel interim dry storage program. 
However, N-Reactor SNF has displayed unexpected pyrophoric behavior during handling and 
processing in the air or water environments and aggressive mechanical handling conditions 
associated with reprocessing operations at Hanford and West Valley (Schulz 1972 
[DIRS 159406]). 
Since uranium hydride inclusions tend to be concentrated near the exposed uranium metal fuel 
surface of damaged SNF (Abrefah et al. 1999 [DIRS 151226]; Marschman et al. 1997 
[DIRS 149429]), a small but finite chance of spontaneous ignition will be considered in this 
analysis to support the screening argument of low consequence of pyrophoric event to the 
postclosure TSPA-LA. Because the fraction of N-Reactor SNF with cladding that is damaged 
enough to expose the metallic uranium core is quantitatively unknown but potentially significant 
(Abrefah et al. 1995 [DIRS 151125]), all the SNF elements will be regarded as damaged for the 
purposes of this analysis. 
April 2004 75 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
If spontaneous ignition of U-hydrides occurs, there must be sufficient oxygen to support 
continuing combustion. This is unlikely, since waste packages containing potentially pyrophoric 
material will be filled with an inert gas prior to emplacement in the repository. Furthermore, the 
N-Reactor SNF will be contained within canisters (known as multicanister overpacks) emplaced 
inside the waste packages that are also filled with an inert gas. The inert environment will 
exclude all but small quantities of water and oxygen, thus reducing the potential for oxidation 
reactions prior to breach of the waste package (DOE 1998 [DIRS 122980], Section 5.1.2.1). If 
breach of the waste package and multicanister overpack occurs, the breach would be expected to 
be small enough that oxygen would leak slowly into the waste package. The slow introduction 
of oxygen would tend to limit the oxidation rate of any remaining hydrides, thus limiting the 
ability of the hydrides to initiate a pyrophoric event. However, there is not sufficient data 
available at this time to absolutely exclude pyrophoric events based on their low probability of 
occurrence. Therefore, the exclusion of this FEP from TSPA-LA is based on the fact that the 
potential adverse effects of a pyrophoric event , are insignificant. 
Attachment I of this analysis report provides a consequence-based analysis that estimates the 
impact a pyrophoric event would have on repository performance. The analysis conservatively 
(but non-mechanistically) assumes that all N-Reactor SNF-containing waste packages that 
breach during the regulatory period undergo a pyrophoric event within one TSPA time step. It is 
also assumed that (1) breach of an N-Reactor SNF-containing waste package results in a 
pyrophoric condition that causes the two adjacent waste packages to fail, (2) the various types of 
waste packages are evenly distributed throughout the repository, and (3) the probability of failure 
of a waste package containing pyrophoric (i.e., N-Reactor) SNF is the same as that of any other 
waste package (Section 5.2.13 of this analysis report). The analysis demonstrates that a 
pyrophoric event involving waste packages containing N-Reactor SNF would not have a 
significant impact on postclosure repository performance in terms of impact on radionuclide 
release and peak offsite dose. This is primarily due to the fact that only a small fraction of the 
emplaced waste packages will contain N-Reactor SNF. In terms of the potential increase in total 
radionuclides released over the repositorys lifetime, the impact of pyrophoric events would not 
exceed about a one percent increase in the total amount of radionuclides released. In terms of the 
peak offsite dose that could result from a single pyrophoric event, a pyrophoric event would, at 
most, cause a few percent increase in peak offsite dose above the dose that would be obtained if 
pyrophoric events were not possible. As for clustered events, unrealistic scenarios involving 
noncredible mechanisms would be required to generate more than a few percent increase in peak 
offsite dose. Therefore, pyrophoricity of DSNF can be excluded from further consideration 
based on low consequence to radiological exposures and radionuclide releases. 
TSPA Disposition: 
N/A 
Supporting Documents: 
N/A 
Related FEPs: 
2.1.11.01.0A, Heat generation in EBS 
2.1.12.29.0A, Flammable Gas Generation from DSNF 
April 2004 76 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
6.2.14 Chemical Effects of Void Space in Waste Package 
FEP Number: 
2.1.02.09.0A 
FEP Description: If waste packages and/or DSNF canisters are not completely filled, then the 
unfilled inert-gas or air-filled volume could influence water-chemistry calculations. 
Descriptor Phrases: 
Chemical effects of void space in waste package 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: The base-case conceptual model for in-package chemistry is composed of 
two conceptual models, the no-drip model and the seepage dripping model, which are 
collectively referred to as the In-Package Chemistry Model (IPCM) in In-Package Chemistry 
Abstraction (BSC 2003 [DIRS 161962]). The water vapor condensation model addresses the 
situation in which water vapor enters the breached waste package and condenses. The seepage 
model addresses the situation in which liquid water seeps or drips into the breached waste 
package. These models differ from each other basically in the chemistry of the water to which 
the waste forms in a breached waste package are exposed. In evaluating the in-package water 
chemistry, the entire void volume in the waste package is taken to be saturated with liquid and 
gas and includes any void volume in a canister and any time-varying void volume in the waste 
form. The IPCM accounts for the unfilled void space in the waste package, and, thus, its impact 
on the in-package chemistry. 
The conceptual model is an unsaturated model in which the void spaces within a waste package 
are partially filled by liquid water in equilibrium with atmospheric gases. Void space in the 
waste package is thus partially gas filled. This has the effect of putting the gas phase in close 
contact with the liquid- and solid-phase reactants in the system, and is in fact the basis of the 
conceptual model implemented in In-Package Chemistry Abstraction (BSC 2003 [DIRS 
161962]). Thus, the increased void space in the waste package decreases the distance that gas 
must diffuse to maintain equilibrium with the liquid phase; this is contrast to the bathtub model 
(BSC 2003 [DIRS 161962], Section 6.3.1) where all of the void space is liquid-filled and the gas 
phase must diffuse much longer distances to maintain equilibrium. 
The effects of the unfilled void space on the in-package chemistry are passed to TSPA explicitly 
via the oxygen and carbon dioxide equilibrium within In-Package Chemistry Abstraction 
(BSC 2003 [DIRS 161962]). The parameters abstracted in that document are pH, ionic strength, 
total carbonate, Eh, chloride, and fluoride. The uncertainty in the in-package chemistry due to 
different void volumes is small compared with that due to uncertainty in the waste package water 
inflow rates. 
April 2004 77 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Supporting Documents: 
BSC 2003 [DIRS 161962] 
Related FEPs: 
2.1.09.01.0B, Chemical Characteristics of Water in Waste Package 
6.2.15 Organic/Cellulosic Materials in Waste 
FEP Number: 
2.1.02.10.0A 
FEP Description: Degradation of cellulose in the waste could affect the long-term performance 
of the disposal system. 
Descriptor Phrases: 
Organic/cellulosic materials in waste 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: Chemically reactive materials, such as cellulosic materials, are excluded 
from the repository (DOE 2002 [DIRS 158873], Sections 4.2.11, 4.3.1.B, 4.3.7.B, and 4.3.17) 
and only minor quantities of organic materials might appear as part of the waste (BSC 2003 
[DIRS 161962]). Organic complexants, such as humic acid in groundwater, can cause actinide 
complexation at concentrations as low as one ppm. However, the quantities of humic materials 
in groundwater flowing into the drifts are not expected to be significant. In any case, actinide 
complexation in the drifts will not be significant (see FEP 2.1.09.13.0A). The minor amounts of 
organic materials in the emplaced waste and the insignificant complexation in the drifts will not 
result in any significant effects on repository performance. Therefore, this FEP can be excluded 
on the basis of low consequence to radiological exposures and radionuclide releases. 
TSPA Disposition: 
N/A 
Supporting Documents: 
BSC 2003 [DIRS 161962] 
Related FEPs: 
2.1.10.01.0A, Biologic Activity 
2.1.12.04.0A, Gas Generation (CO2, CH4, H2S) from Microbial Degradation 
6.2.16 DSNF Cladding 
FEP Number: 
2.1.02.25.0A 
FEP Description: DSNF to be disposed of in Yucca Mountain has a variety of fuel types that 
may not be similar to the CSNF to be disposed. Some of the fuel types may have initial 
April 2004 78 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
cladding-degradation characteristics that are different from those for the CSNF. Therefore, the 
effectiveness of DSNF cladding as a barrier to radionuclide mobilization might be different from 
CSNF. This FEP addresses all types of DSNF cladding except naval SNF cladding. 
Descriptor Phrases: 
Degradation of DSNF cladding 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: More than 80 percent by weight of the DSNF is from the N-Reactor, a 
significant but unquantified fraction of which will have damaged cladding at the time of 
emplacement in their canisters (Abrefah et al. 1995 [DIRS 151125]). Up to 50 percent may be 
already perforated (Rechard 1995 [DIRS 101084], Section 11.3.1, pp. 11 to 24). There has been 
insufficient characterization of the condition of the DSNF cladding (other than the observations 
of extensive damage to the N-Reactor SNF cladding) to establish the effectiveness of the 
cladding as a barrier. Because the cladding integrity of most DSNF will not be extensively 
characterized (BSC 2003 [DIRS 163693], Section 6.4.3), the TSPA takes no credit for 
radionuclide containment by the cladding . Also, the TSPA takes no credit for canister integrity 
(i.e., once the waste package has leaked or failed); it is conservatively assumed that the DSNF is 
directly exposed to repository water or air environment. Because no credit will be taken for the 
cladding, mechanisms that might enhance cladding degradation do not impact the predicted 
consequences (CRWMS M&O 2000 [DIRS 153595]; Thornton 1998 [DIRS 107796]). 
Additionally, postclosure site-boundary dose-sensitivity analyses have indicated that the increase 
of the dose rate at the site boundary resulting from the complete release of the radionuclide 
inventory in waste packages containing N-Reactor SNF during one performance assessment time 
step is insignificant (Thornton 1998 [DIRS 107796]). Since this represents over 80 percent of 
the DSNF inventory, not taking performance credit for the cladding the DSNF will provide a 
conservative estimate of the boundary dose. Also, the rate of unzipping of failed cladding can be 
conservatively assumed to be instantaneous in the TSPA-LA since the DSNF radionuclide 
inventory is small compared to the rest of the waste and the cladding is assumed not to provide a 
barrier anyway. Inclusion of the protective performance of DSNF cladding in the TSPA-SR 
calculation would reduce calculated doses. 
Therefore, DSNF cladding degradation can be excluded from the TSPA-LA based on low 
consequence because it has no adverse effects on performance. 
TSPA Disposition: 
N/A 
Supporting Documents: 
BSC 2003 [DIRS 163693] 
Related FEPs: 
2.1.02.01.0A, DSNF Degradation (Alteration, Dissolution, and Radionuclide Release) 
2.1.02.08.0A, Pyrophoricity From DSNF 
April 2004 79 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
6.2.17 Grouping of DSNF Waste Types into Categories 
FEP Number: 
2.1.02.28.0A 
FEP Description: Several hundred distinct types of DOE-owned SNF (DSNF) may potentially 
be stored at YMP. These represent many more types than can viably be examined for their 
individual effect on the repository. A limited number of representative or bounding degradation 
models must be selected and/or abstracted. As a result, the effects on repository performance of 
the details related to the many distinct types of DSNF can not be evaluated. 
Descriptor Phrases: 
N-Reactor fuel bounding model 
Naval fuel 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: DSNF to be disposed of in Yucca Mountain is composed of a variety of fuel 
types. The various DSNF types have been classified into eleven groups for the purpose of 
TSPA-LA analyses (DOE 2000 [DIRS 118968]). The largest single DSNF type is the N-Reactor 
SNF, which comprises approximately 84 percent by weight of the total DSNF. N-Reactor SNF 
uranium metal (Group 7) models are adopted as a surrogate for the entire DSNF inventory, 
except for the naval SNF (Group 1) (BSC 2003 [DIRS 163693]). The CSNF degradation model 
is used as the surrogate for naval SNF. This grouping and implementation simplifies the 
TSPA-LA model for computational efficiency while ensuring that the great majority of DSNF 
degradation is captured in a bounding fashion. 
In addition to being the largest DSNF inventory type by weight, N-Reactor SNF degradation can 
be used to represent degradation of the entire DSNF inventory (Section 5.2.17 of this analysis 
report) because (1) a significant fraction of the N-Reactor SNF will be damaged at the time of 
emplacement in their canisters and (2) the N-Reactor SNF degradation model predicts 
dissolution rates greater than most other groups (BSC 2003 [DIRS 163693], Section 6.2 and 
Table 9). The group that potentially has greater dissolution rates than N-Reactor SNF is the 
mixed-carbide-fissile fuel waste particles in a nongraphite matrix (Group 3). Group 3 waste 
would not serve as an appropriate DSNF degradation surrogate because the total inventory of the 
Group 3 SNF is less than 0.001 percent of DSNF waste (DOE 2002 [DIRS 158405], 
Appendix D). The immobilized plutonium ceramic waste form is not included in this discussion 
because the DOE Office of Environmental Management has recently decided to use the majority 
of the excess defense plutonium as MOX fuel in commercial reactors as the preferred disposition 
path rather than immobilization in a ceramic waste form (67 FR 19432 [DIRS 162618]). 
As discussed in DSNF and Other Waste Form Degradation Abstraction (BSC 2003 
[DIRS 163693], Section 6.3), even when upper bound assumptions are used for N-Reactor SNF 
degradation rates (i.e., complete dissolution of the waste form during a single code time step 
April 2004 80 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
upon exposure of the waste form to groundwater) overall repository performance is insensitive to 
the DSNF degradation rate (Thornton 1998 [DIRS 107796]). Even under pyrophoric conditions 
(Thornton 1998 [DIRS 125082]), the contribution from DSNF is small because it is a small 
percent of the total waste, therefore waste stream plume chemistry is dominated by the CSNF 
waste even when bounding DSNF degradation models are used. Therefore, the categorization of 
DSNF waste into 11 groups and the use of N-Reactor SNF (Group 7) to represent, in a bounding 
fashion, the degradation rate of all DSNF (except naval SNF) is justified. 
Supporting Documents: 
BSC 2003 [DIRS 163693] 
Related FEPs: 
2.1.02.01.0A, DSNF Degradation (Alteration, Dissolution, and Radionuclide Release) 
2.1.02.25.0A, DSNF Cladding 
6.2.18 Flammable Gas Generation from DSNF 
FEP Number: 
2.1.02.29.0A 
FEP Description: DOE-owned Spent Nuclear Fuel (DSNF) to be disposed of in Yucca 
Mountain will have a small percent of carbide fuel. When carbide is exposed to water, 
flammable gases such as methane and its minor constituents ethane, ethylene, and acetylene 
(referred to as ethyne by the International Union of Pure and Applied Chemistry) are generated. 
If these gases ignite, localized increases in temperature can occur, which might affect fuel 
degradation. The area around the ignition point may be mechanically and/or thermally 
perturbed, which could affect waste container or host-rock properties in the area of the EBS. 
Descriptor Phrases: 
Flammable gas (CH4) 
Flammable gas (ethane, ethylene, acetylene) 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: Flammable-gas generation from DSNF can be excluded based on low 
consequence for the following reasons: 
1. The only fuel waste types capable of producing organic flammable gases such as 
methane, ethane, etc., are the U/Th carbide and the Pu/U carbide DSNF waste fuels 
because they are the only spent fuels containing more than trace quantities of carbon. 
These gases are formed by the reaction of the carbides with liquid water or water 
vapor. Only a small percentage (approximately 1 percent or 25 MTHM) of DSNF fuel 
contains U/Th carbide fuels and there is only about 100 kg of Pu/U carbide DSNF 
(DOE 2002 [DIRS 158405], Table D-1). Since DSNF comprises only about 4 percent 
of the total waste packages in the repository (DOE 2002 [DIRS 158405], Section 1), 
the carbide spent nuclear fuels will amount to less than 0.04 percent of the waste 
April 2004 81 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
packages. Additionally, all DNSF waste will be widely dispersed throughout the 
repository as will any produced flammable gas. 
2. Additionally, the repository is located in fractured and unsaturated tuffaceous rock 
(DOE 2002 [DIRS 155943], Section 1.4.2.2.2). This rock is overlaid by 
approximately 250 meters of unsaturated welded and nonwelded tuffs. With the 
exception of intermittent perched zones, both the matrix and the fractures within the 
host rock above and below the repository horizon have a relatively high degree of gas 
saturation (generally greater than 50 percent), making these areas more gas- than 
liquid-permeable. Within the repository horizon, ambient gas saturation in the matrix 
ranges between 10 to 20 percent, while gas saturation in fractures tends to be much 
higher. These conditions will promote a dispersive gas flow path between the 
repository and host rock, thus diluting any potential flammable-gas concentrations to 
levels below the ignition point. 
3. However, even if flammable gases were to accumulate and such an explosion were to 
occur, it would be small resulting in an insignificant release of radionuclides from the 
waste form. Consequently, the effect on radiological exposures to the reasonably 
maximally exposed individual and radionuclide releases to the accessible environment 
would not be significant. 
The chemical reaction of U/Th carbide with water produces carbon dioxide and hydrocarbon 
gases. The major constituent of the hydrocarbon gases is methane, with minor constituents being 
ethane, ethylene, and acetylene. In sufficiently high concentrations, all of the hydrocarbon gases 
are flammable in air. 
The Fort St. Vrain and Peach-Bottom Core-2 and Core 1 gas-cooled reactors used U/Th carbide 
fuel. Only the Peach-Bottom Core-1 reactor pellets are of concern, since pellets from this reactor 
are not encased in the tough, corrosion-resistant, silicon-carbide (SiC) shells, used in the other 
two reactors. Consequently many Peach-Bottom Core-1 pellets have broken open, with an 
estimated cladding failure rate of 35 percent (DOE 1998 [DIRS 122980], Appendix A, Section 
6.9, p. 41). These pellets can possibly produce flammable hydrocarbons. Only 1.663 metric tons 
heavy metal (MTHM) of Peach Bottom Core 1 pellets (DOE 1998 [DIRS 122980], Appendix B, 
p. 14) contained in 103 waste packages (DOE 1998 [DIRS 122980], Table 1-1, p.1-8) will be 
disposed in the repository. This is small compared to the 24.667 MTHM of high-integrity fuel 
(contained in 545 waste packages) from the Fort St. Vrain and Peach-Bottom Core 2 reactors 
(DOE 1998 [DIRS 122980], Appendix B, p. 14 and Table 1-1, p.1-8). Thus, less than 7 percent 
{1.663/(1.663 + 24.667)} of the carbide fuel is low-integrity pellets, of that only 2.5 percent 
(0.35 x 7 percent) is susceptible to gas production due to reaction with water. Taken all together 
only 0.001 percent (2.5 percent x 0.04 percent) of the waste to be disposed of in the repository is 
capable of producing flammable gases. Even if the cladding failure rate were 100 percent, only 
approximately 0.003 would be susceptible to gas production. 
All DNSF waste will be codisposed among some 3,910 waste packages (DOE 2002 [DIRS 
158405]; BSC 2003 [DIRS 161961], Table I-1), with one DSNF canister packaged per five 
CSNF canisters in each waste package. So the carbide fuel is inherently dispersed and any gas 
produced from carbide fuel would be dispersed as well. Furthermore, because the percent of 
April 2004 82 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
U/Th carbide waste is so small, any explosion from its gas product would be small resulting in an 
insignificant release of radionuclides from the waste form. This argument is further supported 
by an analysis of hydrocarbon production from carbide fuels (Propp 1998 [DIRS 149395]), 
concluding that hydrocarbon production is not a problem. 
Changes to the fluid-flow characteristics of the flow system in the mountain could be produced 
by thermally driven mechanical and chemical processes. This could produce a condensation cap 
in some areas of the drift, reducing gas and water permeabilities, causing gas to possibly 
accumulate within the drift area. However, the probability of an ignition source being present is 
low. The possibility of reduced permeability to gas would also limit the availability of oxygen 
for combustion and greatly reduce the corrosion of containers, thereby reducing the number of 
containers potentially producing flammable or explosive gases. A condensation cap would 
reduce the influx of water contacting carbide U/Th waste, thus limit gas production from this 
source. 
Additionally, flammable gases are not likely to accumulate in concentrations high enough to 
cause an explosion simply due to the physical setting of the repository. The repository is located 
in the unsaturated portion of the Topopah Spring Tuff, a fractured and welded unit. This unit is 
over- and underlain by other tuffaceous unsaturated units. Unsaturated Zone Flow and 
Transport Model Process Model Report (CRWMS M&O 2000 [DIRS 151940]) is a process 
model report generated for the Site Recommendation, and was cancelled for direct use in the 
TSPA-LA along with all Site Recommendation process model reports. However, that reports 
analyses of gas saturation are pertinent to this analysis. This is further supported by the 
unsaturated zone gas flow analyses by Simmons (2004 [DIRS 166960], Section 7.9.1.8). The 
matrix and the fractures above and below the repository horizon have a gas saturation generally 
greater than 50 percent (CRWMS M&O 2000 [DIRS 151940], Section 3.6), making these areas 
more gas- than liquid-permeable. In the repository horizon, gas saturation in the matrix varies 
from 10 to 30 percent (CRWMS M&O 2000 [DIRS 151940], Section 3.6), gas saturation in the 
fractures is much higher, causing gas to preferentially flow in the fractures. Consequently, any 
gas produced in the repository will tend to flow through the fracture complex with some gas 
diverted to the matrix at unit interfaces or where fracture networks pinch out (CRWMS M&O 
2000 [DIRS 151940], Section 3.3). These flow paths will disperse flammable gas concentrations 
to levels below the ignition point. Additionally, airflow through Yucca Mountain maintains 
repository pressures within only a few pascals of atmospheric pressure. Given these conditions, 
any flammable gas produced, once released, will quickly disperse and be diluted due to the 
presence of in situ inert gases (nitrogen in the air, water vapor, and CO2 and CH4 from carbide 
reactions) in the repository. 
TSPA Disposition: 
N/A 
Supporting Documents: 
CRWMS M&O 2000 [DIRS 151940] 
Related FEPs: 
2.1.02.08.0A, Pyrophoricity From DSNF 
April 2004 83 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
6.2.19 Internal Corrosion of Waste Packages Prior To Breach 
FEP Number: 
2.1.03.06.0A 
FEP Description: Aggressive chemical conditions within the waste package could contribute to 
corrosion from the inside out. Effects of different waste forms, including CSNF and DSNF, are 
considered in this FEP. 
Descriptor Phrases: 
Internal corrosion of CSNF 
Internal corrosion of DSNF 
Chemical effects from radiolysis in-package 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: It can be assumed that only a limited amount of water will be allowed by 
waste acceptance criteria inside a waste package (Section 5.2.19 of this analysis report), and 
there will be an inert gas backfill of waste packages and containers. Therefore, internal 
corrosion of waste packages prior to waste package breach can be excluded based on low 
consequence because only minor corrosion can occur within the low humidity environment of an 
unbreached package. Therefore, omission of minor internal corrosion of the waste package prior 
to breach will not significantly change the calculated radiological exposures and radionuclide 
releases. Radiolysis products could contribute to the interior corrosion, but they have been 
excluded based on low consequence. See FEP 2.1.13.01.0A. 
Limited water is allowed in the CSNF and HLW-glass waste packages and that which exists 
initially will be rapidly consumed by reaction with the waste form. Consequently, prior to the 
breach of the waste packages, there should be little corrosion in the CSNF, DSNF, and HLW 
waste packages from normal chemical and water vapor mechanisms. Some DSNF-containing 
waste packages, such as those containing N-Reactor fuel, may have more water, but this water 
would be scavenged by the waste form due to the rapid corrosion rate of the metallic uranium 
that is the matrix of N-Reactor SNF compared with other SNF (BSC 2003 [DIRS 163693], 
Table 9; Gray and Einziger 1998 [DIRS 109691], Section 4), as well. After breach of the waste 
package, inside-out corrosion can be expected to add little to the net area through which fluids 
might enter the WF. 
In general, no credit will be taken in the TSPA for DSNF canisters (within the waste package) as 
a barrier to fuel degradation and radionuclide mobilization. This decision was made because the 
canisters will be constructed of stainless steel, which will degrade relatively quickly once the 
waste package fails. Because no credit will be taken for the canisters, mechanisms that might 
enhance canister degradation do not impact the predicted consequences from the DSNF. 
TSPA Disposition: 
N/A 
April 2004 84 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Supporting Documents: 
BSC 2003 [DIRS 163693] 
Related FEPs: 
2.1.09.01.0B, Chemical characteristics of water in waste package 
6.2.20 Chemical Characteristics of Water in Waste Package 
FEP Number: 
2.1.09.01.0B 
FEP Description: Chemical characteristics of the water in the waste packages (pH and 
dissolved species) may be affected by interactions with materials or steel used in the waste 
packages or waste forms and by the inflowing water from the drifts and near-field host rock. 
The in-package chemistry, in turn may influence dissolution and transport as contaminants move 
through the waste, EBS, and down into the unsaturated zone. 
Descriptor Phrases: 
Chemical effects from corrosion products in-package 
Chemical effects from waste package degradation 
Chemical effects from waste form degradation 
Time-dependent THC characteristics in-package (temperature, pH, water concentration, 
gas concentration) 
Inflowing water composition (into waste package) 
Spatial heterogeneity of HLW glass degradation 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: The chemical characteristics of water in the waste package are included in 
In-Package Chemistry Abstraction (BSC 2003 [DIRS 161962]). The IPCM is a fully coupled 
reaction-path chemical model which included the effects of waste-form dissolution, metal alloy 
corrosion/dissolution, precipitation of metal oxide corrosion products, precipitation of complex 
mineral phases, reaction kinetics, thermal effects, interior waste package void space, interactions 
of codisposed waste forms, oxidationreduction reactions, heterogeneous chemical reactions, and 
seepage composition on the resulting fluid chemistry. 
The variability in the initial chemistry of incoming water is included by considering three 
incoming water compositions in the IPCM. These compositions were then altered based on the 
corrosion rates of the steel components of the waste package internals and the seepage flux. 
These analyses resulted in a set of independent variables (steel corrosion rates, sulfur content of 
the WP inner barrier alloy A516, and the partial pressure of carbon dioxide) that were sampled to 
establish an uncertainty band for the determination of in-package chemistry. The outputs of the 
IPCM are time-varying aqueous inventories of fluid composition, reactant abundances, product 
April 2004 85 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
abundances, and solution properties. Solution parameters are abstracted into a TSPA-compatible 
format in In-Package Chemistry Abstraction (BSC 2003 [DIRS 161962]); therefore, all of the 
processes included in the IPCM are reflected in In-Package Chemistry Abstraction (BSC 2003 
[DIRS 161962]) and thus in the TSPA. The IPCM parameters include pH, ionic strength, total 
carbonate, Eh, chloride, and fluoride. 
Supporting Documents: 
BSC 2003 [DIRS 161962] 
Related FEPs: 
2.1.03.06.0A, Internal Corrosion of Waste Packages Prior to Breach 
2.1.09.02.0A, Chemical Interaction With Corrosion Products 
2.1.09.04.0A, Radionuclide Solubility, Solubility Limits, and Speciation in the Waste 
Form and EBS 
2.1.09.06.0A, ReductionOxidation Potential in Waste Package 
6.2.21 Chemical Interaction with Corrosion Products 
FEP Number: 
2.1.09.02.0A 
FEP Description: Corrosion products produced during degradation of the waste form, metallic 
portions of the waste package, and metals in the drift (rock bolts, steel in invert, gantry rails) may 
affect the mobilization and transport of radionuclides. Corrosion products may facilitate 
sorption/desorption and coprecipitation/dissolution processes. Corrosion products may form a 
rind around the fuel that could (1) restrict the availability of water for dissolution of 
radionuclides or (2) inhibit advective or diffusive transport of water and radionuclides from the 
waste form to the EBS. Corrosion products also have the potential to retard the transport of 
radionuclides to the EBS. Finally, corrosion products may alter the local chemistry, possibly 
enhancing dissolution rates for specific waste forms, or altering radionuclide solubility. 
Descriptor Phrases: 
Colloid formation 
Effects of corrosion products on waste form degradation 
Time-dependent THC characteristics in-drift (solubility) 
Chemical effects from corrosion products in-drift 
Effects of corrosion products on waste package degradation 
Chemical effects from corrosion products in-package 
Time-dependent THC characteristics in-package (solubility) 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: The IPCM model developed in In-Package Chemistry Abstraction 
(BSC 2003 [DIRS 161962]) addresses in-package corrosion products and their effect on in- 
April 2004 86 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
package chemistry. The corrosion products of the steel and aluminum alloys in the waste 
package and their control on the concentration of aqueous species are of primary importance in 
determining the pH and ionic strength of the solution. If these corrosion products were not 
allowed to form during the simulations, then the resulting pH and ionic strength values would be 
much different than the results presented throughout Section 6.6 of In-Package Chemistry 
Abstraction (BSC 2003 [DIRS 161962]). Section 6.8 of In-Package Chemistry Abstraction 
(BSC 2003 [DIRS 161962]) examines the effect of surface complexation of aqueous species with 
waste package corrosion products and provides a pH range for in-package fluids to be used in 
TSPA-LA. The effects of interactions of corrosion products with the in-package chemistry are 
implicitly included in the abstractions passed to TSPA as part of In-Package Chemistry 
Abstraction (BSC 2003 [DIRS 161962]). The IPCM parameters include pH, ionic strength, total 
carbonate, Eh, chloride, and fluoride. 
As described in Waste Form and In-Drift Colloids-Associated Radionuclide Concentrations: 
Abstraction and Summary (BSC 2003 [DIRS 166845]), fixed and suspended colloidal corrosion 
products are modeled in the waste package. Suspended colloidal corrosion products are modeled 
in the EBS. Corrosion colloids are assumed to form and are subject to concentration and 
stability constraints controlled by the aqueous chemistry. The potential development of rinds on 
fuel and glass waste surfaces has been implicitly included in the development of the colloid 
model by incorporating laboratory data derived from fuel and glass waste corrosion experiments. 
Clogging of waste package breached zones by corrosion products, as described in Section 6.3.1.3 
of Waste Form and In-Drift Colloids-Associated Radionuclide Concentrations: Abstraction and 
Summary (BSC 2003 [DIRS 166845]), is addressed in FEP 2.1.03.10.0A (Healing of waste 
packages) in FEPs Screening of Processes and Issues in Drip Shield and Waste Package 
Degradation (BSC 2004 [DIRS 162200]). 
The conceptual model and TSPA implementation for the colloidal radionuclide source term, 
developed from YMP-relevant data, include specific treatment of radionuclides associated with 
corrosion-product colloids formed during corrosion of waste package material and other metallic 
materials. In the TSPA-LA model, mass concentrations of corrosion product colloids are based 
on experimental observations on the degradation of miniature waste packages and iron oxide 
colloid concentrations observed in natural groundwaters within high iron-bearing geologic strata 
(BSC 2003 [DIRS 166845], Section 6.3.1.3). Aqueous chemical parameters, specifically ionic 
strength and pH, control the stability of the colloids. The model includes a provision for 
irreversible attachment of a fraction of Pu to corrosion-generated iron oxyhyrodroxide colloids 
and fixed corrosion mass (BSC 2003 [DIRS 166845], Section 6.3.3.2). 
Model implementation of the following phenomena related to colloids and steel corrosion are 
described in Waste Form and In-Drift Colloids-Associated Radionuclide Concentrations: 
Abstraction and Summary (BSC 2003 [DIRS 166845]): 
 Formation of iron oxyhydroxide colloids and fixed corrosion products in the waste 
package (Section 6.3.1.3); 
 Estimation of stability and mass concentration of iron oxyhydroxide colloids from 
experimental results and calculated ionic strength and pH of in-package and in-drift 
fluids (Section 6.3.2.3); 
April 2004 87 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
 Irreversible sorption of Pu and Am onto iron oxyhydroxide colloids and fixed corrosion 
products (Section 6.3.3.2); 
 Reversible sorption of dissolved Pu, Am, Pa, Th, and Cs onto iron oxyhydroxide 
colloids using developed Kd values and the estimated colloid mass concentrations 
(Section 6.3.3.1). 
The effects of dissolved iron on the glass dissolution rates are addressed by using the results of 
tests in which iron products were added to the test solution when determining model parameter 
values for pH dependence () in acidic and alkaline solutions (BSC 2004 [DIRS 167619], 
Section 6.5.4.1). The effects of glass alteration phases on the glass dissolution rate are addressed 
by including the results of tests in which alteration phases and/or alteration product rinds form 
when determining model parameter values for kE (BSC 2004 [DIRS 167619], Section 6.5.4.3). 
Supporting Documents: 
BSC 2003 [DIRS 161962] 
BSC 2003 [DIRS 166845] 
BSC 2004 [DIRS 167619] 
Related FEPs: 
2.1.03.06.0A, Internal Corrosion of Waste Packages Prior to Breach 
2.1.09.04.0A, Radionuclide Solubility, Solubility Limits, and Speciation in the Waste 
Form and EBS 
2.1.09.06.0A, Reduction-Oxidation Potential in Waste Package 
2.1.09.01.0B, Chemical Characteristics of Water in Waste Package 
6.2.22 Radionuclide Solubility, Solubility Limits, and Speciation in the Waste Form and 
EBS 
FEP Number: 
2.1.09.04.0A 
FEP Description: Degradation of the waste form will mobilize radionuclides in the aqueous 
phase. Factors to be considered in this FEP include the initial radionuclide inventory, 
justification of the limited inventory included in evaluations of aqueous concentrations, and the 
solubility limits for those radionuclides. 
Descriptor Phrases: 
Radionuclide solubility in-package 
Radionuclide solubility (concentration) limits 
Radionuclide speciation 
Radionuclide solubility in-drift 
Screening Decision: 
Included 
Screening Argument: 
N/A 
April 2004 88 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
TSPA Disposition: Radionuclide solubility limits are also called dissolved concentration limits. 
Solubility models for 14 elements (plutonium, neptunium, uranium, thorium, americium, 
actinium, protactinium, radium, technetium, iodine, strontium, carbon, lead, and cesium) were 
developed, based on geochemical model calculations, where radionuclide speciation is 
considered, and documented in Dissolved Concentration Limits of Radioactive Elements 
(BSC 2003 [DIRS 163152]). Solubility models for these elements, except for lead, are included 
in the TSPA-LA model and used to constrain the maximum radionuclide concentrations in waste 
packages and in the invert. In the TSPA model, inventory concentrations of radioactive elements 
released from the waste forms (CSNF, DSNF, and HLW glass) are calculated according to the 
dissolution or degradation rates of waste forms and the volume of water within the (breached) 
waste package, and radionuclide inventory. Then, the inventory concentrations are compared 
against their solubility limits. The real concentrations available for transport are the lesser 
between the inventory concentrations and the solubility limits. If the inventory concentration is 
greater than its solubility limit, then the amount of that radioactive element in excess of the 
solubility limit will be kept in the inventory, and potentially available for transport at a later time. 
For plutonium, neptunium, uranium, thorium, americium, actinium, and protactinium, solubilities 
are tabulated as functions of pH and log fCO2 plus two or more uncertainty terms accounting for 
uncertainties associated with thermodynamic data, variations in water chemistry, and 
identification of solubility controlling solids (BSC 2003 [DIRS 163152], Sections 6.5 through 
6.11). For radium and lead, solubilities are given as distributions or segmental constants (BSC 
2003 [DIRS 163152]) Sections 6.12 and 6.13). 
Under repository conditions, technetium, carbon, iodine and cesium are considered soluble and 
no solubility controlling solids are expected to form. Consequently, their releases are considered 
to be controlled by the dissolution rate of waste forms and the waste inventory (BSC 2003 [DIRS 
163152], Sections 6.14 through 6.17). Strontium carbonate or sulfate may form under repository 
conditions, but they are quite soluble. Moreover, strontium does not contribute to dose 
significantly. Therefore, it is conservatively assumed that no strontium solubility-controlling 
solid exists and its release is considered to be controlled by the dissolution rate of waste forms 
and the waste inventory (BSC 2003 [DIRS 163152], Section 6.18). 
Supporting Documents: 
BSC 2003 [DIRS 163152] 
Related FEPs: 
2.1.03.06.0A, Internal Corrosion of Waste Packages Prior to Breach 
2.1.09.06.0A, ReductionOxidation Potential in Waste Package 
2.1.09.01.0B, Chemical Characteristics of Water in Waste Package 
2.1.09.02.0A, Chemical Interaction with Corrosion Products 
2.1.09.08.0A, Diffusion of Dissolved Radionuclides in EBS 
2.1.09.08.0B, Advection of Dissolved Radionuclides in EBS 
6.2.23 ReductionOxidation Potential in Waste Package 
FEP Number: 
2.1.09.06.0A 
April 2004 89 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
FEP Description: The redox potential in the waste package influences the oxidation of waste 
form materials and the in-package solubility of radionuclide species. Local variations in the inpackage 
redox potential can occur. 
Descriptor Phrases: 
Time-dependent THC characteristics in-package (temperature, pH, water concentration, 
gas concentration) 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: In-Package Chemistry Abstraction (BSC 2003 [DIRS 161962]) bounds the 
redox potential of the groundwater seepage in the same equilibrium model that addresses the inpackage 
pH. The oxidationreduction processes inside of the waste package are explicitly 
modeled in the IPCM and thus implicitly included in In-Package Chemistry Abstraction 
(BSC 2003 [DIRS 161962]). Oxidation of the metal waste package components is the primary 
process by which the waste packages corrode and the initial water composition is altered. In the 
absence of redox reactions there would be little alteration of the water inside of a failed waste 
package (BSC 2003 [DIRS 161962]). 
The in-package chemistry abstraction shows that for CSNF waste packages there is some 
variability in the minerals and aqueous species in the reacted water chemistry of the water 
exiting a failed waste package for three different seepage compositions; J-13, Na-porewater, 
and Ca-porewater. (BSC 2003 [DIRS 161962], Table 2 and Section 6.6.1.2). However, this 
variability is small, and so the pH and ionic strength (redox potential) of the reacted solutions are 
unaffected by the variations in composition of the seepage water. Thus variability of the input 
water compositions, while included in TSPA-LA, are expected to have little or no effect on 
CSNF in-package chemistry. 
For codisposed N-Reactor waste packages the mineral assemblages formed for the three seepage 
compositions indicate that the composition of the seepage water has little influence over which 
minerals will form and their amounts. Therefore, variability in the composition of water entering 
failed codisposed waste packages, while included in TSPA-LA, is expected to have little or no 
effect on the in-package chemistry (BSC 2003 [DIRS 161962], Section 6.6.2.2 and Figure 13). 
The redox potential inside the waste package and its effect on in-package chemistry is implicitly 
included in the abstractions passed to TSPA as part of In-Package Chemistry Abstraction 
(BSC 2003 [DIRS 161962]). The IPCM parameters are pH, ionic strength, total carbonate, Eh, 
chloride, and fluoride. 
Supporting Documents: 
BSC 2003 [DIRS 161962] 
April 2004 90 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Related FEPs: 
2.1.09.01.0B, Chemical Characteristics of Water in Waste Package 
2.1.09.02.0A, Chemical Interaction with Corrosion Products 
6.2.24 Reaction Kinetics in Waste Package 
FEP Number: 
2.1.09.07.0A 
FEP Description: Chemical reactions, such as radionuclide dissolution/precipitation reactions 
and reactions controlling the reduction-oxidation state, may not be at equilibrium within the 
waste package. 
Descriptor Phrases: 
Reaction path analysis using precipitation/dissolution rates 
Reaction path analysis using waste form degradation rates 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: Reaction kinetics and precipitation/dissolution rates are included in the 
TSPA-LA in-package chemistry model abstraction. The IPCM uses kinetic reactants to 
represent the SNF (BSC 2003 [DIRS 161962], Section 4.1.2) and the waste package components 
(BSC 2003 [DIRS 161962], Section 4.1.3.3). The kinetic rates used in the model were either 
linear, where a fixed amount of reactant is added at each time step, or a transition-state rate law 
where the amount of reactant added to the system depends on chemical properties of the aqueous 
phase. The effect of varying the kinetics on the in-package chemistry was examined (BSC 2003 
[DIRS 161962], Section 6.7) where the rates were decreased to assess the contribution to 
uncertainty in pH and ionic strength for inclusion in the abstractions of pH for TSPA. 
The variability in the kinetics of the reactants is included in the abstractions passed to TSPA both 
implicitly by their use in the IPCM, and explicitly via the contribution of the kinetics to output 
uncertainty. The effects of reaction kinetics on in-package chemistry are passed to TSPA as part 
of In-Package Chemistry Abstraction (BSC 2003 [DIRS 161962]). The IPCM parameters are 
pH, ionic strength, total carbonate, Eh, chloride, and fluoride. 
Supporting Documents: 
BSC 2003 [DIRS 161962] 
Related FEPs: 
2.1.09.06.0A Reduction-Oxidation Potential in Waste Package 
2.1.09.01.0B, Chemical Characteristics of Water in Waste Package 
2.1.09.02.0A, Chemical Interaction with Corrosion Products 
April 2004 91 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
6.2.25 Secondary Phase Effects on Dissolved Radionuclide Concentrations 
FEP Number: 
2.1.09.10.0A 
FEP Description: Inclusion of radionuclides in secondary uranium mineral phases, such as 
neptunium in schoepite and uranium silicates, could affect radionuclide concentrations in water 
in contact with the waste form. During radionuclide alteration, the radionuclides could be 
chemically bound to immobile compounds and result in a reduction of radionuclides available 
for mobilization. 
Descriptor Phrases: 
Secondary phase effects on dissolved radionuclide concentrations in-package 
Secondary phase effects on dissolved radionuclide concentrations in-drift 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: Incorporation of certain radionuclide(s) into corrosion products formed 
during the alteration of spent nuclear fuel may reduce radionuclide concentrations in waters that 
have contacted fuel and its corrosion products (BSC 2003 [DIRS 163152], Section 6.3.2). It is 
well known that measured actinide concentrations in spent nuclear fuel corrosion experiments 
are several orders of magnitude lower than calculated solubilities assuming pure phase control 
(e.g., Werme and Spahiu 1998 [DIRS 113466], Chen 2001 [DIRS 161997]). This large 
discrepancy was attributed to the incorporation of actinides into secondary uranium phases (e.g., 
Werme and Spahiu 1998 [DIRS 113466]; Quinones et al. 1996 [DIRS 161925]; Chen 2001 
[DIRS 161997]; Chen et al. 2002 [DIRS 161996]), which was not accounted for in conventional 
solubility calculations. 
It can be assumed (Section 5.2.25 of this analysis report) that the constant ratio of neptunium to 
uranium in the leachate is the result of the formation of solid solutions in the secondary uranium 
phases. Three lines of evidence support the incorporation explanation. First, Burns et al. (1997 
[DIRS 100389]) predicted that the substitutions Pu6+  U6+ and (Np5+, Pu5+)  U6+ are likely 
to occur in most U6+ structures based on an analysis of the crystal.chemical properties of the 
U.O bond, Np.O bond, and Pu.O bond. 
The previous theoretical prediction has been confirmed by direct observations of neptunium in 
uranyl minerals. Buck et al. (1998 [DIRS 100388]) examined corrosion products of spent 
nuclear fuel drip tests by electron energy loss spectroscopy analyses in a transmission electron 
microscope. As reported in their study, neptunium was to be incorporated into dehydrated 
schoepite (UO3  0.8H2O) or metaschoepite (UO3  2H2O). Finch et al. (2002 [DIRS 161979]) 
also reported experimental results on neptunium incorporation into dehydrated schoepite and 
estimated that the amount of neptunium incorporated in dehydrated schoepite can be as high as 
2 percent of the host solid based on electron energy loss spectroscopy measurement. However, 
Finch has since questioned the reliability of the electron energy loss spectroscopy measurement 
(Adkins 2003 [DIRS 163864]). 
April 2004 92 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
More significant progress in this direction is the successful synthesis of neptunium-bearing 
uranium minerals. Burns et al. (1997 [DIRS 100389]) reported the synthesis of uranophane 
(Ca(UO2SiO3OH)2  5H2O) and Na-compreignacite (Na2[(UO2)3O2(OH)3]2(H2O)5) containing 
neptunium ranging up to 497 ppm (Buck et al. 2003 [DIRS 168411]). Furthermore, they found 
that there is a linear relationship between the neptunium content of -uranophane and Nacompreignacite 
and the Np5+ concentration in the initial solution, and Np is incorporated into 
crystals of alpha-uranophane and Np-compreignacite in approximately the same concentrations 
as presented in the mother solutions. Buck (2003 [DIRS 168411]), Friese et al. (2003 [DIRS 
162634]), and Friese (2003 [DIRS 163911]) also reported the successful synthesis of neptuniumbearing 
studtite (UO44H2O). In the neptunium incorporation experiments conducted by Burns et 
al. (Buck et al. 2003 [DIRS 168411]), it is found that schoepite, the major secondary uranyl 
mineral that forms in the process of spent nuclear fuel degradation and believed to be critical for 
neptunium immobilization, contains only a small amount of neptunium (a few parts per million). 
Burns et al. attribute this to the lack of suitable low-valence cations in their experiments (Buck et 
al. 2003 [DIRS 168411]). This may have prevented a charge-balance mechanism from 
happening, which is a prerequisite for neptunium incorporation into uranyl minerals. Burns et al. 
planned to conduct experiments using various counter ions (potassium, cesium, and calcium, 
etc.) to examine their effects on neptunium incorporation into schoepite (Buck et al. 2003 [DIRS 
168411]). 
Finally, Chen (2001 [DIRS 161997]) and Chen et al. (2002 [DIRS 161996]) analyzed the relative 
release rates of neptunium and uranium in the fuel degradation experiments referred to above and 
found a close correlation between neptunium and uranium concentrations in the leachates. In 
other words, the ratio of neptunium to uranium in the leachate is equal to the ratio in the spent 
nuclear fuel used in the experiments. Bruno et al. (1998 [DIRS 101565]) have also observed a 
coherent relation between neptunium and uranium in spent nuclear fuel dissolution experiments 
in the Spanish Nuclear Waste Program. These results are corroborated by the linear relationship 
between neptunium contents in uranophane and Na-compreignacite and the mother solution that 
was found by Burns et al. (Buck et al. 2003 [DIRS 168411]). 
There are several conceivable explanations for this coherent relation between neptunium and 
uranium. First, it is the result of congruent dissolution of fuel matrix with few effects of 
secondary phases. The second explanation is that neptunium and uranium behave coherently, 
not only in the dissolution process but also in the process of secondary phase precipitation. The 
third entails uranium and neptunium self-formed solid phases controlling their associated 
solubilities at a ratio equal to that of uranium to neptunium in the fuel. However, only one 
explanation is reasonable. The first can be eliminated because a large portion of reacted uranium 
precipitated as secondary uranium minerals; therefoe, they are incongruent reactions. The third 
hypothesis relies on coincidence and, as stated earlier, no neptunium pure phases have been 
reported in fuel dissolution experiments; thus, it is unsubstantiated. Therefore, the second 
explanation is the most reasonable. In other words, the coherent relation between neptunium and 
uranium is the result of formation of solid solutions between neptunium and uranium. 
A mechanistic model was developed (CRWMS M&O 1998 [DIRS 100362]; Chen et al. 1999 
[DIRS 145441]; Chen 2003 [DIRS 162709]) that assumed neptunium is incorporated into 
secondary uranyl minerals (Section 5.2.25 of this analysis report), based on experimental 
observations that the neptunium and uranium behave coherently in the release from the waste 
April 2004 93 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
form and in the secondary phase formation process. The reactive-transport model considers 
equilibrium reactions and kinetic dissolution or precipitation reactions of solids under flow 
conditions. The modeling results were compared against laboratory measurements and 
observations of natural analogs. For repository conditions, the model predicted neptunium 
solubility ranges from 1.0  10-9 to 1.0  10-6 mol/L, about two orders of magnitude lower than 
the value used for the TSPA-VA base case. A TSPA sensitivity study used this solubility range 
and showed that the peak neptunium dose rate was reduced by a factor of 45 from the base case 
(DOE 1998 [DIRS 100550], Section 5.5.3). 
An alternative model of accounting for incorporation of neptunium into uranyl minerals was 
generated based on experimental data. The results produced by the alternative model captures 
experimental results well (BSC 2003 [DIRS 163152], Section 6.6.4.5). It eliminates the built-in 
conservatism in the conventional pure phase solubility approach and, thus, gives more realistic 
source terms. However, experimental studies on whether schoepite, the critical secondary uranyl 
phase, can incorporate and immobilize sufficient neptunium during spent nuclear fuel corrosion 
do not provide a solid basis for recommending this model to be used in the TSPA-LA model. 
Experimental work leading to data bearing on this model is under continuous review and may 
lead to an amended recommendation to adopt neptunium concentrations based on the 
coprecipitation model. 
In summary, although the inclusion of secondary-mineral phases in the TSPA-LA would reduce 
calculated doses because these solids have the ability to sequester radionuclides, no credit is 
taken for this process in the TSPA-LA. Therefore, secondary-phase effects on dissolved 
radionuclide concentrations are excluded from the TSPA-LA based on low consequence because 
they have no adverse effects on performance. 
TSPA Disposition: 
N/A 
Supporting Documents: 
BSC 2003 [DIRS 163152] 
Related FEPs: 
2.1.09.06.0A, Reduction-Oxidation Potential in Waste Package 
2.1.09.01.0B, Chemical Characteristics of Water in Waste Package 
2.1.09.02.0A, Chemical Interaction with Corrosion Products 
6.2.26 Complexation in EBS 
FEP Number: 
2.1.09.13.0A 
FEP Description: The presence of organic complexants in water in the EBS could augment 
radionuclide transport by providing a transport mechanism in addition to simple diffusion and 
advection of dissolved material. Organic complexants may include materials found in natural 
groundwater such as humates and fulvates, or materials introduced with the waste or engineered 
materials. 
April 2004 94 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Descriptor Phrases: 
Complexation in EBS 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: Repository temperatures are expected to remain sufficiently high to 
prevent water condensation and consequent formation of media for organic activity for several 
thousand years. Once lower temperatures recur, microbial populations will not be sufficient to 
generate significant concentrations of radionuclide-chelating organics. Organic complexants are 
typically found in sufficient quantity to move radionuclides long distances only in organic rich 
soils and groundwaters  or downgradient from dumps containing high levels of organic matter. 
Bounding biomass estimates for the repository are approximately 100 grams (dry) per meter of 
drift (CRWMS M&O 2000 [DIRS 151561], Table 71 and Section 7.2), several orders of 
magnitude below the cases cited above and insufficient to produce high chelate levels. Note, 
moreover, that many organic complexants also complex with multivalent nonradioactive metal 
cations, such as Fe(III) and Ni(II), will be present in much higher concentrations than the 
actinides (BSC 2001 [DIRS 162733]). This would also inhibit wide-scale complexation of 
actinides by organic matter. Also, the estimates of microbe masses growing in the repository 
system suggest that the effects on the bulk in-drift geochemistry should be small 
(CRWMS M&O 2000 [DIRS 151561], p. 154), and that these colloids are accounted for as well 
in the dissolved concentrations model report (BSC 2003 [DIRS 163152]). In conclusion, the 
magnitude and time of the resulting radiological exposures to the reasonably maximally exposed 
individual, or radionuclide releases to the accessible environment, are not significantly changed 
by the omission of this FEP from the TSPA-LA. 
TSPA Disposition: 
N/A 
Supporting Documents: 
CRWMS M&O 2000 [DIRS 151561] 
BSC 2001 [DIRS 162733] 
BSC 2003 [DIRS 163152] 
Related FEPs: 
2.1.09.06.0A, ReductionOxidation Potential in Waste Package 
2.1.09.01.0B, Chemical Characteristics of Water in Waste Package 
2.1.09.02.0A, Chemical Interaction with Corrosion Products 
6.2.27 Formation of True (Intrinsic) Colloids in EBS 
FEP Number: 
2.1.09.15.0A 
FEP Description: True colloids are colloidal-sized assemblages (between approximately 1 
nanometer and 1 micrometer in diameter) consisting of hydrolyzed and polymerized 
radionuclides. They may form in the waste package and EBS during waste-form degradation 
and radionuclide transport. True colloids are also called primary colloids, real colloids, Type I 
April 2004 95 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
colloids, Eigenkolloide, and intrinsic colloids (or actinide intrinsic colloids, for those including 
actinide elements). 
Descriptor Phrases: 
Formation of true (intrinsic) colloids in EBS 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: Saturated solutions in close proximity to degrading waste may form true 
colloids (BSC 2003 [DIRS 166845], Section 6.3.1). However, true colloids may not be routinely 
detectable, as they are difficult to distinguish from dissolved radionuclides. Further, they would 
be likely to dissociate in dilute fluids typical of the repository as well as sorb to substrates. 
Uranium colloids that have been observed to form are meta-schoepite or other oxyhydroxides 
that are soluble in the dilute oxic waters expected in the unsaturated and saturated zone 
environments. Consequently, these colloids will eventually dissolve, in part because of their 
small size and large reactive surface area, and in part because of increasingly dilute groundwater 
conditions away from the near field (BSC 2003 [DIRS 166845], Sections 6.3.1.2 and 6.3.1.2.1 
through 6.3.1.2.4). 
Uranium is modeled in the EBS as U-aqueous complexes not associated with colloids (BSC 2003 
[DIRS 166845], Sections 6.3.1.2.1, 6.3.1.2.2, and 6.3.1.2.4). Any uranium that might form 
metastable true colloids and then subsequently dissolve is already considered as part of these 
U-aqueous complexes. The mass of uranium that might potentially form metastable true colloids 
is not expected to be significant with respect to the total mass of radionuclides, nor is the 
enhanced colloid transport in the EBS of those potential true colloids significant with respect to 
the total transport time to the accessible environment. Therefore, the impact of true colloids on 
radiological exposures to the reasonably maximally exposed individual and radionuclide releases 
to the accessible environment would not be significant. 
In summary, true colloids, even if they were to form, would be metastable and would dissolve in 
the dilute waters away from the waste package. In their metastable form in the EBS, they would 
not have a significant effect on transport. Therefore, the formation of true colloids in the EBS 
has been excluded from the TSPA-LA on the basis of low consequence. 
TSPA Disposition: 
N/A 
Supporting Documents: 
BSC 2003 [DIRS 166845] 
ANL-WIS-MD-000009 REV 01 April 2004 96 Miscellaneous Waste-Form FEPs 
Related FEPs: 
2.1.09.16.0A, Formation of Pseudocolloids (Natural) in EBS 
2.1.09.17.0A, Formation of Pseudocolloids (Corrosion Product) in EBS 
2.1.09.18.0A, Formation of Microbial Colloids in EBS 
2.1.09.23.0A, Stability of Colloids in EBS 
2.1.09.25.0A, Formation of Colloids (Waste-Form) by Coprecipitation in EBS 
2.1.09.19.0B, Advection of Colloids in EBS 
2.1.09.24.0A, Diffusion of Colloids in EBS 
6.2.28 Formation of Pseudo-Colloids (Natural) in EBS 
FEP Number: 
2.1.09.16.0A 
FEP Description: Pseudo-colloids are colloidal-sized assemblages (between approximately 1 
nanometer and 1 micrometer in diameter) of nonradioactive material that have radionuclides 
bound or sorbed to them. Natural pseudo-colloids include microbial colloids, mineral fragments 
(clay, silica, iron oxyhydroxides), and humic and fulvic acids. This FEP addresses radionuclidebearing 
pseudocolloids formed from host-rock materials and all interactions of the waste and 
EBS with the host rock environment except corrosion. 
Descriptor Phrases: 
Natural groundwater colloids (clay, silica, iron oxyhydroxide, invert/tuff ) 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: Natural colloids are modeled as smectite clay colloids as discussed in Waste 
Form and In-Drift Colloids-Associated Radionuclide Concentrations: Abstraction and Summary 
(BSC 2003 [DIRS 166845], Section 6.3.1.4). Pseudocolloids generally form as a result of 
dissolved (aqueous) radionuclides sorbing to existing colloids (BSC 2003 [DIRS 166845], 
Section 6.3.1). Kd values are developed to model reversible sorption of Pu, Am, Th, Pa, and Cs 
onto smectite groundwater colloids to form pseudocolloids (BSC 2003 [DIRS 166845], 
Section 6.3.3.1). Stability of the colloids is determined based on ionic strength and pH; 
concentration is modeled as a sampled parameter based on field observations in the Yucca 
Mountain vicinity (BSC 2003 [DIRS 166845], Sections 6.3.2.2 and 6.3.2.5). Sorption to iron 
oxyhydroxide colloids is also modeled, with implementation in TSPA-LA simplified by using 
one set of Kd values. Glass waste form colloids are also modeled as smectite clay colloids, based 
on experimental observations at ANL, and are modeled to respond to ionic strength and pH 
conditions in the same manner as natural groundwater colloids (see FEP 2.1.09.25.0A). 
Supporting Documents: 
BSC 2003 [DIRS 166845] 
April 2004 97 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Related FEPs: 
2.1.09.15.0A, Formation of True (Intrinsic) Colloids in EBS 
2.1.09.17.0A, Formation of Pseudocolloids (Corrosion Product) in EBS 
2.1.09.18.0A, Formation of Microbial Colloids in EBS 
2.1.09.23.0A, Stability of Colloids in EBS 
2.1.09.25.0A, Formation of Colloids (Waste-Form) by Coprecipitation in EBS 
2.1.09.19.0B, Advection of Colloids in EBS 
2.1.09.24.0A, Diffusion of Colloids in EBS 
6.2.29 Formation of Pseudocolloids (Corrosion Product) in EBS 
FEP Number: 
2.1.09.17.0A 
FEP Description: Pseudocolloids are colloidal-sized assemblages (between approximately 1 
nanometer and 1 micrometer in diameter) of nonradioactive material that have radionuclides 
bound or sorbed to them. Corrosion product pseudocolloids include iron oxyhydroxides from 
corrosion and degradation of the metals in the EBS and silica from degradation of cementitious 
materials. 
Descriptor Phrases: 
Corrosion product colloids (metal oxyhydroxide, iron oxides, and cement grout silica) 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: The TSPA-LA model abstraction (BSC 2003 [DIRS 166845]) includes 
treatment of radionuclides associated with corrosion of steel components of the waste package 
and surrounding drift materials. Fixed and colloidal corrosion products are modeled in the waste 
package (BSC 2003 [DIRS 166845], Section 6.3.3.2). Colloids formed from the corrosion of 
steel in the EBS are modeled as iron oxyhydroxide colloids, with estimated mass concentrations 
based on corrosion studies of miniature waste packages under YMP-relevant conditions and iron 
oxide colloid concentrations observed in groundwaters associated with high iron-bearing 
geologic strata (BSC 2003 [DIRS 166845], Section 6.3.1.3). 
Pseudocolloids generally form as a result of dissolved (aqueous) radionuclides sorbing to 
existing colloids (Section 6.3.1). Kd values were developed to model reversible sorption of Pu, 
Am, Th, Pa, and Cs onto iron oxyhydroxide corrosion colloids to form pseudocolloids 
(BSC 2003 [DIRS 166845], Section 6.3.3.1). In addition, Pu and Am are modeled as sorbing 
irreversibly to iron oxyhydroxide corrosion colloids (BSC 2003 [DIRS 166845], Section 6.3.3.2). 
Iron oxyhydroxide corrosion colloids are subject to concentration and stability constraints 
controlled by the aqueous chemistry, chiefly determined from the ionic strength and pH of inpackage 
and in-drift fluids (BSC 2003 [DIRS 166845], Section 6.3.2.3). 
April 2004 98 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
In the TSPA-LA model, a major fraction (approximately 0.9 or higher) of sorbed Pu is assumed 
to become irreversibly attached to corrosion products (stationary corrosion materuals and iron 
oxyhydroxide colloids). The remainder is assumed to become reversibly attached according to 
an appropriate Kd (BSC 2003 [DIRS 166845], Section 6.3.3.1, Table 10). This capability was 
included to acknowledge laboratory and field observations that the attachment of a large fraction 
of Pu to iron oxyhydroxide persists over the time of the observations and dissolved Pu 
concentrations are small, as observed in natural groundwater systems, because Pu is strongly 
sorbed onto stationary geologic media and colloids. 
Supporting Documents: 
BSC 2003 [DIRS 166845] 
Related FEPs: 
2.1.09.15.0A, Formation of True (Intrinsic) Colloids in EBS 
2.1.09.16.0A, Formation of Pseudocolloids (Natural) in EBS 
2.1.09.18.0A, Formation of Microbial Colloids in EBS 
2.1.09.23.0A, Stability of Colloids in EBS 
2.1.09.25.0A, Formation of Colloids (Waste-Form) by Coprecipitation in EBS 
2.1.09.19.0B, Advection of Colloids in EBS 
2.1.09.24.0A, Diffusion of Colloids in EBS 
6.2.30 Formation of Microbial Colloids in EBS 
FEP Number: 
2.1.09.18.0A 
FEP Description: This FEP addresses the formation and transport of microbial colloids in the 
waste and EBS. 
Descriptor Phrases: 
Microbial colloids 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: The effects of colloid-facilitated radionuclide transport stemming from 
biological activity have been excluded from the TSPA-LA model analysis. The potential 
influence of organic complexants from microbial metabolism on radionuclide transport is 
assumed (Section 5.2.30) not to affect the stability of inorganic (mineral) colloids (BSC 2003 
[DIRS 166845], Sections 5.9 and 6.3.4) (i.e., inorganic colloid stability is determined by fluid 
chemistry [ionic strength and pH]). Because inorganic aqueous conditions dominate DHLWG 
degradation rates, and microbial degradation is secondary, this process is excluded on the basis 
of low consequence. 
Microbe-induced reduction of radionuclide, metal, and sulfate contaminants would decrease their 
mobility (BSC 2003 [DIRS 166845], Section 6.3.4), which is beneficial to performance. 
Biofilms would tend to retard radionuclides, which is also beneficial. These beneficial processes 
can be excluded from the TSPA-LA on the basis of low consequence because they have no 
April 2004 99 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
adverse effects on performance (i.e., their influence would not lead to increased potential dose as 
their activity has been deemed to promote colloid instability and reduced transport of 
radionuclides in association with colloids). 
For microbes to impact near-field performance, microbes must be present and sufficient energy 
sources and nutrients must be available. For microbes to have an unfavorable impact on 
performance, they must accumulate radionuclides and must be transportable. Because of the 
relatively large microbes, they will be readily filtered during transport, along with any 
accumulated radionuclides, which is beneficial. Inclusion of microbial colloid transport in the 
waste and EBS would reduce calculated doses, and thus be potentially nonconservative. It is 
reasonable to conclude that the presence of organic polymers such as those produced by 
microbes and their byproducts will generally reduce the stability of inorganic colloids and result 
in enlargement of particles through heteroaggregation. It is also reasonable to conclude that 
microbes in the repository environment would likewise decrease the mobility and stability of 
colloids. 
Consequently, this FEP can be excluded based on low consequence. 
TSPA Disposition: 
N/A 
Supporting Documents: 
BSC 2003 [DIRS 166845] 
Related FEPs: 
2.1.09.15.0A, Formation of True (Intrinsic) Colloids in EBS 
2.1.09.16.0A, Formation of Pseudocolloids (Natural) in EBS 
2.1.09.17.0A, Formation of Pseudocolloids (Corrosion Product) in EBS 
2.1.09.23.0A, Stability of Colloids in EBS 
2.1.09.25.0A, Formation of Colloids (Waste-Form) by Coprecipitation in EBS 
2.1.09.19.0B, Advection of Colloids in EBS 
2.1.09.24.0A, Diffusion of Colloids in EBS 
6.2.31 Stability of Colloids in EBS 
FEP Number: 
2.1.09.23.0A 
FEP Description: For radionuclide-bearing colloids to affect repository performance, the 
colloids in the dispersion must remain suspended, that is, be stable, for time scales that are long 
relative to time required for groundwater transport. Further, they must carry significant 
concentrations of radionuclides. The stability of smectite colloids, which is applicable for YMP 
groundwater colloids and waste-form colloids, is determined primarily by ionic strength but also 
to an extent by pH. The stability of iron oxyhydroxide colloids, which is applicable to corrosionproduct 
colloids, is determined by both ionic strength and pH. 
Descriptor Phrases: 
Time-dependent THC characteristics in EBS (pH, ionic strength) 
April 2004 100 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: Colloids in the EBS modeled in Waste Form and In-Drift Colloids- 
Associated Radionuclide Concentrations: Abstraction and Summary (BSC 2003 
[DIRS 166845]) are smectite (waste form and groundwater colloids) and iron oxyhydroxide 
(steel corrosion colloids). Their stabilities are determined from ionic strength and pH of the 
in-package and in-drift fluids, as calculated in the TSPA-LA model calculations. 
The stabilities of both smectite and iron hydroxide colloids are determined at each time step 
executed in the TSPA-LA calculations (BSC 2003 [DIRS 166845], Section 6.3.2). These 
determinations are then used to calculate concentrations of radionuclides associated with the 
colloids (BSC 2003 [DIRS 166845], Section 6.3.3). Ionic strength, pH values, and dissolved 
radionuclide concentrations are taken from intermediate TSPA-LA calculation values from 
submodels described in the in-package chemistry waste package (ionic strength and pH), 
solubility limits (dissolved radionuclides), and near-field chemical environment (drift ionic 
strength and pH) model reports (BSC 2003 [DIRS 166845], Section 6.5). 
Supporting Documents: 
BSC 2003 [DIRS 166845] 
Related FEPs: 
2.1.09.15.0A, Formation of True (Intrinsic) Colloids in EBS 
2.1.09.16.0A, Formation of Pseudocolloids (Natural) in EBS 
2.1.09.17.0A, Formation of Pseudocolloids (Corrosion Product) in EBS 
2.1.09.18.0A, Formation of Microbial Colloids in EBS 
2.1.09.25.0A, Formation of Colloids (Waste-Form) by Coprecipitation in EBS 
2.1.09.19.0B, Advection of Colloids in EBS 
2.1.09.24.0A, Diffusion of Colloids in EBS 
6.2.32 Formation of Colloids (Waste-Form) by Coprecipitation in EBS 
FEP Number: 
2.1.09.25.0A 
FEP Description: Dissolved radionuclides and other ions may coprecipitate to form colloids. 
Coprecipitates may consist of radionuclides bound in the crystal lattice of a dominating mineral 
phase or may consist of radionuclides engulfed by a dominating mineral phase. 
Descriptor Phrases: 
Waste form colloids (HLW glass) 
Screening Decision: 
Included 
April 2004 101 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Screening Argument: 
N/A 
TSPA Disposition: Colloids apparently formed from coprecipitation of smectite and embedded 
radionuclide-bearing phases were observed in the DHLWG degradation tests (BSC 2003 
[DIRS 166845], Section 6.3.1.1). Colloids produced from degradation of DHLWG are modeled 
as smectite colloids with embedded (assumed permanently attached) radionuclides, Pu and 
Am. These may, in a broad sense, be considered coprecipitates. The concentrations of 
radionuclides associated with those colloids are based on empirical results from YMP-relevant 
DHLWG corrosion experiments (BSC 2003 [DIRS 166845], Sections 6.3.1.1 and 6.3.2.4). Mass 
concentrations of the colloids are based on those experiments with consideration of colloid 
mineralogy and the effects of ionic strength and pH on the stability of the colloids (BSC 2003 
[DIRS 166845], Sections 6.3.1.1, 6.3.2.2, and 6.3.2.4). Radionuclides associated with these 
colloids are modeled in the TSPA-LA model as irreversibly attached because they are 
embedded within the colloid matrix and can only be released upon the dissolution of the colloid. 
However, radionuclides within the aqueous environment can reversibly attach to the surfaces of 
these colloids and these colloid-radionuclide complexes can, therefore, be subject to transport in 
the TSPA-LA model as pseudocolloids. Radionuclide sorption onto these pseudocolloids is 
modeled by assigning Kd values developed to model reversible sorption of Pu, Am, Th, Pa, and 
Cs onto smectite colloids (BSC 2003 [DIRS 166845], Section 6.3.3.1). Similar to natural 
groundwater smectite colloids, the pseudocolloid stability is controlled by ionic strength and pH. 
Supporting Documents: 
BSC 2003 [DIRS 166845] 
Related FEPs: 
2.1.09.15.0A, Formation of True (Intrinsic) Colloids in EBS 
2.1.09.16.0A, Formation of Pseudocolloids (Natural) in EBS 
2.1.09.17.0A, Formation of Pseudocolloids (Corrosion Product) in EBS 
2.1.09.18.0A, Formation of Microbial Colloids in EBS 
2.1.09.23.0A, Stability of Colloids in EBS 
2.1.09.25.0A, Formation of Colloids (Waste-Form) by Coprecipitation in EBS 
2.1.09.19.0B, Advection of Colloids in EBS 
2.1.09.24.0A, Diffusion of Colloids in EBS 
6.2.33 Exothermic Reactions in EBS 
FEP Number: 
2.1.11.03.0A 
FEP Description: Exothermic reactions liberate heat and will alter the temperature of the 
disposal system and affect the properties of the repository and surrounding materials. Oxidation 
of uranium metal fuels such as represented by N-Reactor fuels is one example of a possible 
exothermic reaction. Hydration of concrete used in the underground environment is another 
example of a possible exothermic reaction in the EBS. 
April 2004 102 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Descriptor Phrases: 
Oxidation of uranium metal fuel 
Hydration of concrete 
Pyrophoricity 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: Exothermic reactions, and other thermal reactions, that liberate heat in 
the waste and EBS are excluded from the TSPA because the possible temperature rise in a 
disposal container from an exothermic degradation of waste such as N-Reactor uranium metal in 
DSNF, or in the EBS from the hydration of concrete, is negligible in comparison to the 
substantial heat generated by radioactive decay. For example, the heat of reaction by oxidizing 
all N-Reactor fuel uranium metal is roughly 1.3 x 1013J, or 1/100th the heat energy produced in 
one year by all DSNF and HLW (where DSNF and HLW amount to approximately 
7,000 MTHM, or one-tenth of the total repository mass of 70,000 MTHM) (DOE 1998 [DIRS 
122980]), Section ES.3.4). 
In addition, the conservative degradation rates used for DSNF already bound any additional 
thermally induced degradation effects that might result from exothermic reactions in the DSNF 
(see Section 6.2.6, FEP 2.1.02.01.0A). Because there is no temperature dependence in the 
recommended upper-limit DSNF model and because the degradation model is complete 
degradation within the first time step after contact by water, an increase in degradation rate due 
to temperature increase will have no impact on dose calculations. TSPA calculations will not be 
sensitive to differences within a TSPA time step. Note as well that the related effects of 
pyrophoric reactions are also excluded (See Section 6.2.13, FEP 2.1.02.08.0A). 
Therefore, this FEP can be excluded based on low consequence to radiological exposures to the 
reasonably maximally exposed individual and radionuclide releases to the accessible 
environment. 
TSPA Disposition: 
N/A 
Supporting Documents: 
N/A 
Related FEPs: 
2.1.02.01.0A, DSNF Degradation (Alteration, Dissolution, and Radionuclide Release) 
2.1.02.08.0A, Pyrophoricity From DSNF 
6.2.34 Thermal Effects on Chemistry and Microbial Activity in the EBS 
FEP Number: 
2.1.11.08.0A 
FEP Description: Temperature changes may affect chemical and microbial processes in the 
waste and EBS. 
April 2004 103 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Descriptor Phrases: 
Thermal effects on in-drift chemistry 
Thermal effects on in-package chemistry 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: Thermal effects may have an important influence on the rate of water contact 
with either the waste package, drip shield, or both (thereby influencing the corrosion rate), and 
on the rate of water influx into a failed waste package (thereby influencing the rate of waste form 
dissolution). Examples include localized dryout in the immediate vicinity of the waste packages 
during the early thermal phase of the repository history, as well as evaporation condensation on 
the underside of the drip shield providing an additional source term of water for dripping onto 
the waste package. Thermal effects on gaseous, liquid, or two-phase flow within the waste and 
EBS are explicitly accounted for in Multiscale Thermohydrologic Model (BSC 2004 [DIRS 
163056]). This includes consideration of dryout near the waste packages, as well as evaporation 
or condensation within the drift and underneath the drip shield. The results of these analyses 
then feed EBS Radionuclide Transport Abstraction (BSC 2003 [DIRS 162728]). 
As discussed in In-Package Chemistry Abstraction (BSC 2003 [DIRS 161962]), the thermal 
effects on the in-package chemistry are examined in the seepage drift model wherein EQ6 runs 
were performed at various temperatures to examine the temperature effect on the in-package 
chemistry. In these runs the kinetic reaction rates for the SNF were recalculated for runs at 15 to 
90C, and temperature-appropriate thermodynamic data in In-Package Chemistry Abstraction 
(BSC 2003 [DIRS 161962], Section 4.1.1.1) were used in the simulations. 
For TSPA-LA the in-package chemistry abstractions of pH, ionic strength, total carbonate, and 
Eh are cast in terms of temperature (BSC 2003 [DIRS 161962], Sections 8.2.3, 8.2.6, 8.4.3, 
8.4.6, 8.5, and 8.6; CRWMS M&O 2000 [DIRS 151561]), which captures the thermal effects on 
in-package chemistry. 
Supporting Documents: 
BSC 2003 [DIRS 161962] 
Related FEPs: 
2.1.08.04.0A, Condensation (Cold Traps) on Drift Roof 
2.1.08.07.0A, Unsaturated Flow in EBS 
2.1.08.14.0A, Condensation on Underside of Drip Shield 
2.1.11.09.0A, Thermal Effects on Flow in EBS 
2.1.11.03.0A, Exothermic Reactions in the EBS 
2.1.12.04.0A, Gas Generation (CO2, CH4, H2S) From Microbial Degradation 
April 2004 104 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
6.2.35 Thermally Driven Flow (Convection) in the EBS 
FEP Number: 
2.1.11.09.0B 
FEP Description: Temperature differentials may result in convective flow in the EBS. 
Convective flow within the waste and EBS could influence in-package chemistry. 
Descriptor Phrases: 
Thermally driven flow (convection) in-package 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: Locally saturated conditions that could occur in a failed waste package 
containing liquid water could result in convective flow as a result of local thermal gradients. 
Such convective flow could serve to disperse corrosion by-products or dissolved radionuclides 
and result in an acceleration of those degradation processes. Thermally driven flow is not 
included in the TSPA-LA modeling, but its exclusion is conservative. The transport modeling as 
employed in EBS Radionuclide Transport Abstraction (BSC 2003 [DIRS 162728], Table 20 and 
Section 7.5.1) includes a single mixing cell in the waste package (Section 5.2.35 of this analysis 
report). Thermal convection inside the waste package would tend to erase concentration 
gradients and distribute radionuclides uniformly throughout the package. Also, the analysis of 
an alternative model for in-package chemistry (BSC 2003 [DIRS 161962], Section 6.4.1) treated 
the waste package as a one-dimensional assemblage of mixing cells with varying compositions. 
It was concluded therein that the single-mixing-cell primary (water vapor condensation) model 
was preferable because the effect of the materials within the waste package on the water 
chemistry overwhelmed the effect of any spatial variation due to temperature. 
Furthermore, the breach of a waste package required for the admission of water is most likely to 
occur after 1000 years, by which time the temperature gradients are likely to be small (BSC 2004 
[DIRS 165057], Appendix A). 
Therefore, omission of convective flow within the waste package will not significantly change 
the radiological exposures to the reasonably maximally exposed individual and radionuclide 
releases to the accessible environment and this FEP can be excluded from the TSPA-LA based 
on low consequence. 
TSPA Disposition: 
N/A 
Supporting Documents: 
BSC 2003 [DIRS 161962] 
BSC 2003 [DIRS 162728] 
BSC 2004 [DIRS 165057] 
ANL-WIS-MD-000009 REV 01 April 2004 105 Miscellaneous Waste-Form FEPs 
Related FEPs: 
2.1.08.04.0A, Condensation (Cold Traps) on Drift Roof 
2.1.08.07.0A, Unsaturated Flow in EBS 
2.1.08.14.0A, Condensation on Underside of Drip Shield 
2.1.11.09.0A, Thermal Effects on Flow in EBS 
2.1.09.01.0B, Chemical Characteristics of Water in the Waste Package 
2.1.09.01.0A, Chemical Interaction with Corrosion Products 
6.2.36 Gas Generation (CO2, CH4, H2S) From Microbial Degradation 
FEP Number: 
2.1.12.04.0A 
FEP Description: Microbes are known to produce inorganic acids, methane, organic 
byproducts, carbon dioxide, and other chemical species that could change the longevity of 
materials in the repository and the transport of radionuclides from the near field. The rate of 
microbial gas production will depend on the nature of the microbial populations established, the 
prevailing conditions (temperature, pressure, geochemical conditions), and the organic or 
inorganic substrates present. Initial analysis indicates the most important source of nutrient in 
the YMP repository will be metals. Other possible nutrients include cellulosic material, plastics, 
and synthetic materials. Minimal amounts of organics are mandated by regulation. 
Descriptor Phrases: 
Internal gas pressure from CO2, CH4, H2S 
Gas accumulation under biofilms 
Chemical effects from CO2, CH4, H2S generation 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: Gas generation from microbial degradation is excluded from TSPA-LA 
on the basis of low consequence. Omission of the effects of gas generation from microbial 
degradation on in-package waste chemistry will not significantly change radiological exposures 
to the reasonably maximally exposed individual or radionuclide releases to the environment for 
the following reasons. Generally speaking, the YMP system is considered to be an open system, 
and any gas produced will not accumulate in significant amounts to affect the system. Also, 
biological activity inside the waste package has been excluded (FEP 2.1.10.01.0A) because 
organic material will be present in only minor amounts in the waste package. Note also that 
methanogenesis and H2S generation are anaerobic processes. These will not occur under the 
oxidizing conditions that are expected to occur at Yucca Mountain. Finally, there may be some 
localized effects such as gas accumulation under biofilms. The possibility of local depression of 
pH by microbes is addressed in FEP 2.1.02.14.0A. 
The sealed waste packages at emplacement will contain little oxygen or water, suggesting a low 
probability of gas generation from microbial degradation within the package. Once the package 
is breached, it may be reasonably concluded that any gas produced will migrate to the outside of 
the package. Gas produced from microbial degradation, because of the repository host rock 
April 2004 106 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
lithology and structure and physical properties of gas as it flows through these structures, will 
preferentially flow through stress-induced fractures and in situ fracture networks to eventually 
reach the mountain surface. Gas concentrations will become more dispersed and less 
concentrated as gas moves away from the repository; thus, gas concentrations become diluted 
before gas could accumulate in quantities of concern (see FEP 2.1.12.01.0A). 
The primary potential impact of gas generation inside a breached waste package is likely to be 
associated with the generation of CO2 by microorganisms. High levels of CO2 might affect the 
in-package pH, as well as the solubilities of those radionuclides that complex appreciably with 
carbonate. Microbial respiration is potentially able to increase in-drift CO2 production by two 
orders of magnitude (CRWMS M&O 2000 [DIRS 151561]). Absolute levels of carbon dioxide 
will depend not only on the production rate, but on the rate at which it leaves the drift as well. If 
diffusion of carbon dioxide is slow, high carbon dioxide levels might result 
(CRWMS M&O 2000 [DIRS 151561]). Sensitivity calculations in In-Package Chemistry 
Abstraction (BSC 2003 [DIRS 161962]) show that variations in carbon dioxide partial pressures 
from 10-4 to 10-2 have no effect on the outputs of the in-package chemistry model (10-3.5 is the 
baseline atmospheric level of CO2). It is therefore reasonable to expect little impact of carbon 
dioxide production on in-package pH and ionic strength and to assume no impact from increased 
CO2 levels. 
Effects of carbon dioxide from microbial degradation on radioactive element solubilities will be 
small, as shown in the analysis of dissolved concentrations of radioactive elements in Finch and 
Fortner (2002 [DIRS 162004]), which encompassed carbon dioxide partial pressures as high as 
1.0 to 1.5 atmospheres (see also Section 5.2.36 of this analysis report). Dissolved Concentration 
Limits of Radioactive Elements (BSC 2003 [DIRS 163152]) parameterizes for dissolved 
concentrations of radionuclides as a function of in-package and EBS solution chemistry over a 
range of carbon dioxide partial pressures of 10-5 to 10-1.5in effect to two orders of magnitude 
greater than atmospheric pressure. The CO2 ranges, over which TSPA calculates probability 
distributions for radionuclide solubilities, bound high carbon dioxide levels. Therefore, any 
additional effects on radionuclide solubilities, from high carbon dioxide levels from microbial 
degradation would not be significant and can be excluded due to low consequence. 
TSPA Disposition: 
N/A 
Supporting Documents: 
BSC 2003 [DIRS 161962] 
Related FEPs: 
2.1.02.14.0A, Microbiologically Influenced Corrosion (MIC) of Cladding 
2.1.10.01.0A, Microbial Activity in the EBS 
2.1.12.01.0A, Gas generation (Repository Pressurization) 
6.2.37 Effects of Radioactive Gases in EBS 
FEP Number: 
2.1.12.07.0A 
April 2004 107 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
FEP Description: Radioactive gases may exist or be produced in the repository. These gases 
may subsequently escape from the repository. Typical radioactive gases include 14C (in 14CO2 
and 14CH4) produced during microbial degradation, tritium, fission gases (Ar, Xe, Kr), and 
radon. 
Descriptor Phrases: 
Gas generation in EBS 
Gas exsolution in EBS 
Repository pressurization 
Effects of gas on waste package degradation 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: The major gas constituents trapped inside a waste package prior to 
disposal will be the result of fission-product and neutron-activated stable and radioactive noble 
gasespredominantly isotopes of Ar, Kr, and Xe. There are several arguments that support the 
exclusion of radioactive gases from the TSPA-LA calculations, based on low consequence to 
waste package internal pressures and low consequence with respect to radionuclide releases from 
the EBS. 
Radioactive gases residing in fuel rods and waste packages prior to postclosure or produced after 
postclosure times will either decay rapidly or quickly become negligible in terms of mass and 
volume. Specifically: 
A. Xenon (135Xe), a fission yield-product, is short-lived (9.2 hours) and will not be 
produced during the regulatory time periods. It will undergo decay to its long-lived 
daughter 135Cs by the time of waste emplacement. 
B. Argon (39Ar) has a low activity and is screened out due to low consequence. It will not 
be produced during the regulatory time period. 
C. Radon gas (222Rn and 219Rn) is short-lived (approximately 3.8 days and approximately 
4 seconds, respectively), as are its gaseous daughters (218Po, half-life approximately 
3.1 minutes and 215Po, half-life of microseconds) and, thus, will be in secular 
equilibrium with its parents, radium and thorium (229Th). Radon has been screened out 
of the TSPA-LA due to low consequence. 
14 D. 14CH4 and CO2 generation from microbial activity (CRWMS M&O 2000 
[DIRS 151561], Sections 6.3 and 7.2) should not be significant because organic levels 
in the waste are expected to be low. 
E. Once a waste package is breached, an oxidizing environment may be created within 
the waste package, thus enabling 14CO2 to be formed. The limited amount of 14C 
remaining in the inventory, coupled with the low package-failure rate, limits the 
potential creation of 14CO2 to small amounts. Furthermore, the amounts of 14C 
transported as a gas to the accessible environment will be extremely small, so the risk 
April 2004 108 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
consequence is negligible. See FEP 2.2.11.03.0A in Features, Events, and Processes 
in UZ Flow and Transport (BSC 2004 [DIRS 168782], Section 6.1.39). 
F. Of the noble gases, 85Kr has a significant initial inventory, but because of its short halflife 
(approximately 10 years), its concentration rapidly becomes insignificant. 
Partial pressures from radioactive gas will not be great enough to compromise waste package 
integrity via a waste package breach (BSC 2004 [DIRS 162200], Section 6.2.13.5). This 
argument is supported by several numerical studies investigating the sensitivity of internal waste 
package pressures presuming fuel rods rupture, thus causing a release of gases from the fuel 
matrix to the waste package interior (CRWMS M&O 1997 [DIRS 100264]; CRWMS M&O 
2000 [DIRS 144128]; BSC 2001 [DIRS 153755]). These studies conservatively estimate that the 
maximum pressures within the fuel-rod cladding would rise to 8.4 MPa (CRWMS M&O 1997 
[DIRS 100264], Section 7.2.2.7), thus inducing clad failure. Once cladding ruptured, all 
radioactive and fission gases migrated from the fuel matrix to the waste package voids. Waste 
package interior pressures were determined for cases presuming 1, 10, and 100 percent of all fuel 
rods rupture (CRWMS M&O 1997 [DIRS 100264], Table 7.2.2.7-1), and rupture occurs at 
temperatures ranging from 20 to 600C. An additional case presumes all fuel rods rupture, and 
all fuel pellets have a 50-percent higher burnup. Consequently, internal pressures would be 
greater than assemblies experiencing normal burnup conditions by a factor of 1.5 (CRWMS 
M&O 1997 [DIRS 100264], Section 7.2.2.7). Analyses of stresses imposed on the waste 
package interior shells by internal pressurization resulting from fuel rod rupture (CRWMS M&O 
2000 [DIRS 144128], Table 21) determined the maximum allowable waste package interior 
pressures for various waste package designs (BSC 2001 [DIRS 153755], Table 6-1). 
Results from these studies show that even assuming 100 percent fuel rod failure, expected 
interior waste package pressures for normal burnup rates (CRWMS M&O 1997 [DIRS 100264], 
Table 7.2.2.7-1) are consistently and significantly lower than the maximum allowable for the 
various waste package loadings (CRWMS M&O 1997 [DIRS 100264], Table 7.2.2.7-1). For 
example, in the high burnup cases (using the 1.5 multiplier), the waste package pressure at 
200C is 0.62 MPa, much lower than the pressure of 1.02 to 1.81 MPa required to cause the 
waste package outer shell to fail (BSC 2001 [DIRS 153755], Table 6-1). The above studies are 
upper-limit bounded for several reasons. It is estimated that nominal waste package 
temperatures for the base case will not exceed 350C, well below the 500C used in the analysis 
performed in Waste Package Design Basis Events (CRWMS M&O 1997 [DIRS 100264]). Not 
all the fuel rods will rupture within a waste package within the regulatory time frame. 
Furthermore, fission and radioactive gas release fractions from the fuel matrix are typically low 
(CRWMS M&O 2000 [DIRS 151659], Section 6.3.3). Therefore, fission gas released from the 
matrix will not all migrate to free void space, especially at temperatures approaching 27C, 
where migration of fission gases, including radioactive gases, out of the matrix approaches 0.0 
(CRWMS M&O 1997 [DIRS 100264]). This FEP may therefore be screened as excluded due to 
low consequence. 
TSPA Disposition: 
N/A 
April 2004 109 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Supporting Documents: 
CRWMS M&O 2000 [DIRS 144128] 
CRWMS M&O 2000 [DIRS 151659] 
BSC 2001 [DIRS 153755] 
Related FEPs: 
N/A 
6.2.38 Radiolysis 
FEP Number: 
2.1.13.01.0A 
FEP Description: Alpha, beta, gamma, and neutron irradiation of water can cause 
disassociation of molecules, leading to gas production and changes in chemical conditions (Eh, 
pH, concentration of reactive radicals). 
Descriptor Phrases: 
Gas generation from radiolysis 
Chemical effects from radiolysis in-package 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: The effects of radiolysis for gas production and waste form chemistry 
can be excluded from consideration in the TSPA based on low consequence. Alpha, beta, 
gamma, and neutron irradiation of water leads to formation of highly reactive excited and 
ionized species. In pure water, the final products are hydrogen and oxidants. In addition, the 
oxidants formed may react with dissolved iron (+2) which will decrease the net yield of oxidants. 
However, water is not expected to affect the fuel until all, except possibly alpha radiolysis, have 
become significantly reduced. The effects of beta irradiation are expected to be minimal because 
1) almost all beta emitters disappear due to radioactive decay after a few hundred years and 2) 
beta radiation is stopped in the fuel matrix or clad. Recent calculations by Tang have shown that 
neutron irradiation is negligible and gamma dose has been significantly reduced (BSC 2002 
[DIRS 161912]). Intact cladding will stop alpha particles so alpha radiolysis will not occur 
during the early periods of highest alpha activity. Additionally, the rate of corrosion effects of 
used UO2 fuel due to alpha radiolysis, taking no credit for cladding, can be predicted (based on 
semiempirical methods) to have low consequence. 
Interior to the waste package, water will not intrude (i.e., the waste container will not fail) until 
gamma and beta emitters have decayed to low concentrations (Sunder and Shoesmith 1991 
[DIRS 143815]; Shoesmith and Sunder 1992 [DIRS 113368]). According to Sunder and 
Shoesmith (1991 [DIRS 143815]), strong gamma and beta fields associated with the used fuel 
will decrease by a factor greater than 103 in the first few hundred years after disposal. 
Arguments addressing the highly improbable adverse or inconsequential impact of nitric acid and 
hydrogen peroxide production, and other potential products, of gamma radiolysis on corrosion 
are presented in In-Package Chemistry for Waste Forms (BSC 2003 [DIRS 168795], 
Attachment II) and In-Package Chemistry Abstraction (BSC 2003 [DIRS 161962]). In this 
April 2004 110 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
analysis, the production of nitric acid and hydrogen peroxide was increased by a factor of ten. 
The effect of in-package chemistry was to change the chemical compositions in the second 
significant figure; however, this effect is insignificant. 
The use of presently available used fuels to study the effects of alpha radiolysis on fuel 
dissolution is inappropriate because of their associated strong beta and gamma fields . However, 
Sunder et al. (1997 [DIRS 143860]) describe an experimental strategy for determining fuel 
dissolution rates as a function of alpha-source strength, and they show how the evolution of 
corrosion behavior can be predicted as a function of the age of the fuel. Sunder et al. (1997 
[DIRS 143860]) conclude that predictions presentedsuggest the effects of alpha radiolysis on 
fuel corrosion (dissolution) will be transitory and will become minor as alpha dose rates 
decrease. 
During the periods of highest alpha activity, it is expected most of the commercial fuel cladding 
will remain intact and should substantially reduce alpha dose rates to groundwater (Kaplan 1963 
[DIRS 149367], p. 307). The stopping power of metals is at least three orders of magnitude 
greater than air; thus clad of thickness of a few microns would stop alpha particles. 
Additionally, there are these two net findings: 
1. The -radiolysis-enhanced corrosion rate is three orders of magnitude less than the 
dissolution rate. 
2. The -radiolysis-enhanced corrosion rates for HLW metallic carbide, and ceramic 
spent nuclear fuels are much lower than their dissolution rates. 
Since no credit is being taken for the DSNF cladding in the TSPA, the effect of radiolysis on the 
DSNF cladding is not a factor. However, the argument by Sunder et al. (1997 [DIRS 143860]) 
does not include the protection of clad against dissolution. 
This FEP does not have a significant effect on radiological exposures to the reasonably 
maximally exposed individual and radionuclide releases to the accessible environment and may 
therefore be screened as excluded due to low consequence. 
TSPA Disposition: 
N/A 
Supporting Documents: 
BSC 2003 [DIRS 168795] 
BSC 2003 [DIRS 161962] 
Related FEPs: 
2.1.13.02.0A 
6.2.39 Radiation Damage in EBS 
FEP Number: 
2.1.13.02.0A 
ANL-WIS-MD-000009 REV 01 April 2004 111 Miscellaneous Waste-Form FEPs 
FEP Description: Radiolysis due to the alpha, beta, gamma ray, and neutron irradiation of 
water could result in the enhancement for the movement of the radionuclides from the surface of 
a degraded waste form into groundwater flow. When radionuclides decay, the emitted highenergy 
particle could result in the production of radicals in the water or air surrounding the spent 
nuclear fuel. If these radicals migrate (diffuse) to the surface of the fuel, they may then enhance 
the degradation/corrosion rate of the fuel (UO2). This effect would increase the dissolution rate 
for radionuclides from the fuel material (fuel meat) into the groundwater flow. Strong radiation 
fields could lead to radiation damage to the waste forms and containers (CSNF, DSNF, and 
DHLW), drip shield, seals and surrounding rock. 
Descriptor Phrases: 
Enhanced waste form degradation from radiation 
Radiation damage in-package (waste form) 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: Radiation damage can be caused by low linear energy transfer (i.e., beta 
particles and gamma rays) and high linear energy transfer (i.e., alpha) radiation. Low linear 
energy transfer radiation (mostly gamma rays) effects are relatively reduced (in comparison to 
high linear energy transfer effects) because the radiation fields decrease dramatically over the 
first 1,000 years after the fuel is outside the reactor (Sunder et al. 1997 [DIRS 143860], p. 67). 
The effect of high linear energy transfer radiation on fuel degradation will depend on the time at 
which cladding integrity is lost for a given fuel rod (the high linear energy transfer radiation 
would not be able to penetrate past the cladding material while intact cladding material is 
present). Before the cladding is significantly compromised (by cracking, pitting, unzipping, 
etc.), the damage due to radiolysis will be dominated by low linear energy transfer radiation. 
After cladding material is lost, the radiolysis rate will be dominated by high linear energy 
transfer radiation. High linear energy transfer alpha radiation fields will persist for tens of 
thousands of years. 
Of the various modes of radioactive decay (i.e., alpha [-], beta [-], gamma [-], spontaneous 
fission, isomeric transition, etc.), the most important for fissile materials (Attachment IV of this 
analysis report) is -decay. Radionuclides that decay by other modes have correspondingly 
relatively short half-lives (i.e., tens to hundreds of years); thus, they would decay away prior to 
the estimated time to corrosion breakthrough of the fuel cladding. The dominant decay mode for 
heavy radionuclides is -decay. While many of the heavy radionuclides emit alpha particles 
with energies greater than 4.0 MeV (Parrington et al. 1996 [DIRS 103896]), there are no gamma 
releases or beta particles emitted with energies greater than 4.0 MeV and only a few with 
energies greater than 1.0 MeV. Thus, -decay linear energy transfer values will bound the 
effects due to beta particles and gamma rays. Also the majority of the long-term radionuclides 
for YMP are alpha emitters. Other special decay modes such as isomeric transition and 
spontaneous fission decay have probabilities of occurrence that are orders of magnitude less than 
that for alpha decay. 
Radioactive decay of radionuclides after they have entered the groundwater is not of concern 
since they will not, to any measurable quantity, increase the release of radionuclides from the 
April 2004 112 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
waste forms into the groundwater. The decay within the groundwater will only transmute the 
specific radionuclide inventory already being transported by the groundwater (i.e., due to 
colloids, dissolution, etc.), and the subsequent decay chains from the transported radionuclides 
should be modeled within the radionuclide transport computational codes. 
This FEP does not have a significant effect on radiological exposures to the reasonably 
maximally exposed individual and radionuclide releases to the accessible environment and may 
therefore be screened as excluded due to low consequence. Related processes of alpha recoil 
(see FEP 2.1.02.04.0A) and radiolysis (see FEP 2.1.13.01.0A) are also excluded based on low 
consequence. 
TSPA Disposition: 
N/A 
Supporting Documents: 
N/A 
Related FEPs: 
2) From Metal Corrosion 2.1.12.03.0A, Gas Generation (H 
2.1.02.04.0A, Alpha Recoil Enhances Dissolution 
2.1.13.01.0A, Radiolysis 
3.1.01.01.0A, Radioactive Decay and In-Growth 
6.2.40 Chemistry of Water Flowing into Waste Package 
FEP Number: 
2.2.08.12.0B 
FEP Description: Inflowing water chemistry may be used in analysis or modeling that requires 
initial water chemistry in the waste package. 
Descriptor Phrases: 
Inflowing water composition (into waste package) 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: Three different initial water compositions were used to represent the 
chemistry of the water flowing into the waste package. This inflowing water chemistry is one of 
the inputs used to determine the in-package chemistry in the IPCM. Therefore, the variability of 
the incoming water composition is implicitly included in the in-package chemistry and in the 
abstractions passed to TSPA in In-Package Chemistry Abstraction (BSC 2003 [DIRS 161962]). 
In TSPA it then feeds the solubility, colloid, CSNF, and HLW glass submodels. The effects of 
the inflowing water chemistry are captured within the IPCM parameters: pH, ionic strength, total 
carbonate, Eh, chloride, and fluoride. 
April 2004 113 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
The seepage dripping model examined various input water chemistries (BSC 2003 
[DIRS 161962], Section 4.1.1.2) and their effect on the in-package chemistry. The results 
showed the parameters passed to TSPA-LA were unaffected by changes in the input water 
composition. Thus, while the composition of the water flowing into the waste package is 
implicitly included in the TSPA-LA model, variations in the composition do not significantly 
affect the model results. 
Supporting Documents: 
BSC 2003 [DIRS 161962]. (Note that while FEP 2.2.08.12.0B is not in the included FEP 
table (BSC 2003 [DIRS 161962], Table 10), it is addressed in the report.) 
Related FEPs: 
2.1.09.01B, Chemical characteristics of water in the waste package 
6.2.41 Radioactive Decay and In-growth 
FEP Number: 
3.1.01.01.0A 
FEP Description: Radioactivity is the spontaneous disintegration of an unstable atomic nucleus 
that results in the emission of subatomic particles. Radioactive isotopes are known as 
radionuclides. Radioactive decay of the fuel in the repository changes the radionuclide content 
in the fuel with time and generates heat. Radionuclide quantities in the system at any time are 
the result of the radioactive decay and the growth of daughter products as a consequence of that 
decay (i.e., ingrowth). Over a 10,000-year performance period, these processes will produce 
daughter products that need to be considered in order to adequately evaluate the release and 
transport of radionuclides to the accessible environment. 
Descriptor Phrases: 
Radioactive decay and ingrowth (in package) 
Radioactive decay and ingrowth (in drift) 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: Radioactive decay and in-growth were considered in the selection of 
isotopes of importance to TSPA-LA and are included as standard features of the GoldSim code. 
Once the isotopes most important to dose were identified (BSC 2002 [DIRS 160059], Table 13), 
the parents of these isotopes were examined to determine if decay and in-growth could 
significantly affect the amount of the important isotopes during the regulatory period. Seven 
parent isotopes were identified whose decay significantly increased the amount of their progeny 
(BSC 2003 [DIRS 161961], Table 16): 
. 245Cm  241Pu  241Am 
. 235U  231Pa 
April 2004 114 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
. 230Th  226Ra 
. 232Th  228Ra 
. 236U  232Th 
. 242Pu  238U 
These seven isotopes were added to the list of isotopes to be tracked in the TSPA-LA GoldSim 
model. During execution, the GoldSim model automatically calculates decay and in-growth of 
the included isotopes within the waste package and drift modeling cells. 
Supporting Documents: 
BSC 2003 [DIRS 161961] 
Related FEPs: 
2.1.01.01.0A, Waste inventory 
April 2004 115 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
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April 2004 116 Miscellaneous Waste-Form FEPs 
7. CONCLUSIONS 
This scientific analysis report addresses 41 waste form and waste form colloid FEPs. These 
FEPs were identified and examined for inclusion or exclusion in TSPA-LA. Table 1.2-1 lists 
these FEPs and their associated screening decisions. 
By default, a FEP is included in the TSPA if it cannot be excluded based on the screening criteria 
described in Section 4.2.1. For included FEPs, the TSPA-LA Dispositions provided in 
Section 6.2 describe how each FEP is included in the TSPA-LA (i.e., through a parameter or 
TSPA model or sub-model). For excluded FEPs, the screening decision is based on the 
screening criteria (by regulation, low probability, or low consequence) and the technical basis for 
exclusion is elaborated in the screening arguments provided in Section 6.2. 
The analyses developed in this report are qualitative, based on available project information and 
the open technical literature, and so any uncertainties associated with the FEPs screening 
decisions are also qualitative. 
The following conclusions may be drawn: 
 This analysis model report contributes to Yucca Mountain FEP analysis by screening 
41 miscellaneous WF FEPs. This report extracts the TSPA-disposition statements for 
included FEPs from the pertinent model reports and presents the screening arguments 
for excluded FEPs. This analysis model report provides both information for the YMP 
FEP database and guidance to TSPA analyses. 
 By providing references to appropriate other analysis model reports, this report provides 
a link between WF issues and the research directed at their resolution. 
April 2004 117 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
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8. INPUTS AND REFERENCES 
8.1 DOCUMENTS CITED 
151125 Abrefah, J.; Gray, W.J.; Ketner, G.L.; Marschman, S.C.; Pyecha, T.D.; and Thornton, 
T.A. 1995. K-Basin Spent Nuclear Fuel Characterization Data Report. PNL-10778. 
Richland, Washington: Pacific Northwest National Laboratory. TIC: 243197. 
151226 Abrefah, J.; Huang, F.H.; Gerry, W.M.; Gray, W.J.; Marschman, S.C.; and Thornton, 
T.A. 1999. Analysis of Ignition Testing on K-West Basin Fuel. PNNL-11816. 
Richland, Washington: Pacific Northwest National Laboratory. TIC: 248558. 
163864 Adkins, H.E. 2003. Transmittal of Draft Version of Validation of Dissolved 
Concentrations, Argonne National Laboratory. Interoffice memorandum from H.E. 
Adkins (BSC) to Y. Chen, June 19, 2003, 0617037708, with attachment. 
ACC: MOL.20030619.0100. TBV-5177 
101565 Bruno, J.; Cera, E.; Eklund, U.-B.; Eriksen, T.E.; and Grive, M. 1998. Some 
Preliminary Modeling Results of Long Term Radiolytic Mass Balance Experiments. 
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ACC: DOC.20040225.0011. 
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 4. Scientific Analyses. Washington, D.C.: U.S. Department 
of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20040407.0003. 
AP-SV.1Q, Rev. 1, ICN 1. Control of the Electronic Management of Information. 
Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste 
Management. ACC: DOC.20040308.0001. 
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. 
April 2004 128 Miscellaneous Waste-Form FEPs 
8.3 SOURCE DATA, LISTED BY DATA TRACKING NUMBER 
164527 MO0307SEPFEPS4.000. LA FEP List. Submittal date: 07/31/2003. 
168762 MO0402SPAHWCIG.002. Heat and Water Chemistry Output from Igneous 
Intrusion. Submittal date: 02/23/2004. 
April 2004 129 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
ANL-WIS-MD-000009 REV 01 
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April 2004 130 Miscellaneous Waste-Form FEPs 
A list of the attachments can be found in Table 9-1. 
Attachment Title 
I 
II 
III 
IV 
V 
VI 
Table 9-1. List of Attachments 
Analysis of the Potential for Pyrophoric Behavior in DSNF 
Radionuclide Inventory for Final Waste Forms 
Alpha Recoil Mechanics 
Radionuclide Inventory for Pressurized-Water Reactor Spent Nuclear Fuels and Boiling- 
Water Reactor Spent Nuclear Fuels 
Data Qualification For Direct Inputs To Miscellaneous Waste-Form FEPs 
Data Qualification Plan (Scanned Images) 
ANL-WIS-MD-000009 REV 01 
9. ATTACHMENTS 
April 2004 131 Miscellaneous Waste-Form FEPs 
INTENTIONALLY LEFT BLANK 
April 2004 132 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
ATTACHMENT I 
ANALYSIS OF THE POTENTIAL FOR PYROPHORIC BEHAVIOR IN DSNF 
The DOE Office of Environmental Management/Office of Civilian Radioactive Waste 
Management Repository Task Team (DOE 1997 [DIRS 149432], pp. 20-21) has addressed three 
fuel types with regard to their potential for pyrophoric reactions: oxide, metal, and carbide fuels. 
A fourth (metallic sodium-bonded fuel) has been eliminated from further consideration because 
it is a listed RCRA hazard and will be treated prior to repository emplacement. DSNF is only a 
small part of the total waste inventory in the repository. The characteristics and potential for 
pyrophoric behavior of these three fuel types are discussed in the following paragraphs. 
Oxide fuels consist of U-oxide fuels, Th/U-oxide fuels, and U/Pu-oxide fuels of varying 
enrichments and concentrations. Bulk UO2 fuels have been exposed in oxidizing environments 
at temperatures considerably higher than expected repository temperatures with no indication of 
ignition. It is not expected that the addition of significant amounts of ThO2 or PuO2 in the mixed 
oxide fuels will make the fuel pyrophoric as long as the matrix form is composed of heavy-metal 
oxides (DOE 1997 [DIRS 149432], p. 20). 
Metallic uranium-based fuels will oxidize in an air or water environment and can be pyrophoric, 
particularly when in a fine particulate form or when a significant amount of hydride has formed 
in the uranium metal matrix. The corrosion of uranium metal results in the formation of uranium 
hydrides as finely dispersed inclusions in the uranium metal. Although examination and testing 
of damaged N-Reactor zirconium-clad uranium metal fuel showed only small amounts of 
uranium hydride formed by corrosion as precipitates within the metal and in thin coatings on the 
internal crack surfaces (Marschman et al. 1997 [DIRS 149429], Section 3.4.2), their presence is 
believed to be responsible for the observed decreased ignition temperature observed during 
ignition testing of damaged or corroded N-Reactor SNF samples compared to unirradiated or 
undamaged samples (Abrefah et al. 1999 [DIRS 151226], Tables 4.2 and 4.3). Additionally, the 
possibility exists that additional U-hydride will form during interim storage (Reilly 1998 [DIRS 
149433], p. 30), which could potentially act as an ignition source if concentrated in a small area. 
TRIGA fuel, predominantly composed of U-Zr-hydride, is not as susceptible to pyrophoric 
reactions as uranium metal fuel because this fuel does not display the extensive damage or 
corrosion evident in the N-Reactor SNF. 
Most aluminum-clad uranium metal DSNF (i.e., fuel located at the Savannah River Site) does 
not have the amount of damaged/corroded uranium metal (and extent of uranium hydride 
formation and consequent potential for pyrophoricity) shown by the N-Reactor SNF (Lam et al. 
1997 [DIRS 152482], Section 8). The small fraction (e.g., Single-Pass Reactor SNF) of the total 
aluminum-clad uranium-metal SNF inventory that is currently located in the K-Basins at 
Hanford could display damage or corrosion characteristics similar to N-Reactor SNF due to its 
uranium metal core and similar handling and wet-storage history. This aluminum-clad uranium 
metal SNF will be disposed in the N-Reactor SNF canisters. The TSPA-LA will (conservatively) 
not differentiate waste packages with MCOs containing this SNF from MCOs containing the NReactor 
SNF. 
April 2004 I-1 of I-6 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
The U-Mo fuels should behave similarly to the Al-U SNF. The uranium/aluminum alloy matrix 
fuel is much less reactive than U-metal (DOE 1997 [DIRS 149432], p. 21; DOE 2002 
[DIRS 158405], Section 6.2). Because of the low U loading of these U-Mo and most Al-U fuels 
(less than 9 percent of their total mass is U), it is anticipated that less UH3 would be present than 
in the N-Reactor SNF, although this has not been tested experimentally. 
Carbide fuels are not a pyrophoric hazard in a dry atmosphere except as high surface area 
powders. In a moist atmosphere (as would be required to corrode through a waste package), 
uranium carbide (UC2) will react to form UO2, carbon dioxide (CO2), and hydrocarbon gases. 
The major constituent of the hydrocarbon gases is methane with minor constituents being ethane, 
ethylene, and acetylene. These gases are flammable in air, but only acetylene is autocatalytically 
explosive. An analysis of carbide fuels (Propp 1998 [DIRS 149395]) indicated that the oxidizing 
environment of the repository would inhibit the formation of hydrocarbons. It should also be 
noted that less than 1 percent of the DSNF inventory is carbide SNF. 
The following conditions were used in this analysis to create a worst case scenario. This worst 
case scenario will then be used to determine the maximum impact pyrophoricity can have on 
repository performance. A pyrophoric event cannot occur unless or until: 
 The waste package and SNF canister have been breached; 
 There is sufficient oxygen available to support a pyrophoric event; 
 An ignition source (overheating due to oxidation, spark, mechanical impact, etc.) is 
available. 
The greatest risk of pyrophoric behavior is from the N-Reactor SNF because a significant 
fraction of the fuel elements have damaged cladding, resulting in the exposure of the damaged 
and corroded uranium metal cores. Both the N-Reactor and similar types of uranium-metalbased 
materials have shown pyrophoric behavior in the past (Schulz 1972 [DIRS 159406]; 
Wood et al. 1994 [DIRS 165884]; Solbrig et al. 1994 [DIRS 165881]). N-Reactor SNF 
constitutes the large majority of the DSNF (85 percent), but less than approximately 1.3 percent 
of the total waste to be stored at Yucca Mountain (BSC 2003 [DIRS 163693], Sections 6.4 
and 8). A pyrophoric event could result in the nearly instantaneous release of the soluble and 
volatile radionuclide inventory in the affected packages and the breach of the two adjacent waste 
packages. 
A pyrophoric event can impact a number of repository components: the directly affected waste 
packages (cladding, fuel, and in-package chemistry), nearby waste packages, and nearby 
geohydrology. It can be postulated that changes in local percolation rates could result from 
changes in the surrounding rock strata caused by a pyrophoric events thermal energy. However, 
a postclosure pyrophoric event is not possible unless the waste package and the canister holding 
the N-Reactor SNF (i.e., the multicanister overpack) breach. Therefore, changes in local 
percolation rates would not impact waste package failure rates. Also, if the postclosure analysis 
assumes that the complete dissolution of the waste form within a breached waste package occurs 
within one time step following its breach, changes in local percolation rates would not 
significantly impact SNF dissolution rates. The use of titanium drip shields will tend to shield 
April 2004 I-2 of I-6 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
the surrounding geohydrologic system from any sudden release of thermal energy and distribute 
it along the drift. 
The following argument addresses pyrophoricity in terms of both the total radionuclides that 
could be released due to a pyrophoric event and the effect a pyrophoric event could have on the 
peak offsite dose. A sensitivity analysis is included to evaluate the effects of clustering. 
b and the number of waste 
nr, the number of failed waste packages with N-Reactor SNF 
b  Nnr. If the probability that a waste package next to a waste package with N-Reactor SNF 
Impact on Radionuclide ReleaseDuring the regulatory period following repository closure, 
some percentage of the approximately 160 N-Reactor and 12,000 total waste packages 
(DOE 1998 [DIRS 122980], Table I-1) will fail. For the purpose of estimating the maximum 
effect on radionuclide release of pyrophoric events in N-Reactor SNF-containing waste 
packages, it may be conservatively assumed (see Section 5.2.13) that (1) breach of a waste 
package containing N-Reactor SNF will result in a pyrophoric event that causes the two adjacent 
waste packages to fail, (2) the various types of waste packages are evenly distributed throughout 
the repository, and (3) the probability of failure of a waste package containing pyrophoric (i.e., 
N-Reactor) SNF is the same as that of any other waste package. Therefore, the additional 
number of waste package failures due to pyrophoric events during the regulatory period can be 
estimated as follows. If the fraction of failed waste packages is F 
packages with N-Reactor SNF is N 
is F 
has failed prior to a pyrophoric event is Pb, then the number of additional waste packages that fail 
due to pyrophoric events in the waste packages with N-Reactor SNF can be expressed as: 
Nadd = Nnr  Fb  (1-Pb) (Eq. I-1) 
add is the number of additionally failed packages and the term (1-Pb) gives the probability 
b can be 
b. For a fraction of failed packages Fb, 2 out of 2/Fb packages are failed. 
b = 0.001, then 2 out of every 2/(0.001) = 2,000 packages will have failed. The 
where N 
that a neighboring package had not failed prior to the pyrophoric event. The value of P 
expressed in terms of F 
For example, if F 
probability that any particular waste package has failed is 2/(2/Fb) = Fb. The probability that a 
neighboring waste package fails is: 
Pb = 1/[(2/Fb)-1] (Eq. I-2) 
b 
for each neighbor. Since there are two neighboring packages, the probability that 1 of the 2 
packages failed is approximately 2/[(2/Fb)-1]. (The possibility that both neighbors have failed is 
approximately P 2 and is neglected.) The probability that both a waste package and 1 or both of 
its neighboring packages are failed is: 
Pb = Fb {2/[(2/Fb)-1]} (Eq. I-3) 
The probability that a neighboring package has not failed prior to failure of the package with 
N-Reactor SNF is: 
(1-Pb) = 1-Fb  {2/[(2/Fb)-1]} (Eq. I-4) 
The probability that a failed waste package with N-Reactor SNF neighbors 2 waste packages that 
have not failed can be used to calculate the number of additional failures possible from 
April 2004 I-3 of I-6 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
pyrophoric events. The probability that a waste package with N-Reactor SNF has failed is 
(Nnr/Ntot)  Fb, where Ntot is the total number of waste packages. The probability that the two 
neighboring packages have not failed is given in Eq. I-4. The fractional increase in failures is: 
Fadd = (Nnr/Ntot)  Fb x 2  (1-Fb  {2/[(2/Fb)-1]}) (Eq. I-5) 
where a factor of 2 is added to account for the induced failure of 2 neighboring waste packages 
for every failed package with N-Reactor SNF. 
Equation I-5 can be used to calculate fractional increases in the number of waste packages failing 
due to a pyrophoric event for different fractions of failures. There will be approximately 
160 waste packages with N-Reactor SNF and 12,000 packages in total. For Nnr = 160 and 
Ntot = 12,000. For Fb = 0.001, 0.01, 0.1, 0.2, 0.5, and 0.75, Equation I-5 predicts Fadd = 0.00003, 
0.00027, 0.0026, 0.0051, 0.0089, and 0.0020, respectively. The maximum fractional increase 
predicted by Eq. I-5 is Fadd = 0.0089 for Fb = 0.48. That is, if 48 percent of all waste packages 
fail, the increase in failed packages due to pyrophoric events is less than 1 percent. Because the 
various waste package types were considered to be evenly distributed, corresponding increases 
occur in the total amounts of radionuclides released over the regulatory period due to pyrophoric 
events for the N-Reactor SNF-containing waste packages (i.e., the maximum increase in 
radionuclide release is less than 1 percent). Therefore, even under the extremely conservative 
scenario that all failed waste packages with N-Reactor SNF undergo pyrophoric reaction and 
cause the two adjacent waste packages to fail, pyrophoricity in N-Reactor SNF-containing waste 
packages has an insignificant effect on radionuclide release. This supports the exclusion of 
FEP 2.1.02.08.0A due to low consequence. 
Impact of Pyrophoric Events on Peak Offsite DoseThe impact a single pyrophoric event 
would have on peak offsite dose can be bounded by probabilistically assessing the potential 
worst case release from a failed waste package, in terms of the potential percent increase in 
radionuclide release. It is independent of any particular repository model. 
A pyrophoric event involving a single N-Reactor waste package with the simultaneous failure of 
the two adjacent CSNF waste packages could result in a peak offsite dose that is equivalent to 
approximately three times the value of a single waste package failure. 
In the event a waste package containing N-Reactor SNF fails and subsequently undergoes a 
pyrophoric event, the initial release would be approximately three waste package equivalents of 
radionuclides rather than one waste package equivalent. However, because whether or when a 
waste package fails is independent of the type of SNF it contains, the probability that any given 
failed waste package contains N-Reactor SNF is equal to: 
(Eq. 1-2) Nnr/Ntot = 160/12,000 = 0.013 
Therefore, the average increase in radionuclide release from the repository due to a pyrophoric 
event involving a waste package containing N-Reactor SNF can be represented as the probability 
that the failed waste package contains N-Reactor SNF multiplied by the probability that waste 
package undergoes a pyrophoric event multiplied by three waste package equivalents (WPE). In 
the worst case, where every breach of a waste package containing N-Reactor fuel results in a 
pyrophoric event, the increase is: 
April 2004 I-4 of I-6 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
(0.013 N-Reactor WP/total WP)(1pyrophoric event per failed N-Reactor WP)(3 WPE) . 0.04 WPE(Eq. I-3) 
This represents a 4 percent increase in peak offsite dose above that which would result if 
pyrophoric events were not possible. This upper bound reflects the assumption that a pyrophoric 
event will always occur in a breached N-Reactor SNF-containing waste package, and that the 
adjacent waste packages contain CSNF. Because a pyrophoric event will most likely not occur 
following an N-Reactor waste package breach, the actual increase is estimated to be much 
smaller. 
Additionally, preliminary postclosure site boundary dose sensitivity analyses have indicated that 
the increase of the dose rate at the site boundary resulting from the complete release of the 
radionuclide inventory in waste packages containing N-Reactor SNF during one performance 
assessment time step is insignificant (Thornton 1998 [DIRS 107796]). Thus, even the postulated 
instantaneous release due to a pyrophoric event does not have a significant effect on overall 
radionuclide releases and thus supports the exclusion of FEP 2.1.02.08.0A due to low 
consequence. 
Clustering Sensitivity Model.Clustering can be defined as multiple waste packages failing in 
a short time period. They can be postulated as being either induced by some initiating event that 
is not associated with pyrophoricity of DSNF (nonpyrophoric-induced cluster) or induced by an 
initiating pyrophoric event that results in subsequent pyrophoric events (pyrophoric-induced 
cluster). Regarding nonpyrophoric-induced clusters, an argument can be made similar to that in 
the preceding section. Regardless of the number of waste packages involved in some random 
event that results in clustering of nonpyrophoric event-induced waste package failures, each 
waste package involved has approximately 0.013 probability of being an N-Reactor waste 
package. Therefore, the maximum impact pyrophoric events associated with a 
nonpyrophoric-induced clustering event would be a 4 percent increase in the peak dose reflected 
in Eq. I-3. 
The effect of pyrophoric event-induced clustering event will be insignificant based on the 
expected separation that will exist between emplaced N-Reactor waste packages. Also, there 
will likely be insufficient oxygen available in a drift to support multiple pyrophoric events 
occurring simultaneously. 
April 2004 I-5 of I-6 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
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I-6 of I-6 April 2004 Miscellaneous Waste-Form FEPs 
ATTACHMENT II 
RADIONUCLIDE INVENTORY FOR FINAL WASTE FORMS 
 Of the radionuclide mass to be disposed of in the Yucca Mountain repository, 
95.6 percent is anticipated to be due to CSNF. The remaining 4.4 percent of the mass 
will be comprised of DSNF and HLW. 
236 
 The four radionuclides that dominate the total mass inventory are 238U, 235U, 239Pu, and 
U. Of these, 238U is the main contributor with 63.9 million kilograms of mass. This 
corresponds to nearly 97 percent of the total mass from all the radionuclides to be 
disposed of in the repository. 
129I 
Sanchez et al. (1998 [DIRS 149368]) estimated the masses of the radionuclides to be disposed of 
in the Yucca Mountain repository. These values were calculated with use of radioactivity values 
from Waste Form Degradation and Radionuclide Mobilization Expert Elicitation Project 
(CRWMS M&O 1998 [DIRS 100374]) and are presented below in Table II-1. The significant 
findings of this table are: 
227Ac 
93Mo  
94Nb 
59Ni 
 When ranking DOE-owned materials by themselves, 232Th is third behind 238U and 235U. 
This radionuclide, however, is not significant to total inventory. Its presence is due to the 
Nuclide 
ID 
Total 
(kg) 
4.7610-4 
2.1710-5 
existence of Thorium fuel-cycle materials that are unique to DOE-owned fuel. 
Table II-1. 
SNF 
kg (%) 
4.24E-04 (8.92E+01) 
3.09E+02 (5.07E-01) 
8.27E-02 (5.68E-02) 
9.08E+00 (1.08E-01) 
2.06E-01 (9.76E-01) 
1.32E-01 (5.86E-01) 
9.54E-01 (6.75E-02) 
1.77E-02 (1.34E-02) 
1.68E-02 (1.09E-01) 
1.77E+02 (5.74E-01) 
3.73E-06 (7.85E-01) 
108mAg  0.00E+00 (0.00E+00) 0.00E+00 (0.00E+00) 0.00E+00 (0.00E+00) 0.00E+00 (0.00E+00) 0 
241Am 
242mAm 
243Am 
14C 
36Cl 
243Cm  0.00E+00 (0.00E+00) 0.00E+00 (0.00E+00) 0.00E+00 (0.00E+00) 0.00E+00 (0.00E+00) 0 
244Cm 
245Cm 
246Cm 
135Cs 
137Cs 3.94E+02 (7.11E-01) 
9.28E+01 (7.02E-01) 
0.00E+00 (0.00E+00) 0.00E+00 (0.00E+00) 0.00E+00 (0.00E+00) 0.00E+00 (0.00E+00) 0 
93mNb  1.17E-07 (5.37E-01) 
5.52E-02 (1.87E-02) 
6.04E+00 (2.87E-01) 
YMP-Scale Source Term Mass Inventory (Calendar Year = 2035, Time = 0 yr) (a) 
DOE-Owned Commercial 
BWR 
kg (%) 
PWR 
kg (%) 
HLW 
kg (%) 
8.39E-06 (1.76E+00) 3.91E-05 (8.23E+00) 
1.19E+02 (1.94E-01) 4.14E+04 (6.79E+01) 1.92E+04 (3.14E+01) 6.1104 
9.64E+01 (6.62E+01) 4.92E+01 (3.37E+01) 146 9.25E-03 (6.35E-03) 
5.87E+03 (6.96E+01) 2.55E+03 (3.03E+01) 8430 8.68E-01 (1.03E-02) 
0.00E+00 (0.00E+00) 1.33E+01 (6.32E+01) 7.54E+00 (3.58E+01) 21.1 
0.00E+00 (0.00E+00) 1.50E+01 (6.66E+01) 7.39E+00 (3.28E+01) 22.5 
1.01E+03 (7.16E+01) 3.99E+02 (2.83E+01) 1410 6.63E-01 (4.69E-02) 
9.79E+01 (7.38E+01) 3.47E+01 (2.62E+01) 133 1.54E-03 (1.16E-03) 
1.16E+01 (7.53E+01) 3.78E+00 (2.46E+01) 15.4 9.80E-05 (6.38E-04) 
4.70E+02 (1.52E+00) 2.07E+04 (6.72E+01) 9.46E+03 (3.07E+01) 3.08104 
7.13E+02 (1.29E+00) 3.84E+04 (6.92E+01) 1.60E+04 (2.88E+01) 5.54104 
5.06E-02 (3.83E-04) 9.23E+03 (6.98E+01) 3.90E+03 (2.95E+01) 1.32104 
6.96E-06 (3.21E+01) 1.41E-05 (6.50E+01) 5.28E-07 (2.43E+00) 
2.84E+02 (9.63E+01) 1.09E+01 (3.71E+00) 295 7.46E-04 (2.53E-04) 
1.58E+03 (7.51E+01) 5.17E+02 (2.45E+01) 2110 1.67E+00 (7.95E-02) 
ANL-WIS-MD-000009 REV 01 April 2004 II-1 of II-4 99 
229 
230 
232 
232 
99 
Miscellaneous Waste-Form FEPs 
Table II-1. 
Nuclide 
ID 
63Ni 
237Np 
231Pa 
210Pb 
107Pd 
238Pu 
239Pu 
240Pu 
241Pu 
242Pu 
226Ra 
228Ra  
79Se 
151Sm 
6.49E+00 (1.75E+00) 0.00E+00 (0.00E+00) 2.79E+02 (7.50E+01) 8.62E+01 (2.32E+01) 371 
2.30E+02 (5.55E-01) 3.01E+04 (7.26E+01) 1.10E+04 (2.64E+01) 4.15104 1.89E+02 (4.55E-01) 
1.84E+00 (9.29E+01) 9.70E-02 (4.91E+00) 2.94E-02 (1.49E+00) 1.31E-02 (6.65E-01) 1.98 
1.03E-07 (3.15E+01) 3.2810-7 1.68E-09 (5.11E-01) 1.53E-07 (4.67E+01) 6.97E-08 (2.12E+01) 
5.52E+01 (3.33E-01) 0.00E+00 (0.00E+00) 1.15E+04 (6.96E+01) 4.97E+03 (3.01E+01) 1.65104 
3.51E+01 (2.65E-01) 9.58E+03 (7.23E+01) 3.52E+03 (2.66E+01) 1.33104 1.10E+02 (8.29E-01) 
6.98E+03 (1.77E+00) 3.59E+02 (9.08E-02) 2.75E+05 (6.96E+01) 1.13E+05 (2.85E+01) 3.95105 
1.38E+03 (8.76E-01) 1.10E+05 (6.98E+01) 4.61E+04 (2.93E+01) 1.57105 6.80E+01 (4.33E-02) 
6.02E+01 (1.67E-01) 2.46E+04 (6.83E+01) 1.14E+04 (3.16E+01) 3.60104 6.76E+00 (1.88E-02) 
6.73E+01 (1.90E-01) 2.41E+04 (6.81E+01) 1.12E+04 (3.17E+01) 3.54104 6.18E+00 (1.74E-02) 
7.97E-06 (8.73E+00) 9.1310-5 4.46E-07 (4.89E-01) 5.71E-05 (6.25E+01) 2.58E-05 (2.83E+01) 
3.63E-05 (1.00E+02) 3.6310-5 0.00E+00 (0.00E+00) 3.77E-11 (1.04E-04) 1.58E-11 (4.34E-05) 
3.40E+00 (7.77E-01) 6.19E+00 (1.42E+00) 3.03E+02 (6.91E+01) 1.25E+02 (2.87E+01) 438 
1.93E+01 (1.93E+00) 0.00E+00 (0.00E+00) 6.88E+02 (6.88E+01) 2.93E+02 (2.93E+01) 1000 
121mSn  0.00E+00 (0.00E+00) 0.00E+00 (0.00E+00) 0.00E+00 (0.00E+00) 0.00E+00 (0.00E+00) 0 
126Sn 
90Sr 
Tc 
Th 
Th 
Th  
232U  
233U 
234U 
235U 
236U 
238U 
93Zr 
1.11E+01 (5.46E-01) 0.00E+00 (0.00E+00) 1.43E+03 (7.02E+01) 5.95E+02 (2.93E+01) 2030 
2.12E+02 (8.80E-01) 3.13E+02 (1.30E+00) 1.68E+04 (6.95E+01) 6.85E+03 (2.84E+01) 2.41104 
4.36E+02 (7.62E-01) 9.13E+02 (1.60E+00) 3.92E+04 (6.86E+01) 1.66E+04 (2.91E+01) 5.72104 
1.28E-01 (9.97E+01) 3.34E-04 (2.59E-01) 6.96E-05 (5.40E-02) 2.61E-05 (2.03E-02) 0.129 
7.23E-02 (8.12E+00) 2.89E-03 (3.24E-01) 5.65E-01 (6.34E+01) 2.51E-01 (2.81E+01) 0.891 
7.94E+04 (9.46E+01) 4.51E+03 (5.37E+00) 1.32E-01 (1.58E-04) 8.39104 5.47E-02 (6.52E-05) 
0.00E+00 (0.00E+00) 0.00E+00 (0.00E+00) 0.00E+00 (0.00E+00) 0.00E+00 (0.00E+00) 0 
1.24E+03 (9.99E+01) 2.84E-01 (2.29E-02) 2.66E-01 (2.14E-02) 9.57E-02 (7.71E-03) 1240 
2.67E+02 (1.92E+00) 3.76E+01 (2.70E-01) 9.49E+03 (6.82E+01) 4.13E+03 (2.97E+01) 1.39104 
1.37E+05 (2.06E+01) 1.73E+02 (2.59E-02) 3.64E+05 (5.46E+01) 1.65E+05 (2.48E+01) 6.66105 
1.15E+04 (3.87E+00) 3.17E+01 (1.07E-02) 2.03E+05 (6.82E+01) 8.30E+04 (2.79E+01) 2.97105 
2.60E+06 (4.06E+00) 5.30E+04 (8.29E-02) 4.02E+07 (6.28E+01) 2.11E+07 (3.30E+01) 6.39107 
6.24E+02 (9.47E-01) 1.31E+03 (1.99E+00) 4.29E+04 (6.52E+01) 2.10E+04 (3.19E+01) 6.59104 
6.23E+04 4.14E+07 2.84E+06 2.16E+07 
(4.30E+00) b (6.28E+01) (3.28E+01) Total 
2.90E+06 (4.39%) 
238U 
235U 
239Pu 
236U 
240Pu 
Th  
93Zr 
241Am 
Tc 
2.60E+06 (4.06E+00) 5.30E+04 (8.29E-02) 
1.37E+05 (2.06E+01) 1.73E+02 (2.59E-02) 
6.98E+03 (1.77E+00) 3.59E+02 (9.08E-02) 
1.15E+04 (3.87E+00) 3.17E+01 (1.07E-02) 
1.38E+03 (8.76E-01) 6.80E+01 (4.33E-02) 
7.94E+04 (9.46E+01) 4.51E+03 (5.37E+00) 1.32E-01 (1.58E-04) 
6.24E+02 (9.47E-01) 
3.09E+02 (5.07E-01) 1.19E+02 (1.94E-01) 
4.36E+02 (7.62E-01) 
YMP-Scale Source Term Mass Inventory (Calendar Year = 2035, Time = 0 yr) (a) (Continued) 
DOE-Owned Commercial 
SNF 
kg (%) 
Top 10 Radionuclides (Ranked on Total Inventory for DOE-Owned and Commercial) 
1.31E+03 (1.99E+00) 4.29E+04 (6.52E+01) 2.10E+04 (3.19E+01) 6.59104 
9.13E+02 (1.60E+00) 3.92E+04 (6.86E+01) 1.66E+04 (2.91E+01) 5.72104 
ANL-WIS-MD-000009 REV 01 
Total 
(kg) 
BWR 
kg (%) 
PWR 
kg (%) 
HLW 
kg (%) 
(9.44E-02) c 6.60107 
6.31E+07 (95.6%) 
4.02E+07 (6.28E+01) 2.11E+07 (3.30E+01) 6.39107 
3.64E+05 (5.46E+01) 1.65E+05 (2.48E+01) 6.66105 
2.75E+05 (6.96E+01) 1.13E+05 (2.85E+01) 3.95105 
2.03E+05 (6.82E+01) 8.30E+04 (2.79E+01) 2.97105 
1.10E+05 (6.98E+01) 4.61E+04 (2.93E+01) 1.57105 
5.47E-02 (6.52E-05) 8.39104 
4.14E+04 (6.79E+01) 1.92E+04 (3.14E+01) 6.10104 
April 2004 II-2 of II-4 232 
99 
99 
Miscellaneous Waste-Form FEPs 
Table II-1. 
Nuclide 
ID 
137Cs 
238U 
235U 
Th  
236U 
239Pu 
93Zr 
240Pu 
Tc 
233U 
137Cs 
238U 
235U 
239Pu 
236U 
240Pu 
93Zr 
241Am 
Tc 
137Cs 
241Pu 
NOTES:  Data Values for radionuclides were previously reported in Wilson et al. (1994 [DIRS 100191]). 
 Data Values for radionuclides were not previously reported in Wilson et al. (1994 [DIRS 100191]). 
YMP-Scale Source Term Mass Inventory (Calendar Year = 2035, Time = 0 yr) (a) (Continued) 
SNF 
kg (%) 
3.94E+02 (7.11E-01) 
Top 10 Radionuclides (Ranked on Inventory for DOE-Owned Wastes Only) 
2.60E+06 (4.06E+00) 5.30E+04 (8.29E-02) 
1.37E+05 (2.06E+01) 1.73E+02 (2.59E-02) 
7.94E+04 (9.46E+01) 4.51E+03 (5.37E+00) 
1.15E+04 (3.87E+00) 3.17E+01 (1.07E-02) 
6.98E+03 (1.77E+00) 3.59E+02 (9.08E-02) 
6.24E+02 (9.47E-01) 
1.38E+03 (8.76E-01) 
4.36E+02 (7.62E-01) 
1.24E+03 (9.99E+01) 2.84E-01 (2.29E-02) 
3.94E+02 (7.11E-01) 
Top 10 Radionuclides (Ranked on Inventory for Commercial SNFS Only) 
 
 
 
 
 
 
 
 
 
 
a Table after Sanchez et al. (1998 [DIRS 149368]), data for spent nuclear fuel (SNF) and high-level 
radioactive waste (HLW) inventory data originally taken from INEEL/PA Parameters Database (DOE 1998 
[DIRS 122980]) (values represent intermediate database values, upgraded values can be found in 
Attachment II). (In total 41 radionuclides are inventoried in DOE (1998 [DIRS 122980]). 
b Mass inventory values calculated using half-lives from the Decay Libraries from ORIGEN2 (Croff 1980 
[DIRS 142613]; Croff 1980 [DIRS 101554]). 
c Note, the total DOE-owned mass load (due to radionuclides) is only 2.90E+06 kg. Thus only 4.39% of the 
total mass load (due to radionuclides) in YMP is due to DOE-Owned inventory. 
d Note, the total commercial mass load (due to radionuclides) is 6.30E+07 kg. Thus 95.6% of the total mass 
load (due to radionuclides) in YMP is due to commercial inventory. 
DOE-Owned Commercial 
7.13E+02 (1.29E+00) 3.84E+04 (6.92E+01) 1.60E+04 (2.88E+01) 5.54104 
1.31E+03 (1.99E+00) 
6.80E+01 (4.33E-02) 
9.13E+02 (1.60E+00) 
7.13E+02 (1.29E+00) 
ANL-WIS-MD-000009 REV 01 
HLW 
kg (%) 
 
 
 
 
 
 
 
 
 
 
PWR 
kg (%) 
 
 
 
 
 
 
 
 
 
 
4.02E+07 (6.28E+01) 2.11E+07 (3.30E+01) 6.13107 
3.64E+05 (5.46E+01) 1.65E+05 (2.48E+01) 5.29105 
2.75E+05 (6.96E+01) 1.13E+05 (2.85E+01) 3.88105 
2.03E+05 (6.82E+01) 8.30E+04 (2.79E+01) 2.86105 
1.10E+05 (6.98E+01) 4.61E+04 (2.93E+01) 1.56105 
4.29E+04 (6.52E+01) 2.10E+04 (3.19E+01) 6.39104 
4.14E+04 (6.79E+01) 1.92E+04 (3.14E+01) 6.06104 
3.92E+04 (6.86E+01) 1.66E+04 (2.91E+01) 5.58104 
3.84E+04 (6.92E+01) 1.60E+04 (2.88E+01) 5.44104 
2.46E+04 (6.83E+01) 1.14E+04 (3.16E+01) 3.60104 
II-3 of II-4 
Total 
(kg) 
BWR 
kg (%) 
 2.65106 
 1.37105 
 8.39104 
 1.15104 
7340  
1930  
1450  
1350  
1240  
1110  
April 2004 Miscellaneous Waste-Form FEPs 
INTENTIONALLY LEFT BLANK 
April 2004 II-4 of II-4 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
ATTACHMENT III 
ALPHA RECOIL MECHANICS 
The alpha recoil is analyzed per conservation of momentum in the center-of-mass (COM) frame 
of reference. Consider Figure III-1; a radionuclide X (e.g., 238U), at rest in the lab-system frame 
of reference (and also in the COM frame), undergoes an alpha decay. The radionuclide X decays 
to radionuclide Y (e.g., 234Th) by emitting an alpha particle, (see Figure III-2 for COM frame). 
A 
X 
z 
Initial Center-of-Mass (COM) Frame Conditions for Alpha Emitting Radionuclide 
A-4 4 
Y  
Z-2 2 
y V
v 
 vv V =Velocity of nuclide Y 
vr =Velocity of  particles 
Figure III-1. 
v v
y 
Figure III-2. 
where 
Final Center-of-Mass (COM) Frame Conditions after Alpha Decay Resulting in Two 
Decay Products, an Alpha Particle and the Decay Daughter 
Applying the conservation of momentum in the COM system leads to Equations III-1 and III-2. 
Momentum Before = Momentum After 
m V M 0 + v y 
My = Mass of recoil nucleus 
m  = Mass of  particles 
ANL-WIS-MD-000009 REV 01 
(Eq. III-1) 
(Eq. III-2)  
April 2004 
= 
III-1 of III-2 
z=Atomic Number 
A=molecular wt. Miscellaneous Waste-Form FEPs 
Therefore, the velocity of the recoil nucleus (in terms of the velocity of the alpha particle) is: 
 (Eq. III-3) v Vy 
m
My 
where 
the emitted alpha particle as: 
 2 =
= 
m  = mass of the alpha particle 
My = mass of the recoil nucleus. 
The kinetic energy of the recoil nucleus can now be determined in terms of the kinetic energy of 
m
M 
 
.
.
2
 
. .
.
.
m = KE( ) V KE(Y) M V  Y Y 
m
M

Y 
1
2 
1
2 
= . 
. 
Y (Eq. III-4) 
.  
.. 
. 
Thus, the kinetic energy of the recoil nucleus is a small fraction of that given to the alpha 
particle. The energy of the alpha particle is dependent upon the proper mass defect value (the 
amount of mass converted into energy). For the case of 238U decaying to 234Th, there is: KE() = 
4.196 MeV (Lederer and Shirley 1978 [DIRS 142133]), m  . 4.0, MTh . 234.0. This results in a 
kinetic energy value of 0.072 MeV for the recoil nucleus. 
It is important to note that 1) the velocity of the recoil nucleus is in the opposite direction of the 
alpha particle, and 2) the velocities of both the alpha particle and recoil nucleus are not a 
function of direction (i.e. they are isotropic in direction in the COM frame). Because the COM is 
not moving with respect to the lab-system frame, the velocities are also isotropic in the labsystem 
frame. Thus, for any recoil nucleus moving in a given direction (towards the subsurface 
groundwater, etc.), there is an equal probability that another recoil nucleus is moving in the 
opposite direction with the same speed. This isotropy means that only one half of the recoil 
nuclei are initially moving away from the fuel and into the groundwater. 
III-2 of III-2 April 2004 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
ATTACHMENT IV 
RADIONUCLIDE INVENTORY FOR PRESSURIZED-WATER REACTOR SPENT 
NUCLEAR FUELS AND BOILING-WATER REACTOR SPENT NUCLEAR FUELS 
This attachment contains time-dependent radionuclide information for CSNF with average 
burnup histories. Project Requirements Document (Canori and Leitner 2003 [DIRS 166275], 
p. 2-53) states that the Yucca Mountain Project shall comply with the relevant requirements 
contained in Yucca Mountain Site Characterization Project Requirements Document (YMP-RD) 
(YMP 2001 [DIRS 156713]). Table 1.3-1 of YMP-RD (YMP 2001 [DIRS 156713], p. 1.3-6) 
lists the amount of SNL/HLW to be accepted in the first MGR: 63,000 MTHM CSNF, 
640 MTHM commercial HLW, 4,027 MTHM DHLW, and 2,333 MTHM DSNF. 
CSNF has considerably higher burnup than that expected for DSNF (e.g., DOE N-Reactor fuels 
typically have less than one-twentieth of the burnup of CSNF). Table IV-1 presents radionuclide 
inventories for pressurized-water reactors (PWRs) at various decay times (time after burnup). 
Table IV-2 presents radionuclide inventories for boiling-water reactors (BWRs) at the same 
decay times used in Table IV-1. The data for both tables was obtained from PWR Source Term 
Generation and Evaluation (CRWMS M&O 1999 [DIRS 136429]) and BWR Source Term 
Generation and Evaluation (CRWMS M&O 1999 [DIRS 136428]). In each table only a limited 
portion of the original data (timeframes from 125 years up to 10,025 years) is reproduced. The 
only additions to the tables are: 1) presentation of decay mode and half-lives for the 
radionuclides in columns 2 and 3, and 2) the generation of the total radionuclide inventories at 
the bottom of each table. The decay modes and the half-lives were obtained from 
Parrington et al. (1996 [DIRS 103896]), which demonstrates qualitatively that a significant 
portion of the radionuclides in the expected inventory are alpha emitters. For simplicity, the net 
results presented in Table IV-3 present bounding calculations. That is, the calculations consider 
100 percent of the radionuclides to be alpha emitters, which are the worst case radiolysis 
generators. The only numerical values from these tables that are used in Table IV-3 are those for 
the total radionuclide inventory for the PWR and BWR fuel assemblies (see bottom on 
Tables IV-1 and IV-2). As can be identified when comparing the final results in Table IV-3 
(corrosion rates due to alpha radiolysis) to the dissolution rates in Figure 2, the radiolysis rates 
are insignificant. This means that if future estimates for the quantities of radionuclides in Tables 
IV-1 and IV-2 result in slightly different inventories, their differences will not significantly 
impact the screening argument for FEP 2.1.13.02.0A, Radiation damage in EBS. 
April 2004 IV-1 of IV-10 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table IV-1. Yucca Mountain Project Radionuclide Inventory for Pressurized-Water Reactor Fuels 
Radioactivity Inventory [Ci/Assembly] Decay 
Mode(a) 
Nuclide 
ID Half-Life(b) 
10,025 yr 5,025 yr 2,025 yr 1,025 yr 525 yr 425 yr 325 yr 225 yr 125 yr 
227Ac , .,  
., , Ã+ 
, Ã+, e. 
, ,  
., , e. 
7.14E-05 8.70E-05 1.03E-04 1.85E-04 3.41E-04 8.09E-04 1.59E-03 5.59E-05 4.00E-05 21.77 years 
108Ag 7.48E-05 4.34E-05 2.51E-05 1.64E-06 6.99E-09 5.41E-16 7.61E-28 1.29E-04 2.23E-04 2.39 minutes 
108mAg 8.60E-04 4.98E-04 2.89E-04 1.88E-05 8.03E-08 6.22E-15 8.74E-27 1.48E-03 2.56E-03 130 years 
241Am 2.40E+03 2.05E+03 1.75E+03 1.49E+03 1.27E+03 5.69E+02 1.15E+02 1.15E+00 1.36E-01 432.7 years 
242Am 3.89E+00 2.38E+00 1.45E+00 8.90E-01 5.44E-01 4.66E-02 3.42E-04 1.34E-10 2.84E-21 16.02 hours 
242mAm , , e ., SF 141 years 3.91E+00 2.39E+00 1.46E+00 8.94E-01 5.47E-01 4.68E-02 3.43E-04 1.35E-10 2.86E-21 
243Am , , SF 
. 
 
, .,  
, .,  
. 
 
.,  
.,  
, , SF 
, , SF 
., +,  
, , SF 
2.18E+01 2.16E+01 2.14E+01 2.12E+01 2.10E+01 2.00E+01 1.82E+01 1.37E+01 8.57E+00 7370 years 
39Ar 2.41E-05 1.86E-05 1.44E-05 3.97E-06 3.02E-07 1.32E-10 3.36E-16 3.12E-05 4.04E-05 269 years 
137mBa 3.85E+03 3.82E+02 3.79E+01 3.76E+00 3.73E-01 3.58E-06 3.30E-16 0.00E+00 0.00E+00 2.552 minutes 
211Bi 7.14E-05 8.70E-05 1.03E-04 1.85E-04 3.41E-04 8.09E-04 1.59E-03 5.59E-05 4.00E-05 2.14 minutes 
212Bi 1.07E-03 3.98E-04 1.48E-04 1.91E-06 8.94E-07 9.08E-07 9.37E-07 2.89E-03 7.81E-03 1.009 hours 
14C 3.21E-01 3.17E-01 3.13E-01 2.95E-01 2.61E-01 1.82E-01 9.91E-02 3.24E-01 3.28E-01 5730 years 
1.03~105 years 41Ca 9.23E-05 9.22E-05 9.22E-05 9.19E-05 9.12E-05 8.94E-05 8.65E-05 9.24E-05 9.24E-05 
113mCd 3.02E-06 2.21E-08 1.62E-10 3.43E-21 0.00E+00 0.00E+00 0.00E+00 4.12E-04 5.61E-02 14.1 years 
142Ce Stable 1.86E-0 1.86E-0 5 1.86E-0 5 5 1.86E-05 1.86E-05 1.86E-05 1.86E-05 1.86E-05 1.86E-05 
144Ce 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 284.6 days 
249Cf 4.27E-05 3.50E-05 2.87E-05 1.07E-05 1.48E-06 3.92E-09 1.99E-13 5.20E-05 6.34E-05 351 years 
250Cf 1.06E-11 1.55E-13 1.03E-13 1.01E-13 9.67E-14 8.58E-14 7.03E-14 2.10E-09 4.20E-07 13.1 years 
3.01~105 years 36Cl 6.80E-03 6.79E-03 6.79E-03 6.79E-03 6.77E-03 6.72E-03 6.65E-03 6.80E-03 6.80E-03 
1.2~103 years 242Cm , , SF,  3.22E+00 1.97E+00 1.20E+00 7.36E-01 4.50E-01 3.85E-02 2.82E-04 1.12E-10 2.36E-21 
243Cm , , SF 
, , SF 
, , SF 
.,  
., ,  
. 
.,  
,  
6.95E-03 6.10E-04 5.36E-05 2.80E-10 7.66E-21 0.00E+00 0.00E+00 7.91E-02 9.01E-01 162.8 days 
244Cm 1.40E-02 3.03E-04 6.58E-06 3.18E-14 7.30E-31 0.00E+00 0.00E+00 6.42E-01 2.96E+01 29.1 years 
245Cm 2.99E-01 2.97E-01 2.94E-01 2.83E-01 2.60E-01 2.04E-01 1.36E-01 3.02E-01 3.04E-01 18.1 years 
8.5~103 ye ars 1.03E-01 1.01E-01 9.98E-02 9.83E-02 9.69E-02 9.01E-02 7.78E-02 5.01E-02 2.41E-02 246Cm 
4.76~103 years 60Co 2.27E-15 4.40E-21 8.52E-27 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.17E-09 6.06E-04 
134Cs 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.59E-28 6.33E-14 5.271 years 
135Cs 3.50E-01 3.50E-01 3.50E-01 3.50E-01 3.50E-01 3.49E-01 3.49E-01 3.50E-01 3.50E-01 2.065 years 
137Cs 2.3~106 years 4.08E+03 4.05E+02 4.01E+01 3.98E+00 3.95E-01 3.79E-06 3.49E-16 0.00E+00 0.00E+00 
April 2004 IV-2 of IV-10 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table IV-1. Yucca Mountain Project Radionuclide Inventory for Pressurized-Water Reactor Fuels (Continued) 
Radioactivity Inventory [Ci/Assembly] Decay 
Mode(a) 
Nuclide 
ID Half-Life(b) 
10,025 yr 5,025 yr 2,025 yr 1,025 yr 525 yr 425 yr 325 yr 225 yr 125 yr 
150Eu ., , Ã+ 
., , Ã 
.,  
 
. 
.,  
.,  
.,  
, , . 
+,  
e. 
.,  
 
. 
,  
.,  
,.,  
.,  
,  
.,  
.,  
., ,  
.,  
.,  
. 
, ,  
., , Ã 
6.23E-08 8.98E-09 1.30E-09 8.11E-14 3.17E-22 0.00E+00 0.00E+00 4.31E-07 2.99E-06 36 years 
152Eu 2.20E-07 1.21E-09 6.67E-12 3.39E-23 0.00E+00 0.00E+00 0.00E+00 3.98E-05 7.23E-03 13.48 years 
154Eu 2.06E-08 6.44E-12 2.01E-15 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6.57E-05 2.10E-01 8.59 years 
155Eu 2.60E-18 9.61E-25 3.65E-31 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7.04E-12 1.90E-05 4.71 years 
55Fe 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.04E-22 3.25E-11 2.73 years 
3H 5.37E-06 1.94E-08 7.02E-11 4.35E-23 0.00E+00 0.00E+00 0.00E+00 1.48E-03 4.10E-01 12.3 years 
166mHo 1.2~103 ye ars 5.08E-04 4.79E-04 4.52E-04 4.27E-04 4.03E-04 3.02E-04 1.69E-04 2.99E-05 1.67E-06 
1.57~107 years 129I 2.19E-02 2.19E-02 2.19E-02 2.19E-02 2.19E-02 2.19E-02 2.19E-02 2.19E-02 2.19E-02 
85Kr 4.26E-06 6.63E-09 1.03E-11 9.36E-26 0.00E+00 0.00E+00 0.00E+00 2.74E-03 1.76E+00 10.73 years 
~3.5~103 y ears 4.06E-02 3.98E-02 3.90E-02 3.83E-02 3.75E-02 3.40E-02 2.79E-02 1.54E-02 5.71E-03 93Mo 
7~102 ye ars 91Nb 1.77E-05 1.60E-05 1.44E-05 1.30E-05 1.18E-05 7.06E-06 2.55E-06 1.20E-07 7.32E-10 
93mNb 9.27E-01 9.26E-01 9.25E-01 9.22E-01 9.16E-01 9.05E-01 8.95E-01 9.29E-01 1.09E+00 16.1 years 
2.0~104 ye ars 8.36E-01 8.33E-01 8.30E-01 8.28E-01 8.25E-01 8.11E-01 7.84E-01 7.07E-01 5.96E-01 94Nb 
59Ni 7.6~104 years 2.09E+00 2.09E+00 2.08E+00 2.08E+00 2.08E+00 2.07E+00 2.05E+00 2.00E+00 1.90E+00 
63Ni 1.26E+02 6.30E+01 3.15E+01 1.58E+01 7.89E+00 2.47E-01 2.43E-04 2.31E-13 2.12E-28 100 years 
237Np 2.14~106 years 4.60E-01 5.12E-01 5.57E-01 6.98E-01 7.90E-01 8.12E-01 8.11E-01 3.98E-01 3.26E-01 
238Np 6.58E-03 4.02E-03 2.46E-03 2.11E-04 1.54E-06 6.08E-13 1.29E-23 1.08E-02 1.76E-02 2.117 days 
239Np 1.2~103 years 2.18E+01 2.16E+01 2.14E+01 2.12E+01 2.10E+01 2.00E+01 1.82E+01 1.37E+01 8.57E+00 
231Pa 3.28~104 years 7.63E-05 9.18E-05 1.07E-04 1.85E-04 3.41E-04 8.09E-04 1.59E-03 6.08E-05 4.53E-05 
233Pa 4.60E-01 5.12E-01 5.57E-01 6.98E-01 7.90E-01 8.12E-01 8.11E-01 3.98E-01 3.26E-01 27.0 days 
234Pa 1.92E-04 1.92E-04 1.92E-04 1.92E-04 1.92E-04 1.92E-04 1.92E-04 1.92E-04 1.92E-04 6.69 hours 
234mPa 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.17 minutes 
211Pb 7.14E-05 8.70E-05 1.03E-04 1.85E-04 3.41E-04 8.09E-04 1.59E-03 5.59E-05 4.00E-05 36.1 minutes 
212Pb 1.07E-03 3.98E-04 1.48E-04 1.91E-06 8.94E-07 9.08E-07 9.37E-07 2.89E-03 7.81E-03 10.64 hours 
6.5~106 ye ars 8.41E-02 8.41E-02 8.41E-02 8.41E-02 8.41E-02 8.41E-02 8.41E-02 8.41E-02 8.40E-02 107Pd 
145Pm 8.45E-09 1.68E-10 3.35E-12 1.05E-20 0.00E+00 0.00E+00 0.00E+00 4.24E-07 2.13E-05 17.7 years 
146Pm 5.45E-18 1.96E-23 7.03E-29 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.52E-12 4.22E-07 5.53 years 
April 2004 IV-3 of IV-10 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table IV-1. Yucca Mountain Project Radionuclide Inventory for Pressurized-Water Reactor Fuels (Continued) 
Radioactivity Inventory [Ci/Assembly] Decay 
Mode(a) 
Nuclide 
ID Half-Life(b) 
10,025 yr 5,025 yr 2,025 yr 1,025 yr 525 yr 425 yr 325 yr 225 yr 125 yr 
147Pm .,  
 
, .,  
,  
.,  
, , SF 
, , SF 
, , SF 
, , SF 
, .,  
, , SF 
,  
,  
. 
., , Ã+ 
.,  
,  
,  
. 
.,  
.,  
, e. 
. 
.,  
. 
., , e. 
.,  
. 
0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
6.87E-04 2.55E-04 9.47E-05 1.22E-06 5.73E-07 5.81E-07 6.00E-07 
7.14E-05 8.70E-05 1.03E-04 1.85E-04 3.41E-04 8.09E-04 1.59E-03 
1.07E-03 3.98E-04 1.48E-04 1.91E-06 8.94E-07 9.08E-07 9.37E-07 
0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
8.85E-07 8.84E-07 8.84E-07 8.81E-07 8.76E-07 8.60E-07 8.35E-07 
1.04E+03 4.73E+02 2.16E+02 9.83E+01 4.49E+01 9.43E-01 6.33E-04 2.97E-10 6.27E-21 
8.86E-07 
2.6234 years 3.99E-10 1.33E-21 
0.298 ms 
1.780 ms 
0.145 seconds 
17.28 minutes 
2.87 years 
87.7 years 
2.410~104 1.77E+02 1.77E+02 1.76E+02 1.76E+02 1.75E+02 1.73E+02 1.69E+02 1.56E+02 1.37E+02 
6.56 3.15E+ ~103 3.18E+02 3.12E+ 02 3.09E+ 02 02 3.05E+02 2.90E+02 2.61E+02 1.90E+02 1.12E+02 
14.4 years 
3.75~105 years 
11.435 days 
3.66 days 
4.88~1010 ye ars 1.39E-05 1.39E-05 1.39E-05 1.39E-05 1.39E-05 1.39E-05 1.39E-05 1.39E-05 1.39E-05 
207 days 
29.9 seconds 
3.96 seconds 
55.6 seconds 
1.02 years 
2.758 years 
12.4 days 
11 seconds 
6.5~104 ye ars 4.57E-02 4.57E-02 4.57E-02 4.57E-02 4.56E-02 4.56E-02 4.55E-02 4.52E-02 4.47E-02 
90 years 
1.128 days 
55 years 
1.0~105 ye ars 3.85E-01 3.85E-01 3.84E-01 3.84E-01 3.84E-01 3.82E-01 3.80E-01 3.72E-01 3.59E-01 
29.1 years 
212Po 1.85E-03 5.00E-03 
215Po 5.59E-05 4.00E-05 
216Po 2.89E-03 7.81E-03 
144Pr 
236Pu 8.85E-07 
238Pu 
239Pu 
240Pu 
241Pu 1.97E+02 1.87E+00 3.12E-01 2.97E-01 2.95E-01 2.83E-01 2.61E-01 2.04E-01 1.36E-01 
1.64E+00 1.64E+00 1.64E+00 1.64E+00 1.64E+00 1.64E+00 1.63E+00 1.62E+00 1.61E+00 
7.14E-05 8.70E-05 1.03E-04 1.85E-04 3.41E-04 8.09E-04 1.59E-03 4.00E-05 5.59E-05 
1.07E-03 3.98E-04 1.48E-04 1.91E-06 8.94E-07 9.08E-07 9.37E-07 7.81E-03 2.89E-03 
242Pu 
223Ra 
224Ra 
87Rb 
102Rh 1.26E-13 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5.26E-24 
0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
7.14E-05 8.70E-05 1.03E-04 1.85E-04 3.41E-04 8.09E-04 1.59E-03 
1.07E-03 3.98E-04 1.48E-04 1.91E-06 8.94E-07 9.08E-07 9.37E-07 
0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
5.38E-02 5.38E-02 5.37E-02 5.35E-02 5.32E-02 5.21E-02 5.03E-02 
3.84E-01 3.84E-01 3.84E-01 3.82E-01 3.80E-01 3.72E-01 3.59E-01 
106Rh 
219Rn 5.59E-05 4.00E-05 
220Rn 2.89E-03 7.81E-03 
106Ru 
125Sb 8.53E-22 9.11E-11 
126Sb 5.38E-02 5.39E-02 
126mSb 3.85E-01 3.85E-01 
79Se 
151Sm 9.75E+01 4.51E+01 2.09E+01 9.67E+00 4.48E+00 9.52E-02 4.30E-05 3.96E-15 6.85E-32 
121Sn 9.95E-02 3.51E-01 2.82E-02 8.00E-03 2.27E-03 4.16E-06 1.40E-11 5.36E-28 0.00E+00 
121mSn 1.28E-01 4.52E-01 3.64E-02 1.03E-02 2.92E-03 5.36E-06 1.81E-11 6.90E-28 0.00E+00 
126Sn 
90Sr 2.32E+03 1.97E+02 1.68E+01 1.43E+00 1.22E-01 5.48E-07 1.11E-17 0.00E+00 0.00E+00 
April 2004 IV-4 of IV-10 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table IV-1. Yucca Mountain Project Radionuclide Inventory for Pressurized-Water Reactor Fuels (Continued) 
Radioactivity Inventory [Ci/Assembly] Decay 
Mode(a) 
Nuclide 
ID Half-Life(b) 
10,025 yr 5,025 yr 2,025 yr 1,025 yr 525 yr 425 yr 325 yr 225 yr 125 yr 
99 2.13~105 years .,  
, e. 
, .,  
,  
, .,  
.,  
.,  
. 
. 
 
 
 
 
 
. 
 
. 
. 
8.98E+00 8.98E+00 8.98E+00 8.97E+00 8.97E+00 8.96E+00 8.93E+00 8.84E+00 8.69E+00 Tc 
125mTe 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.08E-22 2.23E-11 58 days 
227Th 7.05E-05 8.58E-05 1.01E-04 1.83E-04 3.36E-04 7.98E-04 1.57E-03 5.51E-05 3.95E-05 18.72 days 
228Th 1.07E-03 3.98E-04 1.48E-04 1.91E-06 8.94E-07 9.08E-07 9.37E-07 2.89E-03 7.81E-03 1.913 years 
230Th 7.54~104 years 3.40E-03 4.73E-03 6.08E-03 1.29E-02 2.64E-02 6.61E-02 1.29E-01 2.14E-03 1.00E-03 
231Th 7.42E-03 7.43E+00 7.45E-03 7.54E-03 7.71E-03 8.19E-03 8.91E-03 7.40E-03 7.38E-03 1.063 days 
234Th 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.48E-01 24.10 days 
207Tl 7.12E-05 8.67E-05 1.02E-04 1.85E-04 3.40E-04 8.07E-04 1.58E-03 5.57E-05 3.99E-05 4.77 minutes 
208Tl 3.85E-04 1.43E-04 5.32E-05 6.86E-07 3.21E-07 3.26E-07 3.37E-07 1.04E-03 2.81E-03 3.053 minutes 
232U 1.04E-03 3.87E-04 1.44E-04 1.87E-06 8.76E-07 8.60E-07 8.35E-07 2.81E-03 7.59E-03 70 years 
233U 1.592~105 ye ars 1.62E-04 3.21E-04 5.08E-04 7.19E-04 9.52E-04 2.33E-03 5.61E-03 1.60E-02 3.32E-02 
234U 2.46~105 years 1.12E+00 1.33E+00 1.42E+00 1.46E+00 1.48E+00 1.49E+00 1.49E+00 1.48E+00 1.46E+00 
235U 7.04~108 years 7.42E-03 7.43E-03 7.45E-03 7.54E-03 7.71E-03 8.19E-03 8.91E-03 7.40E-03 7.38E-03 
236U 2.342~107 ye ars 1.73E-01 1.74E-01 1.74E-01 1.75E-01 1.76E-01 1.81E-01 1.89E-01 2.09E-01 2.30E-01 
237U 7.46E-06 7.11E-06 7.05E-06 6.77E-06 6.24E-06 4.88E-06 3.25E-06 4.48E-05 4.71E-03 6.75 days 
238U 4.47~109 years 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.48E-01 1.48E-01 
90Y 2.32E+03 1.97E+02 1.68E+01 1.43E+00 1.22E-01 5.48E-07 1.11E-17 0.00E+00 0.00E+00 2.67 days 
1.5~106 ye ars 8.94E-01 8.94E-01 8.94E-01 8.94E-01 8.94E-01 8.94E-01 8.93E-01 8.92E-01 8.90E-01 93Zr 
Total(c) = 1.70E+04 4.38E+03 2.69E+03 2.18E+03 1.87E+03 1.09E+03 6.03E+02 3.95E+02 2.86E+02 
Source: CRWMS M&O 1999 [DIRS 136429] Attachment X 
NOTES: a Decay modes identified from Parrington et al. (1996 [DIRS 103896]). 
b Half-lives identified from Parrington et al. (1996 [DIRS 103896]). 
c Total radionuclide activities are calculated here and are not part of original reference. 
Decay times listed in Table IV-1 include the 25-year decay time of the fuels prior to its emplacement into the YMP repository. Thus, fuel with a date of 525 
years corresponds to fuel that has been stored for 500 years (calendar year 2535) in the repository. 
April 2004 IV-5 of IV-10 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table IV-2. Yucca Mountain Project Radionuclide Inventory for Boiling-Water Reactor Fuels 
Radioactivity Inventory [Ci/Assembly] Nuclide 
ID 
Decay 
Mode(a) Half-Life(b) 10,025 yr 5,025 yr 2,025 yr 1,025 yr 525 yr 425 yr 325 yr 225 yr 125 yr 
108Ag ., , Ã+ 
, Ã+, e. 
, , SF 
., , e. 
, , e., SF 
, , SF 
. 
 
, .,  
. 
 
.,  
.,  
., +,  
, , SF 
, , SF , 
, , SF 
, , SF 
, , SF 
.,  
., , 
. 
.,  
., , Ã+ 
., , Ã 
.,  
 
. 
.,  
2.39m 1.93E-0 1.12E-0 4 6.48E-0 4 5 3.76E-05 2.18E-05 1.42E-06 6.06E-09 4.68E-16 6.59E-28 
130 years 2.22E-03 1.29E-03 
432.7 years 6.72E+02 5.74E+02 4.89E+02 4.17E+02 3.55E+02 1.59E+02 3.21E+01 2.90E-01 1.80E-02 
16.02 hours 4.94E-01 3.02E-01 1.85E-01 1.58E-02 1.16E-04 4.56E-11 9.64E-22 1.32E+00 8.07E-01 
141 years 4.96E-01 3.03E-01 1.85E-01 1.59E-02 1.16E-04 4.58E-11 9.69E-22 1.33E+00 8.10E-01 
7370 years 5.29E+00 5.24E+00 5.19E+00 5.14E+00 5.09E+00 4.86E+00 4.42E+00 3.34E+00 2.08E+00 
269 years 7.41E-06 5.73E-06 4.43E-06 1.22E-06 9.28E-08 4.07E-11 1.03E-16 
2.552 minutes 
1.009 hours 
5730 years 
1.03~105 years 
14.1 years 
284.6 days 
3.01~105 years 
1.2~103 ye ars 1.09E+00 6.67E-01 4.08E-01 2.50E-01 1.53E-01 1.31E-02 9.58E-05 3.78E-11 8.00E-22 
162.8 days 
29.1 years 
18.1 years 
8.5~103 ye ars 1.43E-02 1.41E-02 1.39E-02 1.37E-02 1.35E-02 1.25E-02 1.08E-02 6.97E-03 3.35E-03 
4.76~103 years 
5.271 years 
2.065 years 
2.3~106 years 
13.48 years 
8.59 years 
4.71 years 
2.73 years 
12.3 years 
1.2~103 ye ars 1.11E-03 1.05E-03 9.93E-04 9.37E-04 8.84E-04 6.62E-04 3.72E-04 6.58E-05 3.66E-06 
108mAg 7.46E-04 4.32E-04 2.50E-04 1.63E-05 6.96E-08 5.39E-15 7.58E-27 
241Am 
242Am 
242mAm 
243Am 
39Ar 1.24E-05 9.59E-06 
137mBa 1.30E+03 1.29E+02 1.28E+01 1.27E+00 1.26E-01 1.21E-06 1.11E-16 0.00E+00 0.00E+00 
212Bi 1.77E-03 6.55E-04 2.43E-04 9.01E-05 3.35E-05 3.98E-07 1.69E-07 1.77E-07 1.92E-07 
1.69E-01 1.67E-01 1.65E-01 1.56E-01 1.38E-01 9.59E-02 5.24E-02 
3.72E-05 3.72E-05 3.72E-05 3.70E-05 3.68E-05 3.60E-05 3.49E-05 
8.91E-07 6.53E-09 4.79E-11 1.01E-21 0.00E+00 0.00E+00 0.00E+00 
0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
1.66E-02 1.22E-04 
14C 1.73E-01 1.71E-01 
41Ca 3.73E-05 3.72E-05 
113mCd 
144Ce 
36Cl 2.93E-03 2.93E-03 2.93E-03 2.92E-03 2.92E-03 2.90E+00 2.86E+00 2.93E-03 2.93E-03 
242Cm 
243Cm 2.17E-01 1.91E-02 1.68E-03 1.47E-04 1.29E-05 6.76E-11 1.85E-21 0.00E+00 0.00E+00 
244Cm 2.61E-03 5.67E-05 1.23E-06 5.94E-15 1.37E-31 0.00E+00 0.00E+00 5.54E+00 1.20E-01 
245Cm 4.00E-02 3.97E-02 3.93E-02 3.90E-02 3.87E-02 3.71E-02 3.42E-02 2.68E-02 1.78E-02 
246Cm 
60Co 8.51E-05 1.65E-10 3.19E-16 6.18E-22 1.20E-27 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
1.39E-01 1.39E-01 1.39E-01 1.39E-01 1.39E-01 1.39E-01 1.38E-01 
134Cs 1.59E-14 3.98E-29 
135Cs 1.39E-01 1.39E-01 
137Cs 1.37E+03 1.36E+02 1.35E+01 1.34E+00 1.33E-01 1.28E-06 1.18E-16 0.00E+00 0.00E+00 
152Eu 2.92E-03 1.61E-05 8.86E-08 4.88E-10 2.69E-12 1.37E-23 0.00E+00 0.00E+00 0.00E+00 
5.34E-09 1.68E-12 5.23E-16 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
8.07E-19 2.98E-25 1.14E-31 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
1.87E-06 6.75E-09 2.45E-11 1.51E-23 0.00E+00 0.00E+00 0.00E+00 
154Eu 5.45E-02 1.71E-05 
155Eu 5.91E-06 2.19E-12 
55Fe 1.02E-11 9.56E-23 
3H 1.43E-01 5.17E-04 
166mHo 
April 2004 IV-6 of IV-10 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table IV-2. Yucca Mountain Project Radionuclide Inventory for Boiling-Water Reactor Fuels (Continued) 
Radioactivity Inventory [Ci/Assembly] Nuclide 
ID 
Decay 
Mode(a) Half-Life(b) 125 yr 225 yr 325 yr 425 yr 525 yr 1,025 yr 2,025 yr 5,025 yr 10,025 yr 
1.57~107 years 7.42E-03 7.42E-03 
10.73 years 5.93E-01 9.21E-04 
~3.5~103 y ears 2.17E-04 2.13E-04 2.09E-04 2.05E-04 2.01E-04 1.82E-04 1.49E-04 8.24E-05 3.06E-05 
16.1 years 3.39E-01 3.39E-01 3.39E-01 3.39E-01 3.38E-01 3.38E-01 3.36E-01 3.41E-01 3.39E-01 
2.0~104 ye ars 1.86E-02 1.86E-02 1.85E-02 1.84E-02 1.84E-02 1.81E-02 1.75E-02 1.58E-02 1.33E-02 
7.6~104 ye ars 5.02E-01 5.01E-01 5.01E-01 5.00E-01 5.00E-01 4.97E-01 4.93E-01 4.79E-01 4.58E-01 
100 years 2.93E+01 1.47E+01 7.34E+00 3.67E+00 1.84E+00 5.76E-02 5.66E-05 5.38E-14 4.94E-29 
2.14~106 years 1.28E-01 1.43E-01 1.55E-01 1.95E-01 2.21E-01 2.27E-01 2.26E-01 9.09E-02 1.11E-01 
2.117 days 2.23E-03 1.36E-03 8.35E-04 7.15E-05 5.24E-07 2.06E-13 4.36E-24 5.96E-03 3.65E-03 
1.2~104 years 5.29E+00 5.24E+00 5.19E+00 5.14E+00 5.09E+00 4.86E+00 4.42E+00 3.34E+00 2.08E+00 
3.28~104 years 3.05E-05 3.60E-05 4.15E-05 6.91E-05 1.24E-04 2.89E-04 5.59E-04 1.95E-05 2.50E-05 
27.0 days 1.28E-01 1.43E-01 1.55E-01 1.95E-01 2.21E-01 2.27E-01 2.26E-01 9.09E-02 1.11E-01 
6.69 hours 8.21E-05 8.21E-05 8.21E-05 8.21E-05 8.21E-05 8.21E-05 8.21E-05 8.21E-05 8.21E-05 
1.17 minutes 6.32E-02 6.32E-02 6.32E-02 6.32E-02 6.32E-02 6.32E-02 6.32E-02 6.32E-02 6.32E-02 
10.64 hours 2.43E-04 9.01E-05 3.35E-05 3.98E-07 1.69E-07 1.77E-07 1.92E-07 
6.5~106 ye ars 2.65E-02 2.65E-02 2.65E-02 2.65E-02 2.65E-02 2.65E-02 2.65E-02 2.65E-02 2.65E-02 
17.7 years 
5.53 years 
2.6234 years 
0.298 ms 
0.145 seconds 
17.28 minutes 
2.87 years 
87.7 years 
2.410~104 5.33E+01 5.32E+01 5.31E+01 5.29E+01 5.28E+01 5.21E+01 5.08E+01 4.69E+01 4.10E+01 
6.56 1.12E+ ~103 1.13E+02 02 1.11E+02 1.10E+02 1.09E+02 1.03E+02 9.28E+01 6.76E+01 3.98E+01 
14.4 years 
3.75~105 years 
3.66 days 
207 days 
29.9 seconds 
129I 
85Kr 
93Mo 
93mNb 
94Nb 
59Ni 
63Ni 
237Np 
238Np 
239Np 
231Pa 
233Pa 
234Pa 
.,  
.,  
, , e. 
e. 
.,  
 
. 
,  
.,  
., , SF 
,  
.,  
.,  
., ,  
.,  
. 
, ,  
., , Ã 
.,  
 
,  
.,  
, , SF 
, , SF 
, , SF 
, , SF 
, .,  
, , SF 
,  
., , Ã+ 
.,  
7.42E-03 7.42E-03 7.42E-03 7.42E-03 7.42E-03 7.42E-03 7.42E-03 
1.43E-06 2.23E-09 3.46E-12 3.15E-26 0.00E+00 0.00E+00 0.00E+00 
234mPa 
212Pb 
107Pd 
145Pm 
146Pm 
147Pm 
212Po 
216Po 
144Pr 
236Pu 
238Pu 
239Pu 
240Pu 
241Pu 
242Pu 
224Ra 
102Rh 
106Rh 
1.77E-03 6.55E-04 
5.20E-06 1.03E-07 2.06E-09 4.10E-11 8.18E-13 2.56E-21 0.00E+00 0.00E+00 0.00E+00 
1.16E-18 4.16E-24 1.49E-29 0.00E+00 0.00E+00 0.00E+00 0.00E+00 8.95E-08 3.22E-13 
0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.33E-10 4.45E-22 
1.56E-04 5.77E-05 2.14E-05 2.55E-07 1.08E-07 1.13E-07 1.23E-07 1.13E-03 4.20E-04 
2.43E-04 9.01E-05 3.35E-05 3.98E-07 1.69E-07 1.77E-07 1.92E-07 1.77E-03 6.55E-04 
0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
1.64E-07 1.64E-07 1.64E-07 1.63E-07 1.62E-07 1.59E-07 1.55E-07 1.64E-07 1.64E-07 
2.66E+02 1.21E+02 5.53E+01 2.52E+01 1.16E+01 2.49E-01 1.88E-04 1.01E-10 2.13E-21 
5.41E+01 4.71E-01 4.28E-02 3.91E-02 3.88E-02 3.72E-02 3.43E-02 2.69E-02 1.79E-02 
5.08E-01 5.08E-01 5.08E-01 5.07E-01 5.06E-01 5.04E-01 4.99E-01 5.08E-01 5.08E-01 
2.43E-04 9.01E-05 3.35E-05 3.98E-07 1.69E-07 1.77E-07 1.92E-07 1.77E-03 6.55E-04 
0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.72E-14 1.13E-24 
0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
April 2004 IV-7 of IV-10 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table IV-2. Yucca Mountain Project Radionuclide Inventory for Boiling-Water Reactor Fuels (Continued) 
Radioactivity Inventory [Ci/Assembly] Nuclide 
ID 
Decay 
Mode(a) Half-Life(b) 10,025 yr 5,025 yr 2,025 yr 1,025 yr 525 yr 425 yr 325 yr 225 yr 125 yr 
220Rn 
106Ru 
,  
. 
.,  
.,  
, e. 
. 
.,  
. 
., , e. 
.,  
. 
.,  
, e. 
,  
,  
.,  
.,  
.,  
, , SF 
, , SF 
, , SF 
, , SF 
, , SF 
2.43E-04 9.01E-05 3.35E-05 3.98E-07 1.69E-07 1.77E-07 1.92E-07 1.77E-03 6.55E-04 55.6 seconds 
0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.02 years 
125Sb 
126Sb 
0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.71E-11 2.54E-22 2.758 years 
1.77E-02 1.77E-02 1.77E-02 1.76E-02 1.75E-02 1.71E-02 1.66E-02 1.77E-02 1.77E-02 12.4 days 
126mSb 11 seconds 1.27E-01 1.27E-01 1.26E-01 1.26E-01 1.26E-01 1.26E-01 1.25E-01 1.22E-01 1.18E-01 
6.5~104 ye ars 1.59E-02 1.59E-02 1.59E-02 1.59E-02 1.59E-02 1.59E-02 1.58E-02 1.57E-02 1.56E-02 79Se 
151Sm 2.50E+01 1.16E+01 5.35E+00 2.48E+00 1.15E+00 2.43E-02 1.10E-05 1.01E-15 2.28E-32 90 years 
121Sn 1.06E-02 3.00E-03 8.52E-04 1.56E-06 5.27E-12 2.01E-28 0.00E+00 1.32E-01 3.74E-02 1.128 days 
121mSn 55 years 1.70E-01 4.81E-02 1.37E-02 3.87E-03 1.10E-03 2.02E-06 6.78E-12 2.59E-28 0.00E+00 
1.0~105 ye ars 1.27E-01 1.27E-01 1.26E-01 1.26E-01 1.26E-01 1.26E-01 1.25E-01 1.22E-01 1.18E-01 126Sn 
90Sr 29.1 years 
2.13~105 years 
8.13E+02 6.92E+01 5.90E+00 5.03E-01 4.28E-02 1.92E-07 3.88E-18 0.00E+00 0.00E+00 
99 3.20E+00 3.20E+00 3.20E+00 3.20E+00 3.20E+00 3.19E+00 3.18E+00 3.15E+00 3.10E+00 Tc 
125mTe 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6.61E-12 6.21E-23 58 days 
228 1.913 years 
7.54~104 years 
2.43E-04 9.01E-05 3.35E-05 3.98E-07 1.69E-07 1.77E-07 1.92E-07 1.77E-03 6.55E-04 Th 
230 1.10E-03 1.51E-03 1.93E-03 4.02E-03 8.17E-03 2.03E-02 3.97E-02 3.49E-04 7.10E-04 Th 
231 2.64E-03 2.64E-03 2.65E-03 2.68E-03 2.73E-03 2.87E-03 3.09E-03 2.63E-03 2.63E-03 1.063 days Th 
234 6.32E-02 6.32E-02 6.32E-02 6.32E-02 6.32E-02 6.32E-02 6.32E-02 6.32E-02 6.32E-02 24.10 days Th 
208Tl 
232U 
233U 
234U 
235U 
236U 
8.73E-05 3.24E-05 1.20E-05 1.43E-07 6.06E-08 6.34E-08 6.88E-08 6.35E-04 2.35E-04 3.053 minutes 
70 years 1.72E-03 6.37E-04 2.36E-04 8.76E-05 3.25E-05 3.88E-07 1.62E-07 1.59E-07 1.55E-07 
1.592~105 ye ars 4.53E-05 8.94E-05 1.42E-04 2.01E-04 2.66E-04 6.51E-04 1.57E-03 4.47E-03 9.26E-03 
2.46~105 years 
7.04~108 years 
4.39E-01 4.50E-01 4.55E-01 4.58E-01 4.57E-01 4.54E-01 4.48E-01 3.64E-01 4.16E-01 
2.63E-03 2.63E-03 2.64E-03 2.64E-03 2.65E-03 2.68E-03 2.73E-03 2.87E-03 3.09E-03 
2.342~107 ye ars 6.30E-02 6.33E-02 6.36E-02 6.39E-02 6.43E-02 6.58E-02 6.87E-02 7.58E-02 8.35E-02 
Total (c) = 4.72E+03 1.24E+03 7.71E+02 6.32E+02 5.49E+02 3.31E+02 1.91E+02 1.28E+02 9.15E+01 
Source: CRWMS M&O 1999 [DIRS 136429] Attachment X 
NOTES: a Decay modes identified from Parrington et al. (1996 [DIRS 103896]). 
b Half-lives identified from Parrington et al. (1996 [DIRS 103896]). 
c Total radionuclide activities are calculated here and are not part of original reference. 
Decay times listed in Table IV-2 include the 25-year decay time of the fuels prior to its emplacement into the YMP repository. Thus, fuel with a date of 525 
years corresponds to fuel that has been stored for 500 years (calendar year 2535) in the repository. 
April 2004 IV-8 of IV-10 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste-Form FEPs 
Table IV-3. Alpha Recoil Enhanced (from Both  and -Recoil Atom) Dissolution Rates Due to the Major 
Mass-Based Constituents of SNF and HLW to be Disposed in the Yucca Mountain 
Repository 
Nuclide 
ID Decay Mode 
238U , , SF 
235U , , SF 
, , SF 239Pu 
236U , , SF 
NOTES: a Half-life values in Sanchez et al. (1998 [DIRS 149368]). 
b The fraction decay rate, also known as the decay constant, is given by =ln(2)/, where  
is the radionuclide half-life given by values in column 3. 
c Each mono-layer thickness is 3.0 A (3.0x10-10m), and the density is upper bounded at 19.86 g/cm3 
(theoretical density of pure plutonium metal) (Wick 1980 [DIRS 143651]). 
ANL-WIS-MD-000009 REV 01 
Half.Life (a) Fraction Decay 
Rate (b) 
(1/yr) (yr) 
1.55 x 10 .10 4.47 x 109 
9.85 x 10-10 3.84 x 10-11 7.04 x 108 
2.88 x 10-5 1.12 x 10-6 2.410 x 104 
2.342 x 107 2.96 x 10-8 1.15 x 10-9 
IV-9 of IV-10 
.Decay Rate in 
13 Mono-Layers(c) 
(g/m2.yr) 
6.05 x 10-12 
April 2004 Miscellaneous Waste-Form FEPs 
INTENTIONALLY LEFT BLANK 
April 2004 IV-10 of IV-10 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
ATTACHMENT V - DATA QUALIFICATION FOR DIRECT INPUTS TO 
MISCELLANEOUS WASTE-FORM FEPS 
This attachment is related to the miscellaneous waste-form FEPs as discussed in Section 6.2. 
This attachment provides the results of data qualification activities for previously unqualified 
project data and documents per the data qualification tasks as required by AP-SIII.2Q, 
Qualification of Unqualified Data, and for data obtained from outside sources that are not 
established fact per the requirements found in AP-SIII.9Q, Scientific Analyses. It is not 
intended as stand-alone documentation separate from the main document. The data qualified 
herein is intended for use only for FEPs screening and, more specifically, for use within this 
work product. The data qualification activity was detailed in the qualification plan 
(Attachment VI). 
The miscellaneous waste-form FEPs analyses require the use of input, cited from journal papers, 
compendiums, proceedings, internet citations, and other sources, to represent the nature, 
magnitude, and potential consequence of the miscellaneous waste-form FEPs. The use and 
classification of such input is subject to classification per Attachment 3 of AP-3.15Q, Managing 
Technical Product Inputs, because the information satisfies the definition of direct input. 
Project-generated information directly used in the analysis is identified as data because it is the 
result of activities such as sample collection, physical measurements, testing, and analysis, both 
in the field and in the laboratory, that are not site-specific and do not meet the definition of 
Established Fact. Because this data does not initially satisfy the definition in Attachment 3 as 
qualified, it is thus subject to the requirements of AP-SIII.2Q, which allows data qualification 
to be conducted within the work product, so long as 1) the qualification is providing a desired 
level of confidence that the data are suitable for their intended use and 2) the intended use the 
data is only for that work product. For data obtained from outside sources that are not 
established facts, the requirements identified in AP-SIII.9Q are used to qualify the data for use 
within the technical product (i.e., this analysis report). Given these conditions, then no action is 
required under AP-3.15Q, Managing Technical Product Inputs. The direct input qualified here, 
along with the direct input from YMP-controlled sources, is listed in Section 4 per requirements 
of AP-SIII.9Q, Scientific Analyses. 
The information sources in this attachment are to be qualified in accordance with Attachment 3 
of AP-SIII.2Q, Qualification of Unqualified Data, which states: 
The Data Qualification Report includes, as applicable, a discussion of the following items: 
1. The data set(s) for qualification 
2. The method(s) of qualification selected and rationale 
3. Evaluation criteria 
4. An evaluation of the technical correctness of the data, as applicable 
5. Data generated by the evaluation, if applicable 
April 2004 V-1 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
6. The evaluation results 
7. A conclusion for/against changing the qualification status of the data based on the 
teams judgments in response to the evaluation criteria and the evaluation results. 
Refer to the Technical Work Plan or Attachment 2, Data Qualification Plan, as 
appropriate. 
8. The rationale for abandoning any of the qualification methods, if appropriate 
9. A discussion of any limits or caveats that should be considered by potential users of 
the data 
10. Identification of any supporting information used in the qualification effort by the 
appropriate reference identifier (Data Tracking Number [DTN], accession number, 
Technical Information Center catalog number, etc.) 
11. Reference to the Technical Work Plan or Attachment 2, Data Qualification Plan. 
Deviations to the plan should be documented and justified in the report. 
Additionally, the non-project sources of information in this report are to be qualified in 
accordance with AP-SIII.9Q, Scientific Analyses, (Rev. 1/ICN 4, Section 5.2.1(l)) which uses 
the following factors in qualifying data as suitable for its intended use: 
 Reliability of data source. 
 Qualification of personnel or organizations. 
 Extent to which the data demonstrate the properties of interest. 
 Prior uses of the data. 
 Availability of corroborating data. 
This attachment has been organized in accordance with the above list of data qualification items. 
Each item listed above is given an individual section. Accordingly, Section 4 provides the 
evaluation of the data, and contains the discussion wherein the direct inputs are corroborated. 
V.1 DATA SOURCES FOR QUALIFICATION 
The direct inputs requiring qualification for use in this analysis report were identified in the Data 
Qualification Plan (Attachment VI) and are shown in Table V-1. Subsequent to the development 
of the initial plan, FEPs Screening of Processes and Issues in Drip Shield and Waste Package 
Degradation (BSC 2004 [DIRS 162200]) was identified as a more appropriate direct input 
source to Section 6.2.37 than the reference previously used (CRWMS M&O 1997 [DIRS 
100264]). Accordingly, there was no need to qualify Waste Package Design Basis Events 
(CRWMS M&O 1997 [DIRS 100264]) in this data qualification report. 
Each of the remaining three sources to be qualified is identified by the section used in the report 
and FEP number, and will be treated separately within Section V.4. Table V-2 also addresses the 
corroborating information used in the FEPs screening arguments (for excluded FEPs) and TSPA 
disposition descriptions (for included FEPs) in Section 6.2 of this document (currently, all three 
of the direct inputs to be qualified are used in screening arguments for excluded FEPs). 
April 2004 V-2 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
The source column in Table V-1 provides the citation as it appears in the DIRS database and 
provides traceability through the TIC number or DIRS numbers as required by AP-SIII.2Q, 
Qualification of Unqualified Data. The description column in the Table V-1 provides a brief 
description of the data being qualified. This information is also provided in Table 4-1. 
Item Source Description of Input Governing Where Used 
Procedure 
1 
2 
3 
4 
Table V-1. Data Sources for Qualification within This Work Product 
Sunder, S.; Shoesmith, D.W.; and Miller, N.H. 
1997. "Oxidation and Dissolution of Nuclear 
Fuel (UO2) by the Products of the Alpha 
Radiolysis of Water." Journal of Nuclear 
Materials, 244, 66-74. Amsterdam, The 
Netherlands: Elsevier. TIC: 246914 
[DIRS143860] 
Thornton, T.A. 1998. "HPPP Issue 1; 
Preliminary TSPA for a Pyrophoric Event 
Involving N-Reactor SNF Waste Packages." 
Interoffice correspondence from T.A. Thornton 
(CRWMS M&O) to J.S. Clouet, S.S. Sareen, 
and D. Stahl, September 21, 1998, 
LV.WP.TAT.09/98-179, with attachment. 
MOL.19981019.0001 [DIRS 125082] 
Rechard, R.P., ed. 1995. Methodology and 
Results. Volume 2 of Performance 
Assessment of the Direct Disposal in 
Unsaturated Tuff of Spent Nuclear Fuel and 
High-Level Waste Owned by U.S. Department 
of Energy. SAND94-2563/2. Albuquerque, 
New Mexico: Sandia National Laboratories. 
TIC: 237102 [DIRS 101084] 
CRWMS M&O 1997. Waste Package Design 
Basis Events. BBA000000-01717-0200-00037 
REV 00. Las Vegas, Nevada: CRWMS M&O. 
ACC: MOL.19971006.0075 [DIRS 100264] 
AP-SIII.9Q 
Rev. 1/ICN 4 
Section 
6.2.38; FEP 
2.1.13.01.0A 
Being Qualified 
The effects of radiolysis on 
spent nuclear fuel 
corrosion (dissolution) will 
be transitory and will 
become minor as alpha 
dose rates decrease. 
AP-SIII.2Q 
Rev 1/ICN 2 
Section 
6.2.13; FEP 
2.1.02.08.0A 
Uranium metal-based 
spent nuclear fuel, 
particularly the N-Reactor 
spent nuclear fuel has the 
potential for pyrophoric 
behavior. 
Section 
6.2.16; FEP 
2.1.02.25.0A 
Significant amounts of AP-SIII.9Q 
DSNF cladding are likely to Rev. 1/ICN 4 
be damaged. 
N/A Will not be 
qualified; 
superceded 
by BSC 2004 
[DIRS 
162200] 
Maximum allowable waste 
package interior pressures 
due to the release of 
radioactive gases to the 
interior of the waste 
package will be too low to 
compromise waste 
package integrity. 
The information used for direct input is also previously identified in Section 4, and supporting 
references are clearly identified within Section 6.2, as needed, to provide the technical basis for 
exclusion of the FEP. 
V.2 THE METHOD OF QUALIFICATION SELECTED AND RATIONALE 
The data to be qualified have been extracted from two non-project sources (items 1 and 3) and 
one product source (item 2). The product source will be qualified by the corroborating data 
approach outlined in Attachment 3 of AP-SIII.2Q, Qualification of Unqualified Data. The data 
will be qualified for use within this document per Section 5.2.1(l) of AP-SIII.9Q. Per the cited 
procedure, the corroborating data approach may be used when subject matter data comparisons 
can be shown to substantiate or confirm parameter values and may include comparisons of 
unqualified to unqualified data. The use of the corroborative data approach seems most feasible 
ANL-WIS-MD-000009 REV 01 V-3 of V-22 April 2004 Miscellaneous Waste Form FEPs: Data Qualification 
for judging correctness and reliability by comparing independently developed, but related, data 
sets. 
Other possible qualification methods listed below were reviewed for applicability and will not be 
used for the following reasons: 
 Equivalent QA ProgramNo QA documentation is available for review. Because of the 
diverse nature and sources of the information used, and because much of the data results 
from primary research activities, it is unlikely that a program equivalency can be 
established or that any available records, practices, or procedures would satisfy YMP 
QA requirements. 
 Confirmatory TestingThis testing is not viable because the data relationships and 
presentations are not amenable to field or laboratory confirmation due to scale, testing 
resources, and schedule. 
 Peer ReviewThis is not justified because the analysis uses multiple approaches to the 
problem and there are no critical interpretations of the data. 
 Technical AssessmentTechnical assessment is not applicable. Most of the cited 
sources come from peer reviewed journals, or from edited and refereed compendiums of 
work and, as a result, are assumed a priori to be suitable and reliable with regard to 
representing the reported work, and there is some attribution of a level of confidence for 
information from these types of sources. However, because the methods used to acquire 
the data are only summarily treated in these sources, the documentation or proof of 
proper data acquisition is unavailable for review. 
The procedure allows comparisons of unqualified to unqualified data. In this attachment and 
Table 6-1, the indirect input information is also referred to as corroborative/supportive 
information to help differentiate it from information being qualified, which is identified as direct 
input. In some cases, a single source may provide direct and indirect input to the individual FEP 
analyses in Section 6.2. 
V.3 EVALUATION CRITERIA 
The following criteria are established for qualifying these data through corroboration. 
1. Is there a sufficient quantity of corroborating data available for comparison? 
Table V-1 is organized by information type to be qualified and lists each of the sources 
of data to be qualified. For each subject area, at least two independent sources of 
information were considered for corroboration. 
2. Can inferences drawn to corroborate these data be clearly identified, justified, and 
documented? 
For each source of information to be qualified, the discussion includes a brief 
statement regarding the original purpose of the study, the method used to acquire the 
April 2004 V-4 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
data, and any limitations germane to the corroborative qualification of the data. 
Additionally, the basis for assuming adequacy for comparison (e.g. similar type study, 
update to previous study, compared to previous studies in related fields) will be stated. 
For quantitative inputs, corroboration is shown either by graphical representation of the various 
data sets, or in table or text format, comparing the various values from the various sources. 
For qualitative inputs addressing key concepts of a feature, event, or process, one or more 
corroborative information sources will be used to substantiate the direct input. The source(s) 
should not conflict with the direct input, and should be in general agreement. This standard may 
also be used when corroborating boundary conditions that define the conditions necessary for the 
initiation of a feature, event, or process. For cases with only one available source of information, 
its appropriateness for use as direct input will be discussed and used only upon mutual agreement 
of the data qualification team members. 
The qualification team consisted of two members: Thomas Thornton, a nuclear fuel expert and 
the analysis originator (the team chair) and Norman Graves, who is technically competent in 
radioactive waste management. As the analysis originator, Thornton is not considered to be 
independent of the information to be qualified (AP-SIII.2Q, Section 5.1.2 (b)(3). However, the 
procedure requires that the analysis report originator chair the team when the qualification is 
performed within the Analysis or Model Report (AP-SIII.2Q, Section 5.1.1). Norman Graves 
did not participate in the acquisition or development of the information, and was, therefore, an 
independent reviewer of the analysis. The data qualification team considered that the intended 
use of the data is for FEP screening arguments (for excluded FEPs) and TSPA Disposition 
discussions (for included FEPs). Therefore, the use of broad acceptance criteria is justified, 
and in lieu of corroboration, a bounding or conservative value (with respect to inclusion of the 
FEP) may be recommended and considered as qualified under this exercise. Consistent with the 
intended use, the data qualification team may exercise some latitude in applying these criteria if 
an adequate explanation or justification for variance from the above criteria can be provided. 
For each FEP-specific data set within Section V.2, the evaluation criteria to be applied 
(i.e., quantitative or qualitative) will be identified. 
V.4 EVALUATION OF THE TECHNICAL CORRECTNESS OF THE DATA 
The technical correctness of the data/information used as direct input (and the corroborating 
information) was evaluated based on the data attributes listed in Attachment 4 of AP-SIII.2Q, 
Qualification of Unqualified Data. Since the information being qualified is qualitative rather 
than quantitative, planning, collecting, and analyzing the data consisted of identifying the 
literature sources and assessing the credibility of the corroborating source via the accepted 
scientific/engineering practice of using the open literature in combination with the team 
members technical expertise as shown in sections V.4.x.3. The attributes considered during 
qualification included: 
3. The extent to which the data demonstrate the properties of interest (e.g., 
physical, chemical, geologic, mechanical) 
7. Prior uses of the data and associated verification processes 
April 2004 V-5 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
8. Prior peer or other professional reviews of the data and their results 
9. Extent and reliability of the documentation associated with the data 
10. The extent and quality of corroborating data or confirmatory testing results 
12. The importance of the data to showing that the proposed U.S. Department of 
Energy repository design meets the performance objectives of 10 CFR 63. 
In the data qualification plan, item 2 (technical adequacy of equipment and procedures) was 
listed as a potentially applicable attribute to be evaluated. However, the technical adequacy of 
equipment and procedures used to collect and analyze the data was not evaluated due to a general 
lack of such information in the cited sources. 
In the data qualification plan, item 8 (prior peer review) was listed as a potentially applicable 
attribute to be evaluated. However, the search of the available literature did not disclose any 
prior peer reviews related to the references to be qualified. 
Section V.4 has been subdivided with direct input for each subject FEP being accorded an 
individual subsection. The discussion of technical correctness for each FEP-specific data set is 
addressed in four parts. 
The first part (Section V.4.x.1) discusses the scope of the literature review. The literature review 
involved a search of the RIS and the DIRS for the extent of usage in the program, and the 
number of citations is given in Table V-2 as relevant use by others. 
The second part (Section V.4.x.2) addresses the technical correctness of both the data being 
qualified, and the corroborating information. Planning, collecting, and analyzing the data 
consisted of identifying the literature sources and assessing the credibility of the corroborating 
source via the accepted scientific/engineering practice of using the open literature in combination 
with the team members technical expertise. The technical correctness of the data and 
corroborating information was evaluated based on the attributes listed in Table V-2. This 
evaluation is presented in table format with attributes 3 through 9, 10, and 12 discussed for the 
direct inputs and corroborative sources (indirect inputs). In the summary tables, direct input 
citations are listed in alphabetical order, followed by the corroborative citations. A single cited 
paper might serve as the source for multiple types of direct input or reference-only information. 
By addressing items 2 through 9 for the corroborative information, item 10 is also addressed. 
However, because the comparison of the direct input to the corroborative information requires 
more detail, it is specifically addressed in the third part (Section V.4.x.3) for each FEP-specific 
data set. The discussions for each FEP may be subdivided by topic to facilitate corroboration of 
the direct inputs. 
The fourth part of each section (Section V.4.x.4) provides a recommendation regarding status of 
the direct input and any associated limitations. 
April 2004 V-6 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
V.4.1 Data Qualification for Sunder et al. 1997 [DIRS 143860] 
The effects of alpha radiolysis could include the production of gas from the waste form and 
changing the chemistry of the contacting water solutions. The relationship being qualified is 
stated as: 
The effects of radiolysis on spent nuclear fuel corrosion (dissolution) will be transitory 
and will become minor as dose rates decrease. 
V.4.1.1 Literature Search 
A literature search was performed using the RIS and DIRS databases. Citations from these 
sources not selected for evaluation were not used if they did not involve waste forms. A 
pertinent corroborating reference (Gray 1988 [DIRS 168471]) was found that had not been 
previously used in the main body of this Analysis Report). 
V.4.1.2 Evaluation of Attributes 
For each of the sources to be used in the corroboration of the direct input (Sunder et al. 1997 
[DIRS 143860]), the pertinent attributes are summarized in tabular form in Table V-2. These 
attributes were selected based on a comparison of the pertinent information in the corroborating 
reference with the reference to be qualified for direct input to the FEPs screening or disposition 
discussions. 
V.4.1.3 Discussion 
The information being qualified is as follows: 
Support for the conclusion that the effects of radiolysis on spent nuclear fuel corrosion 
(dissolution) will be transitory and will become minor as dose rates decrease. 
This is the conclusion obtained from the reference cited as direct input (Sunder et al. 1997 
[DIRS 143860]). The other reports cited in this discussion are corroborative and support the 
conclusions of the direct use reference. 
Analyses by Sunder and Shoesmith (1991 [DIRS 143815]) and Shoesmith and Sunder (1992 
[DIRS 113368]) qualitatively corroborate the information being qualified by showing the waste 
container will not fail such that water will be shielded from the alpha particles emitted from the 
fuel. Water will not intrude until gamma and beta emitters (and consequent nitric acid 
production) have decayed to very low concentrations. In addition, strong gamma and beta fields 
associated with the used fuel will decrease by a factor of about 1000 in the first few hundred 
years after disposal. 
In-Package Chemistry for Waste Forms (BSC 2003 [DIRS 168795], Attachment II) qualitatively 
corroborates the information being qualified by showing that the degradation of the Zircaloy fuel 
cladding will not be significantly enhanced by the products of radiolysis. The study also notes: 
April 2004 V-7 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
the corrosion rate of Zircaloy in water is limited by the slowest of the basic 
corrosion steps: dissociation of water into oxygen and hydrogen ions, diffusion of 
oxygen ions through the oxide film, oxidation of the Zr metal, diffusion of 
electrons through the oxide, and hydrogen-ion reduction by the electrons at the 
water-to-oxide interfaceabove about 230C, diffusion of oxygen ions through 
the oxide film becomes rate limiting. 
If cladding corrosion is controlled by this step, the production of peroxide by radiolysis would 
have no effect on Zircaloy corrosion even if the alpha dose is higher than expected. 
Gray (1988 [DIRS 168471], Summary and Sections 2.0 and 5.1) qualitatively corroborates the 
information being qualified by providing laboratory test results for leaching of unirradiated UO2 
and spent nuclear fuel. The laboratory experiments indicated that there was little effect due to 
alpha radiolysis on the leaching behavior of spent nuclear fuel (i.e., with alpha radiation) and 
UO2 (without alpha radiation) samples. Most of the difference between the leaching results of 
the UO2 and spent nuclear fuel was attributed to surface area difference between them rather than 
alpha radiolysis per se. 
Gamma and Neutron Radiolysis in the 21-PWR Waste Package from Ten to One Million Years 
(BSC 2002 [DIRS 161912], Sections 6.1, and 6.2) provides justification for screening out 
neutron and gamma radiolysis by showing nitric acid production to be small at 1,000 years; 
therefore, it qualitatively corroborates the information being qualified. Gamma and neutron 
doses decrease respectively by approximately four and two orders of magnitude over the 
10,000-year regulatory period. Gamma and Neutron Radiolysis in the 21-PWR Waste Package 
from Ten to One Million Years (BSC 2002 [DIRS 161912], Section 6.3) further showed that 
nitric acid production is dominated by gamma production and is insensitive to neutrons. Since 
the fuel cladding is largely intact, it will isolate water from direct contact with the alpha particles 
emitted from decay of actinides in the fuel. This provides confidence that radiolysis can be 
screened out as significantly affecting fuel degradation. 
V.4.1.4 Data Status and Limitations 
The conclusion by Sunder et al. (1997 [DIRS 143860]) that the effect of radiolysis on spent 
nuclear fuel corrosion (dissolution) will be transitory and will become minor as dose rates 
decrease is qualified for its intended use in this product. Sunder et al. (1997 [DIRS 143860]) as 
an outside-the-project source may be regarded as qualified (per AP-SIII.9Q) for use in this 
analysis report because it is a report in Journal of Nuclear Materials, a highly reputable and 
peer-reviewed technical publication (reliability of data source, and the availabilty of the 
corroborating data). The discussion in Section V.4.1.3 qualitatively substantiated this 
conclusion. There is no limit on the data use in this anlysis report. 
April 2004 V-8 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
V.4.2 Data Qualification for Thornton 1998 [DIRS 125082] 
V.4.2.1 Literature Search 
The literature search identified an additional pertinent corroborating reference that was not 
included in the main body of the analysis report for the qualification of Thornton (1998 
[DIRS 125082] in Swanson et al. (1985 [DIRS 168468]). 
V.4.2.2 Evaluation of Attributes 
Table V-2 summarizes pertinent attributes for each of the sources used to corroborate the direct 
input (Thornton 1998 [DIRS 125082]). These attributes were selected based on comparing the 
pertinent corroborating reference information in the corroborating reference to the reference to 
be qualified for direct input to the FEPs screening or disposition discussions. 
V.4.2.3 Discussion 
The information being qualified is that: 
Uranium metal-based spent nuclear fuel, particularly the N-Reactor spent nuclear fuel, 
has the potential for pyrophoric behavior. 
This is the conclusion obtained from the reference cited as direct input (Thornton 1998 
[DIRS 125082]). The other reports cited in this discussion/analysis are corroborative, and 
support he conclusions of the direct use reference. 
Schulz (1972 [DIRS 159406]) qualitatively corroborates the information being qualified by 
analyzing several unanticipated pyrophoric events that occurred during N-Reactor fuel 
reprocessing at the West Valley reprocessing facility (circa 1967-1968). These uranium metal 
fires occurred primarily as a result of the aggressive mechanical handling of the fuel elements 
during discharge from the N-Reactor (at Hanford), subsequent transport to West Valley (New 
York) in water-filled casks, and loading into the reprocessing dissolver vessels. The fuel often 
sustained damage sufficient to breach the Zircaloy cladding because reactor discharge involved 
pushing the elements from an approximately 15-foot drop into a collector basket. The 
consequent exposure of the chemically active uranium metal that comprised the core of the fuel 
elements made it susceptible to further corrosion and damage. When loaded and processed in the 
dissolvers, the resulting material became susceptible to ignition and burning. Nineteen dissolver 
pyrophoric events (fires) occurred between 1967 and 1968. Fourteen fires occurred inside the 
dissolvers, four in the bins into which the Zircoloy hulls were dumped after the dissolution 
operation, and one on the canyon deck after a hull fell on it. During these reactor discharge, 
transport, and reprocessing operations, the N-Reactor fuel element metallic uranium cores were 
exposed to air and water, thus exposing the chemically active uranium metal surfaces to a strong 
oxidizing environment. It was surmised that sparks resulting from element-to-element contact 
provided the ignition source and that the materials that initially ignited were weld beads 
sensitized by exposure to the highly acidic environment in the dissolvers. 
Swanson et al. (1985 [DIRS 168468]) qualitatively corroborates the information being qualified 
by providing an independent analysis and alternative explanation for the N-Reactor SNF 
April 2004 V-9 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
pyrophoric events that occurred in the dissolvers at the West Valley reprocessing plant. They 
concluded that the material that initially ignited was the metallic uranium, or more particularly 
the uranium hydride inclusions (resulting from corrosion) embedded in the uranium metal. 
Abrefah et al. (1995 [DIRS 151125]) qualitatively corroborate the information being qualified by 
reporting the results of visual examinations of N-Reactor SNF elements that underwent 
characterization in support of the development of a dry intermediate storage facility at Hanford. 
It was evident that a significant fraction of the N-Reactor spent fuel is currently damaged, and so 
will be damaged at the time of emplacement in the repository. The potential for ignition and 
subsequent burning of the uranium metal would be expected to be proportional to the amount of 
exposed and corroded uranium metal and therefore on the fraction of N-Reactor spent nuclear 
fuel assemblies having breached cladding. Taken together with the Schulz (1972 [DIRS 
159406]) analyses of the N-Reactor fuel ignition at West Valley, the observation of extensive 
damaged cladding indicates that the current condition of the N-Reactor spent nuclear fuel makes 
it potentially pyrophoric. 
Abrefah et al. (1999 [DIRS 151226] and Marschman et al. (1997 [DIRS 149429) qualitatively 
corroborate the information being qualified by identifying uranium hydride inclusions in the 
exposed uranium metal surfaces. Uranium hydride is suspected of being the potential source of 
pyrophoric behavior in uranium metal that has corroded in the air or water environment. 
Thornton (1998 [DIRS 125082]) makes the overall case that pyrophoric behavior in metallic 
uranium and uranium metal-based spent nuclear fuel is possible in an oxidizing environment (air 
or water) and under conditions of mechanical trauma. This environment is not expected while 
the waste packages and canisters remain intact, but could occur after waste package failure. 
Attachment I justifies the conclusion that even in the event a pyrophoric event occurs after waste 
package failure, the effect on the boundary dose will be minor. 
V.4.2.4 Data Status and Limitations 
Neither HPPP Issue 1; Preliminary TSPA for a Pyrophoric Event Involving N-Reactor SNF 
Waste Packages (Thornton 1998 [DIRS 125082]) nor its corroborating information enable a 
direct estimate of the precise conditions under which pyrophoric behavior would occur. They 
also do not provide direct answers to questions such as: At what temperature will a damaged 
N-Reactor fuel assembly ignite, or, How hard an impact from an accidental drop of an open 
canister containing damaged N-Reactor fuel elements can be tolerated? Standard Guide for 
Pyrophoricity/Combustibility Testing in Support of Pyrophoricity Analyses of Metallic Uranium 
Spent Nuclear Fuel (ASTM C 1454-00 [DIRS 152779]) provides further guidance in this area. 
However, for the purposes of this analysis report, such estimates are unnecessary since the 
screening decision to exclude DSNF cladding (Section 6.2.13) and the consequence analysis 
(Attachment I) assume a pyrophoric event will occur when a waste package containing NReactor 
SNF fails. Therefore, HPPP Issue 1; Preliminary TSPA for a Pyrophoric Event 
Involving N-Reactor SNF Waste Packages (Thornton 1998 [DIRS 125082]) should be regarded 
as qualified for its use in Section 6.2.13. 
April 2004 V-10 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
V.4.3 Data Qualification for Rechard 1995 [DIRS 101084] 
V.4.3.1 Literature Search 
The literature search identified an additional pertinent corroborating reference (Lawrence 1999 
[DIRS 168475]) not included in the analysis report for the qualification of Methodology and 
Results (Rechard 1995 [DIRS 101084]). 
V.4.3.2 Evaluation of Attributes 
For each of the sources to be used in the corroboration of the direct input (Rechard 1995 
[DIRS 101084]), the pertinent attributes are summarized in tabular form in Table V-2. These 
attributes were selected based on comparing the pertinent information in the corroborating 
reference with the reference to be qualified for direct input to the FEPs screening or disposition 
discussions. 
V.4.3.3 Discussion 
The information being qualified is: 
Significant amounts of DSNF cladding are likely to be damaged. 
This is the conclusion obtained from the reference cited as direct input (Rechard 1995 
[DIRS 101084]). The other reports cited in this discussion are corroborative and support the 
conclusions of the direct use reference. 
Abrefah et al. (1995 [DIRS 151125]) characterized samples of the N-Reactor spent fuel stored in 
the K-East and K-West basins on the Hanford site. N-Reactor fuel consists of cylinders of 
uranium metal clad in Zircaloy. This work was done in support of the plan to remove the 
N-Reactor SNF from the K-East and K-West pools. This included drying the SNF, putting it in 
sealed and helium-inerted canisters (called MCOs or multi-canister overpacks), and placing them 
in dry interim storage at the Hanford site pending ultimate disposal at Yucca Mountain. This 
characterization revealed a significant (but not fully quantified) fraction of the cladding had been 
damaged. In addition, reaction of the exposed uranium with the water in the K-basins had 
resulted in the formation of uranium hydride inclusions in the exposed uranium. As discussed in 
Section 6.2.13 and Attachment I, these uranium hydride inclusions are considered conditions 
supporting pyrophoric behavior in uranium and uranium metal-based SNF. 
DSNF and Other Waste Form Degradation Abstraction (BSC 2003 [DIRS 163693]) discusses 
the grouping of the many types DSNF (see Section 6.2.17). The report notes that few groups 
have had their cladding characterized enough to take credit for intact cladding in their 
performance analyses. Additionally, it notes that the N-Reactor SNF constitutes over 80 percent 
by weight of the total DSNF. The report suggests that no credit should be taken for DSNF 
cladding in the TSPA because the status of the majority of the DSNF groups is unknown and a 
significant fraction of the N-Reactor SNF has damaged or perforated cladding. 
Lawrence (1999 [DIRS 168475], Section 5.1 and Table 1) reports on the results of an effort to 
estimate the fraction of perforated N-Reactor SNF cladding. The perforation types were 
April 2004 V-11 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
described as breached (minor cladding rupture), defected (significant uranium metal exposed), or 
bad (gross failure). The identification of perforated elements and their categorization was done 
by means of either a visual or a video examination of 200 fuel elements from each of the K-East 
and K-West basins. The examination showed 65 (32.5 percent18 breached, 42 defected, and 5 
bad) K-East basin elements and 66 (33 percent53 breached, 0 defected, and 13 bad) K-West 
basin elements were perforated. 
Each of these references shows that the degree of cladding damage to the N-Reactor SNF is 
significant. Additionally, there is an unquantified uncertainty in the estimates of cladding failure 
due to the relatively inexact nature of the visual observations. The information in these 
references sufficiently resembles the estimates of failed fuel by Rechard (1995 [DIRS 101084]) 
to support the conservative decision to take no credit for radionuclide retardation by the DSNF 
cladding in the TSPA. 
Rechard (1995 [DIRS 101084]) as an outside-the-project source may be regarded as qualified 
(per AP-SIII.9Q) for use in this analysis report because it is generated at Sandia National 
Laboratory, and is thereby a reputable and peer-reviewed technical publication (reliability of data 
source), and the availabilty of the corroborating data in the other sources. 
V.4.3.4 Data Status and Limitations 
The N-Reactor SNF characterization reported by Abrefah et al. (1995 [DIRS 151125]) was 
obtained from fuel elements known to be defected. The results of the optical metallographic 
examinations (e.g., identifying uranium hydride inclusions in the corroded uranium surfaces) 
agreed qualitatively with other available information. However, these results could not be used 
to estimate the fraction of failed fuel elements. In addition, these studies did not provide 
information to predict or to estimate the extent of further cladding damage during interim storage 
at Hanford and transport to the repository. However, this uncertainty further supports the 
decision to take no credit for the DSNF cladding in TSPA. Since information provided by 
Rechard (1995 [DIRS 101084]) was used to justify not taking credit for any DSNF cladding in 
the TSPA, this reference may be regarded as qualified for its use in Section 6.2.16. 
April 2004 V-12 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
Table V-2. Sources and Attributes Evaluation for Direct Inputs 
Extent and 
Reliability of 
Qualification 
Attributes Corroborating 
References Documentation Addressed Relevant Use by Others Direct Input Usage Source to be Qualified Item 
Moderate to 3, 9, 10 ANL-EBS-MD-000037, REV 03 BSC (Bechtel SAIC 
Company) 2002. Gamma 
and Neutron Radiolysis in 
the 21-PWR Waste 
Package from Ten to One 
Million Years. 000-00CDSU0-
00700-000-00A. 
Las Vegas, Nevada: 
Bechtel SAIC Company. 
ACC: 
MOL.20021004.0002 
[DIRS 161912] 
Section 6.3 and Figure 5 
Sunder, S. and Shoesmith, 
D.W. 1991. Chemistry of 
UO2 Fuel Dissolution in 
Relation to the Disposal of 
Used Nuclear Fuel. 
AECL-10395. Pinawa, 
Manitoba, Canada: 
Whiteshell Laboratories, 
Atomic Energy of Canada 
Limited. TIC: 246920 
[DIRS 143815] 
Section 6.2.38 Sunder, S.; Shoesmith, 1 
High  000-30R-EBS0-00100-000-000 (FEP 2.1.13.01.0A) D.W.; and Miller, N.H. 
Support for the 1997. "Oxidation and 
conclusion that the Dissolution of Nuclear Fuel 
effects of alpha (UO2) by the Products of 
the Alpha Radiolysis of radiolysis on spent 
Water." Journal of Nuclear 
Materials, 244, 66-74. 
fuel corrosion 
(dissolution) will be 
transitory and will Amsterdam, The 
become minor as Netherlands: Elsevier. 
alpha dose rates TIC: 246914 [DIRS 
decrease. 143860] 
Moderate 3, 10 ANL-WIS-MD-000009, REV 00 
ICN 01 
April 2004 V-13 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
Table V-2. Sources and Attributes Evaluation for Direct Inputs (Continued) 
Extent and 
Reliability of 
Qualification 
Attributes Corroborating 
References Documentation Addressed Relevant Use by Others Direct Input Usage Source to be Qualified Item 
Moderate 3, 10 ANL-EBS-MD-000015 REV 01 Shoesmith, D.W. and 
Sunder, S. 1992. "The 
Prediction of Nuclear Fuel 
(UO2) Dissolution Rates 
Under Waste Disposal 
Conditions." Journal of 
Nuclear Materials, 190, 20- 
35. Amsterdam, The 
Netherlands: Elsevier. 
TIC: 246431 [DIRS 
113368] 
BSC (Bechtel SAIC 
Company) 2003. Errata 
for In-Package Chemistry 
for Waste Forms. ANLEBS-
MD-000056 REV 00. 
Las Vegas, Nevada: 
Bechtel SAIC Company. 
ACC: 
MOL.20010322.0490; 
DOC.20031014.0005. 
[DIRS 168795] 
ICN 01 
ANL-EBS-MD-000019 
REV 00 CN 01 
3,7,9,10,12 High ANL-EBS-GS-000001, REV 00 
ANL-EBS-PA-000007, REV 00 
CAL-EBS-MD-000020, REV 00 
CAL-EBS-PA-000005, REV 00 
MDL-DSU-MD-000001, REV 00 
April 2004 V-14 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
Table V-2. Sources and Attributes Evaluation for Direct Inputs (Continued) 
Extent and 
Reliability of 
Qualification 
Attributes Corroborating 
References Documentation Addressed Relevant Use by Others Direct Input Usage Source to be Qualified Item 
Gray, W.J. 1988. Effect of 
Surface Oxidation, Alpha 
Radiolysis, and Salt Brine 
Composition on Spent 
Fuel and UO2 Leaching 
Performance. 
PNL/SRP/6689 
HQX.19881021.0025 
Summary and Section 
5.1.[DIRS 168471] 
(This reference is not used 
in the body of the report.) 
Schulz, W.W. 1972. 
Shear-Leach Processing 
of N-Reactor Fuel-- 
Cladding Fires. ARH-2351. 
Richland, Washington: 
Atlantic Richfield Hanford. 
TIC: 243079. [DIRS 
159406] 
(pyrophoric behavior of NReactor 
SNF at West 
Valley) 
Moderate-to- N/A 3, 10 
Low 
Moderate-to- 3, 10 ANL-WIS-MD-000004 REV 02 Section 6.2.13 2 Thornton, T.A. 1998. 
"HPPP Issue 1; 
Preliminary TSPA for a 
Pyrophoric Event Involving 
N-Reactor SNF Waste 
Packages." Interoffice 
correspondence from T.A. 
Thornton (CRWMS M&O) 
to J.S. Clouet, S.S. 
Sareen, and D. Stahl, 
September 21, 1998, 
LV.WP.TAT.09/98-179, 
with attachment. 
MOL.19981019.0001 
[DIRS 125082] 
Low ICN 01 (FEP 2.1.02.08.0A) 
Provides analyses 
and examples of the 
potential for 
pyrophoric behavior 
in uranium metalbased 
spent fuel, 
particularly the NReactor 
spent fuel. 
April 2004 V-15 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
Table V-2. Sources and Attributes Evaluation for Direct Inputs (Continued) 
Extent and 
Reliability of 
Qualification 
Attributes Corroborating 
References Documentation Addressed Relevant Use by Others Direct Input Usage Source to be Qualified Item 
Moderate-to- 9, 10 ANL-WIS-MD-000004 REV 02 Abrefah, J.; Gray, W.J.; 
Ketner, G.L.; Marschman, 
S.C.; Pyecha, T.D.; and 
Thornton, T.A. 1995. KBasin 
Spent Nuclear Fuel 
Characterization Data 
Report. PNL-10778. 
Richland, Washington: 
Pacific Northwest National 
Laboratory. TIC: 243197. 
[DIRS 151125] 
(extensive N-Reactor 
SNFcladding damage) 
Abrefah, J.; Huang, F.H.; 
Gerry, W.M.; Gray, W.J.; 
Marschman, S.C.; and 
Thornton, T.A. 1999. 
Analysis of Ignition Testing 
on K-West Basin Fuel. 
PNNL-11816. Richland, 
Washington: Pacific 
Northwest National 
Laboratory. TIC: 248558. 
[DIRS 151226] 
(hydrides concentrate near 
exposed U metal surfaces) 
High ICN 01 
Moderate-to- 3, 9, 10 ANL-WIS-MD-000004 REV 02 
High ICN 01 
April 2004 V-16 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
Table V-2. Sources and Attributes Evaluation for Direct Inputs (Continued) 
Extent and 
Reliability of 
Qualification 
Attributes Corroborating 
References Documentation Addressed Relevant Use by Others Direct Input Usage Source to be Qualified Item 
Moderate-to- 9, 10 ANL-WIS-MD-000004 REV 02 Marschman, S.C.; Pyecha, 
T.D.; and Abrefah, J. 1997. 
Metallographic 
Examination of Damaged 
N Reactor Spent Nuclear 
Fuel Element 
SFEC5,4378. PNNL- 
11438. Richland, 
Washington: Pacific 
Northwest National 
Laboratory. TIC: 248059. 
[DIRS 149429] 
Swanson, J.L.; Bray, L.A 
Kjarmo, H.E.; Ryan, J.L.; 
Matsuzaki, C.L.; Pitman, 
S.G.; and Haberman, J.H. 
1985. Laboratory Studies 
of Shear/Leach Processing 
of Zircaloy Clad Metallic 
Uranium Reactor Fuel. 
PNL-5708. Richland, 
Washington: Pacific 
Northwest Laboratory. 
ACC:MOL.20040325.0049 
[DIRS 168468] 
(Pyrophoricity in damaged 
N-Reactor SNF) 
(this reference is not cited 
in the main body of this 
report) 
High ICN 01 
Moderate 3, 9, 10 N/A 
April 2004 V-17 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
Table V-2. Sources and Attributes Evaluation for Direct Inputs (Continued) 
Extent and 
Reliability of 
Qualification 
Attributes Corroborating 
References Documentation Addressed Relevant Use by Others Direct Input Usage Source to be Qualified Item 
High 3, 10 ANL-WIS-MD-000004 REV 02 Abrefah, J.; Gray, W.J.; 
Ketner, G.L.; Marschman, 
S.C.; Pyecha, T.D.; and 
Thornton, T.A. 1995. KBasin 
Spent Nuclear Fuel 
Characterization Data 
Report. PNL-10778. 
Richland, Washington: 
Pacific Northwest National 
Laboratory. TIC: 243197. 
[DIRS 151125] 
(extensive N-reactor SNF 
cladding damage) 
Section 6.2.16 Rechard, R.P., ed. 1995. 
Methodology and Results. 
3 
ICN 01 (FEP 2.1.02.25.0A) 
Volume 2 of Performance 
Assessment of the Direct 
Estimate of the 
amount of damaged 
Disposal in Unsaturated 
Tuff of Spent Nuclear Fuel 
cladding in DSNF. 
For approximately 
and High-Level Waste 80% of the DSNF 
Owned by U.S. 
Department of Energy. 
that is N-Reactor 
SNF, the cladding 
will be significantly SAND94-2563/2. 
damaged at the time Albuquerque, New 
of emplacement in Mexico: Sandia National 
their canisters. Up to Laboratories. TIC: 237102 
50 percent may be [DIRS 101084] 
already perforated. 
High 3, 7, 10 ANL-EBS-MD-000033 BSC (Bechtel SAIC 
Company) 2003. DSNF 
and Other Waste Form 
Degradation Abstraction. 
ANL-WIS-MD-000004 
REV 02. Las Vegas, 
Nevada: Bechtel SAIC 
Company. ACC: 
DOC.20030711.0002. 
[DIRS 163693] 
Section 6.4.3 
(uncertainty in the extent 
of N-reactor SNF cladding 
damage) 
ANL-EBS-MD-000037 
ANL-EBS-PA-000008 
April 2004 V-18 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
Table V-2. Sources and Attributes Evaluation for Direct Inputs (Continued) 
Extent and 
Reliability of 
Qualification 
Attributes Corroborating 
References Documentation Addressed Relevant Use by Others Direct Input Usage Source to be Qualified Item 
Lawrence, L.A., 1999. K 
Basin Spent Nuclear Fuel 
Characterization. HNF- 
3534, Rev. 0. Richland, 
Washington: Duke 
Engineering & Services 
Hanford. 
ACC:MOL.20040325.0053 
[DIRS 168475], Section 
5.1 and Table 1) 
(estimate of the fraction of 
failed N-reactor SNF from 
the K-East and K-West 
spent fuel pools) 
Moder N/A 3, 10 ate 
April 2004 V-19 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
V.5 DATA GENERATED BY THE EVALUATION 
No data was generated by this qualification effort. Analyses were performed as shown herein to 
allow corroboration of direct input to indirect inputs. However, the results of the analyses are 
not being qualified and are not to be used directly as data or serve as the sole basis for FEP 
screening decisions. 
V.6 THE EVALUATION RESULTS 
The third section for each FEP data set (Section V.4.x.3) provides a comparison of the data being 
qualified to the corroborative information. Any necessary manipulations or calculations for the 
comparison are provided in that section of the discussion. In some case, this may be further 
subdivided by specific data topic. 
V.7 CONCLUSION FOR/AGAINST CHANGING THE QUALIFICATION STATUS 
This is addressed in Section V.4.x.4 for each FEP-specific data set. The fourth section of the 
discussion for each FEP data set provides a statement of recommendation and a discussion of 
any specific limitations that apply. This is addressed for each FEP-specific data set under the 
heading Data Status and Limitation. 
The direct inputs listed and discussed above have been reviewed against the stated review 
criteria. In some cases, limitations have been applied. However, all the data evaluated satisfied 
the respective criteria and the status should be changed from unqualified to qualified. 
V.8 RATIONALE FOR ABANDONING ANY OF THE QUALIFICATION METHODS 
This section is not applicable. All three of the references are qualified using corroboration. The 
basis for not selecting other qualification methods is previously discussed in Section 1.2. 
V.9 LIMITS OR CAVEATS 
The intended use of the data qualified here is for use in FEP screening. For all data discussed in 
Section V.2, the data provide a desired level of confidence that is suitable for the intended use. 
The data are being qualified within this technical product only for this work product and should 
not be used or referenced by others as direct input without qualification external to this technical 
product. The fourth section of the discussion for each FEP data set specifies any additional 
applicable limitations. This is addressed for each FEP-specific data set under the heading Data 
Status and Limitation. 
V.10 IDENTIFICATION OF ANY SUPPORTING INFORMATION 
Supporting information (i.e. corroborating information) is identified in Section V.4.x.2 of each 
FEP-specific data set. The table includes an appropriate reference identifier for each citation, 
such as an accession number, DIRS number, or Technical Information Center catalog number. 
April 2004 V-20 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
V.11 DATA QUALIFICATION PLAN 
The data qualification activities documented herein are outlined in the data qualification plan 
(Attachment VI). 
Deviations from the data qualification plan implemented during its execution include the 
following: 
 Waste Package Design Basis Events (CRWMS M&O 1997 [DIRS 100264]) was not 
qualified per the plan because in the interim a better reference (BSC 2004 [DIRS 
162200]) was found to support the screening argument in Section 6.2.37. 
 Items 1 and 3, being data obtained from outside sources, were additionally qualified to 
the requirements of AP-SIII.9Q Rev.1/ICN 4 for qualifying outside source data. 
 Attribute 2 was not used. 
 Attribute 8 was listed in the Data Qualification Plan (Attachment VI) as being invoked, 
but was not used in this Data Qualification Report. 
 Rev. 1, ICN 3 of AP-SIII.9Q, Scientific Analyses, was identified in the Data 
Qualification Plan (Attachment VI) as the governing procedure for Scientific Analyses. 
It has since been superseded by Rev. 1, ICN 4. 
 On the cover of the Data Qualification Plan, Section 5.3.2 was identified as the section 
of AP-SIII.2Q covering documentation. The correct section is 5.3.1. 
April 2004 V-21 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
INTENTIONALLY LEFT BLANK 
April 2004 V-22 of V-22 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
ATTACHMENT VI  DATA QUALIFICATION PLAN (SCANNED IMAGES) 
April 2004 VI-1 of VI-4 ANL-WIS-MD-000009 REV 01 
Miscellaneous Waste Form FEPs: Data Qualification 
April 2004 VI-2 of VI-4 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
April 2004 VI-3 of VI-4 ANL-WIS-MD-000009 REV 01 Miscellaneous Waste Form FEPs: Data Qualification 
April 2004 VI-4 of VI-4 ANL-WIS-MD-000009 REV 01