Waste Form Features, Events, and Processes  Rev 02, ICN 00
 
ANL-WIS-MD-000009

October 2004

1. PURPOSE 
The purpose of this report is to evaluate and document the inclusion or exclusion of the waste 
form features, events and processes (FEPs) with respect to modeling used to support the Total 
System Performance Assessment for License Application (TSPA-LA). A screening decision, 
either Included or Excluded, is given for each FEP along with the technical bases for screening 
decisions. This information is required by the Nuclear Regulatory Commission (NRC) in 
10 CFR 63.114 (d, e, and f) [DIRS 156605]. 
The FEPs addressed in this report deal with the issues related to the degradation and potential 
failure of the waste form and the migration of the waste form colloids. For included FEPs, this 
analysis summarizes the implementation of the FEP in TSPA-LA, (i.e., how the FEP is 
included). For excluded FEPs, this analysis provides the technical bases for exclusion from 
TSPA-LA (i.e., why the FEP is excluded). This revision addresses the TSPA-LA FEP list 
(DTN: MO0407SEPFEPLA.000 [DIRS 170760]). 
The primary purpose of this report is to identify and document the analyses and resolution of the 
features, events, and processes (FEPs) associated with the waste form performance in the 
repository. Forty FEPs were identified that are associated with the waste form performance. 
This report has been prepared to document the screening methodology used in the process of 
FEP inclusion and exclusion. 
The analyses documented in this report are for the license application (LA) base case design 
(BSC 2004 [DIRS 168489]). In this design, a drip shield is placed over the waste package and 
no backfill is placed over the drip shield (BSC 2004 [DIRS 168489]). Each FEP may include 
one or more specific issues that are collectively described by a FEP name and a FEP description. 
The FEP description may encompass a single feature, process or event, or a few closely related 
or coupled processes if the entire FEP can be addressed by a single specific screening argument 
or TSPA-LA disposition. The FEPs are addressed in associated analysis or model reports. The 
assignments were based on the nature of the FEPs so that the analysis and resolution of screening 
decisions reside with the subject-matter experts in the relevant disciplines. 
1.1 PLANNING AND DOCUMENTATION 
Documentation requirements for this report are described in the technical work plan entitled 
Technical Work Plan For: Regulatory Integration Modeling and Analysis of the Waste Form and 
Waste Package (BSC 2004 [DIRS 171583]). Any changes in the assigned waste form FEP list 
for TSPA-LA that resulted from the planned work scope will be described in Section 6.1. 
1.2 SCOPE 
The scope of this report is to identify the treatment of the FEPs affecting postclosure waste form 
performance. Seismic and igneous effects are not treated within this report. A full discussion of 
seismic effects can be found in Engineered Barrier System Features, Events, and Processes 
(BSC 2004 [DIRS 169898]), whereas the igneous effects can be found in Features, Events, and 
Processes: Disruptive Events (BSC 2004[DIRS 170017]). The FEPs that are deemed potentially 
important to repository postclosure performance are evaluated, either as components of the total 
system performance assessment (TSPA-LA) or as a separate discussion in a scientific analysis 
report. 

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The scope for this activity involves two tasks, namely: 
Task 1: Identify which FEPs are to be considered explicitly in the TSPA-LA (called 
included FEPs) and in which scientific analyses these FEPs are addressed. 
Task 2: Identify FEPs not to be included in the TSPA-LA (called excluded FEPs) and 
provide justification for why these FEPs do not need to be a part of the 
TSPA-LA model. 
The scope of this report is to describe, evaluate, and document screening decisions and technical 
bases for the Waste Form FEPs for TSPA-LA. For FEPs that are included in the TSPA-LA, this 
report provides a TSPA-LA disposition, which is a consolidated summary of how the FEP has 
been included and addressed in the TSPA-LA model, based on the various supporting technical 
analysis reports and model reports that describe the inclusion of the FEP. It also provides a list, 
or reference roadmap, of the specific supporting technical reports that provide more-detailed 
discussions of the FEP. For FEPs that are excluded from the TSPA-LA, this report provides a 
screening argument, which identifies the basis for the screening decision (i.e., low probability, 
low consequence, or by regulation) and discusses the technical basis that supports that decision. 
It also provides appropriate references to project and non-project information that supports the 
exclusion. 
In cases when a FEP covers multiple technical areas and is shared with other FEP reports, this 
analysis report provides only a partial technical basis for the screening decision as it relates to the 
waste form areas of concern. The full technical basis for these shared FEPs is addressed, 
collectively, by all of the sharing FEP reports. The discussion of the waste form FEPs shared 
with other FEPs documents are in their respective reports: Engineered Barrier System Features, 
Events, and Processes (BSC 2004 [DIRS 169898]), FEPs Screening of Processes and Issues in 
Drip Shield and Waste Package Degradation (BSC 2004 [DIRS 169997]), Features, Events, and 
Processes in SZ Flow and Transport (BSC 2004 [DIRS 170013]), Features, Events, and 
Processes in UZ Flow and Transport (BSC 2004 [DIRS 170012]), or Evaluation of Features, 
Events, and Processes (FEP) for the Biosphere Model (BSC 2004 [DIRS 169826]). 
An overview of the FEP analysis and scenario development process is available in The 
Development of the TSPA-LA Features, Events, and Processes (BSC 2004, Sections 2.4, 3, and 4 
[DIRS 168706], which describes the TSPA-LA FEP identification and screening process. As 
part of that process, the TSPA-LA FEP list (DTN: MO0407SEPFEPLA.000 [DIRS 170760]) 
was developed. This DTN was used as an initial input to the waste form FEP analysis. The list 
of waste form TSPA-LA FEPs, presented in Table 1-1, was derived from DTN: 
MO0407SEPFEPLA.000 [DIRS 170760] with subsequent modifications to the FEP list, names, 
or descriptions. These modifications are documented in the FEP History File in the FEP 
database (BSC 2004 [DIRS 168706], Table 6-1) and will be incorporated into a subsequent 
revision of the TSPA-LA FEP list (Section 7). Table 1-1 also includes the designation of shared 
FEPs. 
Direct inputs supporting the screening decisions are listed in Table 4-1. Indirect inputs 
supporting the screening decisions are listed in Table 6.1. The individual FEP discussions 
providing identification (FEP number, name, and description) and screening (screening decision, 
screening argument of TSPA-LA disposition) information are in Section 6.2. 

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Table 1-1. Waste Form FEPs for TSPA-LA 
FEP 
Number FEP Name 
Addressed 
in Section 
Shared 
With 
2.1.01.01.0A Waste Inventory 6.2.1 
2.1.01.02.0A Interactions between colocated waste 6.2.2 
2.1.01.02.0B Interactions between codisposed waste 6.2.3 
2.1.01.03.0A Heterogeneity of waste inventory 6.2.4 
2.1.02.01.0A DSNF degradation (alteration, dissolution, and radionuclide release) 6.2.5 
2.1.02.02.0A CSNF degradation (alteration, dissolution, and radionuclide release) 6.2.6 
2.1.02.03.0A HLW glass degradation (alteration, dissolution, and radionuclide release) 6.2.7 
2.1.02.04.0A Alpha recoil enhances dissolution 6.2.8 
2.1.02.05.0A HLW glass cracking 6.2.9 
2.1.02.06.0A HLW glass recrystallization 6.2.10 
2.1.02.07.0A Radionuclide release from gap and grain boundaries 6.2.11 
2.1.02.08.0A Pyrophoricity from DSNF 6.2.12 
2.1.02.09.0A Chemical effects of void space in waste package 6.2.13 
2.1.02.10.0A Organic/cellulosic materials in waste 6.2.14 
2.1.02.25.0A DSNF cladding 6.2.15 
2.1.02.28.0A Grouping of DSNF waste types into categories 6.2.16 
2.1.02.29.0A Flammable gas generation from DSNF 6.2.17 
2.1.03.06.0A Internal corrosion of waste packages prior to breach 6.2.18 WP 
2.1.09.01.0B Chemical characteristics of water in waste package 6.2.19 
2.1.09.02.0A Chemical interaction with corrosion products 6.2.20 EBS 
2.1.09.04.0A Radionuclide solubility, solubility limits, and speciation in the waste form 
and EBS 
6.2.21 
2.1.09.06.0A Reduction-oxidation potential in waste package 6.2.22 
2.1.09.07.0A Reaction kinetics in waste package 6.2.23 
2.1.09.10.0A Secondary phase effects on dissolved radionuclide concentrations 6.2.24 
2.1.09.13.0A Complexation in EBS 6.2.25 
2.1.09.15.0A Formation of true (intrinsic) colloids in EBS 6.2.26 
2.1.09.16.0A Formation of pseudocolloids (natural) in EBS 6.2.27 
2.1.09.17.0A Formation of pseudocolloids (corrosion product) in EBS 6.2.28 
2.1.09.18.0A Formation of microbial colloids in EBS 6.2.29 
2.1.09.23.0A Stability of colloids in EBS 6.2.30 
2.1.09.25.0A Formation of colloids (waste form) by coprecipitation in the EBS 6.2.31 
2.1.11.03.0A Exothermic reactions in the EBS 6.2.32 EBS 
2.1.11.08.0A Thermal effects on chemistry and microbial activity in the EBS 6.2.33 EBS 
2.1.11.09.0B Thermally-driven flow (convection) in waste packages 6.2.34 
2.1.12.04.0A Gas generation (CO2, CH4, H2S) from microbial degradation 6.2.35 EBS 
2.1.12.07.0A Effects of radioactive gases in EBS 6.2.36 EBS 
2.1.13.01.0A Radiolysis 6.2.37 WP, EBS 
2.1.13.02.0A Radiation damage in EBS 6.2.38 WP, EBS 
2.2.08.12.0B Chemistry of water flowing into the waste package 6.2.39 
3.1.01.01.0A Radioactive decay and ingrowth 6.2.40 SZ, UZ, 
BIO 
NOTES: FEPs = features, events, and processes; BIO = Biosphere; SZ = Saturated Zone; 
UZ = Unsaturated Zone; DE = Disruptive Events; EBS = Engineered Barrier System; 
WP = Waste Package; WF = Waste Form; CLAD = Cladding; CRIT = Criticality; SYS = System Level 

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1.3 SCIENTIFIC ANALYSIS LIMITATION AND USE 
The intended use of this report is to provide FEP screening information for a project specific FEP 
database, and to promote traceability and transparency for both included and excluded waste 
form FEPs. This report is intended for use as the source documentation for inclusion or 
exclusion of waste form FEPs within or from the TSPA-LA model. The following limitations 
apply to this analysis report: 
 Because this report cites other reports and controlled documents as direct input, the 
limitations of this analysis report inherently include any limitations or constraints 
described in the cited reports or controlled documents. 
 In cases where FEPs are shared, the scope of this analysis report is limited to 
performance of the waste form. The full technical basis for these shared FEPs is 
addressed, collectively, by all of the sharing FEP reports. 
 For screening purposes, this analysis report generally uses mean values of probabilities, 
mean amplitude of events, or mean value of consequences (e.g., mean time to waste 
package degradation) as a basis for reaching an include/exclude decision. Mean values 
are determined based on the range of possible values. 
The results of the FEP screening presented herein are specific to the repository design and 
processes available at the time of the License Application. Changes in direct inputs listed in 
Section 4.1, in baseline conditions used for this evaluation, or in other subsurface conditions, will 
need to be evaluated to determine whether the changes are within the limits stated in the FEP 
evaluations. Engineering and design changes are subject to evaluation to determine whether 
there are any adverse impacts to safety, as codified in 10 CFR 63.73 Subparts F and G 
[DIRS 156605] (see also the requirements in 10 CFR 63.44 [DIRS 156605]). 

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2. QUALITY ASSURANCE 
Development of this report and its supporting analyses are subject to the Office of Civilian 
Radioactive Waste Management (OCRWM) quality assurance (QA) program as identified in 
Technical Work Plan For: Regulatory Integration Modeling and Analysis of the Waste Form and 
Waste Package (BSC 2004 [DIRS 171583], Section 8, Work Package ARTM03). 
The report contributes to the analysis and modeling used to support performance assessment. 
The waste form FEPs documented herein involve the investigations of items or barriers 
determined to be important to waste isolation in accordance with AP-2.22Q, Classification 
Analyses and Maintenance of the Q-List, and, therefore, is classified as Safety Category (SC) in 
Q-List (BSC 2004 [DIRS 168361], Appendix A). 
Approved QA procedures identified in the technical work plan (BSC 2004 [DIRS 171583], 
Section 4) have been used to conduct and document the activities described in this analysis 
report. In particular, this work constitutes an analysis report, and the documentation has been 
prepared according to AP-SIII.9Q, Scientific Analyses, and in accordance with related procedures 
and guidance documents as outlined in the technical work plan, which also identifies applicable 
controls for the electronic management of data (BSC 2004 [DIRS 171583], Section 8) during the 
analysis and documentation activities. 

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3. USE OF SOFTWARE 
This report uses no computational software; therefore, this analysis is not subject to software 
controls. The analyses and arguments presented herein are based on guidance and regulatory 
requirements, results of analyses presented and documented in other reports or on other technical 
literature. Software and models used in the supporting documents are cited in this report for 
traceability and transparency purposes, but were not used in its development. 
This analysis report was developed using only commercial-off-the-shelf software 
(Microsoft Word 2000) for word processing, which is controlled under the Software 
Configuration Management System, but is not required to be qualified or documented per LPSI.
11Q-BSC, Software Management. No additional applications (routines or macros) were 
developed using this commercial off-the-shelf software. 

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4. INPUTS 
AP-3.15Q, Managing Technical Product Inputs categorizes technical product input usage as 
either direct or indirect input. Direct input is used to develop the results or conclusions in a 
technical product. Indirect input is used to provide additional information that is not used in the 
development of results or conclusions. Direct inputs are identified in Table 4-1 and indirect 
inputs are identified in Table 6-1. 
Section 4.1 identifies all direct inputs used in this FEP report. The direct inputs were obtained 
from controlled source documents and other appropriate sources in accordance with AP-3.15Q, 
Managing Technical Product Inputs. Section 4.2 identifies the FEP screening criteria described 
in 10 CFR Part 63 [DIRS 156605] along with the regulatorily derived FEP screening criteria. 
Section 4.3 identifies applicable codes, standards, and regulations. 
4.1 DIRECT INPUTS 
The LA FEP list (DTN: MO0407SEPFEPLA.000 [DIRS 170760]) was used as a direct input to 
provide the initial list of waste form FEPs for screening in this analysis report. The LA FEP list 
identifies a FEP report or a set of sharing FEP reports for each FEP. Subsequent additions to or 
changes from that FEP list (numbers, names, or descriptions) are reflected in the information 
provided in Section 6.2 and can be traced through the FEP History File in the FEP database 
(BSC 2004 [DIRS 168706], Table 6-1). 
Other direct inputs used for the FEP screening analysis are listed in Table 4-1. The 
appropriateness of the inputs by DOE (2002 [DIRS 158405]), Rechard (1995 [DIRS 101084]) 
and Sunder et al. (1997 [DIRS 143860]) are discussed in Appendix B. The technical input from 
the cancelled document (BSC 2001 [DIRS 171508]), is discussed in Appendix B. All other 
inputs listed in Table 4-1 are appropriate for use in the FEP screening analysis because they 
represent the latest information specifically developed for use in the TSPA-LA. 

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Table 4-1. Direct Inputs 
Technical Product Input Used From Used In Input Description 
10 CFR 63 [DIRS 156605] Part 114 (d, e, f, g), 
Part 342 
4.1, 4.2, 4.3, 6.1.2, 
6.1.6, 6.2, Appendix 
B, Table 4-1, Table 
4-2 
FEP Screening requirements 
Section 6.2 Table 4-1, 6.2.12 The maximum boundary dose would be less than 5 mrem/yr BSC 2001 [DIRS 171508] 
Section 5.2.6 6.2.32 The energy release from oxidation of U metal in waste package with 2 MCOs 
BSC 2002 [DIRS 160059] Table 2 Table 4-1, 6.2.36 Argon, krypton, and radon were screened out due to short half-lives 
BSC 2004 [DIRS 171744] Table 11 Section 6.2.38, 
Table 4-1 
Thermal neutron flux in BWR 
Tables 4-2 and 4-4 6.2.8 Flow through test results 
Sections 6.2.1, 6.3 6.2.11 Fractional release rate 
BSC 2004 [DIRS 169987] 
Sections 1, 6.1, 6.2, 
6.2.2.3, 6.4.1.5 
Sections 6.2.6, Table 
4-1 
Characteristics of CSNF waste degradation processes and CSNF degradation model. 
BSC 2004 [DIRS 167754] Table 12 Table 4-1, 6.2.32 Time dependent decay heat for Waste Packages in a 7-package segment 
Sections 6.5.1, 
6.5.5.3.2, and Table 
7.1 
Section 6.2.7 General discussion of entire glass model 
Section 6.5.6 Section 6.2.9 Exposure factor incorporates cracking and is part of model 
Table 8-1 Section 6.2.38 Two order magnitude range in dissolution rate 
Section 6.5.2 Sections 6.2.4, Table 
4-1 
Glass compositions are accounted for through the model parameter kE 
APP A, Sections 
6.5.2.4, 6.5.2.5, 
6.5.3.3 
Table 4-1, Section 
6.2.20 
Effects of dissolved iron are accounted for in the model when determining parameter 
values and the effects of glass alteration phases on the rate are included in determining the 
kE value 
Section 8.2.1 Table 4-1, Section 
6.2.3 
HLWG release depends on the surface area. The surface area used to represent a HLWG 
canister is a weighted average of canister sizes. 
BSC 2004 [DIRS 169988] 
Sections 6.5.1, 
6.5.2.2, 6.5.5 
Table 4-1, Section 
6.2.10 
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. 

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Table 4-1. Direct Inputs (Continued) 
Technical Product Input Used From Used In Input Description 
Section 6.5.3.2 Section 6.2.22 Use of adjusted Eh model to account for redox potential in the repository 
Section 6.3.2 Section 6.2.23 Consideration of kinetics in the solubility model 
Sections 6.3.2, 6.5 
through 6.18 
Table 4-1, 
Section 6.2.21 
Abstraction of solubility models for 14 elements (plutonium, neptunium, uranium, thorium, 
americium, actinium, protactinium, radium, technetium, iodine, strontium, carbon, and 
cesium) may be used to constrain the maximum radionuclide concentrations. 
Sections 6.3.2 and 
6.6.4.5 
Table 4-1, 
Sections 5.2, 6.2.24 
Alternative model (not utilized in the base-case) of incorporation of neptunium into uranyl 
minerals that 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. 
Section 6.4 Table 4-1, 
Section 6.2.35 
Parameterizes 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 10-1.5  in 
effect to two orders of magnitude greater than atmospheric. 
BSC 2004 [DIRS 169425] 
Section 6.4 Table 4-1, 
Section 6.2.39 
The dissolved concentration model considers the potential chemistry of water entering the 
waste package. 
BSC 2004 [DIRS 169593] Table 6.4-5 Section 6.2.38, 
Table 4-1 
Neutron flux at the waste package surface 
Section 8.1 Section 6.2.5 Naval fuel is conservatively treated as CSNF 
Section 8.1 Sections 6.2.5 and 
6.2.32 
DSNF is released in a single TSPA-LA timestep 
Table 6-2 Sections 6.2.8 and 
Table 4-1 
Uranium metal corrosion rate 
Section 6.1.12 Section 6.2.3 DSNF is released within one TSPA-LA time step, so is not sensitive to surface area 
Section 6.1 Section 6.2.5, 
Table 4-1 
Development of upper-limit and best estimate degradation models for each of 11 DSNF 
Waste Groups for TSPA-LA. 
Section 6.2 Sections 6.2.16, 
6.2.18, Table 4-1 
Rapid corrosion rate of the metallic uranium that is the matrix of N Reactor DSNF 
compared with other SNF. 
Section 6.4.3 Section 6.2.15, 
Table 4-1, and 
Appendix B 
Lack of sufficient available information to adequately characterize N Reactor spent fuel 
cladding. 
BSC 2004 [DIRS 167618] 
Section 8.1 Section 6.2.3, 
Table 4-1 
The DSNF release rate model has no explicit dependence on water chemistry effects that 
could result from chemical interactions between or within waste packages. 

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Table 4-1. Direct Inputs (Continued) 
Technical Product Input Used From Used In Input Description 
Sections 4.1 and 
6.6.2 
Section 6.2.39 Various water compositions entering the waste package 
Sections 6.3.1.1, 
6.6.4, 6.10.2.3 
Section 6.2.22 Discussion of redox potential in the in-package chemistry mode 
Sections 4.1.3, 
4.1.4, 6.6, 6.3.1.1, 
6.6.3, and 6.10.2.3 
Section 6.2.23 Discussion of kinetics in the in-package chemistry model 
6.4.1 Section 6.2.34 Convection on the in-package chemistry model 
Sections 6.3.1, 
6.6.1, and 6.6.4 
Section 6.2.13 Discussion of modeling voids in the waste package 
Appendix B Section 6.2.37 Radiolysis analysis show negligible impact on In-Package chemistry 
Section 6.5 Table 4-1 and 
Section 6.2.3 
Support for the conclusion that HLWG 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 
Sections 4.1.2, 6.6.2 
and Table 10 
Table 4-1, 
Section 6.2.35 
Support for the conclusion that it is reasonable to expect little impact of carbon dioxide 
production on in-package pH and ionic strength, and that only minor amounts of organic 
materials might appear in the WP. 
Section 6.10 and 
6.3.1 
Table 4-1, 
Section 6.2.33 
Temperature effect on the in-package model 
BSC 2004 [DIRS 167621] 
Entire Table 4-1, 
Sections 6.2.3, 
6.2.13, and 6.2.19 
General reference describing the basis of the conceptual model for in-package chemistry of 
water. 
Section 6.7 and 7, 
Table 6-2, 
Appendices I.4.2 
and I.5, Table 7-1 
Sections 6.2.1, 6.2.2, 
6.2.3, 6.2.4, 6.2.8, 
6.2.16, 6.2.17, 
6.2.40 and Table 4-1 
Provides the inventories of radionuclides for three major categories of waste forms, 
currently planned for disposal at Yucca Mountain: CSNF, DSNF, and DHLW Glass. 
Appendix I (I.4 and 
I.5) 
Section 6.2.1 Crud deposits not significant 
Table 6-3 Section 6.2.3 Some WPs will contain 5-HLWG and no DSNF 
BSC 2004 [DIRS 170022] 
Table 6-1 Section 6.2.40 Radionuclide precursors 

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Table 4-1. Direct Inputs (Continued) 
Technical Product Input Used From Used In Input Description 
BSC 2004 [DIRS 170025] Entire Table 4-1 and 
Sections 5.3, 5.4, 
6.2.20, 6.2.26, 
6.2.27, 6.2.28, 
6.2.29, 6.2.30, 
6.2.31 
Formation and stability of colloids 
BSC 2004 [DIRS 169565] Entire Section 6.2.33 and 
Table 4-1 
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. 
CRWMS M&O 2000 
[DIRS 151561] 
Sections 6.3, 7.2 
and Table 71 
Sections 6.2.14, 
6.2.25, 6.2.36, 
Table 4-1 
Estimates of microbe masses growing in the repository system suggest that the effects on 
the bulk in-drift geochemistry should be small. Bounding biomass estimates are ~100grams 
(dry) per meter of drift; insufficient to produce high chelate levels. 
Entire, Appendix D, 
Section 9 
Table 4-1, 
Sections 6.2.2, 6.2.5, 
6.2.12, 6.2.15, 
6.2.16, Appendix B, 
Table B-1 
N Reactor accounts for about 85% of the total metric tons of heavy metal of DOE spent 
nuclear fuel 
Appendix D, 
Section 1, Table D-1 
Sections 4.1, 6.2.16 
and 6.2.17, Table 4- 
1, Appendix B, 
Table B-1 
The total inventory of the Group 3 (Pu/U carbide) SNF is less than .001 % of DSNF waste 
DOE 2002 [DIRS 158405] 
Section 9 Sections 4.1, 6.2.2, 
6.2.3, 6.2.5, Table 4- 
1, Appendix B, Table 
B-1 
Present plans call for five glass logs and one DSNF canister to be placed in each 
codisposal package 
MO0407SEPFEPLA.000 
[DIRS 170760] 
Entire Sections 4.1 and 
6.1.1, Table 4-1 
FEPS considered for LA 
Rechard 1995 
[DIRS 101084] 
Section 11.3.1, 
p. 11-24 
Table 4-1, 
Sections 4.1, 6.2.5, 
6.2.15, Appendix B, 
Table B-1 
Significant amounts of DSNF cladding are likely to be damaged 
Sunder et al. 1997 
[DIRS 143860] 
pp. 66 to 74 Sections 4.1, 6.2.37, 
Tables 4-1, 
Appendix B, Table 
B-1 
Support for the conclusion that the effects of alpha radiolysis on spent fuel corrosion 
(dissolution) will be transitory and will become minor as alpha dose rates decrease. 

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The regulations in 10 CFR Part 63 [DIRS 156605] also provide direct inputs to the FEP 
screening process. By specifying characteristics, concepts, and definitions, the regulations serve 
as de facto inputs used for screening FEPs. The regulatory definitions and elucidation of the 
regulatory concepts pertaining to the reference biosphere, geologic setting, reasonably maximally 
exposed individual (RMEI), and human intrusion are explained in detail in Section 4.1.3 of The 
Development of the TSPA-LA Features, Events, and Processes (BSC 2004 [DIRS 168706]). 
Specific direct inputs from the regulations addressing these concepts are listed in Table 4-2. 
4.2 CRITERIA 
The technical work plan for this activity (BSC 2004 [DIRS 171583], Table 3-1) has identified 
the acceptance criteria (AC) applicable to this report based on the requirements of Project 
Requirements Document (PRD) (Canori and Leitner 2003 [DIRS 166275]) and Yucca Mountain 
Review Plan, Final Report (NRC 2003 [DIRS 163274]). 
4.2.1 Criteria from the Projects Requirements Document and the Yucca Mountain 
Review Plan 
The licensing criteria for postclosure performance assessment are stated in 10 CFR 63.114 
[DIRS 156605]. The requirements to be satisfied by TSPA-LA are identified in Project 
Requirements Document (PRD) (Canori and Leitner 2003 [DIRS 166275]). The Acceptance 
Criteria that will be used by the NRC to evaluate the adequacy of technical arguments are 
identified in Yucca Mountain Review Plan, Final Report (NRC 2003 [DIRS 163274]). Table 4-2 
provides a crosswalk between the regulatory requirements, the PRD (Canori and Leitner 2003 
[DIRS 166275]) and the acceptance criteria (Scenario Analysis and Event Probability) provided 
in Sections 2.2.1.2.1.3 and 2.2.1.2.2.3 of Yucca Mountain Review Plan, Final Report (NRC 2003 
[DIRS 163274]). 
The PRD (Canori and Leitner 2003 [DIRS 166275]) documents and categorizes the regulatory 
requirements and other project requirements. The regulatory requirements include criteria 
relevant to performance assessment activities, in general, and to FEP-related activities as they 
pertain to TSPA-LA, in particular. In Table 4-2, NRC Acceptance Criteria are correlated to the 
corresponding regulations as they pertain to FEPs-related criteria. 
The basis of the NRC review is described in Yucca Mountain Review Plan, Final Report 
(NRC 2003 [DIRS 163274], Sections 2.2.1.2), and the bases for acceptance are stated as 
acceptance criteria. The following acceptance criteria that apply to this report are based on 
meeting the requirements of 10 CFR 63.114 (a), (b), (d), (e), and (f) [DIRS 156605] from 
Section 2.2.1.2.1.3 of Yucca Mountain Review Plan, Final Report (NRC 2003 [DIRS 163274]). 
 Acceptance Criteria 1 - The Identification of a List of Features, Events, and Processes 
Is Adequate 
 Acceptance Criteria 2 - Screening of the List of Features, Events, and Processes Is 
Appropriate 
Section 7.1 describes how applicable acceptance criteria are addressed. Acceptance criteria for 
FEP screening reiterate the regulatory screening criteria of low probability and low consequence 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 4-7 October 2004 
but also allow for exclusion of a FEP if the process is specifically excluded by the regulations 
(see Section 4.2.2). 
Acceptance criteria listed in Section 2.2.1.2.2.3 of Yucca Mountain Review Plan, Final Report 
(NRC 2003 [DIRS 163274]) pertaining to identification of events with probability greater than 
10-8 per year are not considered because this analysis does not develop probabilities for such 
events. 
Table 4-2. Relationships of Regulations to the Project Requirements and the NRC Acceptance Criteria 
10 CFR Part 63 
[DIRS 156605] 
Canori and Leitner 
2003 [DIRS 166275] Description of the Applicable 
Regulatory Requirement or 
Acceptance Criterion 
Regulatory Citation 
Associated 
PRD 
Associated 
Acceptance Criteria 
NRC 2003 
[DIRS 163274] 
General Requirements and Scope Pertinent to FEP Screening 
Include data related to geology, 
hydrology, geochemistry, and 
geophysics 
63.114(a) PRD-002/T-015 2.2.1.2.1.3 
Acceptance Criterion 1 
Include information of the design of 
the engineered barrier system used 
to define parameters and conceptual 
models 
63.114(a) PRD-002/T-015 2.2.1.2.1.3 
Acceptance Criterion 1 
Account for uncertainties and 
variabilities in parameter values and 
provide the technical basis for 
parameter ranges, probability 
distributions, or bounding values 
63.114(b) PRD-002/T-015 
2.2.1.2.2.3 
Acceptance Criteria 2 
and 5 
FEP Screening Criteria 
Provide the justification and technical 
basis for excluding FEPs specifically 
excluded by regulation. 
Not Applicable Not Applicable 2.2.1.2.1.3 
Acceptance Criterion 2 
Provide the technical basis for either 
inclusion or exclusion of FEPs. 
Provide the justification and technical 
basis for those excluded based on 
probability. 
63.114(d) 
63.342 
PRD-002/T-015 
PRD-002/T-034 
2.2.1.2.1.3 
Acceptance Criterion 2 
2.2.1.2.2.3 
Acceptance Criteria 
1 and 2 
Provide the technical basis for either 
inclusion or exclusion of FEPs. 
Provide the justification and the 
technical basis for those excluded 
based on lack of significant change 
in resulting radiological exposure or 
release to the accessible 
environment. 
63.114 (e and f) 
63.342 
PRD-002/T-015 
PRD-002/T-034 
2.2.1.2.1.3 
Acceptance Criterion 2 
2.2.1.2.2.3 
Acceptance Criteria 1 
and 2 
NOTE: FEPs = features, events, and processes ; PRD = Project Requirements Document. 
4.2.2 FEPs Screening Criteria 
The NRC regulations and guidance specifically allow the exclusion of FEPs from the TSPA-LA 
if they can be shown to be of low probability or of low consequence. Additionally, FEPs can be 
excluded based on the constraints provided within 10 CFR Part 63 [DIRS 156605]. In this 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 4-8 October 2004 
document, this exclusion is called exclusion by regulation. FEPs screening criteria are 
described further in the following three sections. 
4.2.2.1 Exclusion by Low Probability 
The low-probability criterion is stated at 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. 
Furthermore, it is stated at 10 CFR 63.342 [DIRS 156605] that: 
DOEs assessments for the human intrusion and groundwater protection standards 
shall not include consideration of unlikely features, events, or 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. 
4.2.2.2 Exclusion by Low Consequence 
The low consequence criteria are stated at 10 CFR 63.114 (e and f) [DIRS 156605]: 
(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 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. 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 4-9 October 2004 
and supported by 10 CFR 63.342 [DIRS 156605]: 
DOEs performance assessments need not evaluate the impacts resulting from any 
features, events, and processes or sequences of events or processes with a higher 
chance of occurrence if the results of the performance assessments would not be 
changed significantly. 
The terms significantly changed and changed significantly are undefined terms in the NRC 
regulations. The absence of significant change is inferred for FEP screening purposes to be 
equivalent to having no, or negligible, effect. 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. 
4.2.2.3 Exclusion by Regulation 
The provisions and constraints provided within 10 CFR Part 63 [DIRS 156605] pertaining to the 
reference biosphere, receptor, and performance assessment serve as the basis for exclusion of 
some FEPs. This process of screening out the FEPs that fall outside the parameters established 
by 10 CFR Part 63 [DIRS 156605] is described in Yucca Mountain Review Plan, Final Report 
(NRC 2003 [DIRS 163274], Section 2.2.1.2.1.3 Acceptance Criterion 2) together with the 
screening criteria of low probability and low consequence: 
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. 
Exclusion of FEPs by regulation involves consideration of those portions of 10 CFR Part 63 
[DIRS 156605] that define requirements and key concepts for performance assessment. In this 
context, portions of 10 CFR Part 63 [DIRS 156605] serve as criteria for screening related FEPs. 
4.3 CODES, STANDARDS, AND REGULATIONS 
This document was prepared to comply with NRC regulatory requirements presented in 
10 CFR Part 63 [DIRS 156605]. Subparts of this rule that are applicable include Subpart B, 
Section 15 (Site Characterization), Subpart E, Section 114 (Requirements for Performance 
Assessment), Subpart F (Performance Confirmation Program) and Subpart G (Quality 
Assurance). 

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ANL-WIS-MD-000009 REV 02 4-10 October 2004 
INTENTIONALLY LEFT BLANK 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 5-1 October 2004 
5. ASSUMPTIONS 
5.1 ANNUAL-EXCEEDANCE PROBABILITY 
For naturally occurring FEPs, it is assumed that regulations expressed as a probability criterion 
can also be expressed as an annual-exceedance probability, which is defined as the probability 
that a specified value (such as for ground motions or fault displacement) will be exceeded during 
one year. More specifically, a stated probability-screening criterion of one chance in 10,000 in 
10,000 years (10-4/104 yr) criterion is assumed equivalent to a 10-8 annual-exceedance 
probability. 
The definition of annual exceedance probability, and the following justification for this 
assumption is taken from Characterize Framework for Seismicity and Structural Deformation at 
Yucca Mountain, Nevada (BSC 2004 [DIRS 168030], Glossary). 
The assumption of equivalence of annual-exceedance probability is appropriate if the possibility 
of an event is equal for any given year. This satisfies the definition of a Poisson distribution as 
a mathematical model of the number of outcomes obtained in a suitable interval of time and 
space, that has its mean equal to its variance (Merriam-Webster 1993 [DIRS 100468], 
p. 899). This is inferred to mean that naturally occurring, infrequent, and independent events can 
be represented as stochastic processes in which distinct events occur in such a way that the 
number of events occurring in a given period of time depends only on the length of the time 
period. The use of this assumption is justified in Characterize Framework for Seismicity and 
Structural Deformation at Yucca Mountain, Nevada (BSC 2004 [DIRS 168030], Section 6.4.2), 
which indicates that assuming the behavior of the earth is generally Poissonian or random is the 
underlying assumption in all probabilistic hazard analyses. 
This assumption is used throughout. It is particularly germane to FEPs related to processes or 
phenomena that have or are relative to annual-exceedance probability. 
5.2 SECONDARY PHASE EFFECTS ON DISSOLVED RADIONUCLIDE 
CONCENTRATIONS 
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 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 6.2.24 to demonstrate that the use of the mechanistic 
model for the dissolution/precipitation reactions are conservative in terms of predicting 
neptunium transport. This assumption is supported by Dissolved Concentration Limits of 
Radioactive Elements (BSC 2004 [DIRS 169425]) and requires no further confirmation for its 
usage in this document. 
5.3 THERMALLY DRIVEN FLOW (CONVECTION) IN THE EBS 
It is assumed 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 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 5-2 October 2004 
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 6.2.34. This assumption does not require validation because having the concentration of 
dissolved materials dispersed uniformly would slightly overpredict the transport of radionuclides 
out of the waste package, and is, thus, conservative. 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-1 October 2004 
6. SCIENTIFIC ANALYSIS DISCUSSION 
The primary concern of this report is to address and document the screening decisions for the 40 
waste form FEPs listed in Table 1-1. In some cases, where a FEP covers multiple technical areas 
and is shared with other FEP reports, this report provides only partial technical basis for the 
screening of the FEP. The shared FEPs and the appropriate other area of responsibility are 
identified in Table 1-1. The full technical basis for these shared FEPs is addressed collectively 
by all of the sharing FEP reports. 
6.1 METHODS AND APPROACH 
The identification and screening of a comprehensive list of FEPs potentially relevant to the 
repository postclosure performance is an ongoing, iterative process based on site-specific 
information, design, and regulations. FEP analysis uses the following definitions, as taken from 
Yucca Mountain Review Plan, Final Report (NRC 2003 [DIRS 163274], Glossary): 
Feature  An object, structure, or condition that has a potential to affect disposal system 
performance. 
Event  A natural or human-caused phenomenon that has a potential to affect disposal 
system performance and that occurs during an interval that is short compared to 
the period of performance. 
Process  A natural or human-caused phenomenon that has a potential to affect disposal 
system performance and that operates during all or a significant part of the period 
of performance. 
FEP analysis for TSPA-LA is described in The Development of the TSPA-LA Features, Events, 
and Processes (BSC 2004 [DIRS 168706]). It is summarized in the following sections. 
6.1.1 Features, Events, and Processes Identification and Classification 
The first step of FEP analysis is FEP identification and classification, which addresses 
Acceptance Criterion 1 of Yucca Mountain Review Plan, Final Report (NRC 2003 
[DIRS 163274], Section 2.2.1.2.1.3). The TSPA-LA FEP identification and classification 
process is described in The Development of the TSPA-LA Features, Events, and Processes 
(BSC 2004 [DIRS 168706], Section 3). This process produced a version of the LA FEP list 
(DTN: MO0407SEPFEPLA.000 [DIRS 170760], used as initial input in this waste form FEP 
analysis. Subsequent modifications to the FEP list from the information shown in 
DTN: MO0407SEPFEPLA.000 [DIRS 170760], aside from editorial corrections to FEP 
descriptions, are discussed later in this section. All subsequent modifications are also 
documented in the FEP History File in the FEP database (BSC 2004 [DIRS 168706], 
Table 6-1). 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-2 October 2004 
Table 6-1. Indirect Input 
Technical Product Input Used From Used In Input Description 
10 CFR 60 [DIRS 103540] Entire Appendix B, Table 6-1 High-level radioactive wastes in geologic repositories 
10 CFR 63 [DIRS 156605] Entire Sections 1, 1.3, 4, 
Table 6-1 
Guidance for exclusion or inclusion in screening FEPs 
67 FR 19432 [DIRS 162618] Entire Section 6.2.16 and 
Table 6-1 
Proposal by DOE-EM to use most excess defense plutonium for MOX fuel for 
commercial reactors rather than disposal as immobilized Pu ceramic in the 
repository. 
Abrefah et al. 1995 
[DIRS 151125] 
Section 3.1 and 
Figures 3.1 to 3.5 
Table 6-1, 
Sections 6.2.12, 6.2.15 
and Appendix B 
Evidence 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. 
Abrefah et al. 1999 
[DIRS 151226] 
Tables 4.2 & 4.3, 
Appendix C 
Table 6-1, 6.2.12 Observed decrease in ignition temperature observed during ignition testing of 
damaged/corroded N Reactor SNF samples 
ASTM C 1454-00. 2000 
[DIRS 152779] 
Sec. 3.2 Section 6.2.12, Table 6-1 ASTM Standard Guide for evaluating the potential for pyrophoric behavior of 
uranium metal-based spent fuel. 
Benedict et al. 1981 
[DIRS 103523] 
p. 393 Section 6.2.38, Table 6-1 Average thermal-neutron flux for a PWR 
Bruno et al. 1998 
[DIRS 101565] 
Entire Section 6.2.24 and 
Table 6-1 
Observation that a coherent relationship exists between neptunium and uranium in 
spent fuel dissolution experiments in the Spanish Nuclear Waste Program in 
support of the conclusion that the ratio of neptunium to uranium in the leachate is 
equal to the ratio in the spent fuel used in experiments. 
BSC 2001 [DIRS 162733] Entire Table 6-1 and 
Section 6.2.25 
Calculated microbial populations will not be sufficient to generate significant 
concentrations of radionuclide-chelating organics 
BSC 2001 [DIRS 152059] Entire Section 6.2.5 and 
Table 6-1 
Estimated releases from naval SNF waste 
BSC 2004 [DIRS 172017] Sections 6.1, 6.2, 
6.3 
Table 6-1, Appendix B Justification in screening out radiolysis due to gamma LET. Recent calculations by 
Tang have shown that neutron irradiation is negligible and gamma dose has been 
significantly reduced. 
BSC 2002 [DIRS 160059] Table 13, Entire Table 6-1, Sections 6.2.1 
and 6.2.40 
The radionuclides of importance to dose calculations were assessed, and 
considered nominal, human intrusion and igneous events, groundwater and 
atmospheric exposure scenarios, and exposure times. 
BSC 2003 [DIRS 168795] Attachment II Table 6-1 and Appendix B Support for screening out gamma radiation as affecting radiation-induced changes 
in waste. 
BSC 2004 [DIRS 170038] Table 6-6 Section 6.2.17 and 
Table 6-1 
Gas saturation of groundwater 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-3 October 2004 
Table 6-1. Indirect Input (Continued) 
Technical Product Input Used From Used In Input Description 
BSC 2004 [DIRS 168030] Glossary, Section 
6.4.2 
Table 6-1, Section 5.1 Justification of assumption on annual exceedance 
Appendix A Section 6.2.34 Low temperature gradients in waste package at 1,000 years BSC 2004 [DIRS 170019] 
Section 6.2.4 Section 6.2.35 Localized depression of pH due to microbes. 
BSC 2004 [DIRS 168489] Entire Section 1, Table 6-1 Base-case design elements 
BSC 2004 [DIRS 169868] Entire Sections 6.2.33, 6.2.34 
and Table 6-1 
Radionuclide transport 
BSC 2004 [DIRS 169898] Entire, 
Sections 6.2.43, 
6.2.65, and 6.2.77 
Tables 6-1, Sections 1.2, 
6.2.2, 6.2.33, 6.2.35 
Engineer Barrier System FEP discussion 
BSC 2004 [DIRS 169826] Entire Tables 6-1, Section 1.2 Biosphere FEPs discussion 
BSC 2004 [DIRS 169991] 6.4.1 Tables 6-1, 
Section 6.2.25 
Repository temperatures are expected to remain sufficiently high that significant 
microbial activity is possible only after the first 1,000 years, when the drift 
temperature drops below the approximate 96C boiling point 
BSC 2004 [DIRS 170013] Entire Table 6-1, Section 1.2 Discussion on SZ flow and transport FEPS 
BSC 2004 [DIRS 170012] Entire Table 6-1, 1.2, 6.2.36 UZ Flow and Transport FEPS discussion 
BSC 2004 [DIRS 170017] Entire Section 1.2, Table 6-1 Igneous effect FEPS 
Section 6.2.12 Section 6.2.18 Waste package design to preclude any adverse effects. 
Section 6.2.18 Table 6-1 and 
Section 6.2.20 
Analysis of clogging of waste package breached zones by corrosion products 
BSC 2004 [DIRS 169997] 
Section 6.2.13.5 Table 6-1 and 
Section 6.2.36 
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. 
BSC 2004 [DIRS 164327] Section 6.3.7.2.3 Table 6-1, 6.2.2 Underside of drip shield condensation 
BSC 2004 [DIRS 167621] Sections 6.5, 6.6, and 
6.7 
Section 6.2.20 Discussion of impacts of corrosion products on in-package chemistry 
BSC 2004 [DIRS 168361] Appendix A Section 2, Table 6-1 Q-List of system components 
BSC 2004 [DIRS 171583] Entire Sections 1.1 and 2, 
Table 6-1 
Document requirements 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-4 October 2004 
Table 6-1. Indirect Input (Continued) 
Technical Product Input Used From Used In Input Description 
BSC 2004 [DIRS 168706] Entire Sections 1.2, 4.1, 6.1, 
6.1.1, 6.1.2, 6.1.6 and 
Table 6-1 
FEP screening process 
BSC 2004 [DIRS 166695] Table 21 Table 6-1, 6.2.2 The codisposal waste packages are expected to be generally cooler than CSNF 
waste packages at the time of disposal 
Buck et al. 2003 [DIRS 
168411] 
Entire Table 6-1 and 
Section 6.2.24 
Np is incorporated into uranium-based alteration phases in approximately the same 
concentrations as presented in the mother solutions. 
Canori and Leitner 2003 
[DIRS 166275] 
Entire Sections 4.2.1, 6.2.18, 
Tables 4-2 and 6-1 
Project requirements document 
Chen 2001 [DIRS 161997] Entire Table 6-1 and 
Section 6.2.24 
Analyses 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 in the leachates. 
Chen et al. 2002 [DIRS 
161996] 
Entire Table 6-1 and 
Section 6.2.24 
Analyses 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 in the leachates. 
CRWMS M&O 2000 [DIRS 
147651] 
Table 8-1 Table 6.1, Section 6.2.2 Characteristics of DOE spent fuel; HEU 
CRWMS M&O 2000 [DIRS 
147650] 
Table 8-1 Table 6.1, Section 6.2.2 Characteristics of DOE spent fuels; UZrH 
CRWMS M&O 2000 [DIRS 
151561] 
Sections 6.3 and 7.2 Section 6.2.35, 
Table 6-1 
Potential sources for the release of gases from the fuel matrix to the waste package 
interior. 
CRWMS M&O 2000 [DIRS 
125038] 
Entire, 6.6.2 Section 6.2.12 and 
Table 6-1 
TSPA-LA sensitivity analyses 
DOE 2000 [DIRS 118968] Entire, Section 8.1 Table 6-1, 
Sections 6.2.5, 6.2.12 
and 6.2.16 
Classification of DSNF types 
DOE 2002 [DIRS 158405] Appendix D, 
Table D-1 
Sections 6.2.17, 
Table 6-1 
The only fuel waste capable of producing 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 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. 
DOE 2002 [DIRS 171675] Entire Table 6-1, Appendix B DOE audit report 
DOE 2002 [DIRS 161752] Section 1.5 Section 6.2.17 and 
Table 6-1 
Packaging of carbide fuel 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-5 October 2004 
Table 6-1. Indirect Input (Continued) 
Technical Product Input Used From Used In Input Description 
DOE 2004 [DIRS 169354] p. D-580 Table 6-1, 
Section 6.2.17 
About 5% of waste packages contain carbide fuels 
Appendix A, 
Section 6.9, p. 41 
Table 6-1, 
Section 6.2.17 
Broken Peach-Bottom Core-1 pellets have an estimated cladding failure rate of 35%. 
Table 1-1, p.1-8 and 
Appendix B, p. 14 
Section 6.2.17 1.663 MTHM of Peach Bottom Core 1 pellets contained in 103 waste packages will 
be disposed in the repository. 
Appendix B, p. 14 and 
Table 1-1, p.1-8 
Section 6.2.17 24.667 MTHM of high-integrity fuel (contained in 545 waste packages) from the Fort 
St. Vrain and Peach-Bottom Core 2 reactors. 
Section 5.1.2.1 Section 6.2.12 Inert environment reduce potential for oxidation 
DOE 1998 [DIRS 122980] 
Section ES.3.4 Section 6.2.32 Heat from oxidation of N-Reactor fuel is 1/100th of decay heat from DSNF and 
DHLW in first year after disposal 
DOE 2002 [DIRS 158873] Sections 4.2.11, 
4.3.1.B, 4.3.7.B, 
4.3.17 
Table 6-1, 
Section 6.2.14 
Waste Acceptance Criterion that emplaced waste must not contain detectible levels 
of chemically reactive materials, such as cellulosic or organics. 
DOE-HDBK-1081-94 1994 
[DIRS 103327] 
Entire Section 6.2.12 and 
Table 6-1 
Pyrophoric compounds of uranium 
Foster and Wright 1973 
[DIRS 144061] 
pp. 296 to 297 Table 6-1, Section 6.2.8 Basic Nuclear Engineering textbook - it requires a minimum displacement energy, 
Ed, of 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. 
Gray 1988 [DIRS 168471] Summary, 
Sections 2.0 and 5.1 
Table 6-1 and Appendix 
B 
Laboratory experiments show little effect of alpha radiolysis on the leaching behavior 
of spent fuel and UO2 samples. 
Gray and Einziger 1998 
[DIRS 109691] 
Section 4 Table 6-1, 
Section 6.2.18 
Rapid corrosion rate of the metallic uranium that is the matrix of N Reactor SNF 
compared with other SNF. 
Kaplan 1963 [DIRS 149367] p. 307 Table 6-1, 
Section 6.2.37 
Commercial fuel cladding will remain intact and should substantially reduce alpha 
dose rates to groundwater. 
Lawrence 1999 [DIRS 
168475] 
Section 5.1 and 
Table 1 
Table 6-1 and Appendix 
B 
Estimate of approximately 33%(by means of visual survey for the fraction of failed N 
Reactor spent fuel assemblies for both the K-East and K-West basins. 
Lederer and Shirley 1978 
[DIRS 142133] 
p. 1,464 Table 6-1, Section 6.2.8 
and Appendix A 
Table of isotopes 
Marschman et al. 1997 
[DIRS 149429] 
Section 3.4.2 Table 6-1, 
Section 6.2.12 
Uranium hydride inclusions tend to be concentrated near the exposed uranium metal 
fuel surface of damaged N Reactor SNF. 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-6 October 2004 
Table 6-1. Indirect Input (Continued) 
Technical Product Input Used From Used In Input Description 
Merriam-Webster 1993 
[DIRS 100468] 
Entire Entire Definitions 
MO0407SEPFEPLA.000 
[DIRS 170760] 
Entire Section 1, Table 6-1 FEPS considered for LA 
MO0410MWDTLCSC.000 
[DIRS 172057] 
Entire Section 6.2.6 Potential oxidation of fuel has negligible consequences 
NRC 2003 [DIRS 163274] Section 2.2.1.2, 
2.2.1.2.1.3, Table 4-2, 
Glossary 
Sections 4.2, 4.2.1, 
4.2.2, 6.1, 6.1.1, 6.1.2, 
7.1.1, Tables 4-2 & 6-1 
Yucca Mountain Review Plan Guidance to assumptions pertaining to the 
characteristics of the reference biosphere and geologic setting. 
P. 48 Section 6.2.8 & Table 6- 
1 
Significant portion of the radionuclides in the expected YMP inventory are alpha 
emitters. 
Parrington et al. 1996 [DIRS 
103896] 
P. 34-35 Section 6.2.36 The half-life of Xe isotopes very short 
Propp 1998 [DIRS 149395] Entire Table 6-1, Section 
6.2.17 
Support for the conclusion that hydrocarbon production is not a problem with carbide 
fuels. 
Quinones et al. 1996 [DIRS 
161925] 
Entire Table 6-1 & Section 
6.2.24 
Support for the conclusion that the large discrepancy in Np solubility is due to the 
incorporation of actinides into secondary uranium phases, that was not accounted 
for in conventional solubility calculations. 
Reilly 1998 [DIRS 149433] p. 30 Table 6-1, Section 
6.2.12 
It is possible that additional U-hydride will form in N Reactor SNF during interim 
storage. 
Schulz 1972 [DIRS 159406] Entire Table 6-1, Section 
6.2.12 
Description and analysis of the pyrophoric behavior of N Reactor SNF during 
reprocessing operations. 
Shoesmith and Sunder 1992 
[DIRS 113368] 
Entire Table 6-1, Section 
6.2.37 & Appendix B 
Support for the conclusion that water will not intrude until gamma and beta emitters 
have decayed to very low concentrations 
Sunder and Shoesmith 1991 
[DIRS 143815] 
Entire Table 6-1, Section 
6.2.37 & Appendix B 
Support for the conclusion that water will not intrude (i.e., the waste container will 
not fail) until 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. 
Entire Table 6-1, Section 6.2.5 Support for the conclusion that that overall repository performance is very insensitive 
to the DSNF degradation rate. 
Thornton 1998 [DIRS 
125082] 
Entire Section 6.2.16, Table 6- 
1 
Examples of the potential for pyrophoric behavior in uranium metal and uranium 
metal-based spent nuclear fuel. 
Thornton 1998 [DIRS 
107796] 
Entire Table 6-1, Sections 
6.2.5, 6.2.15, 6.2.16 
DSNF contribution to postclosure dose is negligible. 

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Table 6-1. Indirect Input (Continued) 
Technical Product Input Used From Used In Input Description 
Werme and Spahiu 1998 
[DIRS 113466] 
Entire Table 6-1, Section 
6.2.24 
Support for the conclusion that the large discrepancy in Np solubility is due to the 
incorporation of actinides into secondary uranium phases, that was not accounted 
for in conventional solubility calculations. 
Wick 1980 [DIRS 143651] Table 7.1 Tables 6-1, 6.2-1 & 
Section 6.2.8, 
Density of pure plutonium metal 
Wronkiewicz 1993 [DIRS 
171709] 
Abstract Section 6.2.38, Table 6- 
1 
Impacts of radiation on glass dissolution 

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ANL-WIS-MD-000009 REV 02 6-8 October 2004 
6.1.2 Feature, Event, and Processes Screening 
The second step of FEP analysis is screening, which addresses Acceptance Criterion 2 of Yucca 
Mountain Review Plan, Final Report (NRC 2003 [DIRS 163274], Section 2.2.1.2.1.3). The 
TSPA-LA FEP screening process is described in The Development of the TSPA-LA Features, 
Events, and Processes (BSC 2004 [DIRS 168706], Section 4). 
For FEP screening, each FEP is filtered by specified exclusion criteria (Section 4.2.3), as 
summarized in the three following statements: 
1. FEPs having less than one chance in 10,000 of occurring over 10,000 years may be 
excluded (screened out) from the TSPA-LA on the basis of low probability (as per 
10 CFR 63.114(d) [DIRS 156605]). 
2. FEPs whose omission would not significantly change the magnitude and time of the 
resulting radiological exposures to the RMEI, or radionuclide releases to the accessible 
environment, may be excluded (screened out) from the TSPA-LA on the basis of low 
consequence (as per 10 CFR 63.114 (e and f) [DIRS 156605]). 
3. FEPs that are inconsistent with the characteristics, concepts, and definitions specified 
in 10 CFR Part 63 [DIRS 156605] may be excluded (screened out) from the TSPA-LA 
by regulation. 
A FEP need only satisfy one of the exclusion screening criteria to be excluded from TSPA-LA. 
A FEP that does not satisfy any of the exclusion screening criteria must be included (screened in) 
in the TSPA-LA model. 
This report documents the screening decisions for the waste form FEPs. In cases where a FEP 
covers multiple technical areas and is shared with other FEP reports, this report provides only a 
partial technical basis for the screening decision as it relates to waste form issues. The full 
technical basis for these shared FEPs is addressed, collectively, by all of the sharing FEP reports. 
Documentation of the screening for each FEP is provided in Section 6.2. The following 
standardized format is used. 
Section 6.2.x FEP Name (FEP Number) 
FEP Description: This field describes the nature and scope of the FEP under consideration. 
Screening Decision: Identifies the screening decision as one of: 
 Included 
 Excluded  Low Probability 
 Excluded  Low Consequence 

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ANL-WIS-MD-000009 REV 02 6-9 October 2004 
 Excluded  By Regulation. 
In a few cases, a FEP may be excluded by a combination of two criteria (e.g., low probability 
and low consequence), but the final decision can never be listed as included and excluded. 
Screening Argument: This field is used only for excluded FEPs. It provides the discussion for 
why a FEP has been excluded from TSPA-LA. 
TSPA Disposition: This field is used only for included FEPs. It provides the consolidated 
discussion of how a FEP has been included in TSPA-LA, making reference to more detailed 
documentation in other supporting technical reports, as applicable. 
Supporting Reports: This field is only used for included FEPs. It provides the list of 
supporting technical reports that identified the FEP as an included FEP and contain information 
relevant to the implementation of the FEP within the TSPA-LA model. This list of supporting 
technical reports provides traceability of the FEP through the document hierarchy. For excluded 
FEPs, it is indicated as not applicable (N/A). 
6.1.3 Supporting Reports and Inputs 
The direct inputs used for the screening arguments are identified in Table 4-1. Indirect inputs are 
listed in Table 6-1. These are inputs that are primarily used for reference only and do not have 
direct inputs. 
6.1.4 Qualification of Unqualified Direct Inputs 
Direct inputs are listed in Section 4.1. Any data that required qualification are identified in 
Appendix B. 
6.1.5 Assumptions and Simplifications 
For included FEPs, the TSPA-LA dispositions may include statements regarding assumptions 
made to implement the FEP within the TSPA-LA model. Such statements are descriptive of the 
manner in which the FEP has been included and are not used as the basis of the screening 
decision to include the FEP with the TSPA-LA model. 
As the individual FEPs are specific in nature, any discussion of applicable mathematical 
formulations, equations, algorithms, numerical methods, or idealizations or simplifications are 
provided within the individual FEP discussions in Section 6.2. 
6.1.6 Intended Use and Limitations 
The intended use of this report is to provide FEP screening information for a project-specific 
FEP database and to promote traceability and transparency regarding FEP screening. This report 
is to be used as the source documentation for the FEP database described in The Development of 
the TSPA-LA Features, Events, and Processes (BSC 2004, [DIRS 168706]). For included FEPs, 
this report summarizes and consolidates the method of implementation of the FEP in TSPA-LA 
in the form of disposition statements, based on more detailed information in the listed supporting 

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ANL-WIS-MD-000009 REV 02 6-10 October 2004 
technical reports. For excluded FEPs, this report also provides the technical basis for exclusion 
in the form of screening arguments. 
Inherent in this evaluation approach is the limitation that the repository will be constructed, 
operated, and closed according to the design used as the basis for the FEP screening and in 
accordance with NRC license requirements. This is inherent in performance evaluation of any 
engineering project, and design verification and performance confirmation are required as part of 
the construction and operation processes. The results of the FEP screening presented herein are 
specific to the repository design evaluated in this report for TSPA-LA. 
Any changes in direct inputs listed in Section 4.1, in baseline conditions used for this evaluation, 
or in other subsurface conditions, will need to be evaluated to determine if the changes are within 
the limits stated in the FEP evaluations. Engineering and design changes are subject to 
evaluation to determine if there are any adverse manner impacts to safety as codified at 
10 CFR 63.73 Subparts F and G [DIRS 156605]. See also the requirements at 10 CFR 63.44 and 
10 CFR 63.131 ([DIRS 156605]). 
6.2 WASTE FORM FEPS SCREENING AND ANALYSES 
Screening information for each of the 40 waste form FEPs is presented in separate subsections 
addressed in numeric order based on the FEP number. 
For many of the FEPs addressed in this report, it was determined that the probability of a 
consequential event or condition occurring during the 10,000-year period following permanent 
closure is extremely low. However, considering the limited availability of data, analyses of the 
potential impacts of such low probability events or conditions was performed in lieu of a 
quantitative probabilistic evaluation. These analyses resulted in a determination that even in the 
unlikely event that on a probability weighted basis the event or condition should occur, it would 
not have a significant consequential effect on the magnitude and time of the resulting 
radiological exposures to the reasonably maximally exposed individual (RMEI), or radionuclide 
release to the environment. As such, these FEPs have been excluded due to low consequences as 
permitted per 10 CFR 63.114 (e and f) ([DIRS 156605]). In many of the exclusion argument 
discussions provided below, the low probability weighting of the FEP provides additional 
support to the exclusion argument and the FEP is excluded on the basis of low consequence. 
6.2.1 Waste Inventory (2.1.01.01.0A) 
FEP Description: 
The waste inventory includes all potential sources of radio toxicity and chemical toxicity. It 
consists of the radionuclide inventory (typically in units of curies), by specific isotope, and the 
non-radionuclide inventory (typically in units of density or concentration), comprised of physical 
(e.g., CSNF, DNSF, and HLW) and chemical waste constituents. The radionuclide composition 
of the waste will vary due to initial enrichment, burn-up, the number of fuel assemblies per waste 
package, 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. 

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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 2004 [DIRS 170022]) and is reproduced in Table 4-13 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 20,000 years (for TSPA-LA calculations) and from 20,000 to 1,000,000 years (for EIS 
calculations). Nonradioactive waste is not part of the disposal inventory for the repository. 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-LA: 227Ac, 241Am, 243Am, 14C, 245Cm, 135Cs, 137Cs, 
129I, 237Np, 231Pa, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 226Ra, 228Ra, 90Sr, 99Tc, 229Th, 230Th, 232Th, 
232U, 233U, 234U, 235U, 236U, and 238U (BSC 2004 [DIRS 170022], Table 6-2). Four additional 
radionuclides were added to the inventory to retain the capability to perform the million-year EIS 
calculations: 36Cl, 210Pb, 79Se, and 126Sn (BSC 2004 [DIRS 170022], Table 6-2). 
Nominal average inventories of radionuclides for each type of waste are documented in Initial 
Radionuclide Inventories (BSC 2004 [DIRS 170022]). 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 three waste types (CSNF, DSNF, and HLW). 
Table 7-1 of Initial Radionuclide Inventories (BSC 2004 [DIRS 170022]) 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 2004 [DIRS 170022], Appendix I.5). 
The weighted average waste inventory values (grams per package for each radionuclide for each 
waste type) (BSC 2004 [DIRS 170022], Table 7-1) along with the uncertainty multipliers from 
Table 7-2) are input to GoldSim for use in the TSPA-LA model. 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: 
Radionuclide Screening (BSC 2002 [DIRS 160059]) 
Initial Radionuclide Inventories (BSC 2004 [DIRS 170022]) 

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ANL-WIS-MD-000009 REV 02 6-12 October 2004 
6.2.2 Interactions Between Co-located Waste (2.1.01.02.0A) 
FEP Description: 
Co-location refers to the disposal of CSNF, DSNF, HLW, and possibly other wastes in close 
proximity within the repository. Co-location might affect thermal outputs, chemical interactions, 
or radionuclide mobilization. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
The following waste form types are represented in TSPA-LA (BSC 2004 [DIRS 170022], 
Section 7): 
 CSNF from boiling-water and pressurized-water reactors 
 DSNF (including naval SNF), approximately 85 percent by weight MTHM of which is 
the N Reactor SNF currently stored at Hanford (DOE 2002 [DIRS 158405]) 
 HLW in the form of glass logs in stainless steel canisters. 
The HLW and the DSNF are combined in specially designed waste packages (codisposal) for 
emplacement. Codisposal refers to the disposal of different waste types within the same waste 
package. Each codisposal waste package will contain either two glass logs and two multicanister 
overpacks for N Reactor SNF or five glass logs and one DSNF canister for other DOE SNF 
codisposal packages (DOE 2002 [DIRS 158405], Section 9). The codisposal waste 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 (CRWMS M&O 2000 [DIRS 147650], Table 8-1; 
CRWMS M&O 2000 [DIRS 147651], Table 8-1) are expected to be generally cooler than CSNF 
waste packages (BSC 2004 [DIRS 166695], Table 21) at the time of disposal. Condensation due 
to this co-located waste could possibly occur under the drip shield just above the relatively 
cooler co-disposal (DSNF and HLW) waste packages (BSC 2004 [DIRS 164327], Section 
6.3.7.2.3). However, the amount of condensation predicted to form on the underside of the drip 
shield was not determined to be a significant factor and has been excluded from TSPA (BSC 
2004 [DIRS 169898], Section 6.2.43). 
Because the distribution of waste packages is horizontal and the major flow direction in a 
fractured rock is vertical under the influence of gravity, infiltrating groundwater will move 
downward through the repository tunnels and into the underlying fractured rock. Contaminants 
contributed by breached waste packages will move into the underlying rock and will not be 
available to react with other waste packages along the tunnel. Hence, significant changes in the 
rates of waste package degradation and radionuclide mobilization due to chemical reactions 
between colocated waste packages are not expected and chemical interactivity between colocated 
waste packages can, therefore, be excluded based on low consequence. 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-13 October 2004 
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 nondesign 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: 
N/A 
6.2.3 Interactions Between Co-disposed Waste (2.1.01.02.0B) 
FEP Description: 
Co-disposal 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 co-disposal 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. 
Screening Decision: 
Included 
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 codisposal waste packages do not contain navy spent nuclear fuel, 
because naval fuel is emplaced separately. 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 accounts 
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 169988], Section 8.2.1), but the SNF 
upper-limit degradation model assumes complete dissolution in one TSPA-LA time step and is 
thus not sensitive to exposed surface area (BSC 2004 [DIRS 167618], Section 6.1.12). The 
geometric surface area used to represent all HLW glass in the HLW glass model (BSC 2004 
[DIRS 169988], 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. 

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ANL-WIS-MD-000009 REV 02 6-14 October 2004 
Codisposal waste packages are specially designed waste packages that will contain the HLW 
glass and DSNF waste forms. 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 2004 [DIRS 170022], 
Table 6-3). 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 vapor influx or seepage influx with DSNF and HLW glass in a breached 
codisposal waste package were simulated in In-Package Chemistry Abstraction (BSC 2004 
[DIRS 167621]), which uses degradation models for DSNF (BSC 2004 [DIRS 167618] and 
HLW glass (BSC 2004 [DIRS 169988]) 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 co-disposal waste package. 
Degradation of HLW glass tends to generate high pH solutions (BSC 2004 [DIRS 167621], 
Section 6.5), while DSNF degradation generates neutral pH solutions (BSC 2004 
[DIRS 167621], Section 6.5). 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 2004 [DIRS 167621]). 
The resulting in-package water chemistry is provided to TSPA-LA as part of In-Package 
Chemistry Abstraction (BSC 2004 [DIRS 167621]). The abstraction provides TSPA-LA 
parameters for the pH, Eh, ionic strength, and the total carbonate, chloride, and fluoride 
concentrations under vapor influx and water influx 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 2004 [DIRS 167618]). 
The glass degradation rate model contains an explicit pH dependence and will be affected by 
changes in the solution pH within a codisposal package (BSC 2004 [DIRS 169988]). 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: 
In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621]) 
DSNF and Other Waste Form Degradation Abstraction (BSC 2004 [DIRS 167618]) 
Defense HLW Glass Degradation Model (BSC 2004 [DIRS 169988]) 
6.2.4 Heterogeneity of Waste Inventory (2.1.01.03.0A) 
FEP Description: 
CSNF, DSNF, and HLW shipped to the repository may contain quantities of radionuclides that 
vary from waste package to waste package, fuel assembly to fuel assembly, and canister to 
canister. The composition of each of these waste forms may vary due to initial uranium 
enrichment, possible plutonium enrichment, and fuel burn-up, among other factors. The physical 
state within the waste form may also vary. For example, damaged fuel pellets or extremely high

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ANL-WIS-MD-000009 REV 02 6-15 October 2004 
burn-up fuels may have greater surface area exposed to any water penetrating a waste package 
than undamaged, low burn-up 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 
based on waste-package composition. This could affect repository chemistry, breach times, 
dissolution rates, and availability of radionuclides for transport. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
As discussed in Initial Radionuclide Inventories (DOE 2004 [DIRS 170022]) the repository 
waste types are quite heterogeneous in type (SNF versus glass) and in inventory per package 
including uranium enrichment, possible plutonium enrichment, and fuel burn-up. 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-LA, the only simulations that result in doses to 
the RMEI that approach the regulatory limits are those simulations that postulate the premature 
breaching of multiple waste packages. 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 presented in BSC 2003 [DIRS 170022], Tables 7-1 and 7-2). 
As discussed in Defense HLW Glass Degradation Model (BSC 2004 [DIRS 169988]), 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 169988], Section 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 169988]) can be used with the 
average radionuclide concentrations for the entire HLW inventory. 

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ANL-WIS-MD-000009 REV 02 6-16 October 2004 
Supporting Documents: 
Defense HLW Glass Degradation Model (BSC 2004 [DIRS 169988]) 
Initial Radionuclide Inventories (BSC 2003 [DIRS 170022]) 
6.2.5 DSNF Degradation (Alteration, Dissolution, and Radionuclide Release) 
(2.1.02.01.0A) 
FEP Description: 
DSNF to be disposed in Yucca Mountain contains 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 CSNF, and the degradation, alteration, and dissolution may be 
different from the 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 on 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 are classified into eleven groups 
DOE Spent Nuclear Fuel Grouping in Support of Criticality, DBE, TSPA-LA (DOE 2000 
[DIRS 118968] Section 8.1), for TSPA-LA these are being treated as two types, those waste 
packages containing naval fuel and all the rest (BSC 2004 [DIRS 167618], Section 8.1). The 
development of upper-limit and best-estimate degradation models for TSPA-LA are described in 
DSNF and Other Waste Form Degradation Abstraction (BSC 2004 [DIRS 167618] Section 6.1). 
The largest single DSNF type is the N-Reactor SNF, which comprises approximately 85 percent 
by weight MTHM of the total DSNF (DOE 2002 [DIRS 158405]). For TSPA-LA waste 
packages containing naval fuel are conservatively treated as CSNF (BSC 2004 [DIRS 167618], 
Section 8.1). 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 167618], Section 8.1). 
Up to 50 percent of the DSNF cladding may be already perforated (Rechard 1995 
[DIRS 101084], Section 11.3.1, p. 11 to 24). Because the cladding integrity of most DSNF will 
not be extensively characterized, the TSPA-LA takes no credit for delayed degradation of the 
DSNF due to DSNF cladding or delayed release of radionuclide by the DSNF cladding 
(FEP 2.1.02.25.0A). Also, the TSPA-LA takes no credit for DSNF canister integrity (i.e., once 
the waste package has failed, it is conservatively assumed that the DSNF is directly exposed to 
the water or air of the repository environment). 

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ANL-WIS-MD-000009 REV 02 6-17 October 2004 
The recommended upper-limit model to be used in TSPA-LA analyses is instantaneous 
degradation or dissolution of the waste form upon exposure of the waste form to groundwater. 
This is chosen as the upper-limit model in part because TSPA-LA analyses performed for the 
DSNF (Thornton 1998 [DIRS 125082]; 1998 [DIRS 107796]) have shown that the overall effect 
of the failure of DSNF-containing waste packages does not significantly contribute to the 
repository site boundary dose. 
Expected releases from naval SNF waste packages were provided in Performance Assessment of 
U.S. Department of Energy Spent Fuels in Support of Site Recommendation (BSC 2001 
[DIRS 152059]). Because of the robust design of naval SNF, the radionuclide releases from 
naval SNF waste packages are considerably less than releases from commercial light water 
reactor SNF waste packages; accordingly, the treatment of waste packages containing naval fuels 
be treated as CSNF is conservative. Furthermore, naval SNF represents approximately one-tenth 
of one percent of the SNF MTHM inventory. 
Supporting Documents: 
DSNF and Other Waste Form Degradation Abstraction (BSC 2004 [DIRS 167618]) 
6.2.6 CSNF Degradation (Alteration, Dissolution, and Radionuclide Release) 
(2.1.02.02.0A) 
FEP Description: 
Alteration of the original CSNF mineralogy (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, burn-up, 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. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
This FEP considers the CNSF alteration, dissolution, and radionuclide release. The outputs from 
the report (BSC 2004 [DIRS 169987], Section 8.1) are the TSPA-LA models for the 
instantaneous release fractions (fi) and for the matrix fractional release rates (Fi) under acidic and 
basic conditions. The output includes the model parameter values and the associated 
distributions that capture the uncertainty in these parameters. The outputs include the developed 
models, associated model parameter values, and uncertainty distributions for the release of: 
 Gap and grain-boundary inventory fractions of cesium, iodine, technetium, and 
strontium 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-18 October 2004 
 Fuel matrix inventory under basic and acidic conditions. An instantaneous radionuclide 
release rate model is to be used for any fuel that is exposed to humid air at temperatures 
greater than 100C and is subsequently contacted by water. 
The release of the gap and grain-boundary inventories of cesium, iodine, technetium, and 
strontium is modeled as an instantaneous release of the fraction (fi) of the total inventory of each 
of these elements to be in the gap and grain-boundary regions. 
As described in CSNF Waste Form Degradation: Summary Abstraction (BSC 2004 
[DIRS 169987], Sections 6.1 and 6.2), 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 mathematical form of the model has six model 
parameters (A, a0, a1, a2, a3, and a4) and four independent variables (absolute temperature, pCO3, 
pO2, and pH) (BSC 2004 [DIRS 169987], Section 1, 6.4.1). The effects of surface area are 
included through the effective specific surface area parameter (A) (BSC 2004 [DIRS 169987], 
Section 6.4.1.5). Potential fuel oxidation may increase the fuel surface area. The increase in 
surface area is accounted for by modeling an instantaneous release for any fuel in a waste 
package that breaches above 100C (BSC 2004 [DIRS 169987], Section 8.2). Potential 
oxidation of fuel during preclosure handling options has been shown to have negligible 
consequences to the reasonable maximally exposed individual (RMEI) dose 
(DTN: MO0410MWDTLCSC.000 [DIRS 172057]) and is not modeled. 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 169987], 
Section 6.2.2.3). In summary, the burnup has a negligible effect on the CSNF degradation rate. 
Secondary phase formation and the resultant armoring decrease the degradation rate and are not 
included in the conservative analysis. 
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-LA 
simulation. The model is based on groundwater flow-through experiments using a simulated 
J-13 groundwater from the vicinity of Yucca Mountain. 
Supporting Documents: 
CSNF Waste Form Degradation: Summary Abstraction (BSC 2004 [DIRS 169987]) 
6.2.7 HLW Glass Degradation (Alteration, Dissolution, and Radionuclide Release) 
(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, co-precipitation of other minerals (including iron corrosion 
products), and selective leaching. 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-19 October 2004 
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 169988]) 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 (BSC 2004 [DIRS 169988], Section 6.5.1 and Equation 13): 
rateglass = kE 10.pH exp(-Ea/RT) 
where 
kE = effective rate constant 
. = order of reaction with respect to H+ 
Ea = activation energy 
R = universal gas constant 
T = temperature. 
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-LA calculations, including pH, temperature, and relative 
humidity. 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 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 and those values are 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 glass 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 169988]) 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 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-20 October 2004 
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 169988], 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 surface area of glass that is available 
for corrosion, and the average radionuclide inventory (BSC 2004 [DIRS 169988], Table 7.1), are 
calculated based on the characteristics and predicted numbers of glass logs to be received from 
different production facilities. 
In TSPA-LA, the HLW glass degradation rate is used along with the in-package chemistry to 
calculate a HLW glass degradation rate at each model time step. The degradation rate model 
parameters are determined by sampling from parameter distributions that incorporate 
uncertainty, while the in-package chemistry parameters are calculated as part of the TSPA-LA 
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: 
Defense HLW Glass Degradation Model (BSC 2004 [DIRS 169988]) 
6.2.8 Alpha Recoil Enhances Dissolution (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. 
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 are shown in the following discussion to be 
much less than the dissolution rate due to chemical processes. Even when conservatively 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-21 October 2004 
assuming that all radioactive decay results in alpha recoil, it does not cause significant increases 
in the dissolution rate of the different waste forms. 
Of the various radioactive decay modes (i.e., alpha (a), 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, but are orders of magnitude less than that of alpha decay. Also, information from 
Lederer and Shirley (1978 [DIRS 142133], p. 1,464) indicates that the spontaneous fission halflives 
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: 
0 
m
d 
T 
E 
Disp(E) P(T )K(E,T ) (T )dT P(T )K(E,T ) (T )dT 8 = . = . . . (Eq. 1) 
where 
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. Because 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 0
1 
d 
d 
d 
if T E 
P(T ) T E 
if T E 
< . 
= - =.
= . 
(Eq. 2) 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-22 October 2004 
To simplify the analysis, the probability for kinetic energy transfer is treated as being a uniform 
distribution over the applicable energy range: 
1 
m d 
K( E,T ) 
T E 
= 
- 
(Eq. 3) 
The Kinchin-Pease Model (Foster and Wright 1973 [DIRS 144061], p. 297) describes the total 
number of displacements that originate from a primary knock-on: 
0 2
2 2 2 
d i i 
i 
i d i d d 
T / E if T E E T T (T ) T E 
E / E if T E E E 
. 
= . - 
= + - =.
> . 
(Eq. 4) 
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 
reasonably concludes there is an ionization threshold (EI  1,000A) 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: 
2 
2 2 
2 2 
2 2 2 2 
m
d 
i m 
d i 
T 
m i E 
d m d 
m 
E T d 
i E E 
d m d d m d 
T dT if T E 
E T E 
Disp( t ) 
E T dT dT if T E 
E T E E T E 
. = . - . = ..
+ > . - - . 
. 
. . 
2 2 2 
4 
4 2 
m d 
m i 
d 
i d im i 
m i 
d m d d m d 
T E if T E 
E 
E E ET E if T E 
E (T E ) E (T E ) 
. + 
= ..
= . 
- - . + > . - - . 
(Eq. 5) 
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 
Appendix I, 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 1,000A equal 
to 0.234 MeV). This results in Dispmax(a) 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 which 
recoil nuclei (due to alpha decay) could pass through and enter the surrounding seepage water is 
approximately 10. This is only for recoil atoms traveling in the direction of the water. 
Appendix A indicates that only half of the recoil could be in the proper direction. This also does 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-23 October 2004 
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 Table of Isotopes (Lederer and 
Shirley 1978 [DIRS 142133]), it can be identified that the maximum energy for alpha particles 
emitted from 238U 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( a). 
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 
monolayer thickness of approximately 3.0  (3.0  10-10m). The combination of these two 
parameters results in 0.0387 grams of SNF material within the first thirteen half-monolayers of a 
surface area of 1.0 m2 that are in the direction of the material surface (19.84 g/cm3  106 cm3/m3 
 13/2 monolayers  3  10-10 m). When this surface density (0.0387 g/m2) is multiplied by the 
fractional rate at which the SNF material experiences radioactive a-decay, it can be expressed in 
the units used in the CSNF (BSC 2004 [DIRS 169987] and DSNF (BSC 2004 [DIRS 167618]) 
degradation analysis reports for dissolution of CSNF and DSNF material due to chemical 
processes. The fractional rate was determined from radiological half-lives along with inventory 
data from Initial Radionuclide Inventories (BSC 2004 [DIRS 170022], Table 7-1). The 
inventory data identifies the major constituents on a mass basis of the final waste forms (SNF 
and HLW): 238U, 235U, 239Pu, 236U, and 240Pu. These major radionuclides 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 (decay rate = 0.0387 g/m2  half-life) by noting that 
the fractional decay rate for the radionuclides is given by their decay constant (. = ln(2)/t). To 
put these values from Table 6.2-1 into proper perspective, a comparison to the dissolution rates 
of different waste forms due to chemical (non-nuclear) processes must be made. CSNF flowthrough 
test dissolution data show rates ranging from 0.58 to 109 mg/m2day (BSC 2004 
[DIRS 169987], Tables 4-2 and 4-4). When converted to the same units as Table 6.2-1 by 
using 365.25 days/yr, the rates are 0.21 to 39.8 g/m2y. For DSNF, the major type of SNF is 
uranium-based metal, which has a corrosion rate of 1.1  105 mg/m2day (40,200 g/m2y) 
(BSC 2004 [DIRS 167618], Table 6-2). As demonstrated above, 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 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-24 October 2004 
processes has been eliminated from performance assessment calculations, on the basis of low 
consequence. 
Table 6.2-1. Alpha Recoil Enhanced (From a and a-Recoil Atom) Dissolution Rates Due to the Major 
Mass-Based Constituents of SNF and HLW to be Disposed of in the Yucca Mountain 
Repository 
Nuclide 
ID Decay Mode 
HalfLife (a) 
(years) 
Fraction Decay Rate (b) 
(1/yr) 
aDecay Rate in 
13 Half-Mono- 
Layers(c) (g/m2yr) 
238U 
235U 
239Pu 
236U 
240Pu 
a, ., SF 
a, ., SF 
a, ., SF 
a, ., SF 
a, ., SF 
4.47 x 109 
7.04 x 108 
2.410 x 104 
2.342 x 107 
6.56 x 103 
1.55 x 10 10 
9.85 x 10-10 
2.88 x 10-5 
2.96 x 10-8 
1.06 x 10-4 
6.00 x 10-12 
3.81 x 10-11 
1.11 x 10-6 
1.15 x 10-9 
4.09 x 10-6 
NOTES: a Data obtained from Parrington et al. 1996 [DIRS 103896], p. 48. 
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  (3.0x10-10m), and the density is upper bounded at 
19.84 g/cm3 (theoretical density of pure plutonium metal ([Wick 1980 [DIRS 143651]). 
TSPA Disposition: 
N/A 
Supporting Documents: 
None 
6.2.9 HLW Glass Cracking (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. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
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 169988]). 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. 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-25 October 2004 
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 169988], Section 6.5.6, 
Equation 44): 
S0 = fexposure  (2p ro
2 + 2p ro  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 169988], Section 6.5.6, Equation 48): 
S = fexposure  2.70  10-3 m2/kg  (2,710 kg - S Mt kg) (Eq. 7) 
where S M gives the mass loss during all previous time steps (and resulting loss of surface area). 
The value of S 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: 
Defense HLW Glass Degradation Model (BSC 2004 [DIRS 169988]) 
6.2.10 HLW Glass Recrystallization (2.1.02.06.0A) 
FEP Description: 
HLW glass recrystallization could occur and would lead to a less corrosion-resistant waste form. 
Recrystallization is a slow process and typically occurs only if a high glass temperature is 
maintained over a prolonged period. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
The glass degradation model developed in Defense HLW Glass Degradation Model (BSC 2004 
[DIRS 169988]) utilizes a range of model parameter values determined from experimentally 
measured dissolution rates of glasses having compositions similar to HLW glasses (BSC 2004 
[DIRS 169988], Section 6.5.2.2). 
Because of the uncertainties associated with the different glass compositions that will be 
disposed of in the repository, the value of kE (the effective dissolution constant) takes into 
account the effects of the glass composition, including the heterogeneity of the waste inventory, 
as well as the effect of the solution compositions (BSC 2004 [DIRS 169988], Section 6.5.1). 
As part of model validation for the glass degradation model, the glass degradation rate was 
shown to be insensitive to the presence of crystalline phase either formed by devitrification of 
the glass melt during manufacture or intentionally added to the waste form. This is based on 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-26 October 2004 
literature results that show the effects of devitrification on glass dissolution rate are small and 
near the uncertainty in the measurement of the rate (BSC 2004 [DIRS 169988], Section 6.5.5). 
This process has no affect on the overall dissolution rates and is excluded from TSPA-LA based 
on low consequence. 
TSPA Disposition: 
N/A 
Supporting Documents: 
N/A 
6.2.11 Radionuclide Release from Gap and Grain Boundaries (2.1.02.07.0A) 
FEP Description: 
While in the reactor at 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 fuel assembly or waste package infill such as I and Pb may affect this 
process. 
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 and-grain-boundary inventory is 
assumed in the TSPA-LA 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 169987]), 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 169987], 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-LA 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 radionuclides to the gap and grain boundaries occurs. 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-27 October 2004 
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 169987], Section 6.2.1). 
Supporting Documents: 
CSNF Waste Form Degradation: Summary Abstraction (BSC 2004 [DIRS 169987]) 
6.2.12 Pyrophoricity from DSNF (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 behavior of the 
system and could contribute to degradation of the waste package, waste form, and cladding. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
DSNF may potentially include pyrophoric material, particularly the uranium metal-based 
N Reactor SNF currently being removed from the K-basin and placed in dry storage in the 
Canister Storage Building at Hanford. There is extensive literature concerning the pyrophoric 
behavior of uranium-based SNF in general, and N Reactor SNF in particular; there are no known 
instances of pyrophoric behavior in oxide or carbide-based spent fuels. Thus, although the 
DSNF inventory consists of spent fuel groups (DOE 2000 [DIRS 118968]) that include metal-, 
oxide-, and carbide-based fuels, only the uranium metal-based N Reactor spent fuel group 
(which accounts for about 85 percent by weight MTHM of the total DSNF inventory (DOE 2002 
[DIRS 158405], Appendix D)) need be considered pyrophoric for the purposes of this analysis. 
Uranium metal and uranium hydride are, but uranium oxides and carbide are not, regarded as 
pyrophoric materials in the DOE handbook, Primer on Spontaneous Heating and Pyrophoricity 
(DOE-HDBK-1081-94 [DIRS 103327]). 
Metallic uranium-based fuels will oxidize in an air or water environment, producing uranium 
hydride and oxide as corrosion products. The uranium hydride corrosion product can be 
pyrophoric, particularly when in a fine particulate form or when a significant amount of hydride 
has formed in the uranium metal matrix. Examination and testing conducted at PNNL on 
damaged N Reactor zirconium-clad uranium metal fuel showed small but discernable amounts of 
uranium hydride formed as precipitates within the metal (Marschman et al. 1997 [DIRS 149429], 
Section 3.4.2). Their presence is believed to be responsible for the decreased ignition 
temperature observed during ignition testing of damaged/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. 
N Reactor SNF is composed primarily of zirconium-clad metallic uranium. However, a 
significant amount of the cladding has been breached, thereby exposing the metallic uranium to 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-28 October 2004 
the K-basin water (and any other subsequent canister or dry storage) environment. Metallic 
uranium-based SNF has at times shown pyrophoric behavior when exposed to similar water or 
air environments in the past (Abrefah et al 1999 [DIRS 151226], Appendix C). The N Reactor 
SNF must, therefore, be assessed for its potential for, and consequences of, pyrophoric behavior 
under repository conditions. 
Pyrophoric materials are materials that can spontaneously ignite due to overheating as a result of 
oxidation, sparks, or mechanical trauma. This can 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, 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 result of a pyrophoric event could be the formation of colloids and 
subsequent transport that is discussed in Section 6.2.31. 
Uranium hydride inclusions in the metallic uranium matrix of damaged N Reactor SNF (i.e., fuel 
elements with damaged cladding) provides 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-basin 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]), the 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. 
In order for spontaneous ignition of metallic uranium or U-hydride, or both, inclusions to occur, 
there must be sufficient oxygen to support continuing combustion. This is unlikely, since waste 
packages containing potentially pyrophoric material have been dried and filled with an inert gas 
prior to emplacement in the repository. Furthermore, the N Reactor SNF contained within 
canisters (known as multicanister overpacks) emplaced inside the waste packages are themselves 
also dried and 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 initial breach would be expected to be small 
enough that oxygen would leak slowly into the waste package. The slow introduction of oxygen 

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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 because the potential adverse effects of a 
pyrophoric event are insignificant. 
Analyses have been performed in Performance Assessment Sensitivity Analyses of Selected U.S. 
Department of Energy Spent Fuels (CRWMS M&O 2000 [DIRS 125038]) concerning the 
sensitivity of the TSPA and in Performance Assessment of a Potential Post-Closure Pyrophoric 
Event Involving Uranium Metal Spent Fuel (BSC 2001 [DIRS 171508]) concerning the 
consequences to the boundary dose of a postclosure pyrophoric event. 
Performance Assessment Sensitivity Analyses of Selected U.S. Department of Energy Spent Fuels 
(CRWMS M&O 2000 [DIRS 125038]) documented the results of TSPA-VA sensitivity analyses 
for failed co-disposal waste packages, and included studies of the sensitivity of the boundary 
dose to the rate of degradation of DSNF. The study examined both a realistic degradation model 
and an instantaneous release model that would be typical of a waste package pyrophoric event. 
The analyses were performed using the waste package degradation (WAPDEG) analysis in the 
repository integration model (RIP) code. The results of the analysis showed that the total dose 
rates from the instantaneous release of the DSNF of the waste package are below that of the total 
dose rate from the TSPA-VA base case (CRWMS M&O 2000 [DIRS 125038], Section 6.6.2). It 
was concluded that there was no significant difference in the dose history between assuming 
instantaneous release or using a surrogate uranium metal-based degradation model for the DSNF. 
Performance Assessment of a Potential Post-Closure Pyrophoric Event Involving Uranium 
Metal Spent Fuel (BSC 2001 [DIRS 171508]) documented the results of a calculation (in support 
of the TSPA-SR) of the potential for a postclosure pyrophoric event in a failed codisposal waste 
to damage adjacent waste packages and consequently release radionuclides. The dose 
calculation was performed with the qualified GoldSim Version 6.04.007 code. The assumption 
used in the calculation was that the oxidation (due to the pyrophoric event) of the uranium metal 
in a failed N Reactor codisposal waste package could occur in a matter of days (as little as one 
day) after waste package failure and could involve up to two N Reactor SNF-containing canisters 
(called multicanister overpacks). The pyrophoric event was set to occur at 10,000 years after 
closure. It was also conservatively assumed that the drip shield also failed at 10,000 years. In 
the most severe scenario examined  that of a pyrophoric event in a codisposal waste package 
simultaneously failing two adjacent CSNF-containing waste packages  the maximum boundary 
dose (at about 40,000 years) still was less than 5 mrem/yr (BSC 2001 [DIRS 171508], 
Figures 6.2-1 through 6.2-7). 
Both documents (CRWMS M&O 2000 [DIRS 125038]; BSC 2001 [DIRS 171508]) have been 
cancelled in the TDMS since they were strictly applicable only to, respectively, the TSPA-VA 
and TSPA-SR rather than the TSPA-LA. Also, the N Reactor SNF degradation models (other 
than the instantaneous release model) used in these analyses were not the models used for 
TSPA-LA. However, the use of these documents in this waste form FEPs modeling report is 
qualitative and so they may be regarded as validated for this purpose. 

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These analyses show 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 because 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 cause, at most, 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 
6.2.13 Chemical Effects of Void Space in Waste Package (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. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
This FEP is to evaluate the effects of the unfilled void space on the in-package chemistry. The 
base-case conceptual model for in-package chemistry is composed of two conceptual models, the 
vapor flux model and the water flux model, which are collectively referred to as the in-package 
chemistry model in In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621] Section 6.3). 
The vapor influx model addresses the situation in which water vapor enters the breached waste 
package and condenses. The water influx 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 void 
volume in the waste form. The in-package chemistry model accounts for the unfilled void space 
in the waste package and, thus, its impact on the in-package chemistry. 
In the conceptual model, the void spaces within a waste package are partially filled by liquid 
water in equilibrium with atmospheric gasses. Void space in the waste package is, thus, partially 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-31 October 2004 
gas filled. This has the effect of putting the gas phase in close contact with the liquid- and solidphase 
reactants in the system, and is the basis of the conceptual model implemented in 
In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621]). The in-package chemistry 
model uses a batch reactor model to model bulk water. The inputs to this model are scaled 
according to their surface area, and the base case considers the waste package void space to be 
approximately half filled with water. The other half of the void space is filled with air in 
equilibrium with the drift atmosphere (Section 6.3.1). Sensitivity studies were conducted that 
varied the ratio of water to air (Section 6.6.1) and that varied the composition of the air in contact 
with the water (Section 6.6.4). Therefore, the chemical effects of the void space, both the 
composition and quantity of void space, have been investigated and incorporated into the model. 
The effects of the unfilled void space on the in-package chemistry are passed to TSPA-LA 
explicitly via the oxygen and carbon dioxide equilibrium within In-Package Chemistry 
Abstraction (BSC 2004 [DIRS 167621]). 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 (examined in Section 6.6.1) compared with that 
due to uncertainty in the waste package water inflow rates. 
Supporting Documents: 
In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621]) 
6.2.14 Organic/Cellulosic Materials in Waste (2.1.02.10.0A) 
FEP Description: 
Degradation of cellulose in the waste could affect the long-term performance of the disposal 
system. 
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). 
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 (FEP 2.1.09.13.0A). The estimates of microbe masses growing 
in the potential repository system suggest that effects to the bulk in-drift geochemistry should be 
small (CRWMS M&O 2000 [DIRS 151561], Section 7.2). The minor amounts of organic 
materials in the emplaced waste, the low levels of microbial masses, 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: 
N/A 

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ANL-WIS-MD-000009 REV 02 6-32 October 2004 
6.2.15 DSNF Cladding (2.1.02.25.0A) 
FEP Description: 
DSNF to be disposed in Yucca Mountain contains a variety of fuel types that may not be similar 
to CSNF. Some of the fuel types may have initial cladding-degradation characteristics that are 
different from those for 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. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: About 85 percent by weight MTHM of the DSNF is from the N Reactor 
(DOE 2002 [DIRS 158405], Appendix D), 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 2004 [DIRS 167618], 
Section 6.4.3), the TSPA-LA takes no credit for radionuclide containment by the cladding. Also, 
the TSPA-LA takes no credit for canister integrity (i.e., once the waste package has leaked or 
failed); it is conservatively modeled 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. 
Including the influence of cladding would delay the release of DSNF radionuclides; however, 
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, rather 
than over multiple time steps, is insignificant (Thornton 1998 [DIRS 107796]). 
Therefore, because all DSNF cladding is treated as failed, 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: 
N/A 
6.2.16 Grouping of DSNF Waste Types into Categories (2.1.02.28.0A) 
FEP Description: 
Several hundred distinct types of DSNF may potentially be stored in the repository. These 
represent many more types than can viably be examined for their individual effect on the 

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ANL-WIS-MD-000009 REV 02 6-33 October 2004 
repository. A limited number of representative or bounding degradation models must be selected 
and/or abstracted. 
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. Initial 
Radionuclide Inventories (BSC 2004 [DIRS 170022]) considered all fuel types and the quantities 
of each fuel type when compiling the nominal radionuclide inventory in grams per package and 
uncertainty distributions for radionuclides important to dose calculations for TSPA-LA. 
For degradation rate determination, 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 85 percent by weight MTHM 
of the total DSNF (DOE 2002 [DIRS 158405], Appendix D). N Reactor SNF uranium metal 
(Group 7) degradation models are adopted as a surrogate for the entire DSNF inventory, except 
for the naval SNF (Group 1) (BSC 2004 [DIRS 167618]). Waste packages containing naval fuel 
are conservatively treated as CSNF for the TSPA-LA model. 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 TSPA-LA, the in-package pH and ionic strength abstraction for the DSNF waste packages are 
based on N Reactor SNF degradation calculations. 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 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 2004 [DIRS 167618], 
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 2004 
[DIRS 167618], 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 
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 

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ANL-WIS-MD-000009 REV 02 6-34 October 2004 
percent of the total waste; therefore, waste stream plume chemistry is dominated by the CSNF 
waste even when bounding DSNF degradation models are used. 
Supporting Documents: 
DSNF and Other Waste Form Degradation Abstraction (BSC 2003 [DIRS 167618]) 
Initial Radionuclide Inventories (BSC 2004 [DIRS 170022]) 
6.2.17 Flammable Gas Generation from DSNF (2.1.02.29.0A) 
FEP Description: 
DSNF to be disposed in Yucca Mountain will contain a small percentage 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 package or host-rock properties in the 
adjacent area of the EBS. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: Flammable-gas generation based on the reaction of DSNF carbide fuel 
with liquid water or water vapor can be excluded based on low consequence for the following 
reasons: 
1. The only fuel waste types capable of producing organic flammable gasses such as 
methane, ethane, etc., are the uranium-thorium carbide and the plutonium-uranium 
carbide DOE SNF waste fuels because they are the only spent nuclear fuels containing 
more than trace quantities of carbon. These gasses 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). The 
carbide spent nuclear fuels will be present in about 5 percent of the waste packages 
(DOE 2004 [DIRS 169354], p. D-580). Additionally DSNF will be widely dispersed 
throughout the repository, as will any produced flammable gas (BSC 2004 [DIRS 
170020], Section 6.2.17). 
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 permeable 
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 

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ANL-WIS-MD-000009 REV 02 6-35 October 2004 
repository and host rock, thus diluting any potential flammable-gas concentrations to 
levels below the ignition point. 
3. 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. Considering the low probability of waste package breach, the small 
number of waste packages capable of generating flammable gas, the low probability of 
ignition due to high rock permeability, and the fact that DSNF is already modeled as 
degrading instantaneously, this FEP is excluded due to low consequence per 
10 CFR 63.114 (e)(f) [DIRS 156605]. 
The two types of U/Th carbide fuel were used in the Fort St. Vrain and Peach Bottom Unit 1, 
reactor cores 1 and 2. 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 reactor cores. 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 cores (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 amount, only 2.5 percent (0.35  7 percent) is susceptible to gas production due 
to reaction with water. Taken all together, only 0.001 percent (2.5 percent  5 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 percent (7 percent  5 percent) 
would be susceptible to gas production. 
All DSNF waste (except for naval fuel) will be codisposed among some 3,412 waste packages 
(BSC 2004 [DIRS 170022], Table 6-3). Like most of the DSNF waste packages, the carbide fuel 
will be placed in a DSNF canister and packaged with five HLW Glass canisters in each waste 
package (DOE 2002 [DIRS 161752], Section 1.5). So the carbide fuel is inherently dispersed 
and any gas produced from carbide fuel would be dispersed as well. Furthermore, because the 
percent of 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 

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ANL-WIS-MD-000009 REV 02 6-36 October 2004 
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 rock matrix at the 
repository horizon has gas saturation generally greater than 50 percent (BSC 2004 
[DIRS 170038], Table 6-6), making these areas more gas- than liquid-permeable. These flow 
paths will disperse flammable gas concentrations to levels below the ignition point. 
Additionally, airflow through Yucca Mountain maintains repository pressures close to 
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: 
N/A 
6.2.18 Internal Corrosion of Waste Packages Prior To Breach (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. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: Project Requirements Document (Canori and Leitner 2003 
[DIRS 166275], PRD-013/T-036 and PRD-013/T-045) states that waste packages will be 
designed to preclude internal corrosion of the waste package and contained material. The waste 
package design will also preclude chemical, electrochemical, or other reactions (such as internal 
corrosion) of the waste package such that there will be no adverse effect on normal handling, 
transportation, storage, emplacement, containment, or isolation, or on abnormal occurrences such 
as a waste package drop accident and premature breach in the repository. Further discussion is 
contained in FEPs Screening of Processes and Issues in Drip Shield and Waste Package 
Degradation (BSC 2004 [DIRS 169997], Section 6.2.12). 
The small amount of water in the CSNF and HLW-glass waste packages that exists initially will 
be rapidly consumed by reaction with the waste form and other waste package internal materials. 
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 

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ANL-WIS-MD-000009 REV 02 6-37 October 2004 
(BSC 2004 [DIRS 167618], Table 6-1; Gray and Einziger 1998 [DIRS 109691], Section 4), as 
well. 
In general, no credit is taken in the TSPA-LA 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 prior to breach do not impact the predicted releases from waste 
packages containing DSNF. 
TSPA Disposition: 
N/A 
Supporting Documents: 
N/A 
6.2.19 Chemical Characteristics of Water in Waste Package (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 steel and other materials used in the waste packages or waste forms, 
as well as 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. 
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 2004 [DIRS 167621]). The IPCM is a fully coupled reaction-path chemical 
model that includes the effects of waste form dissolution, metal alloy corrosion or 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 base-case IPCM and several other water compositions in the 
sensitivity analyses. These compositions were combined with various corrosion rates of the steel 
components of the waste package internals and the seepage flux. These analyses resulted in a set 
of independent variables (water composition, steel and fuel corrosion rates, and temperature) that 
were sampled to establish an uncertainty band for the determination of in-package chemistry. 
Uncertainty bands were modified as necessary to incorporate results of sensitivity runs that 
varied carbon dioxide and oxygen fugacity, oxidation state of chromium, type of iron-oxide 

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ANL-WIS-MD-000009 REV 02 6-38 October 2004 
mineral formed, composition of HLW glass and alloys. Solution parameters are abstracted into a 
TSPA-LA-compatible format in In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621]); 
therefore, all of the processes included in the IPCM are reflected in In-Package Chemistry 
Abstraction (BSC 2004 [DIRS 167621]) and thus in the TSPA-LA. The direct IPCM outputs 
include pH, ionic strength, total carbonate, Eh, chloride, and fluoride. Additional outputs of the 
IPCM include time-varying aqueous inventories of fluid composition, reactant abundances, 
product abundances, and solution properties. 
Supporting Documents: 
In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621]) 
6.2.20 Chemical Interaction with Corrosion Products (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 (i.e., rock bolts, steel in the invert, gantry rails) may affect 
the mobilization and transport of radionuclides. Corrosion products may facilitate 
sorption/desorption and co-precipitation/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. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
The IPCM model developed in In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621]) 
addresses in-package corrosion products (including gasses) and their effect on in-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 Sections 6.5 and 6.6 of In-Package Chemistry Abstraction 
(BSC 2004 [DIRS 167621]). Section 6.7 of In-Package Chemistry Abstraction (BSC 2004 
[DIRS 167621]) 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-LA as part of In-Package Chemistry 
Abstraction (BSC 2004 [DIRS 167621]). The IPCM output includes pH, ionic strength, total 
carbonate, Eh, chloride, and fluoride. 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-39 October 2004 
As described in Waste Form and In-Drift Colloids-Associated Radionuclide Concentrations: 
Abstraction and Summary (BSC 2004 [DIRS 170025]), 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 2004 [DIRS 170025]), is addressed in FEPs Screening of Processes and Issues 
in Drip Shield and Waste Package Degradation (BSC 2004 [DIRS 169997], Section 6.2.18). 
The conceptual model and TSPA-LA 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 2004 [DIRS 170025], 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 oxyhydroxides colloids 
and fixed corrosion mass (BSC 2004 [DIRS 170025], 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 2004 [DIRS 170025]): 
 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 6.3.2.2) 
 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 (BSC 2004 [DIRS 169988], 
Appendix A) when determining model parameter values for pH dependence (.) in acidic and 
alkaline solutions (BSC 2004 [DIRS 169988], Sections 6.5.2.4 and 6.5.2.5). The effects of glass 
alteration phases on the glass dissolution rate are addressed by including the results of tests in 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-40 October 2004 
which alteration phases and/or alteration product rinds form when determining model parameter 
values for kE (BSC 2004 [DIRS 169988], Section 6.5.3.3). 
Supporting Documents: 
In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621]) 
Waste Form and In-Drift Colloids-Associated Radionuclide Concentration: Abstraction 
and Summary (BSC 2004 [DIRS 170025]) 
Defense HLW Glass Degradation Model (BSC 2004 [DIRS 169988]) 
6.2.21 Radionuclide Solubility, Solubility Limits, and Speciation in the Waste Form and 
EBS (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. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
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 2004 [DIRS 169425]). 
TSPA-LA includes the solubility model for actinium from Revision 2 of Dissolved 
Concentration Limits of Radioactive Elements (BSC 2003 [DIRS 163152]). Solubility models 
for the rest of the elements, except for lead, are included from Revision 3 of Dissolved 
Concentration Limits of Radioactive Elements (BSC 2004 [DIRS 169425]) in the TSPA-LA 
model and used to constrain the maximum radionuclide concentrations in waste packages and in 
the invert. In the TSPA-LA model, the masses 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. The resulting concentration in the rind surrounding the waste form depends 
on the volume of water the rind can hold and the dissolved concentration limits. If the water 
quantity is not great enough to dissolve all the radionuclides, and still stay beneath the dissolved 
concentration limit, then the radionuclides precipitate and are available for future dissolution. 
By diffusion or advection, depending on the TSPA-LA scenario, the radionuclides can be 
transported from the waste form to areas of the waste package containing corrosion products and 
transported out of the waste package to the invert. The dissolved concentration limits are also 
applied to radionuclides collected in the corrosion products and in the invert. The neutron 
poisons (such as B and Gd) are taken into account in In-Package Chemistry Abstraction (BSC 
2004 [DIRS 167621]), where their impact on the overall in-package chemistry is assessed. 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-41 October 2004 
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 2004 [DIRS 169425], Sections 6.5 
through 6.11). For radium and lead, solubilities are given as distributions or segmental constants 
(BSC 2004 [DIRS 169425]) 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 2004 
[DIRS 169425], 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 solubilitycontrolling 
solid exists and its release is considered to be controlled by the dissolution rate of 
waste forms and the waste inventory (BSC 2004 [DIRS 169425], Section 6.18). 
Supporting Documents: 
Dissolved Concentration Limits of Radioactive Elements (BSC 2004 [DIRS 169425] 
6.2.22 ReductionOxidation Potential in Waste Package (2.1.09.06.0A) 
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 in-package redox 
potential can occur. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621]) includes the redox potential of 
the groundwater seepage in the same equilibrium model that addresses the in-package 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 167621]). 
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 2004 
[DIRS 167621]). 
The in-package chemistry model generally considers the redox potential to be fully oxidizing, 
and this modeling assumption is discussed in Section 6.3.1.1 of In-Package Chemistry 
Abstraction (BSC 2004 [DIRS 167621]). The redox potential is varied in a sensitivity study in 
Section 6.6.4 of In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621]) by using lower 
fixed fO2 values, including the adjusted Eh model used in the dissolved concentration model, all 
the way down to log fO2 = -69. The sensitivity shows that the pH and ionic strength results are 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-42 October 2004 
sensitive to the fO2 values, so the results of this sensitivity were used to develop the pH limits 
and the uncertainty to be applied to the ionic strength abstraction (BSC 2004 [DIRS 167621], 
Section 6.10.2.3). 
Also, the dissolved concentration model examines redox potential as it pertains to repository 
conditions (BSC 2004 [DIRS 169425], Section 6.5.3.5) and develops an adjusted Eh model 
based on experimental data and natural analog. This adjusted Eh model is used for the solubility 
calculations and accounts for the fact that the solutions in contact with air are never fully at 
equilibrium with the atmosphere. 
The redox potential inside the waste package and its effect on in-package chemistry is implicitly 
included in the abstractions passed to TSPA-LA as part of In-Package Chemistry Abstraction 
(BSC 2004 [DIRS 167621]). The IPCM parameters are pH, ionic strength, total carbonate, Eh, 
chloride, and fluoride. 
Supporting Documents: 
In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621]) 
Dissolved Concentration Limits of Radioactive Elements (BSC 2004 [DIRS 169425]) 
6.2.23 Reaction Kinetics in Waste Package (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. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
Reaction kinetics and precipitation or dissolution rates are included in the TSPA-LA in-package 
chemistry model abstraction. The in-package chemistry model uses kinetic reactants to represent 
the SNF (BSC 2004 [DIRS 167621], Section 4.1.3) and the waste package components (BSC 
2004 [DIRS 167621], Section 4.1.4). 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 2004 [DIRS 
167621], Section 6.6) 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-LA. 
The variability in the kinetics of the reactants is included in the abstractions passed to TSPA-LA 
implicitly by their use in the in-package chemistry model and explicitly via the contribution of 
the kinetics to output uncertainty. The effects of reaction kinetics on in-package chemistry are 
passed to TSPA-LA as part of In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621]). 
The in-package chemistry model parameters are pH, ionic strength, total carbonate, Eh, chloride, 
and fluoride. 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-43 October 2004 
The in-package chemistry model generally allows reactions to reach thermodynamic equilibrium, 
and this modeling assumption is discussed in Section 6.3.1.1 of In-Package Chemistry 
Abstraction (BSC 2004 [DIRS 167621]) and the most important species, chromium species, are 
examined in more detail in Section 6.6.3 of In-Package Chemistry Abstraction (BSC 2004 
[DIRS 167621]). This sensitivity examines the impact of allowing kinetics to control the 
oxidation state, allowing only Cr(III) species to form. The sensitivity shows that the ionic 
strength results are sensitive to the oxidation state, so the results of this sensitivity were used to 
develop the uncertainty to be applied to the ionic strength abstraction (BSC 2004 
[DIRS 167621], Section 6.10.2.3). 
The dissolved concentration model examines kinetic as well as thermodynamic considerations 
when selecting the solubility-controlling phase for each radionuclide. This report uses the 
Ostwald Step Rule when selecting the controlling phase for radionuclide solubilities, which 
states that unstable minerals form first, followed by progressively more stable minerals. The 
lowest concentrations occur when the most stable (thermodynamically favored) minerals are 
controlling the solubilities. The kinetics of the system may not allow the most stable minerals to 
form in the time period of interest. So, a conservative approach is employed in which the least 
stable minerals are chosen as the solubility controlling phases (BSC 2004 [DIRS 169425], 
Section 6.3.2). In addition, the adjusted Eh model used for the solubility calculations takes into 
account the fact that water in contact with air is not at complete thermodynamic equilibrium with 
the atmosphere due to kinetic factors. 
Supporting Documents: 
In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621]) 
Dissolved Concentration Limits of Radioactive Elements (BSC 2004 [DIRS 169425]) 
6.2.24 Secondary Phase Effects on Dissolved Radionuclide Concentrations (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 available radionuclides for mobilization. 
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 2004 [DIRS 169425], 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. 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-44 October 2004 
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 Buck et al. (2003 [DIRS 168411]). 
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 capture 
experimental results well (BSC 2004 [DIRS 169425], 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: 
N/A 
6.2.25 Complexation in EBS (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. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: Repository temperatures are expected to remain sufficiently high that 
significant microbial activity is possible only after the first 1,000 years, when the drift 
temperature drops below the approximate 96C boiling point (BSC 2004 [DIRS 169991], 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-45 October 2004 
Section 6.4.1). Once lower temperatures recur, microbial populations will not be sufficient to 
generate significant concentrations of radionuclide-chelating organics. Organic complexants, 
which could be a transport mechanism, are typically found in sufficient quantity to transport 
radionuclides long distances only in organic rich soils and groundwaters  or downgradient from 
disposal sites 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), based on low abundance of the microbes being 
generated in the EBS, the effects on bulk geochemistry are expected to be negligible 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), which 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. 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: 
N/A 
6.2.26 Formation of True (Intrinsic) Colloids in EBS (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 colloids, 
Eigenkolloide, and intrinsic colloids (or actinide intrinsic colloids, for those including actinide 
elements). 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: Saturated solutions in close proximity to degrading waste may form true 
colloids (BSC 2004 [DIRS 170025], 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 2004 [DIRS 170025], Sections 6.3.1 and 6.3.2.6). The 
discussion on the reversible and irreversible sorption of radionuclides into colloids is contained 
in Section 6.2.20. 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-46 October 2004 
Uranium in the EBS is modeled as U-aqueous complexes not associated with colloids (BSC 2004 
[DIRS 170025], Sections 6.3.1 and 6.3.2.6). 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: 
N/A 
6.2.27 Formation of Pseudo-Colloids (Natural) in EBS (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 (i.e., clay, 
silica, iron oxyhydroxides), and humic and fulvic acids. This FEP addresses radionuclidebearing 
pseudo-colloids formed from host-rock materials and all interactions of the waste and 
EBS with the host rock environment except corrosion. 
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 2004 
[DIRS 170025], Section 6.3.1.4). Pseudocolloids generally form as a result of dissolved 
(aqueous) radionuclides sorbing to existing colloids (BSC 2004 [DIRS 170025], 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 2004 [DIRS 170025], Section 6.3.3.1). 
Colloid stability 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 2004 
[DIRS 170025], Section 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 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-47 October 2004 
are also modeled, as smectite clay colloids, and to respond to ionic strength and pH conditions in 
the same manner as natural groundwater colloids (Section 6.2.31). 
Supporting Documents: 
Waste Form and In-Drift Colloids-Associated Radionuclide Concentration: Abstraction 
and Summary (BSC 2004 [DIRS 170025]) 
6.2.28 Formation of Pseudocolloids (Corrosion Product) in EBS (2.1.09.17.0A) 
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. Corrosion product pseudo-colloids include iron oxyhydroxides from corrosion 
and degradation of the metals in the EBS and silica from degradation of cementitious materials. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
The TSPA-LA model abstraction identified in Waste Form and In-Drift Colloids-Associated 
Radionuclide Concentration: Abstraction and Summary (BSC 2004 [DIRS 170025]) includes 
treatment of colloids associated with corrosion of steel components of the waste package, 
surrounding drift materials, and silica from degradation of cementitious materials. Fixed and 
colloidal corrosion products are modeled in the waste package (BSC 2004 [DIRS 170025], 
Section 6.3.1). 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 2004 
[DIRS 170025], 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 2004 [DIRS 170025], Section 6.3.3.1). In addition, Pu and Am are modeled as sorbing 
irreversibly to iron oxyhydroxide corrosion colloids (BSC 2004 [DIRS 170025], 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 
in-package and in-drift fluids (BSC 2004 [DIRS 170025], Section 6.3.2.2). 
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 materials and iron 
oxyhydroxide colloids). The remainder is assumed to become reversibly attached according to 
an appropriate Kd (BSC 2004 [DIRS 170025], Section 6.3.3.1, Table 6.8). This capability was 
included to acknowledge laboratory and field observations that the attachment of a large fraction 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-48 October 2004 
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: 
Waste Form and In-Drift Colloids-Associated Radionuclide Concentration: Abstraction 
and Summary (BSC 2004 [DIRS 170025]) 
6.2.29 Formation of Microbial Colloids in EBS (2.1.09.18.0A) 
FEP Description: 
This FEP addresses the formation and transport of microbial colloids in the waste and 
EBS. 
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. Since the inorganic colloid stability is 
determined by fluid chemistry (pH and ionic strength) the potential influence of organic 
complexants from microbial metabolism is not expected to affect the stability of inorganic 
(mineral) colloids (BSC 2004 [DIRS 170025], Sections 5.9 and 6.3.4). Because inorganic 
aqueous conditions dominate DHLW glass 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 2004 [DIRS 170025], 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 
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 the 
microbes are relatively large, they will be readily filtered during transport, along with any 
accumulated radionuclides, which is beneficial in reducing the potential for transportation of 
radionuclides. Inclusion of microbial colloid transport from the waste and EBS environment 
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 is excluded based on low consequence. 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 6-49 October 2004 
TSPA Disposition: 
N/A 
Supporting Documents: 
N/A 
6.2.30 Stability of Colloids in EBS (2.1.09.23.0A) 
FEP Description: 
For radionuclide-bearing colloids to affect repository performance, they must remain suspended 
in the groundwater (i.e., be stable) for time scales that are long relative to the time required for 
groundwater travel. Further, they must carry significant concentrations of radionuclides. The 
stability of smectite colloids (applicable for natural groundwater colloids and waste form 
colloids) is determined primarily by ionic strength but also to an extent by pH. The stability of 
iron-(hydr)oxide colloids (applicable to corrosion-product colloids) is determined by both ionic 
strength and pH. 
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 2004 [DIRS 170025]) 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 2004 [DIRS 170025], Section 6.3.2). These 
determinations are then combined with the sorption analysis to calculate concentrations of 
radionuclides associated with the colloids (BSC 2004 [DIRS 170025], 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 2004 [DIRS 170025], 
Section 6.5). 
Supporting Documents: 
Waste Form and In-Drift Colloids-Associated Radionuclide Concentrations: Abstraction 
and Summary (BSC 2004 [DIRS 170025]) 

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ANL-WIS-MD-000009 REV 02 6-50 October 2004 
6.2.31 Formation of Colloids (Waste Form) by Co-precipitation in EBS (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. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
Colloids apparently formed from coprecipitation of smectite and embedded radionuclide-bearing 
phases were observed in the DHLW glass degradation tests (BSC 2004 [DIRS 170025], 
Section 6.3.1.1). Colloids produced from degradation of DHLW glass 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 DHLW glass 
corrosion experiments (BSC 2004 [DIRS 170025], 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 2004 [DIRS 170025], 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 2004 [DIRS 170025], Section 6.3.3.1). Similar 
to natural groundwater smectite colloids, the pseudocolloid stability is controlled by ionic 
strength and pH. 
Supporting Documents: 
Waste Form and In-Drift Colloids-Associated Radionuclide (BSC 2004 [DIRS 170025]) 
6.2.32 Exothermic Reactions in EBS (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. Examples of possible 
exothermic reactions include oxidation of uranium metal fuels such as represented by N-Reactor 
fuels and hydration of concrete used in the underground environment. 
Screening Decision: 
Excluded (low consequence) 

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ANL-WIS-MD-000009 REV 02 6-51 October 2004 
Screening Argument: Exothermic reactions, and other thermal reactions, that liberate heat in the 
waste and EBS are excluded from the TSPA-LA because the possible temperature rise in a waste 
package 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. The energy released from the oxidation of the uranium metal in 
a codisposal waste package containing two multicanister overpacks filled with N-Reactor fuel 
was found to be 5.29  107 kJ/package (BSC 2001 [DIRS 171508], Section 5.2.6). This value is 
only 3% of 1.62  109 kJ, which is the heat generation over the first year of emplacement due to 
radioactive decay in a segment of the repository containing seven waste packages (1.62  109 kJ 
= 51.2 kW (BSC 2004 [DIRS 167754], Table 12)  3,600 s/day  24 hr/day  365.25 day/yr). In 
a similar calculation, the heat of reaction by oxidizing all N Reactor fuel uranium metal is 
roughly 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 
(Section 6.2.5). 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 (BSC 2004 [DIRS 167618]), an increase in degradation rate due to 
temperature increase will have no impact on dose calculations. TSPA-LA calculations will not 
be sensitive to differences within a TSPA-LA time step. Note as well that the related effects of 
pyrophoric reactions are also excluded (Section 6.2.12). 
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 
6.2.33 Thermal Effects on Chemistry and Microbial Activity in the EBS (2.1.11.08.0A) 
FEP Description: 
Temperature changes may affect chemical and microbial processes in the waste and EBS. 
Screening Decision: 
Included 
Screening Argument: 
N/A 

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ANL-WIS-MD-000009 REV 02 6-52 October 2004 
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 package 
and EBS are explicitly accounted for in Multiscale Thermohydrologic Model (BSC 2004 
[DIRS 169565]). 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 2004 [DIRS 169868]). 
Thermal effects on microbial activity in the EBS are discussed in Section 6.2.25 and Engineered 
Barrier System Features, Events, and Processes (BSC 2004 [DIRS 169898], Section 6.2.65). 
As discussed in In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621]), the thermal 
effects on the in-package chemistry are examined in the batch reactor model wherein EQ6 runs 
were performed at various temperatures to examine the temperature effect on the in-package 
chemistry (Section 6.3.1). In these runs the kinetic reaction rates for the SNF were recalculated 
for runs at 25C to 90C, and a temperature-appropriate thermodynamic database was used in the 
simulations. In addition, temperature effects were considered in the surface complexation model 
(Section 6.3.2). 
For TSPA-LA the in-package chemistry abstractions of pH, ionic strength, total carbonate, and 
Eh are cast in terms of temperature (BSC 2004 [DIRS 167621], Section 6.10), which captures the 
thermal effects on in-package chemistry. 
Supporting Documents: 
In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621]) 
Multiscale Thermohydrologic Model (BSC 2004 [DIRS 169565]). 
6.2.34 Thermally Driven Flow (Convection) in Waste Packages (2.1.11.09.0B) 
FEP Description: 
Temperature differentials may result in convective flow in the EBS. Convective flow within the 
waste packages could influence in-package chemistry. 
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 as the waste package is well mixed. 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 2004 

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ANL-WIS-MD-000009 REV 02 6-53 October 2004 
[DIRS 169868], Table 20 and Section 7.5.1) includes a single mixing cell in the waste package 
(Section 5.3). 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 2004 [DIRS 167621], Section 6.4.1) treated 
the waste package as a one-dimensional assemblage of mixing cells with varying compositions. 
It was concluded 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 will be small (BSC 2004 
[DIRS 170019], 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: 
N/A 
6.2.35 Gas Generation (CO2, CH4, H2S) From Microbial Degradation (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. 
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 Yucca Mountain 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 (BSC 2004 
[DIRS 169898], Section 6.2.65) because organic material will be present in only minor amounts 
in the waste package. Methanogenesis and H2S generation are anaerobic processes, which will 

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ANL-WIS-MD-000009 REV 02 6-54 October 2004 
not occur under the oxidizing conditions that are expected to occur at Yucca Mountain. The 
discussion of the effects of radiation, pressure, temperature, and biofilm on microbial gas 
generation is contained in (BSC 2004 [DIRS 169898], Sections 6.2.65 and 6.2.77). The 
possibility of local depression of pH by microbes is addressed in Clad Degradation  FEPs 
Screening Arguments (BSC 2004 [DIRS 170019], Section 6.2.4). 
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 
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 (BSC 2004 [DIRS 169898], Section 
6.2.77). 
The 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. 
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 2004 [DIRS 167621]) cover concentration changes of CO2 by 
several orders of magnitude above and below ambient levels. The results show that variations in 
carbon dioxide partial pressures from 10-5 to 10-1.5 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. 
Dissolved Concentration Limits of Radioactive Elements (BSC 2004 [DIRS 169425]) 
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 10-1.5 (which covers 
two orders of magnitude greater than atmospheric pressure). The CO2 ranges, over which TSPALA 
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: 
N/A 

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6.2.36 Effects of Radioactive Gases in EBS (2.1.12.07.0A) 
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. 
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 argon, xenon, and krypton (Manaktala 1993 [DIRS 101719], 
Section 3.3.6). 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 gasses 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. Argon, krypton, and radon were screened out due to short half-lives or low activity 
(BSC 2002 [DIRS 160059], Table 2). Likewise, xenon is also negligible, since the half-lives of 
the isotopes are on the order of just a few days at most (Parrington et al. 1996 [DIRS 103896, 
p. 35). 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. 
D. 14CH4 and 
14CO2 generation from microbial activity (CRWMS M&O 2000 
[DIRS 151561], Sections 6.4 and 7.1) should not be significant because organic levels 
in the waste are expected to be low. 
E. Of the noble gases, 85Kr has a significant initial inventory, but because of its short 
half-life (approximately 10 years), its concentration rapidly becomes insignificant. 
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 consequence is negligible (FEP 2.2.11.03.0A; 

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ANL-WIS-MD-000009 REV 02 6-56 October 2004 
BSC 2004 [DIRS 170012]). Therefore, the effect of radioactive gasses in the EBS is excluded 
due to low consequence. 
TSPA Disposition: 
N/A 
Supporting Documents: 
N/A 
6.2.37 Radiolysis (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 (potential, pH, and concentration 
of reactive radicals). 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: The effects of radiolysis for gas production, the waste form chemistry, 
and fuel dissolution can be excluded from consideration in the TSPA-LA 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. The effects of beta irradiation on fuel degradation 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. 
The waste package 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. An analysis was performed to examine the consequences of 
nitric acid and hydrogen peroxide production, and other potential products, due to radiolysis 
(BSC 2004 [DIRS 167621], Appendix B). In this analysis, the concentration of nitric acid and 
hydrogen peroxide was conservatively increased by a factor of ten to bound the effects of 
radiolysis. The results show (BSC 2004 [DIRS 167621], Figure B-1) that any effect on the 
in-package pH is minimal and there would be no impact to TSPA-LA. Localized effects (small 
areas with reducing conditions and more extreme pH values and ionic strengths) will occur, but 
these will be transient (BSC 2004 [DIRS 167621], Section 6.3.1.1), and because the modeled 
system is an oxidizing environment which is known to be a conservative approach for 
radionuclide release and transport, these conditions are not significant enough to alter model 
results. 
In the event of early failure, the waste package is treated as completely failed and thus this FEP 
does not apply to waste package early failure. 

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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, 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 a few microns thick would stop alpha particles. 
Additionally, there are these two net findings: 
1. The a-radiolysis-enhanced corrosion rate of CSNF fuel is three orders of magnitude 
less than the standard dissolution rate. 
2. The a-radiolysis-enhanced corrosion rates for HLW metallic carbide, and ceramic 
spent nuclear fuels are much lower than their standard dissolution rates. 
Since no credit is being taken for the DSNF cladding in the TSPA-LA, 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: 
N/A 
6.2.38 Radiation Damage in EBS (2.1.13.02.0A) 
FEP Description: 
Radiolysis due to the alpha, beta, gamma-ray, and neutron irradiation of water could result in 
enhancement of the radionuclide migration from the surface of a degraded waste form into 
groundwater. 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 matrix) into the groundwater. Strong radiation fields could lead to radiation 
damage to the waste forms (CSNF, DSNF, DHLW), waste packages, drip shield, seals, and 
surrounding rock. 

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Screening Decision: 
Excluded (low consequence) 
Screening Argument: CSNF and DSNF waste forms that are placed in the repository have 
already been exposed to extensive radiation while in the reactors, and any further radiation that 
may occur while in the repository would be very small. On the waste package surface in the 
repository environment, the peak neutron flux has been estimated to be about 4  105 n/cm2s for 
PWR SNF (BSC 2004 [DIRS 169593], Table 6.4-5). Compared to a thermal-neutron flux of 
approximately 3  1013 n/cm2s (maximum, based on maximum burnup) for both a pressurized 
water reactor (PWR) (Benedict et al. 1981 [DIRS 103523], p. 393) and a boiling water reactor 
(BWR) (BSC 2004 [DIRS 171744], Table 11), the additional radiation applied to the fuel while 
stored in the repository would be insignificant and therefore cause no additional radiation 
damage to the fuel. 
Radiation levels in the repository may produce damage to the HLW glass, such as bond damage 
and atomic displacements. This type of damage has been shown to increase the release rates of 
glass components up to four-fold during corrosion tests (Wronkiewicz 1993 [DIRS 171709], 
Abstract). A four-fold increase in release rates is small compared to the two orders of magnitude 
range in dissolution rates used in the TSPA-LA (BSC 2004 [DIRS 169988], Table 8-1). In 
addition, under actual disposal conditions, glass-annealing processes may negate most of the 
radiation damage effects (Wronkiewicz 1993 [DIRS 171709], Abstract). 
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 
(FEP 2.1.02.04.0A) and radiolysis (FEP 2.1.13.01.0A) are also excluded based on low 
consequence. 
TSPA Disposition: 
N/A 
Supporting Documents: 
N/A 
6.2.39 Chemistry of Water Flowing into Waste Package (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. 
Screening Decision: 
Included 
Screening Argument: 
N/A 

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ANL-WIS-MD-000009 REV 02 6-59 October 2004 
TSPA Disposition: 
Six different initial water compositions were used to represent the chemistry of the water flowing 
into the waste package. The base case considers three water compositions, and a sensitivity study 
(BSC 2004 [DIRS 167621], Section 6.6.2) considers an additional three water compositions. 
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-LA in In-Package Chemistry 
Abstraction (BSC 2004 [DIRS 167621]). TSPA-LA then feeds these results to 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. 
The water influx model examined various input water chemistries (BSC 2004 [DIRS 167621], 
Section 4.1.2) 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. 
The dissolved concentration report considers the chemistry of the water potentially entering a 
waste package (BSC 2004 [DIRS 169425], Section 6.4). Various sensitivities were performed to 
examine the effect of compositions of 25 pore water compositions (Section 6.4.2.5). For the 
species that have an impact on solubility, an uncertainty term was developed (Sections 6.4.3.6 
and 6.4.6.7) to account for the variation of that species in the incoming water. 
Supporting Documents: 
In-Package Chemistry Abstraction (BSC 2004 [DIRS 167621]). While FEP 2.2.08.12.0B 
is not in the included FEP table (BSC 2004 [DIRS 167621], Table 10), it is addressed in 
the report. 
Dissolved Concentration Limits of Radioactive Elements (BSC 2004 [DIRS 169425] 
6.2.40 Radioactive Decay and Ingrowth (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. 
Screening Decision: 
Included 
Screening Argument: 
N/A 

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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 first 10,000 years following repository closure. 
Seven parent isotopes were identified whose decay significantly increased the amount of their 
progeny (BSC 2004 [DIRS 170022], Table 6-1): 
 245Cm . 241Pu . 241Am 
 235U . 231Pa 
 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: 
Initial Radionuclide Inventories (BSC 2004 [DIRS 170022]) 

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7. CONCLUSIONS 
This report addresses 40 waste form FEPs. These FEPs were identified and examined for 
inclusion or exclusion in TSPA-LA. Table 7-1 lists these FEPs and their associated screening 
decisions. 
By default, a FEP is included in the TSPA-LA if it cannot be excluded based on the screening 
criteria described in Section 4.2.3. 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-LA 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. 
This analysis model report contributes to Yucca Mountain FEP analysis by screening 40 waste 
form FEPs. TSPA-LA dispositions 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 FEP database and guidance to TSPA-LA analyses. 
Table 7-1. Summary of Waste Form FEPs 
FEP Number FEP Name Screening 
Decision and 
Basis 
Addressed in 
Section 
2.1.01.01.0A Waste Inventory Included 6.2.1 
2.1.01.02.0A Interactions between co-located waste Excluded (low 
consequence) 
6.2.2 
2.1.01.02.0B Interaction between co-disposed waste Included 6.2.3 
2.1.01.03.0A Heterogeneity of waste inventory Included 6.2.4 
2.1.02.01.0A DSNF degradation (alteration, dissolution, and 
radionuclide release) 
Included 6.2.5 
2.1.02.02.0A CSNF degradation (alteration, dissolution, and 
radionuclide release) 
Included 6.2.6 
2.1.02.03.0A HLW glass degradation (alteration, dissolution, and 
radionuclide release) 
Included 6.2.7 
2.1.02.04.0A Alpha recoil enhances dissolution Excluded (low 
consequence) 
6.2.8 
2.1.02.05.0A HLW glass cracking Included 6.2.9 
2.1.02.06.0A HLW glass recrystallization Excluded (low 
consequence) 
6.2.10 
2.1.02.07.0A Radionuclide release from gap and grain 
boundaries 
Included 6.2.11 
2.1.02.08.0A Pyrophoricity from DSNF Excluded (low 
consequence) 
6.2.12 
2.1.02.09.0A Chemical effects of void space in waste package Included 6.2.13 
2.1.02.10.0A Organic/cellulosic materials in waste Excluded (low 
consequence) 
6.2.14 

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ANL-WIS-MD-000009 REV 02 7-2 October 2004 
FEP Number FEP Name Screening 
Decision and 
Basis 
Addressed in 
Section 
2.1.02.25.0A DSNF cladding Excluded (low 
consequence) 
6.2.15 
2.1.02.28.0A Grouping of DSNF waste types into categories Included 6.2.16 
2.1.02.29.0A Flammable gas generation from DSNF Excluded (low 
consequence) 
6.2.17 
2.1.03.06.0A Internal corrosion of waste packages prior to 
breach 
Excluded (low 
consequence) 
6.2.18 
2.1.09.01.0B Chemical characteristics of water in waste package Included 6.2.19 
2.1.09.02.0A Chemical interaction with corrosion products Included 6.2.20 
2.1.09.04.0A Radionuclide solubility, solubility limits, and 
speciation in the waste form and EBS 
Included 6.2.21 
2.1.09.06.0A Reduction-oxidation potential in waste package Included 6.2.22 
2.1.09.07.0A Reaction kinetics in waste package Included 6.2.23 
2.1.09.10.0A Secondary phase effects on dissolved radionuclide 
concentrations 
Excluded (low 
consequence) 
6.2.24 
2.1.09.13.0A Complexation in EBS Excluded (low 
consequence) 
6.2.25 
2.1.09.15.0A Formation of true (intrinsic) colloids in EBS Excluded (low 
consequence) 
6.2.26 
2.1.09.16.0A Formation of pseudo-colloids (natural) in EBS Included 6.2.27 
2.1.09.17.0A Formation of pseudo-colloids (corrosion product) in 
EBS 
Included 6.2.28 
2.1.09.18.0A Formation of microbial colloids in EBS Excluded (low 
consequence) 
6.2.29 
2.1.09.23.0A Stability of colloids in EBS Included 6.2.30 
2.1.09.25.0A Formation of colloids (waste form) by coprecipitation 
in the EBS 
Included 6.2.31 
2.1.11.03.0A Exothermic reactions in the EBS Excluded (low 
consequence) 
6.2.32 
2.1.11.08.0A Thermal effects on chemistry and microbial activity 
in the EBS 
Included 6.2.33 
2.1.11.09.0B Thermally-driven flow (convection) in waste 
packages 
Excluded (low 
consequence) 
6.2.34 
2.1.12.04.0A Gas generation (CO2, CH4, H2S) from microbial 
degradation 
Excluded (low 
consequence) 
6.2.35 
2.1.12.07.0A Effects of radioactive gases in EBS Excluded (low 
consequence) 
6.2.36 
2.1.13.01.0A Radiolysis Excluded (low 
consequence) 
6.2.37 
2.1.13.02.0A Radiation damage in EBS Excluded (low 
consequence) 
6.2.38 
2.2.08.12.0B Chemistry of water flowing into the waste package Included 6.2.39 
3.1.01.01.0A Radioactive decay and ingrowth Included 6.2.40 
The conclusions from this document (FEP screening decision, TSPA-LA disposition for included 
FEPs, or screening argument for excluded FEPs), along with any modification to the FEP list, 
names, or descriptions, are incorporated in the Yucca Mountain TSPA-LA FEP database. The 
FEP database contains all Yucca Mountain FEPs considered for TSPA-LA with FEP number, 

Waste Form Features, Events, and Processes 
ANL-WIS-MD-000009 REV 02 7-3 October 2004 
name, description, and relevant reports where the documentation of the screening of specific 
FEPs is summarized. The FEP database also contains screening decisions (included or 
excluded), screening arguments, and TSPA-LA dispositions quoted from this and other FEP 
reports. 
7.1 ACCEPTANCE CRITERIA 
The purpose of this report is to evaluate and document the inclusion or exclusion of waste form 
FEPs with respect to modeling used to support the TSPA-LA. A screening decision, either 
Included or Excluded, was given for each FEP along with the corresponding technical basis for 
the excluded FEPs and the descriptions of how the included FEPs were incorporated in the 
TSPA-LA. This information was required by the U.S. Nuclear Regulatory Commission (NRC) 
regulations at 10 CFR 63.114 (d, e, and f) ([DIRS 156605]). 
The acceptance criteria, identified as applicable to this analysis in Section 4.2, that are related to 
the FEPs screening process were addressed in this analysis. The following acceptance criteria 
are based on meeting the requirements at 10 CFR 63.114(e) and (f) ([DIRS 156605]). 
7.1.1 Scenario Analysis Acceptance Criteria 
The following Acceptance Criteria from Yucca Mountain Review Plan, Final Report (NRC 2003 
[DIRS 163274], Section 2.2.1.2.1.3) were addressed in this report: 
Acceptance Criterion 1 The Identification of a List of Features, Events, and Processes 
Is Adequate. 
(1) 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-frequency-low magnitude, and rare largemagnitude 
events); tectonic evolution (slip on existing faults and formation of new 
faults); climatic change (change to pluvial conditions); and criticality. 
Documentation of the origin of the FEPs list is provided in Section 6.1.1; FEP descriptions are 
provided in Section 6.2. This analysis contains a list of waste form-related FEPs (Tables 1-1 and 
7-1). This list of FEPs includes those related to the degradation, deterioration, or alteration of 
engineered barriers. The list of waste form-related FEPs is consistent with the site 
characterization data. 
Acceptance Criterion 2 Screening of the List of Features, Events, and Processes Is Appropriate. 
(1) The U.S. Department of Energy has identified all features, events, and 
processes related to either the geologic setting or to the degradation, 

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ANL-WIS-MD-000009 REV 02 7-4 October 2004 
deterioration, or alteration of engineered barriers (including those processes that 
would affect the performance of natural barriers) that have been excluded; 
The relevant FEPs related to the degradation, deterioration, or alteration of engineered barriers 
were identified. These FEPs were screened for inclusion in the TSPA-LA. See Table 7-1 for a 
list of excluded waste form-related FEPs. 
(2) 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 
For the FEPs that were excluded from TSPA-LA by either low probability, low consequence, or 
by regulation, the justification was provided in the appropriate subsections of Section 6.2, 
documenting the basis for the exclusion. See the method and approach discussion provided in 
Section 6.1.2 for an explanation of the use of various types of justification. 
(3) 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 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. 
For the FEPs that were excluded from TSPA-LA by either low probability, low consequence, or 
by regulation, the appropriate technical basis was provided in the subsections of Section 6.2, 
documenting the basis for the exclusion. See the method and approach discussion provided in 
Section 6.1.2 for an explanation of the use of various types of justification. 

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ANL-WIS-MD-000009 REV 02 8-1 October 2004 
8. INPUTS AND REFERENCES 
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of Nuclear Materials, 244, 66-74. Amsterdam, The Netherlands: Elsevier. 
TIC: 246914. 
125082 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. ACC: MOL.19981019.0001. 
107796 Thornton, T.A. 1998. HPPP Issue 1; 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. ACC: MOL.19980730.0142. 
113466 Werme, L.O. and Spahiu, K. 1998. Direct Disposal of Spent Nuclear 
Fuel: Comparison Between Experimental and Modelled Actinide Solubiluties in 

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Natural Waters. Journal of Alloys and Compounds, 271-273, 194-200. Lausanne, 
Switzerland: Elsevier. TIC: 243085. 
143651 Wick, O.J., ed. 1980. Plutonium Handbook: A Guide to the Technology. Volumes I 
& II. La Grange Park, Illinois: American Nuclear Society. TIC: 248002. 
171709 Wronkiewicz, D.J. 1993. Effects of Radionuclide Decay on Waste Glass Behavior--A 
Critical Review. ANL-93/45. Argonne, Illinois: Argonne National Laboratory. 
TIC: 210799. 
8.2 CODES, STANDARDS, REGULATIONS, AND PROCEDURES 
103540 10 CFR 60. Energy: Disposal of High-Level Radioactive Wastes in Geologic 
Repositories. Readily available. 
156605 10 CFR 63. Energy: Disposal of High-Level Radioactive Wastes in a Geologic 
Repository at Yucca Mountain, Nevada. Readily available. 
162618 67 FR 19432. Surplus Plutonium Disposition Program. Readily available. 
AP-2.22Q, Rev. 1, ICN 1. Classification Analyses and Maintenance of the Q-List. 
Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste 
Management. ACC: DOC.20040714.0002. 
AP-3.15Q, Rev. 4, ICN 5. Managing Technical Product Inputs. Washington, 
D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste 
Management. ACC: DOC.20040812.0004. 
AP-SIII.2Q, Rev. 1, ICN 2. Qualification of Unqualified Data. Washington, D.C.: 
U.S. Department of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20040127.0008. 
AP-SIII.9Q, Rev. 1, ICN 7. Scientific Analyses. Washington, D.C.: U.S. Department 
of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20040920.0001. 
152779 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. 
TIC: 248977. 
103327 DOE-HDBK-1081-94. 1994. Primer on Spontaneous Heating and Pyrophoricity. 
Washington, D.C.: U.S. Department of Energy. TIC: 213578. 
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. 

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8.3 SOURCE DATA, LISTED BY DATA TRACKING NUMBER 
170760 MO0407SEPFEPLA.000. LA FEP List. Submittal date: 07/20/2004. 
172057 MO0410MWDTLCSC.000. TSPA-LA Cladding Sensitivity Cases. Submittal date: 
10/19/2004. 

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APPENDIX A 
ALPHA RECOIL MECHANICS 

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A. ALPHA RECOIL MECHANICS 
The alpha recoil is analyzed per conservation of momentum in the center-of-mass (COM) frame 
of reference. Consider Figure I-1a; 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 I-1b for COM frame). 
Figure I-1. Conceptual Illustration of Alpha Recoil Mechanics 
Applying the conservation of momentum in the COM system leads to Equations I-1 and I-2: 
Momentum Before = Momentum After (Eq. I-1) 
a av m V M y y 
v v 
+ = 0 (Eq. I-2) 
where 
My = Mass of recoil nucleus 
m a = Mass of a particles. 
Therefore, the velocity of the recoil nucleus (in terms of the velocity of the alpha particle) is: 
. .
.
. 
. .
.
. 
- = 
y 
y M
m v V a 
a (Eq. I-3) 
where 
m a = mass of the alpha particle 
My = mass of the recoil nucleus. 

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ANL-WIS-MD-000009 REV 02 A-4 October 2004 
The kinetic energy of the recoil nucleus can now be determined in terms of the kinetic energy of 
the emitted alpha particle as: 
) ( 
2
1 
2
1 ) ( 2 2 a a 
a a 
a KE 
M
m 
V m 
M
m 
V M Y KE 
Y Y 
Y Y = .. 
. 
.. 
. = = 
(Eq. I-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(a ) = 
4.196 MeV (Lederer and Shirley 1978 [DIRS 142133]), m a . 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 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 lab-system 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. 

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APPENDIX B 
DATA QUALIFICATION FOR DIRECT INPUTS TO WASTE FORM FEPS 

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DATA QUALIFICATION FOR DIRECT INPUTS TO WASTE FORM FEPS 
B.1 PURPOSE 
This appendix is related to the direct input data used to evaluate the waste form FEPs screening 
as discussed in Section 4.1. The screening of the waste form FEPs relies on data that may not 
have been collected under an approved quality assurance program that meets the requirements of 
10 CFR Part 63, Subpart G [DIRS 156605] or its predecessor, 10 CFR Part 60 [DIRS 103540]. 
This appendix describes the data qualification activities for previously unqualified project data 
and documents the data qualification tasks as required by AP-SIII.9Q, Scientific Analyses for 
data obtained from outside sources and AP-SIII.2Q, Qualification of Unqualified Data for data 
from project sources that require qualification, and that are not established fact per the 
requirements of those procedures. It is not intended as stand-alone documentation separate from 
the main document. The data qualified herein are for use only for FEPs screening within this 
work product. 
B.2 METHODS AND CRITERIA 
Section 5.2.1(l) of AP-SIII.9Q, Scientific Analyses, specifies that the extent to which the data 
demonstrate the properties of interest shall be addressed and one or more of the following factors 
in qualifying data as suitable for their intended use: 
 Reliability of data source 
 Qualification of personnel or organizations 
 Prior uses of the data 
 Availability of corroborating data. 
The data qualification team consisted of two members: Randolph Schreiner, a FEPs expert and 
the analysis originator, chaired the team and William Downs, Ph.D., a technically competent 
geochemist, assisted. As the analysis originator, Randolph Schreiner is independent of the data 
to be qualified. Neither he nor William Downs participated in the acquisition or development of 
the data, and each was an independent reviewer of this analysis. 
B.3 QUALIFICATION OF DATA 
The procedure does not apply to Established Fact data, or numerical data obtained from an 
established/authoritative data source. Because some of the data comes from authoritative 
sources (other than the U.S. Department of Energy, Office of Civilian Radioactive Waste 
Management), it does not need to be qualified. The remaining information is qualified only for 
its intended use i.e., to justify excluding from performance assessment some of the forty modes 
of waste form tracking and degradation. Some of the exclusionary arguments depend on 
qualitative arguments from the open literature that the analysis cites as data. 
Table B-1 lists sources of data and the criteria used to qualify the information. When the source 
is an authoritative source, the table provides only the sources' names. For the remaining sources, 
the Table designates the attributes or criteria that are used, in conjunction with the technical 
assessment, to qualify the information. 

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Table B-1. Data Sources for Qualification within This Work Product 
Item Source Description of Input Being Qualified 
Criteria Used for 
Qualification 
1 BSC 2001 [DIRS 171508] Maximum boundary dose would increase by, 
at most, a few percent due to pyrophoric event 
involving uranium metal spent fuel. The 
maximum boundary dose would be less than 5 
mrem/yr 
Energy released from oxidation of waste 
package containing 2 multicanister overpacks. 
AP-SIII.9Q 
Rev. 1/ICN 7 
Qualifications of 
personnel or 
organizations 
generating the data 
2 DOE 2002 [DIRS 158405] The total inventory of the Group 3 (Pu/U 
carbide) SNF is less than .001 % of DSNF 
waste. 
N Reactor accounts for about 85% of the total 
metric tons of heavy metal of DOE spent 
nuclear fuel. 
Each codisposal package will contain either 
(1) two glass logs and two multicanister 
overpacks or (2) five glass logs and one DSNF 
canister. 
AP-SIII.2Q, 
Rev. 1/ICN 2 
Identified in 
Appendix C 
3 Rechard 1995 [DIRS 101084] Significant amounts of DSNF cladding are 
likely to be damaged. 
AP-SIII.9Q 
Rev. 1/ICN 7 
Availability of 
Corroborating Data 
4 Sunder et al. 1997 [DIRS143860] The effects of radiolysis on spent nuclear fuel 
corrosion (dissolution) will be transitory and 
will become minor as alpha dose rates 
decrease. 
AP-SIII.9Q 
Rev. 1/ICN 7 
Availability of 
Corroborating Data 
The information used for direct input is also previously identified in Section 4 to provide the 
technical basis for exclusion of the individual FEP. 
B.3.1 DATA QUALIFICATION FOR BSC 2001 [DIRS 171508] 
B.3.1.1 Properties of Interest 
This calculation evaluates the consequences of a post closure pyrophoric event in terms of its 
potential to damage adjacent waste packages and its potential to cause release of radionuclides 
and dose to humans. The energy release from the oxidation of uranium metal in a waste package 
containing two multicanister overpacks was found to be 5.29  107 kJ/package. In addition, the 
report found that the maximum boundary dose would increase by, at most, a few percent due to 
pyrophoric event involving uranium metal spent fuel. The maximum boundary dose would be 
less than 5 mrem/yr. 
B.3.1.2 Qualifications of Personnel or Organizations Generating the Input 
The report was cancelled in the TDMS because the results were developed in support of 
TSPA-SR rather than TSPA-LA. The results of the calculations are still valid for use in the 
screening arguments for FEP 2.1.02.08.0A (Pyrophoricity from DSNF) and FEP 2.1.11.03.0A 

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(Exothermic Reactions in EBS). The results were developed under an approved quality 
assurance program. The calculations were performed with qualified software and following 
qualified procedures. 
B.3.2 DATA QUALIFICATION FOR DOE 2002 [DIRS 158405] 
B.3.2.1 Properties of Interest 
From Appendix D, Table D-1, the MTHM numbers for TSPA groups 3 and 7 were used to 
determine that Pu/U carbide SNF makes up less than .001% of DSNF waste and that the N 
Reactor accounts for about 85% of DOE SNF. Also Section 9 was used to show that the present 
plans for codisposal packages call for five glass logs and one DSNF canister. 
B.3.2.2 Technical Assessment 
The concern with the report is the sources of the data used and the lack of documentation or 
proof of proper data acquisition as stated in AP-SIII.2Q, Attachment 3, Section 5(a). DOE SNF 
includes diverse fuels from various experimental, research, and production reactors and consists 
of several hundred different fuel types that have been stored at several sites over the years. As a 
result, the availability and completeness of the radionuclide inventories and associated 
documentation varies considerably for DOE SNF. The data used is considered existing 
information and is the best that is available. 
The report was produced by the Idaho National Engineering and Environmental Laboratory 
(INEEL) under contract number DE-AC07-99ID13727 to U.S. Department of Energy for use in 
the Yucca Mountain Project. INEEL was a qualified supplier for DOE work at the time of the 
report with a quality assurance program that implemented DOE quality assurance requirements. 
Audits have been performed on the Quality Assurance programs at INEEL by the Office of 
Civilian Radioactive Waste Management (OCRWM). 
OCRWM performed an audit of selected QA program elements of INEELs Quality Assurance 
program September 23 to 26, 2002 (DOE 2002 [DIRS 171675]). Included in the program 
elements evaluated was software. The result of the audit was satisfactory and effective 
implementation of the examined portions of the Office of Civilian Radioactive Waste 
Management (OCRWM) QA program with the exception of deficiencies written during the 
audit. Deficiency Reports EM(O)-03-D-004 through 007 were written as a result of the audit and 
closed by 06/25/03. 
The percentages of the N Reactor SNF and the Pu/U carbide SNF have been derived using the 
MTHM values identified for the TSPA Groups 3 and 7 and the total MTHM for all groups in 
Appendix D, Table D-1. The MTHM values come from a DOE SNF database maintained to 
support TSPA-LA and is considered the best information available. This provides the desired 
level of confidence for the parameters of interest. 
The makeup of the codisposal package, of five glass logs and one DSNF canister, comes from 
design documents for the Yucca Mountain Project and has a good level of confidence. 

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B.3.2.3 Conclusion 
Based on an evaluation of the attributes identified from AP-SIII.2Q, attachment 4 and the 
technical assessment, it has been concluded that the data are qualified for its intended use in this 
report; that the sources for the data and the data themselves provide the desired level of 
confidence needed for this analysis. The plan for the qualification and rationale for the selected 
qualification method are identified in Appendix C, Data Qualification Plan. These data are 
qualified here only for this work product. 
B.3.3 DATA QUALIFICATION FOR RECHARD 1995 [DIRS 101084] 
B.3.3.1 Properties of Interest 
FEP 2.1.02.25.0A specifically concerns the differences in the DSNF and CSNF (the base case 
component) upon receipt at the repository and the effect that these differences will have on the 
performance of the fuels in the repository environment over the regulatory period of 10,000 
years. This report concludes that because DSNF has not been as fully characterized as CSNF, 
the damaged fuel rods may be in greater abundance (up to 50 percent) than found in CSNF. This 
estimate is conservatively high and is used as a bounding condition. 
B.3.3.2 Availability of Corroborating Data 
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 multicanister 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 
Sections 6.2.12 and 6.2.13, 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 167618]) discusses 
the grouping of the many types DSNF (Section 6.2.17). The report notes that few groups have 
had 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 

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TSPA-LA 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 
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-LA. 
As an outside-the-project source, Rechard (1995 [DIRS 101084]) 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 availability of the corroborating data in the other sources. 
B.3.4 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. 
B.3.4.1 Properties of Interest 
FEP 2.1.13.02.0A is concerned with the potential damage to the engineered barrier system due to 
enhanced corrosion via the mechanism of radiolitic generation of active chemical species from 
the atmosphere and groundwater solutions and to the damage of the barrier system materials by 
radionuclide decay products. This document concludes, after extensive analyses, the radiation 
fields from the spent fuel will be much lower that those to which the fuel was exposed and will 
decrease with time. 

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B.3.4.2 Availability Corroborating Data 
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]). Several other reports cited in this discussion are corroborative and support the 
conclusions of the direct use reference. 
Earlier 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: 
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 230 C, 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 2004 [DIRS 172017], 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 

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ANL-WIS-MD-000009 REV 02 B-9 October 2004 
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 2004 [DIRS 172017], 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. 
B.4 CONCLUSION 
In conclusion, the data are qualified for its intended use in this report. 

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APPENDIX C 
DATA QUALIFICATION PLAN 

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C. DATA QUALIFICATION PLAN 

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