Clad Degradation - FEPs Screening Arguments Rev 02, ICN 00

ANL-WIS-MD-000008

October 2004
 
1. PURPOSE 
The purpose of this report is to evaluate and document the screening of the clad degradation 
features, events, and processes (FEPs) with respect to modeling used to support the Total System 
Performance Assessment  License Application (TSPA-LA). This report also addresses the 
effect of certain FEPs on both the cladding and the commercial spent nuclear fuel (CSNF), 
DOE-owned spent nuclear fuel (DSNF), and defense high-level waste (DHLW) waste forms, as 
appropriate to address the effects on multiple materials and both components (FEPs 
2.1.09.09.0A, 2.1.09.11.0A, 2.1.11.05.0A, 2.1.12.02.0A, and 2.1.12.03.0A). These FEPs are 
expected to affect the repository performance during the postclosure regulatory period of 
10,000 years after permanent closure. 
Table 1-1 provides the list of cladding FEPs, including their screening decisions (include or 
exclude). The primary purpose of this report is to identify and document the analysis, screening 
decision, and TSPA-LA disposition (for included FEPs) or screening argument (for excluded 
FEPs) for these FEPs related to clad degradation. In some cases, where a FEP covers multiple 
technical areas and is shared with other FEP reports, this report may provide only a partial 
technical basis for the screening of the FEP. The full technical basis for shared FEPs is 
addressed collectively by the sharing FEP reports. The screening decisions and associated 
TSPA-LA dispositions or screening arguments from all of the FEP reports are cataloged in a 
project-specific FEPs database. 
1.1 PLANNING AND DOCUMENTATION 
Documentation requirements for this report are described in 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 cladding degradation FEP list for TSPA-LA 
that resulted from the planned work scope are described in Section 6.1. 
1.2 SCOPE 
The scope of this report is to describe, evaluate, and document screening decisions and technical 
bases for the cladding degradation FEPs for TSPA-LA. For FEPs included in the TSPA-LA, this 
report provides a consolidated summary of how the FEP has been included and addressed in the 
TSPA-LA model based on the supporting analysis and model reports. It also provides a list of 
the reports that provide detailed discussions of the FEP. For FEPs 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 bases that 
supports that decision. It also provides appropriate references to project and non-project 
documentation that supports the exclusion. This information is provided in Section 6.2 and 
subsequent sections and subsections. 
An overview of the YMP FEP analysis and scenario development process is available in The 
Development of the TSPA-LA Features, Events, and Processes (BSC 2004 [DIRS 168706], 
Sections 2.4, 3, and 4), describing 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 cladding degradation FEP analysis. 

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The list of cladding degradation 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 6-2) 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 or TSPA-LA disposition) information are in Section 6.2. 
This report addresses cladding degradation FEPs, which represent the key phenomena that result 
in degradation of cladding and, in some cases, the CSNF pellets. 
Table 1-1. Cladding Degradation Features, Events, and Processes 
FEP Number FEP Name 
Addressed 
in Section 
Sharing 
FEP With 
2.1.02.11.0A Degradation of cladding from waterlogged rods 6.2.1 
2.1.02.12.0A Degradation of cladding prior to disposal 6.2.2 
2.1.02.13.0A General corrosion of cladding 6.2.3 
2.1.02.14.0A Microbially influenced corrosion (MIC) of cladding 6.2.4 
2.1.02.15.0A Localized (radiolysis enhanced) corrosion of cladding 6.2.5 
2.1.02.16.0A Localized (pitting) corrosion of cladding 6.2.6 
2.1.02.17.0A Localized (crevice) corrosion of cladding 6.2.7 
2.1.02.18.0A Enhanced corrosion of cladding from dissolved silica 6.2.8 
2.1.02.19.0A Creep rupture of cladding 6.2.9 
2.1.02.20.0A Internal pressurization of cladding 6.2.10 
2.1.02.21.0A Stress corrosion cracking (SCC) of cladding 6.2.11 
2.1.02.22.0A Hydride cracking of cladding 6.2.12 
2.1.02.23.0A Cladding unzipping 6.2.13 
2.1.02.24.0A Mechanical impact on cladding 6.2.14 
2.1.02.25.0B Naval SNF Cladding 6.2.15 
2.1.02.26.0A Diffusion-controlled cavity growth in cladding 6.2.16 
2.1.02.27.0A Localized (fluoride enhanced) corrosion of cladding 6.2.17 
2.1.07.01.0A Rockfall 6.2.18 WP, EBS 
2.1.09.03.0A Volume increase of corrosion products impacts cladding 6.2.19 
2.1.09.09.0A Electrochemical effects in EBS 6.2.20 WP 
2.1.09.11.0A Chemical effects of waste-rock contact 6.2.21 
2.1.11.05.0A Thermal expansion/stress of in-package EBS 
components 
6.2.22 
2.1.12.02.0A Gas generation (He) from waste form decay 6.2.23 EBS 
2.1.12.03.0A Gas generation (H2) from waste package corrosion 6.2.24 WP, EBS 
NOTE: WP = waste package EBS = engineered barrier system 

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1.3 SCIENTIFIC ANALYSIS LIMITATIONS 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 included and excluded cladding 
degradation FEPs. It is intended for use as the documentation for inclusion or exclusion of 
cladding degradation FEPs within or from the TSPA-LA model. The following limitations apply 
to this report: 
 Because this report cites other documents as direct input, the limitations of this report 
inherently include any limitations or constraints described in the cited AMRs or 
controlled documents. 
 In cases where FEPs are shared, the scope of this report is limited to the degradation of 
the cladding of the spent nuclear fuel. The full technical basis for these shared FEPs is 
addressed, collectively, by all of the sharing FEP reports. 
 For screening purposes, this 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. 
Changes in direct inputs listed in Section 4.1, in baseline conditions used for this evaluation, or 
in other subsurface conditions, are required 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 required by 
10 CFR 63.73 Subparts F and G [DIRS 156605] (see also the requirements in 10 CFR 63.44 
[DIRS 156605]). 
This report is limited to nuclear fuel exposed to normal operation and anticipated operational 
occurrences (i.e., events that are anticipated to occur within a reactor lifetime) prior to receipt at 
the repository, and is not applicable to fuel that has been exposed to severe accidents. Fuel 
burnup projections have been limited to the current commercial reactor-licensing environment 
with restrictions on fuel enrichment (5 percent for material shipment and fuel manufacturing, 
shipment, and storage), cladding oxide coating thickness, and rod plenum pressures. This is 
consistent with the projections for advanced reloads. The fuel considered has burnup up 
to 75 MWd/kgU and half of the fuel is assumed to have a burn-up of greater than 
above 44.7 MWd/kgU, which is consistent with the typical burn-up range of fuel in use in 2000 
and beyond. Ranges and uncertainties have been defined in Initial Cladding Condition 
(CRWMS M&O 2000 [DIRS 151659], Section 6.2). The information presented in this report is 
used to determine which cladding degradation mechanisms are addressed by TSPA-LA. 
The CSNF cladding performance is used to bound naval fuel cladding behavior. Cladding 
degradation for DOE-owned spent nuclear fuel (DSNF) is not addressed in this report, but is 
addressed in a general waste form screening report Waste-Form Features, Events, and Processes 
(BSC 2004 [DIRS 170020]). 

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This report is directly supported by input from several model and analysis reports: Clad 
Degradation  Summary and Abstraction for LA, CSNF Waste Form Degradation: Summary 
Abstraction, Defense HLW Glass Degradation Model, In-Package Chemistry Abstraction, 
Multiscale Thermohydrologic Model, and Pitting Model for Zirconium-Alloyed Cladding. This 
report does not directly feed any downstream reports. 

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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 8.1, Work Package ARTM03). 
Cladding has been determined to be important to waste isolation in accordance with AP-2.22Q, 
Classification Analysis and Maintenance of the Q-List, and, therefore, is classified as Safety 
Category (SC) in Q-List (BSC 2004 [DIRS 168361], Appendix A). The technical work plan 
contains the Process Control Evaluation used to evaluate the control of electronic management of 
data (BSC 2004 [DIRS 171583], Attachment I) during modeling and documentation activities. 
This evaluation determined that the methods identified in the implementing procedures are 
adequate. There were no deviations from these methods. This report was prepared using 
AP-SIII.9Q, Scientific Analyses. 

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3. USE OF SOFTWARE 
This analysis 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 analysis 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 LP-SI.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 the use of technical product input 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 addressed in this section. Indirect 
inputs are addressed in Section 6.1.3. 
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 
at 10 CFR Part 63 [DIRS 156605] along with the required screening criteria. Section 4.3 
identifies applicable codes, standards, and regulations. 
4.1 DIRECT INPUTS 
The TSPA-LA FEP list (DTN: MO0407SEPFEPLA.000 [DIRS 170760]) was used as a direct 
input to provide the initial list of cladding degradation FEPs for screening in this report. The 
TSPA-LA FEP list identifies a FEP report or a set of sharing reports for each FEP. Subsequent 
additions to or changes from that 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 6-2). 
The direct inputs used for the FEP screening in this report are listed in Table 4-1. The direct 
inputs listed in Table 4-1 have been evaluated and have been determined to be appropriate for 
their intended use of providing the necessary information for the disposition of the cladding 
FEPs. The data from sources requiring justification per the requirements of AP-SIII.9Q or 
AP-SIII.2Q are considered qualified for their intended use per the documentation in Appendix B. 
This report does not utilize any previously developed and validated model to complete this 
report. 

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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 
1.3, 4., 4.2.1, 4.2.2, 4.3, 5.2, 5.3, 
6.1.2, Table 4-1, Table 4-2, Appendix 
B 
FEP Screening requirements 
BSC 2004 [DIRS 170023] Entire, 6.2.1, 
6.2.4, 6.2.3, 6.0, 
6.24 
Table 4-1, 6., 6.2.1, 6.2.2, 6.2.13, 
6.2.15, 6.2.19, 6.2.23, 7. 
Inclusion arguments for cladding degradation and cladding degradation 
mechanisms, fuel rod characteristics, Fuel rods w/breached cladding 
BSC 2004 [DIRS 169987] 6.2.1 Table 4-1, 6.2.22 Fuel pellet fragmentation 
BSC 2004 [DIRS 166692] Entire Table 4-1, 6.2.12, Appendix A Internal Temperature across WP 
BSC 2004 [DIRS 169988] Section 7.5.3 Table 4-1, 6.2.22 Melting temperature of glass 
BSC 2004 [DIRS 169991] Section 7.1 Table 4-1, 6.2.4 Due to the environmental constraints, the microbial activity in the YMP is 
expected to be low, and its impacts on drift chemistry can be insignificant 
BSC 2004 [DIRS 169997] 6.2 Table 4-1, 6.2.14, 6.2.18 Information on WP FEPs 
BSC 2004 [DIRS 167621] Appendix B, 6.3 6.2.3, 6.2.5, 6.2.17, 6.2.20, 6.2.21, 
Table 4-1 
In-package chemistry with radiolysis 
BSC 2004 [DIRS 169565] 6.2, Figure 6.3-67 Table 4-1, 6.2.23, Appendix A Temperature across WP surface 
BSC 2004 [DIRS 170043] Section 8.2, Entire 6.2.3, 6.2.4, 6.2.6, 6.2.20, 6.2.21, 
Table 4-1 
Pitting potentials, chemical environment where pitting is expected 
Greene et al. 2000 [DIRS 145073] Entire, p.7 Table 4-1, 6.2.7, Appendix B Resistance to crevice corrosion 
Hansson 1984 [DIRS 101676] p. 6 Table 4-1, 6.2.20, Appendix B Rapid passivation 
Hayes et al. 1999 [DIRS 164598] Figures 
2,5,6,8,11, Entire 
Table 4-1, 6.2.16, Appendix B Diffusion controlled cavity growth is unlikely if T < 330C to 440C 
Hillner et al. 1998 [DIRS 100455] Entire, p. 6, p. 9, 
p. 25, Fig. 5, 
Table 4, p. 11 
Table 4-1, 6.2.3, 6.2.4 Appendix B General corrosion of zirconium, MIC 
IAEA 1998 [DIRS 150560] p. 92 Table 4-1, 6.2.12, 6.2.24 H2 not absorbed through passive layer 
MO0407SEPFEPLA.000 
[DIRS 170760] 
Entire 1.2, 4.1, 6.1.1, Table 4-1 FEP numbers, titles, and descriptions 
NRC 2002 [DIRS 164593] p. 2 Table 4-1, 6.2.9, Appendix B Little creep damage below 400C 
NRC 1997 [DIRS 101903] Sec. 8.V.1 Table 4-1, 6.2.1 Waterlogging not a problem 
Piron and Pelletier 2001 
[DIRS 165318] 
5.3, Entire, p. 232 Table 4-1, 6.2.10, 6.2.23, Appendix B Helium pressurization 
Wolfram et al. 1996 [DIRS 165268] pp. iii, iv Table 4-1, 6.2.4, Appendix B MIC and Zirconium 
Yau and Webster 1987 [DIRS 
100494] 
Table 15, p. 718, 
Table 6, Entire 
Table 4-1, 6.2.7, 6.2.8, 6.2.11, 6.2.20, 
Appendix B 
Corrosion properties of zirconium, Galvanic corrosion 

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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]). The technical work plan (BSC 2004 
[DIRS 171583], Table 3-1) identifies two additional acceptance criteria (2.2.1.1 and 2.2.1.3.4). 
This report deviates from the technical work plan (BSC 2004 [DIRS 171583]) in that these 
acceptance criteria were deemed to be germane to the individual AMRs and not to this report. 
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 (AC) that will be used by the NRC to evaluate the adequacy of technical arguments are 
identified in Yucca Mountain Review Plan, Final Report (YMRP) (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 provided in the Yucca Mountain 
review plan (YMRP) (NRC 2003 [DIRS 163274], Sections 2.2.1.2.1.3 and 2.2.1.2.2.3). 
Project Requirements Document (Canori and Leitner 2003 [DIRS 166275]) documents and 
categorizes regulatory 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, YMRP acceptance criteria are correlated to 
the corresponding regulations as they pertain to FEPs-related criteria. 
The basis of the NRC review is described in the YMRP (NRC 2003 [DIRS 163274], 
Section 2.2.1.2), and the bases for acceptance are stated as acceptance criteria. The following 
acceptance criteria that apply to this AMR 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 the YMRP 
(NRC 2003 [DIRS 163274]) 
AC 1: The Identification of a List of Features, Events, and Processes Is Adequate 
AC 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 
but also allow for exclusion of a FEP if the process is specifically excluded by the regulations 
(Section 4.2.2). 
Acceptance criteria listed in Section 2.2.1.2.2.3 of the YMRP (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. 

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Table 4-2. Relationships of Regulations to the Project Requirements and the YMRP 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 Criteria 
in the YMRP 
[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 
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 63 [DIRS 156605]. In this document, 
this exclusion is called exclusion by regulation. FEPs screening criteria are described further 
in the following three subsections. 
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. 

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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. 
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 

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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, Review Plan 
(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 
American Society for Testing and Materials (ASTM) Standard C1174-97, Standard Practice for 
Prediction of the Long-Term Behavior of Materials, Including Waste Forms, Used in Engineered 
Barrier Systems (EBS) for Geological Disposal of High-Level Radioactive Waste (ASTM 1998 
[DIRS 105725]) is used to support the degradation model development methodology, categorize 
the model developed with respect to its usage for long-term TSPA, and relate the information or 
data used to develop the model to the requirements of the standard. 
The applicable requirements given in 10 CFR Part 63 [DIRS 156605] are discussed in 
Sections 4.1 and 4.2. 

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5. ASSUMPTIONS 
This section addresses assumptions used in the FEP screening for the cladding degradation FEPs. 
5.1 ANNUAL-EXCEEDANCE PROBABILITY 
Assumption: 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 annualexceedance 
probability. 
Rationale: 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. 
Confirmation Status: Although there may be cases where sufficient data and information exist to 
depart from this assumption, the Poissonian model is generally an effective representation of 
nature and represents a compromise between the complexity of natural processes, availability of 
information, and the sensitivity of results of engineering relevance. Consequently, for geologic 
processes that occur over long time spans, assuming annual equivalence over a 10,000-year 
regulatory period (a relatively short time span for geologic-related events) is reasonable and 
consistent with the basis of probabilistic hazard analyses. Therefore, no further confirmation is 
required. 
Use in the Analysis: 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 NATURALLY OCCURRING EVENTS 
Assumption: It is assumed that potential naturally occurring events, but perhaps of different 
magnitude, have occurred at least once in the past within the geologic record used as the basis for 
determining that factors that could affect the Yucca Mountain disposal system over the 
10,000-year regulatory period. 

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Rationale: This assumption is justified because it is consistent with the regulations used as direct 
input. At 10 CFR 63.305(c) [DIRS 156605], DOE is directed to vary factors related to the 
geology, hydrology, and climate based upon cautious, but reasonable assumptions consistent with 
present knowledge of factors that could affect the Yucca Mountain disposal system over the next 
10,000 years. 
The implication of this assumption is that any discernible impacts or processes related to past 
events on the site setting are reflected in the present knowledge of natural processes that form the 
basis of the TSPA-LA. If the subject FEP phenomena are not reflected or discernible in the data 
used to describe past settings, then they are either of low consequence or of low probability and 
can be excluded from consideration. 
Confirmation Status: Because this assumption is consistent with the regulations, no further 
confirmation is necessary. 
Use in the Analysis: This assumption is used throughout. It is particularly germane to FEPs 
related to processes or phenomena that, speculatively, could affect future states of the system, but 
for which the magnitude and/or coupling to the effect on the repository is not well defined, or for 
which consequences in present time are known to be minor. 
5.3 APPLICABILITY OF REGULATORY REQUIREMENTS 
Assumption: It is assumed that the repository will be constructed, operated, and closed according 
to the regulatory requirements applicable to the construction, operation, and closure period. It is 
further assumed that deviations from the design that is approved by the NRC will be identified 
and corrected. 
Rationale: Inherent in the FEPs evaluation approach is the assumption that the repository will be 
constructed, operated, and closed according to the design used as the basis for this FEP screening 
and that serves as the basis for the NRCs review. 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. Therefore, no further confirmation of the 
assumption is required. 
Engineering and design changes are subject to evaluation to determine if there are any adverse 
impacts to safety as required by 10 CFR 63.73 and in Subparts F and G [DIRS 156605]. See 
also the requirements at 10 CFR 63.32, 10 CFR 63.44, and 10 CFR 63.131 [DIRS 156605]. 
These regulations require the DOE to develop and submit periodic and special reports regarding: 
1. Progress of construction 
2. Any data about the site, obtained during construction, that are not within the predicted 
limits on which the facility design was based 
3. Any deficiencies in design and construction that, if uncorrected, could adversely affect 
safety at any future time 

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4. Results of research and development programs being conducted to resolve safety 
questions. 
Confirmation Status: No confirmation is required for this assumption. 
Use in the Analysis: Any changes in direct inputs listed in Section 4.1, in baseline conditions 
used for this evaluation, or in other subsurface conditions, must be evaluated to determine if the 
changes are within the limits stated in the FEP evaluations. 

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INTENTIONALLY LEFT BLANK 

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6. SCIENTIFIC ANALYSIS DISCUSSION 
This section discusses each of the 24 FEPs dealing with cladding degradation (Table 1-1). The 
purpose of this report is to document the screening of the cladding degradation FEPs. For 
included FEPs, a summary of their implementation in TSPA-LA is presented here. For excluded 
FEPs, the arguments for their exclusion are documented here. 
Appendix A of this report contains an evaluation of the temperatures expected for the cladding 
needed to evaluate some of the FEPs. For FEPs included in the TSPA-LA, alternative 
conceptual models are discussed in the specific report for the included FEP (BSC 2004 
[DIRS 170023]). For excluded FEPs, the discussions of alternative conceptual models, where 
appropriate, are included in this report. The output of this report is used to support the projectspecific 
FEP database and to promote traceability and transparency regarding FEP screening. 
6.1 METHODS AND APPROACH 
The identification and screening of a comprehensive list of FEPs potentially relevant to 
repository postclosure performance are based on site-specific information, design, and 
regulations. This report 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 the version of the LA FEP list 
(DTN: MO0407SEPFEPLA.000 [DIRS 170760] used as initial input in this waste form FEP 
report. 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- 
2). 

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Table 6-1. Indirect Inputs 
Technical Product Input Used From Used In Input Description 
10 CFR 60 [DIRS 103540] Entire Table 6-1, Appendix B QA Requirements 
10 CFR 63 [DIRS 156605] Entire 1.3, 5.2, 5.3, 6.1.6, 6.2, 7.1, 
Table 6-1 
FEPs screening 
p. 4 Table 6-1, 6.2.4 No MIC observed Adler Flitton et al. 2002 [DIRS 161991] 
p. 4 6.2.21, Corrosion and soil contact 
ASTM C 1174-97. 1998 [DIRS 105725] Entire 4.3, Table 6-1 Criteria, Codes, or Standards 
Baker 1992 [DIRS 149104] pp. 2-326, Entire Table 6-1, 6.2.24 Solubility of oxygen in zirconium 
Beckman 2001 [DIRS 156122] p. 103 Table 6-1, 6.2.11 Iodine induced SCC unlikely 
Bradley et al. 1981 [DIRS 101564] p. 38, Entire Table 6-1, 6.2.3 Zircaloy-clad fuel rods before and after in-water 
spent fuel pool storage 
Entire, F3, Table 6-1, 6.2.7, Appendix B Pitting experiments, crevice corrosion Brossia et al. 2002 [DIRS 161988] 
F3 6.2.5, 6.2.6 Corrosion experiments, addition of H2O2 
BSC 2001 [DIRS 155640] Table 18 Table 6-1, 6.2.17 Composition of concentrated J13 
BSC 2001 [DIRS 152059] p. 36, Figure 6.1-2 Table 6-1, 6.2.15 Naval Fuels 
BSC 2004 [DIRS 168030] Entire Table 6-1, 5.1 Seismic information 
BSC 2004 [DIRS 169987] 6.2.1 & 6.2.2 6.2.23 & Table 6-1 CSNF testing, grain and gap inventory 
BSC 2004 [DIRS 169868] Entire Table 6-1, 6.2.14 Melting temperature for glass 
BSC 2004 [DIRS 169898] Sections 6.2.8 through 6.2.10 Table 6-1, 6.2.18 EBS FEPS discussion 
BSC 2004 [DIRS 167621] Figure 6-54 Section 6.2.5 PH profiles 
BSC 2004 [DIRS 169991] Section 7.1 Table 6-1, Appendix B Due to the environmental constraints, the 
microbial activity in YMP is expected to be low, 
and its impacts on drift chemistry can be 
insignificant 
BSC 2004 [DIRS 168960] Entire Table 6-1, 6.2.21 Igneous intrusion on cladding 
BSC 2004 [DIRS 168361] Appendix A Section 2, Table 6-1 List of Q- items 
BSC 2004 [DIRS 167780] Entire Table 6-1, 6.2.14, 6.2.18 Information on Seismic events 
BSC 2004 [DIRS 169183] Section 6.8.2 Table 6-1, Appendix A Discussion on seismic consequence of cladding 
BSC 2004 [DIRS 171583] Entire Entire Work Plan for FEPs documentation 
BSC 2004 [DIRS 168706] Entire 1.2, 4.1, 6.1, 6.1.1, 6.1.2, 6.1.6, 
Table 6-1 
System FEPs 
BSC 2004 [DIRS 170020] Entire 1.3, Table 6-1 Addresses DSNF FEPs 
Canori and Leitner 2003 [DIRS 166275] Entire 4.2.1, Table 4-2, Table 6-1 Project requirements 
Clayton 1989 [DIRS 149208] Tables 1 through 5, p. 270 Table 6-1, 6.2.24 Hydriding of zircaloy 
Cox 1973 [DIRS 152920] Abstract Table 6-1, 6.2.11 Passive layer of oxides on the zirconium surface 
Cox 1990 [DIRS 152778] p. 15, Figure 20 Table 6-1, 6.2.11 Threshold Intensity Factors, SCC 

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Table 6-1. Indirect Inputs (Continued) 
Technical Product Input Used From Used In Input Description 
p.4-13 6.2.20 & Table 6-1 Galvanic corrosion unlikely 
p. 4-15 6.2.11 & Table 6-1 Chloride-induced SCC 
Cragnolino et al. 1999 [DIRS 152354] 
p. 4-27 6.2.11 & Table 6-1 low threshold stress and failure 
CRWMS M&O 1995 [DIRS 102829] p. 15, 16 Table 6-1, 6.2.1 Standard fuel drying 
Entire Table 6-1, 6.2.5, 6.2.17 All corrosion mechanisms for Zircaloy cladding 
4.1.1 6.2.7 Corrosion of zirconium 
6.1.10 6.2.7 Crevice condition 
CRWMS M&O 2000 [DIRS 136058] 
Entire 6.2.20, 6.2.21 Equilibrium-model calculations 
CRWMS M&O 2000 [DIRS 151659] Entire 1.3, Table 6-1, 6.2.6 Initial condition of the cladding 
CRWMS M&O 2001 [DIRS 151662] Entire, 6.2, Table 10B, 6.2.6 Table 6-1, 6.2.9, 6.2.11, 6.2.16 Cladding properties and creep performance 
Debes 1999 [DIRS 161193] Entire Table 6-1, 6.2.18 No rod failures in transportation 
DOE 1996 [DIRS 100320] p. 123 Table 6-1, 6.2.14 Description of assembly 
Einziger 1994 [DIRS 100442] p. 556, eq. 14 Table 6-1, 6.2.3 Oxidation of UO2 is slower in steam than in air 
Entire Table 6-1, 6.2.11 No failures from fuel assemblies and dry 
storage 
EPRI 2002 [DIRS 161421] 
P. xii 6.2.9 Fuel assemblies and exposure 
Farina et al. 2002 [DIRS 163639] p.5 Table 6-1, 6.2.11 Passive layer prevents chloride-induced SCC 
Garde 1991 [DIRS 101652] Pg. 582 Table 6-1, 6.2.22 Fuel Temperatures 
Garzarolli et al. 1979 [DIRS 149256] p. 64, Entire Table 6-1, 6.2.24 Oxygen affects on corrosion of zirconium, gas 
affects 
Greene et al. 2000 [DIRS 145073] F8, F11 6.2.5 & Table 6-1 Effect of adding H2O2 
Gldner and Burtak 1999 [DIRS 171573] Entire Table 6-1, 6.2.3 Information on zircaloy corrosion resistance 
Hansson 1984 [DIRS 101676] Entire Table 6-1, 6.2.8 Corrosion rates in cement pore solutions 
Section 4.2.4 Table 6-1, 6.2.3 Oxidation of zirconium surface IAEA 1998 [DIRS 150560] 
p.220, T8.4 6.2.5 Stability of H2O2 
Jones 1992 [DIRS 169906] Ch. 7 Table 6-1, 6.2.6 Pitting corrosion information 
Knoll and Gilbert 1987 [DIRS 123682] Entire Table 6-1, 6.2.1 Cask drying 
Kohli and Pasupathi 1986 [DIRS 131519] p. iii, Entire Table 6-1, 6.2.1 Water-logged spent fuel storage rods 
Kreyns et al. 1996 [DIRS 100462] p. 767, Figure 5, Entire Table 6-1, 6.2.12 Hydrides could decrease the fracture toughness 
Langmuir 1997 [DIRS 100051] 2.7.2 Table 6-1, 6.2.5 Half-life equation of radioactive decay 

Clad Degradation - FEPs Screening Arguments 
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Table 6-1. Indirect Inputs (Continued) 
Technical Product Input Used From Used In Input Description 
Lanning et al. 1997 [DIRS 101704] V1, p. 3.2, Fig. 3.1) Table 6-1, 6.2.22 Fuel center line temperatures 
Lide 2002 [DIRS 160832] Entire Table 6-1, 6.2.1, 6.2.5 Properties of zirconium 
Little and Wagner 1996 [DIRS 131533] pp. 367 through 368, Entire Table 6-1, 6.2.4 Microbiologically influenced corrosion 
Manaktala 1993 [DIRS 101719] Figure 3-4, pp. 3 to 12 Table 6-1, Appendix B Helium pressure buildup 
McNeil and Odom 1994 [DIRS 131537] p.176 Table 6-1, 6.2.4 Sulfate-reducing bacteria do not affect 
zirconium 
Merriam-Webster 1993 [DIRS 100468] Entire Entire Definitions 
NRC 2002 [DIRS 164593] Entire, Appendix, p. 2 Table 6-1, 6.2.16, Appendix B Temperature limit for Creep and Hydride 
reorientation 
NRC 2003 [DIRS 163274] 2.2.1.2, 2.2.1.2.1.3, Entire 4.2.1, Table 4-2, 4.2.2.3, 6.1, 
6.1.1, 6.1.2, Table 6-1, 7.1.1 
Criteria 
NRC 1997 [DIRS 101903] Sec.8.V.1, Sec.4.V.4.b 6.2.16 & Table 6-1 Omission of water logged rods 
NRC 2000 [DIRS 147797] Entire Table 6-1, 6.2.16 Little creep damage under 400C 
Peehs and Fleisch 1986 [DIRS 102065] pp. 199 to 202, Entire Table 6-1, 6.2.1 Water-logged spent fuel storage rods 
Pescatore and Cowgill 1994 
[DIRS 102066] 
pp. 83 to 85, Entire Table 6-1, 6.2.16 Comparison of creep models for Zircaloy. 
p. 226, p. 312, Entire Table 6-1, 6.2.3, 6.2.24 Stability of water and oxygen in zirconium. 
Ferric ion stability 
Pourbaix 1974 [DIRS 100817] 
P. 583 6.2.17 Stability and solubility of HF 
Puls 1988 [DIRS 102067] pp. 1,507 thru 1,522, Entire Table 6-1, 6.2.12 Series of strain tests on Zircaloy-2 
Ralph et al. 2002 [DIRS 161992] Entire, p.6 Table 6-1, 6.2.4 MIC and galvanic corrosion 
Reed-Hill 1973 [DIRS 121838] Pg. 800 Table 6-1, 6.2.11, 6.2.12 Crack intensity factor 
Table 6 Table 6-1, 6.2.10, Appendix B Helium pressurization 
Entire, p. 6-13 6.2.3 General corrosion of zirconium 
Rothman 1984 [DIRS 100417] 
pp. 33-39 6.2.12 DHC not expected 
Smith 1966 [DIRS 149107] p. 325, Table 3, Entire Table 6-1, 6.2.24 Hydrides, hydrogen absorption and oxygen 
dissolution in zirconium 
Soderman and Jonsson 1996 
[DIRS 149441] 
Entire Table 6-1, 6.2.20 Electrophoresis 
tefanic and LaVerne 2002 
[DIRS 166303] 
Abstract Table 6-1, 6.2.5 Decay constants for H2O2 

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Table 6-1. Indirect Inputs (Continued) 
Technical Product Input Used From Used In Input Description 
Tasooji et al. 1984 [DIRS 102093] Entire, F.12, p. 612 Table 6-1, 6.2.11 Iodine-induced SSC was not a major failure 
mechanism 
Webster 1990 [DIRS 143561] Table 3 p. 665 Table 6-1, 6.2.10 Minimum yield strength of Zircaloy-4 
Wolery and Daveler 1992 [DIRS 100097] Entire Table 6-1, 6.2.5 Description of EQ6 
Wolframet al. 1996 [DIRS 165268] Entire Table 6-1, Appendix B MIC & Zirconium 
Yau 1983 [DIRS 149233] Entire, p. 26/10, p. 26/2 Table 6-1, 6.2.4, 6.2.7, 6.2.8 Discusses pitting and corrosion in chloride 
solutions 
Yau 1984 [DIRS 102050] Entire Table 6-1, 6.2.21 General, crevice, and SCC tests with zirconium 
Yau and Maguire 1990 [DIRS 110761] Entire Table 6-1, 6.2.6 Fluorides and pitting 
Yau and Webster 1987 [DIRS 100494] Entire, Table 6, p. 709 Table 6-1, 6.2.3, 6.2.4, 6.2.5, 
6.2.21 & Appendix B 
Corrosion properties of zirconium 

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6.1.2 Features, Events, 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 FEP screening 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 cladding FEPs. In cases where a FEP 
covers multiple technical areas and is shared with other FEP reports, this report provides a partial 
technical basis for the screening decision as it relates to waste form issues. Collectively, the 
shared FEP reports provide the complete technical basis. 
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 
 Excluded  By Regulation. 
In a few cases, a FEP may be excluded by a combination of criteria (e.g., low probability and 
low consequence), but the final decision can never be listed as both included and excluded. 

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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 inputs 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 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 
document summarizes and consolidates the method of implementation of the FEP in TSPA-LA 
in the form of TSPA disposition statements, based on more detailed implementation information 
in the listed supporting technical reports. For excluded FEPs, this document 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 

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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 CLADDING FEPS SCREENING AND ANALYSES 
Screening information for each of the 24 cladding degradation 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/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 by 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 Degradation of Cladding from Waterlogged Rods (2.1.02.11.0A) 
FEP Description: 
Failed fuel rods (attributed to breaches caused by manufacturing defects and reactor operations) 
comprise a small fraction of the fuel rods that are currently being stored in commercial reactor 
spent fuel pools. Failed fuel contains water in the fuel rod void space that may promote 
degradation of the spent fuel cladding. Such fuel is referred to as waterlogged. The moisture 
remaining in a dried fuel rod is used to determine the extent of degradation of spent fuel 
cladding. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Degradation of cladding and waste form from waterlogged rods is excluded from the TSPA-LA 
because of low consequence. 

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Waterlogged rods are already failed, and further cladding failures of intact rods and waste form 
degradation are not expected because of the small amount of water present. It is assumed that 
only a few rods will be breached at the time they arrive at the repository (Section 6.2.2; 
BSC 2004 [DIRS 170023], Section 6.2.1). The fraction of fuel rods with breached cladding is 
currently estimated to be 0.01 to 1 percent (Section 6.2.2; BSC 2004 [DIRS 170023], 
Section 6.2.1), or about 5.6 rods in an average PWR waste package. The supply of water that 
remains in the fuel after cask drying will not be sufficient to oxidize the fuel to an unacceptable 
level (NRC 1997 [DIRS 101903], Section 8.V.1; Knoll and Gilbert 1986 [DIRS 123682], p. iii). 
Moisture is effectively removed from defective rods during the cask drying operations. Standard 
Review Plan for Dry Cask Storage Systems (NRC 1997 [DIRS 101903], Section 8.V.1) describes 
the cask drying criteria with reference to Knoll and Gilbert 1987 [DIRS 123682]. The residual 
moisture of the cask atmosphere can be estimated based on the drying conditions where the cask 
or waste package is vacuum dried at 5 mbars (maximum water vapor pressure, equivalent to 
7.7 g or 0.43 moles of water). This amount of water produces an insignificant potential for 
corrosion of the cladding during dry storage or during disposal (CRWMS M&O 1995 
[DIRS 102829], p. 16). 
A PWR waste package contains approximately 2,500 kg (2.74  104 moles) of zirconium alloys 
(Lide 2002 [DIRS 160832]). Less than 0.43 mole (7.7 g) of H2O is expected to be present (Knoll 
and Gilbert 1987 [DIRS 123682]) and general corrosion of the zirconium alloys from the water 
would only degrade a small fraction of the zirconium (0.2 mole). Hydrogen uptake would be 
insignificant compared to existing hydrides on the cladding. 
Kohli and Pasupathi (1986 [DIRS 131519], p. iii) discuss removal of water from waterlogged 
fuel rods. Two reactor-breached fuel rods were tested, along with two fuel rods that were 
intentionally damaged after irradiation. A predetermined amount of moisture was added to the 
intentionally damaged rods to enable the extent of the moisture released during the drying to be 
determined. The rods were dried in flowing argon at atmospheric pressure while being heated in 
a furnace. The center 1.8 m of the furnace was heated to 400C; the remainder was heated 
to 200C. The reactor operations breached rods were dried in the as-received condition, then a 
hole was drilled in the cladding, water was injected, and the experiment was repeated. In the 
reactor-breached rods, the bulk of the uncombined water was removed in 1 to 1.3 hours and all 
measurable releases ended after 3.7 hours. This set of experiments demonstrated that standard 
cask drying procedures would remove water from failed rods. 
Peehs and Fleisch (1986 [DIRS 102065], pp. 199 to 202) described the behavior of waterlogged 
PWR fuel rods on heating at 400C in a hot cell. The bulk of the water was released during the 
cask drying operation. Testing results showed that the moisture can be removed from defective 
rods during the cask drying operations and the residual moisture of the cask atmosphere can be 
minimized. 
An alternative conceptual model for waterlogged rods would be that the failed rods contain 
water, which is then released into the waste package to be consumed by the carbon steel rack 
inside the waste package. This alternative conceptual model is possible because of the rate at 
which carbon steel racks react with water vapor to form rust is faster than that for zirconium 
and UO2. The hydrogen that is produced tends not to be absorbed by the zirconium alloys 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-10 October 2004 
(Section 6.2.24). It would lead to no significant damage because of the limited amount of water 
and large volume of steel in a waste package. 
The above results have shown that the amount of water in waterlogged rods is small and that the 
degradation of the cladding is minimal. These results support the conclusion noted above in 
CRWMS-M&O 1995 [DIRS 102829]. 
In conclusion, cladding and waste form degradation from waterlogged fuel rods is excluded from 
TSPA-LA. There are few failed rods in any waste package that could be waterlogged and the 
volume of water inside a rod is quite limited. Because the drying procedure is effective the 
quantity of water is limited. Cladding failure due to rod waterlogging has a low consequence 
and is excluded from further consideration. The magnitude and time of the resulting radiological 
exposures to the reasonably maximally exposed individual, or radionuclide releases to the 
accessible environment, would not be significantly changed by the omission of this FEP 
(degradation from waterlogged rods) from the TSPA-LA model. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.2 Degradation of Cladding Prior to Disposal (2.1.02.12.0A) 
FEP Description: 
Certain aspects of cladding degradation may occur before the spent fuel arrives at Yucca 
Mountain. Possible mechanisms include rod cladding degradation during reactor operation, 
degradation during wet spent fuel pool storage, degradation during dry storage, and rod 
degradation during shipping (i.e., from creep and from vibration and impact) and fuel handling. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
Degradation of cladding prior to disposal is included in TSPA-LA cladding degradation 
abstraction. The failure rate from prior degradation is based on historical data on reactor 
operation. It also includes failure from wet pool storage and transportation (negligible), dry 
storage, and handling (including spent pool events). In the TSPA, this percentage of rods is 
available for radionuclide release through fast release and axial splitting when the waste package 
is breached. It is specified as a 0.01 to 1 percent log uniform distribution (0.1 percent median) 
(BSC 2004 [DIRS 170023], Section 6.2.1). The TSPA-LA abstraction models that all stainless 
steel cladding fails and places the stainless steel cladding into waste packages as it arrives at the 
repository. This results in 3.5 to 7 percent (uniform distribution) of the waste packages 
containing stainless-steel-clad fuel rods (BSC 2004 [DIRS 170023], Section 6.2.2). These waste 
packages contain 15 to 30 percent stainless-steel-clad fuel rods that are failed and available for 

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ANL-WIS-MD-000008 REV 02 6-11 October 2004 
fast release and unzipping (axial splitting) upon waste package failure (BSC 2004 [DIRS 
170023], Section 8.1.2). Stainless steel cladding is modeled for TSPA-LA as failed, therefore, 
this FEP does not apply to stainless steel cladding. 
Supporting Reports: 
BSC 2004 [DIRS 170023] 
6.2.3 General Corrosion of Cladding (2.1.02.13.0A) 
FEP Description: 
General corrosion of cladding could expose large areas of fuel and produce hydrides. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
General corrosion of cladding has been excluded from the TSPA-LA on the basis of low 
consequence. 
The in-package chemistry model (BSC 2004 [DIRS 167621]), which includes equilibrium 
solubilities of phases, solution complexes, and surface adsorption phenomena, estimates a 
minimum pH of 4.5 (lower green dashed line in Figure 6-2) for the solution in contact with the 
cladding. Yau and Webster (1987 [DIR 100494], pp. 709 to 710, Figs. 5, 7, and Table 6) review 
the corrosion potential for zirconium alloys in sulfuric acid. They note that zirconium alloys 
resist attack from H2SO4 at all concentrations up to 70 percent and at temperatures to boiling 
(their Figure 5 is reproduced in this report as Figure 6-1). A concentration of 70 percent H2SO4 
represents a theoretical pH of about 1.15, well below anything expected in the waste package. 
In the range that zirconium alloys show corrosion resistance in H2SO4, a protective film is 
formed on the zirconium that is predominantly cubic zirconium oxide (ZrO2) with only traces of 
monoclinic phases. At concentrations higher than 70 percent, zirconium corrodes because loose 
films form that are zirconium disulfate tetrahydrate and partially zirconium hydrides. In 
concentrations of less than 60 percent H2SO4, zirconium can tolerate some amounts of strong 
oxidizing agents such as 200 ppm Fe3+ and 200 ppm NO3
- (Yau and Webster (1987 
[DIRS 100494], Fig. 7). Moreover, in 20 percent or less H2SO4 (pH > -0.61), zirconium can 
tolerate a great amount of strong oxidizing agents. Consequently, zirconium equipment is often 
used in steel pickling. Zirconium alloys are used in the chemical industry under low pH 
conditions. In the manufacturing of H2O2, zirconium alloys are used to contain up to 65 percent 
H2SO4 at up to 150C. In the manufacturing of HNO3 zirconium alloys are used to contain the 
acid up to 65 percent concentrations and temperatures to 204C. 
At low pHs (below 0.6), the ZrO2 film will start to slowly dissolve. This is discussed in Pitting 
Model for Zirconium-Alloyed Cladding at YMP (BSC 2004 [DIRS 170043]), where there is 
discontinuity in the corrosion potential at about pH = -0.6; below this pH, the corrosion potential 
starts to rapidly decrease. This change in behavior is attributed to dissolution of the general 
oxide surface. Since the in-package chemistry model predicts pHs above 0.6, accelerated 
general corrosion is excluded from TSPA-LA performance models. The Pourbaix diagram for 
zirconium (Pourbaix 1974 [DIRS 100817], p. 226) shows that ZrO2 could start to dissolve at a 

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ANL-WIS-MD-000008 REV 02 6-12 October 2004 
pH of 4. In the text describing the diagram, the author (Pourbaix 1974 [DIRS 100817], pp. 228 
and 229) recognizes that the solubility is much too high (1.2  10-3 moles/liter) and not 
consistent with experimental observations of solubility (10-7 moles/liter). The author attributes 
these high solubilities to not having modeled the dominant form of oxide (such as cubic ZrO2), 
which controls the dissolution. 
General corrosion is synonymous with zirconium oxidation for this repository application. The 
outer surface of the cladding becomes oxidized with a ZrO2 film, which adheres to the surface 
and slows down further oxidation (IAEA 1998 [DIRS 150560], Section 4.2.4). The oxidation 
could be from O2 consumption (dry oxidation) or H2O consumption (wet oxidation). For the fuel 
in the repository, this corrosion does not occur until the waste package is penetrated. 
Source: Yau and Webster 1987 [DIR 100494], Figure 5. 
Figure 6-1. The Iso-Corrosion Diagram for Zirconium in Sulfuric Acid (H2SO4) 
There are three possible effects of surface oxidation: 
1. The oxidation could thin the cladding, contributing to cladding failure by creep 
rupture (Section 6.2.9). Wet oxidation generates hydrogen, and some of the 
hydrogen is absorbed into the cladding to form hydrides. 
2. This hydrogen pick up could lead to delayed hydride cracking (DHC), or general 
hydride embrittlement (Section 6.2.12). 

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ANL-WIS-MD-000008 REV 02 6-13 October 2004 
3. In the extreme, the oxidation could lead to cladding failure and expose the fuel 
pellets to the waste package environment. 
The in-package chemistry model predicts that the pH will remain above 4.5. Under these 
nominal chemical conditions in the repository, general corrosion failures of the cladding are 
unlikely. Waterside Corrosion of Zirconium Alloys in Nuclear Power Plants (IAEA 1998 
[DIRS 150560]) summarizes much of the research on zirconium corrosion. Hillner et al. (1998 
[DIRS 100455], p. 9) studied corrosion of Zircaloy and published a Zircaloy corrosion 
correlation based on Bettis Atomic Power Laboratory experiments. Bettis Atomic Laboratory 
developed Zircaloy for naval reactors in the early 1950s and has an extensive database on 
Zircaloy performance, including continuous autoclave corrosion tests on some samples for 30 
years. Some samples have developed oxide thickness as great as 110 m, greater than those 
expected during repository corrosion based on Hillner et al. calculations. The experiments are 
consistent with diffusion of oxygen ions through the corrosion film being the rate-limiting 
phenomenon. This corrosion film is generated in layers, with the physical characteristics of 
lower layer staying consistent (uniform). The consistency of the lower 2 m of oxide film leads 
to a steady corrosion rate after a transition period. 
Hillner et al. (1998 [DIRS 100455], Figure 5) compare the weight gain of the samples in water, 
which correlates to the corrosion rate in water (the correlation used here) with that of steam. The 
steam corrosion rate is about 30 to 40 percent slower. Einziger (1994 [DIRS 100442], p. 556, 
Equation 14) states that dry oxidation of zirconium is slightly slower than the wet corrosion rate. 
There are experimental corroborating observations about the slow corrosion rates of zirconium 
alloys near ambient temperature (27C). Bradley et al. (1981 [DIRS 101564], p. 38) performed 
metallurgical examinations of Zircaloy-clad fuel rods from two assemblies (0551 and 0074) of 
the Shippingport PWR Core 1 blanket fuel after extended in-water spent fuel pool storage 
(21 years for 0551, and 16 years for 0074). The oxide film thickness on the Shippingport fuel 
rods after reactor operation was reported to be an average cladding oxide film thickness of 1.8 
m (0551) and 2.4 m (0074). After extended in-water spent fuel pool storage, the average 
cladding oxide film thickness was found to be 1.7 m (0551) and 2.3 m (0074) (Bradley et al. 
1981 [DIRS 101564], p. 38). The slight disagreement in these values is attributed to differences 
in measurement technique and experimental error. These results led to the conclusion that no 
significant change in oxide thickness occurred even after 16 to 21 years of pool storage. This 
conclusion is further supported by the observation that Zircaloy tube sheets (that had been cut to 
remove assembly 0551 fuel rods in 1960) stored in water for over 20 years were unblemished 
and showed no evidence of reaction with water. 
Rothman (1984 [DIRS 100417], pp. 6 to 13, Table 3) discusses cladding oxidation in repository 
conditions in great detail and compares the predicted cladding loss using six different oxidation 
correlations and predicts cladding thinning of 4 to 53 m after 10,000 years at 180C 
(a conservative temperature condition as the repository cools after a few hundred years), which 
would not lead to rod failure. 
The cladding corrosion analyses (Hillner et al. 1998 [DIRS 100455], Rothman 1984 
[DIRS 100417]) are based on the performance of Zircaloy-4 (Zirc-4), the cladding used in most 
PWRs before the year 2000. Since then, Zircaloy variants with increased corrosion resistance 

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ANL-WIS-MD-000008 REV 02 6-14 October 2004 
(i.e., zirconium-niobium alloys) have been available for reactor reloads. The improved corrosion 
resistance of these cladding materials yields lower corrosion thickness compared to Zircaloy-4, 
even though the burnup has significantly increased over this time (Guldner and Burtak 1999 
[DIRS 171573]). The hydride content of the cladding is directly proportional to the oxide 
thickness. Therefore, the new cladding is less likely to fail from general corrosion or hydride 
embrittlement (Section 6.2.12). The thinner oxide thickness produces lower fuel pellet 
temperatures and, thus, lowers fission gas release fractions. The remaining metal thickness is 
also greater. Therefore, cladding stresses are lower in advanced alloys and less creep is expected 
(Section 6.2.9). 
Stainless steel cladding is modeled for TSPA-LA as failed, therefore, this FEP does not apply to 
stainless steel cladding. 
Cladding degradation from general corrosion is excluded from TSPA-LA. The small amount of 
corrosion that will occur during the first 10,000 years following permanent closure of the 
repository will not penetrate the cladding and therefore will not affect the release of 
radionuclides. Cladding failure due to general corrosion has a low consequence and is excluded 
from further consideration. The magnitude and time of the resulting radiological exposures to 
the reasonably maximally exposed individual, or radionuclide releases to the accessible 
environment, would not be significantly changed by the omission of this FEP (general corrosion 
of cladding) from the performance assessment (TSPA-LA) model. 
TSPA Disposition: 
N/A 
Support Reports: 
N/A 
6.2.4 Microbially Influenced Corrosion (MIC) of Cladding (2.1.02.14.0A) 
FEP Description: 
Microbially influenced corrosion (MIC) of cladding is a potential localized corrosion mechanism 
where microbes produce a local acidic environment that could produce multiple penetrations 
through the fuel cladding. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Microbially influenced corrosion (MIC) activity is excluded as a cladding failure mechanism 
because of low consequence. Microbes are expected to be present in the repository, but MIC of 
cladding is not expected to cause cladding failure and, therefore, not have a significant effect on 
radionuclide exposures to the RMEI. 
The term microbially influenced corrosion (MIC) is commonly used to designate corrosion 
caused by the presence and activities of microorganisms at the surfaces of metals. 

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ANL-WIS-MD-000008 REV 02 6-15 October 2004 
Two studies of MIC on spent nuclear fuel have been performed. Wolfram et al. (1996 
[DIRS 165268], pp. iii and iv) measured microbial activity in spent fuel pools. They concluded 
that all spent fuel pools tested contained microbial colonies. They also performed a literature 
search and concluded that There was no evidence found in the literature that zirconium or its 
alloys are susceptible to MIC. 
Hillner et al. (1998 [DIRS 100455], p. 11) studied the corrosion of Zircaloy-clad fuels under 
repository conditions. They indicate that there are two major forms of MIC for materials being 
considered for waste packages. They are (1) sulfide attack through the action of sulfate reducing 
bacteria (SRB), and (2) corrosion induced by organic acids secreted from certain bacteria. With 
respect to attack by SRB, Hillner et al. (1998 [DIRS 100455]) reference the work of McNeil and 
Odom (1994 [DIRS 131537], p. 176), which indicates by thermodynamic calculations that SRB 
do not affect zirconium alloys. With respect to corrosion induced by organic acids, 
Hillner et al. (1998 [DIRS 100455], p. 11) noted that it is most unlikely because of zirconiums 
tolerance of a wide range of pHs and it is unlikely that production of weak organic acids will 
have an adverse effect on the passivation of Zircaloy by a ZrO2 film. Yau and Webster (1987 
[DIRS 100494], p. 717) also note that zirconium alloy resists a wide range of organic 
compounds, including acetic acid, acetic anhydride, formic acid, urea, ethylene dichloride, 
formaldehyde, citric acid, lactic acid, oxalic acid, tannic acid, and trichloroethylene. This 
supports the concept that organic solutions produced by MIC are unlikely to cause significant 
acceleration of the corrosion of zirconium alloys. Evaluation of Potential Impacts of Microbial 
Activities on Drift Chemistry (BSC 2004 [DIRS 169991], Section 7.1) evaluated in-drift 
microbial communities and concluded that the estimates of microbial masses growing in the 
potential repository system suggest that the effect to the in-drift geochemistry should be small. 
Little and Wagner (1996 [DIRS 131533], p. 367) corroborate this information in an overview of 
MIC of metals and alloys used in the storage of nuclear wastes. They indicate that MIC is a form 
of localized corrosion that results in pitting, selective leaching, crevice corrosion, under-deposit 
corrosion, and enhanced erosion and corrosion. Little and Wagner (1996 [DIRS 131533], 
pp. 367 and 368) describe several mechanisms for MIC. In addition, various case studies are 
presented that document MIC of alloys of iron, nickel, and copper. However, it should be noted 
that no indication has been found in the literature of MIC occurring on zirconium metal or alloys. 
Yau and Webster (1987 [DIRS 100494], p. 709) report no corrosion of zirconium alloy from 
marine organisms was found during seawater corrosion tests for 129 days. 
MIC is excluded as a component of the localized (pitting) corrosion model where MIC could 
cause a localized suppression of the water pH and permit other aggressive species to attack 
cladding. The zirconium alloy pitting model (BSC 2004 [DIRS 170043], Section 8) shows that 
pitting is dependent on the concentration of chlorides, ferric ions, and hydrogen peroxide (from 
radiolysis). 
One U.S. commercial nuclear plant spent fuel pool experienced a significant MIC event that 
lasted for about four years (Ralph et al. 2002 [DIRS 161992]). After an extended lay-up period, 
the spent fuel pool water was found to contain a significant amount of algae and bacteria. 
Biological agents were purged using controlled additions of chlorine and hydrogen peroxide 
before the pool was returned to normal operating chemistries. 

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ANL-WIS-MD-000008 REV 02 6-16 October 2004 
The assessment revealed that the steel rack corrosion products were up to 2.5 cm thick, and they 
had started to engulf the individual fuel rods or flow channels of the stored assemblies in the 
region where the rack and plates contacted the fuel assemblies. The corrosion product had 
adhered to the fuel. The iron oxide was composed of FeO, Fe2O3, and Fe3O4. Ralph et al. (2002 
[DIRS 161992], p. 6) state: 
One fuel assembly was removed from its storage location and its channel was 
removed. The oxide from contact with the carbon steel rack was removed with a 
water lance utilizing 350 to 700 kg/cm2 of water pressure. A camera with 
resolution of 0.025 mm was used to inspect the channel. The channel surface 
appeared uniform and smooth. No pitting, white discoloration or surface 
anomalies were observed. 
The paper concluded: The fuel cladding was not affected through any type of corrosion. 
Therefore, the corrosion did not change the classification of the fuel as intact or damaged. As 
with the experiments by Yau (1983 [DIRS 149233]), the lack of pitting or stress corrosion 
cracking implies that the corrosion potential (Ecorr) was not elevated to exceed the repassivation 
potential (Erp) and thus cause corrosion, even with Fe3O4 present and adhering to the zirconium 
oxide film. MIC colonies could also have locally suppressed the pH but, again, no localized 
corrosion was observed. 
Adler Flitton et al. (2002 [DIRS 161991], p. 4) buried various metal samples in an arid vadose 
zone environment for three years. They reported indications of pitting from MIC on some of the 
metals, but observed no pitting on the zirconium alloy samples. 
Stainless steel cladding is modeled for TSPA-LA as failed, therefore, this FEP does not apply to 
stainless steel cladding. 
In summary, microbes are expected to be present in the repository, but MIC of cladding is not 
expected to cause cladding failure. The NRC requirements in 10 CFR 63.114 (e and f) allow the 
exclusion of MIC from the TSPA-LA because the omission would not significantly change the 
magnitude and time of the resulting radiological exposures to the reasonably maximally exposed 
individual (RMEI), or radionuclide releases to the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.5 Localized (Radiolysis Enhanced) Corrosion of Cladding (2.1.02.15.0A) 
FEP Description: 
Radiolysis in a nitrogen/oxygen gas mixture with the presence of water film results in the 
formation of nitric acid (HNO3). Hydrogen peroxide (H2O2) is formed in the water from 
radiolysis. These chemicals can enhance corrosion of the fuel cladding. 

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ANL-WIS-MD-000008 REV 02 6-17 October 2004 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Radiolysis, as a cladding failure mechanism, is excluded from the TSPA-LA on the basis of low 
consequence. 
The in-package chemistry model (BSC 2004 [DIRS 167621], Appendix B) addressed the change 
of water chemistry with the inclusion of nitric acid and hydrogen peroxide production from 
radiolysis. In the analysis, all of the nitric acid that can be produced in a moist waste package 
was absorbed into the water film on the cladding surface. The radiation field was modeled as 
being constant at the radiation levels expected at 500 years following repository closure, 
although the level would, in fact, decrease with the passing of time. 
The effect of radiolysis on the in-package chemistry was handled in a nonmechanistic manner 
(i.e., EQ6 does not have the facility to model the radiolysis process directly), therefore, only the 
products of radiolysis were included as inputs in EQ6 simulations. EQ6 is a reaction path code 
that models water, rock or other material interactions or fluid mixing in either a pure reaction 
progress mode or a time mode (Wolery and Daveler 1992 [DIRS 100097]). A series of runs was 
performed where nitric acid and hydrogen peroxide were included as inputs in EQ6 runs and the 
results of these simulations were compared to their nonradiolysis counterparts. Two base-case 
runs (C12C25 and C22C25; BSC 2004 [DIRS 167621], Section 6.3) were used to test the effects 
of radiolysis. These runs were chosen because they represent the median fuel exposure value 
and the low and median water flux values. It would be expected that the effects of nitric acid and 
hydrogen peroxide additions would be greater at the low end of the flux range. Two simulations 
were performed for each file, the first using the base-case nitric acid and hydrogen peroxide rates 
and the second multiplying the HNO3 and H2O2 rates by a factor of ten. In addition, one 
simulation (C12C10x_S20x) was conducted with twenty times the sulfur in carbon steel. 
The results of the simulations are displayed in Figure 6-2, which includes the results of the 
surface complexation pH limits, where the pH profiles for the various runs are displayed versus 
time. These results show that neither the base case nor the 10 base-case generation rates of 
HNO3 and H2O2 had an impact on the in-package pH. Therefore, it may be concluded that if 
radiolysis only affects the chemistry via HNO3 and H2O2 generation, then it will not be a 
significant process with regard to influencing the in-package chemistry. The radiolysis did not 
significantly affect the concentrations of Cl-, Fe3+, or H2O2 and therefore did not change the 
corrosion potential of the passive film on the zirconium alloy. The Naval Nuclear Propulsion 
Program cladding and SNF performance is represented by Figure 6-54 in In-Package Chemistry 
Abstraction (BSC 2004 [DIRS 167621]), and is not indicative of the surface complexation pH 
profiles shown on Figure 6-2. 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-18 October 2004 
Source: BSC 2004 [DIRS 167621], Figure B-1. 
Figure 6-2. pH Profiles Showing Radiolysis Impacts on pH 
Table 6-2 provides the chloride, ferric iron, and hydrogen peroxide maximum concentrations for 
the simulations. Again, runs with HNO3 and H2O2 input show little deviation compared to the 
runs without. 
Table 6-2. Chloride, Ferric Iron, and Hydrogen Peroxide Molality 
Maximum Molalities (m) 
EQ6 Input File Cl- Fe+++ H2O2(aq) 
C22C25 6.8E-04 5.1E-11 3.7E-19 
C22CBC 6.6E-04 8.5E-11 3.7E-19 
C22BC10x 6.5E-04 8.5E-11 3.7E-19 
C12C25 9.7E-04 5.3E-11 3.7E-19 
C12CBC 9.6E-04 5.3E-11 3.7E-19 
C12C10x_S20x 9.7E-04 3.8E-06 3.7E-19 
Source: BSC 2004 [DIRS 167621], Table B-2. 
The Center for Nuclear Waste Regulatory Analysis (CNWRA) performed a series of corrosion 
tests where hydrogen peroxide was added to the ongoing test while the corrosion potential was 
being measured. Greene et al. (2000 [DIRS 145073], Figure 8) shows two experiments where 
H2O2 was added and the corrosion potential was measured. In one test, H2O2 was added two 
different times. In the three cases where H2O2 was added, the effect of the hydrogen peroxide 
rapidly died out. In another test (Greene et al. 2000 [DIRS 145073], Figure 11), a sample that 
was oxidized in air at 200C was exposed to a solution of 1M NaCl. When 5 mM H2O2 was 
added, the corrosion potential increased by 0.275 VSCE (volts, saturated calomel electrode scale) 
and pitting was observed. In this experiment the corrosion potential normally is nominally 0.07 

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ANL-WIS-MD-000008 REV 02 6-19 October 2004 
VSCE, and the repassivation potential is 0.04 VSCE, so the increase in corrosion potential is 
significant. The concentrations of chloride and hydrogen peroxide in this experiment are many 
orders of magnitude higher than expected in the waste package (see concentrations in Table 6-2). 
Therefore, the corrosion potential in the repository should not increase in the same proportion as 
these experiments. 
Brossia et al. (2002 [DIRS 161988], Figure 3) report two experiments where H2O2 was added to 
ongoing corrosion potential tests. The metal samples had oxide coatings of 1.7 m and 3.4 m 
thick. The initial solution contained 0.1 M NaCl at 95C and 5 mM H2O2 was added. In both 
tests the corrosion potential initially increased, but later one test showed decreasing corrosion 
potentials. Pitting was not observed in either experiment. Again, these concentrations are higher 
than expected in the in-package chemistry. 
Greene et al. (2000 [DIRS 145073], Figure 8) discuss the stability of H2O2. International 
Atomic Energy Agency suggests the that H2O2 is known to decompose catalytically on the 
surfaces of various types of materials at ambient temperatures, and that zirconium oxides 
enhance decomposition (IAEA 1998 [DIRS 150560], p. 220). H2O2 decomposes in bulk 
solutions at elevated temperatures with a decay rate, k = 6.5  105  exp(-71 kJ/mol/RT)), in 
units of inverse seconds (tefanic and LaVerne 2002 [DIRS 166303], Abstract). Table 6-3 
provides the half-life of H2O2 at various temperatures, calculated from t1/2 = 0.693/k (Langmuir 
1997 [DIRS 100051], Section 2.7.2) and R = 0.008314 kJ/molK (Lide 2002 [DIRS 160832]), 
and shows that bulk decomposition during the regulatory period will be significant. 
The in-package model, coupled with the pitting experiments and pitting model (Section 6.2.6 and 
Table 6-4), shows that radiolytic production of nitric acid and hydrogen peroxide is not sufficient 
to influence the corrosion potentials significantly at the steady-state concentrations expected in 
the repository and to produce pitting. Zirconium alloys have been shown to be relatively inert in 
nitric acid and hydrogen peroxide as discussed by Yau and Webster (1987 [DIRS 100494]) and 
in Clad DegradationLocal Corrosion of Zirconium and Its Alloys Under Repository Conditions 
(CRWMS M&O 2000 [DIRS 136058], Section 6.1.6 and Appendix II.4). For example, the 
chemical processing industry uses peroxide strengths of 90 percent with zirconium equipment. 
The service life has been increased by an order of magnitude compared to graphite components 
previously used, which were generally considered to be inert. In nitric acid, zirconium and its 
alloys are inert up to acid concentrations of 65 weight percent. Since radiolysis does not produce 
nitric acid and hydrogen peroxide at greater concentrations, there will be no impact (low 
consequence) on the uniform corrosion rate as a result of radiolysis. 
Table 6-3. Half-Life of Hydrogen Peroxide at Various Temperatures 
Temperature (C) Half-Life (Days) 
25 34 
50 3.79 
73 0.65 
95 0.15 
100 0.11 
Source: tefanic and LaVerne 2002 [DIRS 166303], Abstract. 

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ANL-WIS-MD-000008 REV 02 6-20 October 2004 
Stainless steel cladding is modeled for TSPA-LA as failed; therefore, this FEP does not apply to 
stainless steel cladding. Naval Nuclear Propulsion Program cladding and SNF performance is 
discussed in Naval Nuclear Propulsion Program Technical Support Document for the License 
Application, which is a classified document. 
Cladding degradation from radiolysis-enhanced corrosion is excluded from TSPA-LA. 
Radiolytic production of nitric acid and hydrogen peroxide was included in the in-package 
chemistry model (BSC 2004 [DIRS 167621], Appendix B), which showed that radiolysis had a 
small effect on the chemistry. Experiments where hydrogen peroxide was added to tests show 
that in many cases the effect of the hydrogen peroxide quickly becomes negligible. Radiolysis 
by itself is not expected to damage the cladding (low consequence). Cladding failure due to 
radiolysis-enhanced corrosion has a low consequence, and is excluded from further 
consideration. The magnitude and time of the resulting radiological exposures to the reasonably 
maximally exposed individual, or radionuclide releases to the accessible environment, would not 
be significantly changed by the omission of this FEP (radiolysis enhanced corrosion) from the 
performance assessment (TSPA-LA) model. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.6 Localized (Pitting) Corrosion of Cladding (2.1.02.16.0A) 
FEP Description: 
Localized corrosion in pits could produce penetrations of cladding. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Localized corrosion of the cladding from pitting has been excluded on the basis of low 
consequence. A comparison of the expected in-package chemistry to the chemical composition 
where pitting is observed shows that pitting is not expected. Therefore, the inclusion of a pitting 
model in TSPA-LA would not have a significant effect on the magnitude and time of 
radiological exposures or radionuclide releases (low consequence). 
A zirconium-pitting model (BSC 2004 [DIRS 170043]) was developed to investigate the 
chemical conditions at which pitting occurs. Zirconium alloys are susceptible to pitting in a 
particularly aggressive combination of chloride (Cl-) ions, ferric ions (Fe+3), or hydrogen 
peroxide (H2O2). In order to predict cladding failure from chloride pitting, a review of the 
literature for pitting rates and electrochemical data for various zirconium alloys was conducted. 
Based on this review of the literature, failure criteria were constructed based on an 
electrochemical definition of pitting as the condition at which the corrosion potential for aspolished 
metal exceeds repassivation potential (i.e., Ecorr > Erp). Corrosion potential and 
repassivation potential values were obtained for as-polished zirconium alloys in various solution 
concentrations of Cl-, Fe+3, and H2O2 using measurements obtained from various experiments. 

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ANL-WIS-MD-000008 REV 02 6-21 October 2004 
The model to predict repassivation potential depends only on chloride concentration in the 
solution. The corrosion potential for as-polished metal (Ecorr) was modeled by performing a 
regression analysis to fit experimental data with varying molar concentrations of Cl-, Fe+3, and 
H2O2. The model describes the conditions where pitting was observed in experiments. High 
concentrations of chlorides at extremely low pH (below -0.6) can lead to the general dissolution 
of the protective zirconium oxide film. 
This model was evaluated using in-package chemistry, including the production of nitric acid 
and hydrogen peroxide from radiolysis (Section 6.2.5 and Table 6-2). Table 6-4 provides the 
corrosion potential for as-polished metals, repassivation potential, and potential differences for 
the cases described in Table 6-2. Table 6-4 shows that pitting is not expected because the 
repassivation potential exceeds the corrosion potential for as-polished metal. 
Table 6-4. Corrosion and Repassivation Potentials for In-Package Chemistry Conditions 
Case 
# 
Ecorr 
VSCE 
Erp 
VSCE 
Differencea (Erp-Ecorr) 
VSCE 
C22C25 -0.22 0.30 0.52 
C22CBC -0.22 0.30 0.52 
C22BC10x -0.22 0.30 0.52 
C12C25 -0.21 0.29 0.50 
C12CBC -0.21 0.29 0.50 
C12BC10x -0.21 0.29 0.50 
Source: BSC 2004 [DIRS 170043], Table 8-1. 
NOTE: a Negative difference implies pitting is possible. 
The pitting model was generated with data for as-polished and atmospherically oxidized 
samples, but CSNF fuel rods are coated with thick oxides (54 m mean, 5 to 95 percent m 
range = 112 m to 5.3 m; CRWMS M&O 2000 [DIRS 151659], Section 6.4). These oxides 
will affect the measured corrosion potential (open circuit potential) because of the high electrical 
resistance of the coating, but will not increase susceptibility to pitting corrosion. This sensitivity 
to oxide thickness was demonstrated in a series of four tests performed on zirconium samples at 
the Center for Nuclear Waste Regulatory Analyses. Two tests had oxide thickness of 1.7 m and 
two tests had 3.4 m coatings (Brossia et al. 2002 [DIRS 161988], Figure 3). These samples 
were exposed to 0.1M of NaCl. Prior to addition of oxidants to the solution, the highest 
corrosion potential was 0.67 VSCE, well above the repassivation potential of 0.12 VSCE. Two 
other samples also had potentials above the repassivation potential. No pitting was observed in 
the samples. Yau and Maguire (1990 [DIRS 110761], Fig. 4) showed that zirconium annealed in 
air (air oxidized) also had a corrosion potential that was above the as-polished value and the 
repassivation potential. Again, no pitting was observed. These observations show that the 
corrosion potential for oxide-coated material can exceed the repassivation potential and pitting 
will not occur. Ecorr > Erp is a necessary but not sufficient condition for pitting. For the case 
where the corrosion potential is raised above the repassivation potential only due to oxide 
formation, pitting will not occur, as discussed below. 
When pitting occurs on metal surfaces, the pit behaves as an anode and the surrounding metal 
surface behaves as a cathode. The metal in the pit gives up electrons, becomes oxidized, and 

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ANL-WIS-MD-000008 REV 02 6-22 October 2004 
goes into solution (Jones 1992 [DIRS 169906], Chapter 7). To support pitting, the electrons 
must be conducted to the free surface of the surrounding metal and be accepted by an oxidizing 
agent (e.g., H2O2, Fe3+, or O2). The potential for pitting starts when some event creates a site 
with less protection than the surrounding metal. For pitting to then occur, three conditions must 
be satisfied: 
1. The surface of the surrounding metal must be able to transport the electrons into 
the solution 
2. Oxidizing agents in the solution must be present to accept the electrons 
3. The pit site must support transport of the metal ions into solution. 
Zirconium oxide is a poor conductor and a thick oxide coating restricts the current of electrons 
from the metal into the solution (Condition 1 above). This insulation results in the measured 
corrosion potential increasing from the Zircaloy corrosion potential toward the oxide-surface 
redox-reaction potential as the oxide insulation increases. The redox potential is not relevant to 
pitting susceptibility but is a consequence of inerting of the Zircaloy metal. If the oxide has a 
defect, the metal corrosion potential decreases to that of the bare metal, and pitting occurs only if 
oxidizing species exist in solution or an applied potential force the bare-metal Zircaloy corrosion 
potential above the repassivation potential. If there are insufficient oxidizing agents in the 
solution to support a critical current (Condition 2 above), then a new protective oxide surface 
will form on the potential pit surface and prevent dissolution of the metal (Condition 3 above). 
Stainless steel cladding is modeled for TSPA-LA as failed; therefore, this FEP does not apply to 
stainless steel cladding. 
Although the corrosion potential (Ecorr) is greater than the repassivation potential (Erp) for 
zirconium with thick oxide layers, both empirical evidence and understanding of the response to 
local breakdown of the protective oxide layer (e.g., due to oxide layer cracking) indicate that 
pitting does not occur. The empirical evidence discussed in the above paragraph (Brossia et al. 
2002 [DIRS 161988] Figure 3; Yau and Maguire 1990 [DIRS 110761], Fig. 4) showed that 
zirconium with thick oxide layers did not pit. The oxide layer on the zirconium surrounding the 
breakdown region prevents sufficient cathodic current to support pitting. Therefore, the anodic 
and cathodic reactions supporting corrosion occur in the breakdown region, a condition that 
cannot occur for pitting. In addition, upon breakdown of the oxide, the corrosion potential drops 
to about the corrosion potential of the as-polished zirconium. In effect, the breakdown region 
will not be influenced by the surrounding oxidized zirconium surface and will behave as aspolished 
zirconium. Therefore, results obtained using as-polished zirconium coupons apply in 
assessing the corrosion behavior in the breakdown region. For the experiments described by 
Brossia et al. (2002 [DIRS 161988], Figure 3), the corrosion potential for an as-polished surface 
(-0.11 VSCE) is significantly less than the repassivation potential (0.12 VSCE), and pitting was 
neither observed nor expected to occur. 
In conclusion, cladding degradation from localized (pitting) corrosion of the cladding is excluded 
from TSPA-LA. A comparison of the expected in-package chemistry to the chemical 
composition where pitting is observed shows that pitting is not expected. Cladding failure due to 

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ANL-WIS-MD-000008 REV 02 6-23 October 2004 
pitting has a low consequence, and is excluded from further consideration. The magnitude and 
time of the resulting radiological exposures to the reasonably maximally exposed individual, or 
radionuclide releases to the accessible environment, would not be significantly changed by the 
omission of this FEP (localized (pitting) corrosion of the cladding) from the performance 
assessment (TSPA-LA) model. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.7 Localized (Crevice) Corrosion of Cladding (2.1.02.17.0A) 
FEP Description: 
Localized corrosion in crevices could produce penetrations of cladding. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Localized (crevice) corrosion of the cladding is excluded from the TSPA-LA on the basis of low 
consequence. 
Yau and Webster (1987 [DIRS 100494], p. 717) report: Of all the corrosion-resistant structural 
metals, zirconium and tantalum are the most resistant to crevice corrosion. In low-pH chloride 
solutions or chlorine gas, for example, zirconium is not subject to crevice attack. Greene et al. 
(2000 [DIRS 145073]) and Brossia et al. (2002 [DIRS 161988]) performed pitting and crevice 
corrosion tests on Zircaloy-4. They covered temperatures from 25C to 95C, chloride 
concentrations from 0.001 to 4.0 M, and pH from 2.1 to 10.7. The solutions also contained the 
predominant anions in the groundwater. Some of their tests had sufficiently aggressive solutions 
to cause pitting on exposed surfaces. Other tests had voltages applied to the sample to raise the 
corrosion potential above the repassivation potential and cause pitting on exposed surfaces. 
They report that no crevice corrosion is observed under the same environment and 
electrochemical conditions that promote pitting corrosion on exposed surfaces. In summary, 
crevice corrosion is not observed under severe conditions that promote pitting on the exposed 
surfaces. 
Additional corroborating information is also available. More detailed information is provided by 
Yau (1983 [DIRS 149233]) showing that zirconium and Zr-1.5 percent Sn were resistant to 
crevice corrosion after 14 days exposed to boiling (107C), saturated NaCl solution with the pH 
adjusted to 0 by the addition of HCl. Clad Degradation-Local Corrosion of Zirconium and its 
Alloys Under Repository Conditions (CRWMS M&O 2000 [DIRS 136058]) summarizes these 
experiments and shows that zirconium is not susceptible to crevice corrosion. Section 4.1.3 of 
that report (CRWMS M&O 2000 [DIRS 136058]) discusses the crevice corrosion resistance of 
zirconium in various chemical solutions, summarizes seven crevice corrosion tests, and reports 
that crevice corrosion was not observed. The U-bend tests discussed in Section 4.1.4 of Clad 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-24 October 2004 
DegradationLocal Corrosion of Zirconium and Its Alloys Under Repository Conditions 
(CRWMS M&O 2000 [DIRS 136058]) are also designed to produce crevice corrosion under the 
U-bend test washers. In these tests, no crevice corrosion was reported. Section 6.1.10 of Clad 
DegradationLocal Corrosion of Zirconium and Its Alloys Under Repository Conditions 
(CRWMS M&O 2000 [DIRS 136058]) discusses the theoretical reasons why zirconium is 
immune to this type of corrosion. 
Stainless steel cladding is modeled for TSPA-LA as failed; therefore, this FEP does not apply to 
stainless steel cladding. 
In conclusion, cladding degradation from localized (crevice) corrosion is excluded from 
TSPA LA. Crevice corrosion of zirconium under repository in-package chemistry conditions is 
not expected. NRC requirements in 10 CFR 63.114 (e and f) [DIRS 156605] allow the omission 
because crevice corrosion is not expected under repository conditions and therefore will not 
significantly change the magnitude and time of the resulting radiological exposures to the 
reasonably maximally exposed individual (RMEI), or radionuclide releases to the accessible 
environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.8 Enhanced Corrosion of Cladding from Dissolved Silica (2.1.02.18.0A) 
FEP Description: 
High dissolved silica content of waters may enhance corrosion of cladding. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Enhanced corrosion of cladding due to high dissolved silica content in waters is excluded from 
the TSPA-LA on the basis of low consequence. 
The potential for silica itself degrading the cladding is negligible. Hansson (1984 
[DIRS 101676]) reports corrosion tests with concrete pore fluids that normally contain silica. 
Yau (1983 [DIRS 149233]) reports corrosion tests in seawater that also contains silica. Neither 
experimenter reports significant corrosion. Yau and Webster (1987 [DIRS 100494], Table 6) 
review the corrosion potentials for zirconium and report no corrosion with sodium silicate 
concentrations from 0 to 100 weight percent at ambient temperature to 100C. 
Silicon dioxide reacts with hydrofluoric acid to form fluorosilicic acid, which is highly corrosive 
to zirconium in high concentrations (10 weight percent) (Yau and Webster 1987, 
[DIRS 100494], p. 712). This reaction is not expected to occur because the pH is generally too 
high (pH>3.18) for hydrofluoric acid to exist and fluoride concentrations are too low (J13 well 

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ANL-WIS-MD-000008 REV 02 6-25 October 2004 
water contains only 2.2 ppm) to form significant concentrations (less than 2  10-5 M) of 
fluorosilicic acid. The fluoride corrosion itself is addressed in Section 6.2.17 and is excluded. 
Stainless steel cladding is modeled for TSPA-LA as failed; therefore, this FEP does not apply to 
stainless steel cladding. 
In conclusion, enhanced cladding degradation from dissolved silica is excluded from TSPA-LA. 
The NRC requirements in 10 CFR 63.114 (e and f) [DIRS 156605] allow the omission because 
silica will not degrade cladding and therefore will not significantly change the magnitude and 
time of the resulting radiological exposures to the reasonably maximally exposed individual 
(RMEI), or radionuclide releases to the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.9 Creep Rupture of Cladding (2.1.02.19.0A) 
FEP Description: 
At high temperatures (>400 degrees C) for sufficiently long time intervals, creep rupture of 
Zircaloy cladding on spent fuel can occur and produce small perforations in the cladding to 
relieve stress. After the waste package fails, the fuel can react with water and radionuclides can 
escape over time from the fuel rod. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Creep rupture of cladding is excluded from the TSPA-LA on the basis of low consequence. This 
FEP applies to the conditions within the repository. The cladding may have experienced creep 
rupture during its residence in the PWR, but that degradation will be covered in the percentage of 
damaged rods received after use and storage (i.e., degradation of cladding prior to disposal in 
Section 6.2.2). This FEP also addresses cladding containing small perforations that cannot allow 
buildup of internal pressure to cause creep. 
The Spent Fuel Project Office of the NRC issued Interim Staff Guidance  11, Revision 2 
(NRC 2002 [DIRS 164593], p. 2), which set a maximum temperature limit for the cladding 
(400C) to prevent damage from creep or hydride reorientation during dry storage (similar 
temperature histories as repository closure). Since peak cladding temperatures in the repository 
are expected to be less than 300C (Appendix A), damage from creep is expected to be minimal. 
Additional corroborating information is available. A statistical analysis of creep failure was 
performed (CRWMS M&O 2001 [DIRS 151662], Section 6.2), in which a burnup distribution 
(rod average = 44 MWd/kgU, range = 2 to 75 MWd/kgU) was used and a distribution of rod 
properties (including stress) was developed. The Murty creep correlation was selected after 
comparing six different correlations with results from five different experiments. It was then 

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ANL-WIS-MD-000008 REV 02 6-26 October 2004 
modified to better predict irradiated cladding creep data. The fuel rods were exposed to two 
consecutive temperature histories before being placed in the repository. They were exposed to 
24 hours of vacuum drying with a peak temperature of 430C, followed by 20 years dry storage 
with a peak temperature of 350C. 
The sensitivity study considered creep strain for the 1,000 years following repository closure, 
because after that time, the fuel is too cool for creep to occur. Uncertainties in the temperatures 
and strain rate were included. The radial temperature distribution across the waste package was 
also modeled. The waste package surface temperature was shifted upward or downward and the 
fraction of rods that fail from creep was calculated. Failure was predicted when the creep strain 
of a rod exceeds a creep failure criterion. The upper limit was a 1-percent strain failure criterion 
and lower limit was 6-percent strain. The best estimate creep failure criterion was a 
complementary cumulative distribution function (CCDF) based on 52 failure tests. Figure 6-3 
shows the results of this sensitivity study. Rod failures started to occur during repository closure 
at a peak cladding temperature of 400C. Cladding failure reached 1 percent at a peak 
temperature of about 430C (best estimate). In the current repository design, the cladding 
temperatures are below 300C (Appendix A) and rod failures from creep are not expected. 
The Dry Storage Characterization Project (EPRI 2002 [DIRS 161421], p. xii) studied the 
condition of fuel assemblies that were exposed to various thermal transients (peak temperatures 
to 415C) followed by 15 years of dry storage (peak temperature at 342C and slowly 
decreasing). These temperatures are higher than expected at the repository (Appendix A of this 
report). EPRI concluded little or no cladding creep occurred during the thermal benchmark 
testing and dry storage. It is anticipated that the creep would not increase significantly during 
additional storage due to the low temperature after 15 years, the continual decrease in 
temperature from the reduction of decay heat, and the concurrent reduction in pressure and 
stress. 
Over time, helium production from alpha decay could generate pressures that exceed the pressure 
at the peak temperature (see Section 6.2.10). The creep rate has a strong temperature 
dependency (Figures 6.9-1) and the slightly higher stress that occurs at 10,000 years does not 
change the creep results because the temperatures are less than 73C and no observable creep 
occurs at these temperatures and strains. 
Stainless steel cladding is modeled for TSPA-LA as failed, therefore, this FEP does not apply to 
stainless steel cladding. 
In conclusion, creep rupture of cladding is excluded from the TSPA-LA on the basis of low 
consequence. The low repository temperatures are consistent with low to insignificant creep. 
The NRC requirements in 10 CFR 63.114 (e and f) allow the omission because the small amount 
of creep that will occur during the regulatory period will not significantly change the magnitude 
and time of the resulting radiological exposures to the reasonably maximally exposed individual 
(RMEI), or radionuclide releases to the accessible environment. 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-27 October 2004 
Source: CRWMS M&O 2001 [DIRS 151662]. 
Figure 6-3. Creep Failure Fraction as a Function of Peak Cladding Temperature 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.10 Internal Pressurization of Cladding (2.1.02.20.0A) 
FEP Description: 
Increased pressure within the fuel rod due to the production of helium gas could contribute to 
cladding failure. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Internal pressurization of the cladding is excluded from the TSPA-LA on the basis of low 
consequence. Stainless steel cladding is modeled for TSPA-LA as failed; therefore, this FEP 
does not apply to stainless steel cladding. 
Piron and Pelletier (2001 [DIRS 165318], Section 5.3) investigated the pressurization of the fuel 
rods from helium production (alpha decay). They concluded that fuel with a burnup of 
47.5 MWd/kgU would produce 1,171 cm3 (STP) of helium in a rod after 10,000 years. Piron and 
Pelletiers (2001 [DIRS 165318], Section 5.3) values are based on having all of the helium 
released. This provides a total overpressure of 90 bar or 9.10 MPa at 20C in 10,000 years after 
emplacement. This corresponds to 9.3 MPa at 30C. Further discussion on potential pressure 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-28 October 2004 
buildup in the void space between the fuel and cladding is in Section 6.2.23. Another report 
(Rothman 1984 [DIRS 100417], Table 6) calculates a total pressure of 1,530 psi or 10.65 MPa at 
30C for a fuel rod having a burn-up of 36 MWd/kgU after 10,000 years. This latter calculation 
also includes the added pressure from Xe and Kr considered by Prion and Pelletier (2001 [DIRS 
165318]. These results, considering the variability, are within an acceptable tolerance range for 
this evaluation. The minimum yield strength of Zircaloy-4 is about 241 MPa (Webster 1990 
[DIRS 143561], Table 3, p. 665) and, with the internal gas pressure identified above of less than 
11 MPa, there is minimal effect due to the internal pressurization of the cladding. Thus, this FEP 
is excluded based on low consequences. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.11 Stress Corrosion Cracking (SCC) of Cladding (2.1.02.21.0A) 
FEP Description: 
Stress corrosion cracking (SCC) mechanisms can contribute to cladding failure. These 
mechanisms can operate both from the inside out from the action of fission products, or from the 
outside in from the actions of salts or other chemicals within the waste package. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Stress corrosion cracking of cladding is excluded from the TSPA-LA on the basis of low 
consequence. 
SCC is the formation of brittle cracks on a metal surface through the simultaneous action of a 
tensile stress and a corrosive environment. SCC requires a certain chemical environment and 
sufficiently high stresses (stress intensity factors at the crack tip). Chloride-induced SCC could 
occur on the outside of the cladding. Chloride induced stress corrosion cracking requires that the 
passive layer of oxides on the zirconium surface be unstable (Cragnolino et al. 1999 
[DIRS 152354], p. 4-15; Cox 1973 [DIRS 152920], Abstract; Farina et al. 2002 [DIRS 163639], 
p. 5; Yau and Webster 1987 [DIRS 100494], p. 718). These are the same conditions under 
which pitting occurs. As demonstrated in Section 6.2.6, the chemical conditions for pitting and 
SCC do not exist in the waste package. Therefore, chloride-induced SCC is not expected. 
During the times of interest (peak temperature or peak helium pressure) the maximum stress 
intensity factors vary from 0.47 to 2.73 MPa-m0.5. This is less than the threshold stress intensity 
factor for SCC, the value at which crack propagation will start. Table 10b of Clad Degradation 
 Summary and Abstraction (CRWMS M&O 2001 [DIRS 151662]) provides the threshold stress 
intensity factor values for various zirconium alloys in chemical solutions. The reported threshold 
values are greater than or equal to 4 MPa-m0.5. The threshold stress intensity factor (KISCC) of 28 
MPa-m0.5 for moist chlorine is included in this table and pertinent to the case of external cracking 
(Cox 1990 [DIRS 152778], Figure 20, p. 15). This value is for 70C. The threshold stress 

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ANL-WIS-MD-000008 REV 02 6-29 October 2004 
intensity decreases with increasing temperature (Cox 1990 [DIRS 152778], Figure 14). At a 
boiling point of 100C, when water, possibly carrying chlorides, could enter the waste package, 
the threshold stress intensity could be about 24 MPa-m0.5. This value is still well above values 
reported in Table 6.10b of Clad Degradation  Summary and Abstraction (CRWMS M&O 2001 
[DIRS 151662]). Because the estimated stress intensities are less than the threshold values, 
chloride-induced cracking is not expected. 
Stress corrosion cracking on the interior surface of the cladding from iodine is not expected. In 
NRC Issue Resolution Status Report Key Technical Issue: Container Life and Source Term 
(Beckman 2001 [DIRS 156122], p. 103), the NRC concluded: 
The possibility of SCC induced by iodine as discussed in the WF PMR [Process 
Model Report] does not appear so important because it is limited essentially by 
the availability of iodine. The phenomenon as such has been postulated as the 
cause of pellet cladding interaction failure under reactor operating conditions 
following steep power ramps, but it does not seem plausible under repository 
conditions. 
Yau and Webster 1987 [DIRS 100494] identified iodine as a possible agent for SCC of zircaloy. 
However, anticipated iodine concentrations in the repository will be too low for SCC 
(Beckman 2001 [DIRS 156122], p. 103). Failure has not been observed in dry storage tests with 
similar conditions to early repository closure. Stress intensities for a sharp crack, (a limiting 
case, Reed-Hill 1973 [DIRS 121838], p. 800)(up to 2.73 MPa-m0.5) are lower than the range of 
threshold stress intensity factors (4.0 to 15 MPa-m0.5) identified for iodine-induced SCC (Tasooji 
et al. 1984 [DIRS 102093], Figure 12, p. 612). Cragnolino et al. (1999 [DIRS 152354], p. 4-27) 
suggested that lower threshold stress intensity factors could cause failure when considering small 
cracks. The Dry Storage Characterization Project (EPRI 2002 [DIRS 161421]) studied the 
condition of fuel assemblies that were exposed to various thermal transients (peak temperatures 
to 415C) followed by 15 years of dry storage (peak temperature at 342C and slowly 
decreasing). These temperatures are higher than expected at the repository (Appendix A). Rod 
failure during dry storage was not observed, suggesting that iodine induced SCC from small 
cracks did not fail rods. 
Stainless steel cladding is modeled for TSPA-LA as failed, therefore, this FEP does not apply to 
stainless steel cladding. 
Thus, the foregoing discussion has shown that SCC is not likely and hence can be excluded 
based on low consequence. In conclusion, the magnitude and time of the resulting radiological 
exposures to the reasonably maximally exposed individual (RMEI), or radionuclide releases to 
the accessible environment, would not be significantly changed by the omission of this FEP 
(stress corrosion cracking (SCC) of cladding) from the performance assessment (TSPA-LA) 
model. 
TSPA Disposition: 
N/A 

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ANL-WIS-MD-000008 REV 02 6-30 October 2004 
Supporting Reports: 
N/A 
6.2.12 Hydride Cracking of Cladding (2.1.02.22.0A) 
FEP Description: 
Cladding contains hydrogen after reactor operation. The cladding might pick up more hydrogen 
from cladding general corrosion (wet oxidation) after the waste package is breached. The 
hydrogen can exist both as zirconium hydride precipitates and as hydrogen in solid solution with 
zirconium. Hydrides might also form from UO2 oxidation after waste package and cladding 
perforation. In addition, hydrides may dissolve in warmer areas of the cladding and migrate to 
cooler areas. Hydrogen can also move from places of low stress to places of high stress, causing 
hydride reorientation or delayed hydride cracking. The buildup of hydrides can cause existing 
cracks to propagate by delayed hydride cracking or hydride embrittlement. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Hydride cracking and embrittlement of the cladding is excluded from the TSPA-LA on the basis 
of low consequence. 
There are several mechanisms of cladding failure that hydrogen and the formation of hydrides 
can sponsor. Among these are: 
1. Hydride embrittlement from zirconium corrosion of cladding 
2. Hydride embrittlement from galvanic corrosion 
3. Hydride reorientation within cladding 
4. Delayed hydride cracking of cladding 
5. Hydride axial migration of cladding 
6. Hydride embrittlement from fuel reactions. 
As the waste package internals corrode, hydrogen is generated, although little is expected to be 
absorbed directly by the fuel cladding, because H2 molecules do not migrate through the 
high-density ZrO2 fuel cladding layer that is produced during the time the fuel is in the PWR 
(Section 6.2.24). The hydriding of BWR cladding is less likely due to the difference in waste 
chemistry. In addition, available data on zirconium hydriding indicate that corrosion of waste 
package internals will not result in hydriding of fuel cladding, as long as an oxidizing 
environment exists in the waste package (IAEA 1998 [DIRS 150560], p. 92). 
Hydride embrittlement from galvanic corrosion of waste package contacting cladding has been 
excluded based on low consequence. This is justified because the thick, electrically insulating, 
oxide layer prevents direct absorption of hydrogen gas and galvanic coupling to dissimilar 

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ANL-WIS-MD-000008 REV 02 6-31 October 2004 
metals. If the passive film has been mechanically removed, the unprotected cladding oxidizes 
within seconds and forms a passive layer when exposed to water or humid air. 
Delayed hydride cracking will not be a viable mechanism for cladding failure as the stresses 
within the waste package are too low to sponsor crack propagation (Cragnolino et al. 1999 
[DIRS 152354]; Rothman 1984 [DIRS 100417]). Cladding failure by hydride reorientation is 
unlikely, because the maximum temperatures are too low to dissolve much hydrogen and the fuel 
rods have stresses too low for reorientation. Craganolino et al. (1999 [DIRS 152354]) concluded 
that hydride reorientation is likely to occur at stresses less than 100MPa at 290C to 300C. The 
cladding material will maintain sufficient strength, even if hydride reorientation were to occur, 
that failure would not be expected (Puls 1988 [DIRS 102067]). 
Hydrogen axial migration will be limited within the temperature gradients expected (one year 
after closure following 50 years of ventilation) (BSC 2004 [DIRS 166692]): 
Time Centerline Edge Gradient 
45 years 142.8C 46.4C 96.4C 
55 years 103.5C 29.5C 74.0C 
Interpolating gives the approximate temperature gradient across the fuel rod assembly at 51 years 
of around 83C. Failure of the cladding by hydrogen embrittlement is unlikely. Hydrogen 
absorption in the cladding from UO2 fuel corrosion only occurs in fuel with already failed 
cladding. Such reaction, if it should occur, has little consequence. 
Hydrogen embrittlement in metals usually results in a reduced resistance to fracture. In Zircaloy, 
hydrogen embrittlement is caused by precipitation of intergranular zirconium hydride. Because 
Hydride precipitates are quite brittle, a crack can propagate more readily by preferentially 
following the hydrides. Resistance to fracture (fracture toughness KIC) is a measure of resistance 
to crack propagation through the material. Fracture toughness is measured in terms of the critical 
stress intensity factor, the value that will promote the growth of a crack. This factor is 
proportional to the far-field stress times the square root of the crack length. Kreyns et al. (1996 
[DIRS 100462], Figure 5, reproduced here as Figure 6-4) show that for both irradiated and 
unirradiated material, such hydrides decrease the fracture toughness (KIC) from 42 MPam 0.5 to 8 
MPam0.5 as the hydrogen content increases from zero to 4,000 ppm. The stress intensity (KI) 
maxima calculated from the formula from Physical Metallurgy Principles (Reed-Hill 1973 
[DIRS 121838], p. 800) varies from 0.47 to 2.73 MPa m0.5. Thus, failure is not expected, even 
with hydride concentrations as high as 4,000 ppm. The ultimate limit (100 percent hydride and 
no metal) has a fracture toughness of about 1 MPam0.5. The surficial layer of the cladding could 
be fairly brittle (i.e., hydrogen content greater than 800 ppm), but much of the cladding thickness 
will have a reasonable toughness. 
Stainless steel cladding is modeled for TSPA-LA as failed, therefore, this FEP does not apply to 
stainless steel cladding. 
In conclusion, hydride cracking and embrittlement of the cladding is excluded from the TSPALA 
on the basis of low consequence. The NRC requirements in 10 CFR 63.114 (e and f) [DIRS 
156605] allow this omission because the small amount of hydride cracking and embrittlement 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-32 October 2004 
that will occur during the regulatory period will not significantly change the magnitude and time 
of the radiological exposures to the reasonably maximally exposed individual (RMEI), or 
radionuclide releases to the accessible environment. The stresses in the cladding are not 
sufficient to fail the cladding at the repository temperatures and experimental data indicate that 
the in-package environment and cladding stresses are not conducive to hydride cracking and 
embrittlement. 
Source: Kreyns et al. 1996 [DIRS 100462]. 
Figure 6-4. Fracture Toughness vs. Hydrogen Content of Zircaloy-4 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-33 October 2004 
6.2.13 Cladding Unzipping (2.1.02.23.0A) 
FEP Description: 
In either dry or wet oxidizing conditions and with perforated fuel cladding, the UO2 fuel can 
oxidize. The volume increase of the fuel as it oxidizes can create stresses in the cladding that 
may cause gross rupture of the fuel cladding (unzipping). 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
Cladding axial splitting or unzipping is included in the TSPA-LA model by instantly splitting the 
cladding for the full length as soon as the waste package fails and the cladding is perforated. The 
cladding degradation abstraction models all failed rods (after initial perforation) as bare fuel 
pellet fragments for the full length of the fuel rod. Failed fuel rod unzipping (cladding axially 
splits down its length) is caused by the volume increase of corrosion products, Section 6.2.19, 
(fuel or cladding). It is based on experimental observations at Argonne National Laboratory, 
where 2 of 2 rods unzipped in less than 2 years. Unzipping leaves fuel pellets exposed to the 
waste package internal environment. Details of the scientific analysis that describes the 
disposition in greater detail is presented in Clad Degradation  Summary and Abstraction for LA 
(BSC 2004 [DIRS 170023]). 
Unzipping by dry oxidation (oxidation of UO2 to U3O8) of the fuel requires low humidity and 
high temperature conditions. Dry oxidation is expected to occur in the repository if the waste 
package fails at closure and the fuel is exposed to the temperatures presented in Appendix A of 
this report. If it should occur, it also would cause rapid unzipping and is well modeled with the 
instant unzipping model used in TSPA-LA. Unzipping by wet oxidation was not observed in the 
ANL tests because unzipping due to the fuel-side cladding corrosion occurred first. Wet 
oxidation is still fast and can be modeled by the instant unzipping model where necessary. 
Supporting Reports: 
BSC 2004 [DIRS 170023] 
6.2.14 Mechanical Impact on Cladding (2.1.02.24.0A) 
FEP Description: 
Mechanical failure of cladding may result from external stresses, such as rockfall or impact from 
waste package internals. Seismic-induced impacts are addressed in a separate FEP. 
Screening Decision: 
Excluded  Low Probability 
Screening Argument: 
Mechanical failure of the cladding from external stresses originating outside the waste package 
has been excluded from the TSPA-LA on the basis of low probability. According to FEPs 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-34 October 2004 
Screening of Processes and Issues in Drip Shield and Waste Package Degradation (BSC 2004 
[DIRS 169997], Section 6.2.23), a nominal rockfall could damage the drip shield but will not 
deflect the drip shield sufficiently to contact the waste package. Waste package damage from 
rockfall was screened out based on low probability of occurrence. Therefore, during the 
10,000-year regulatory period, no damage to the CSNF is expected, as long as the waste package 
remains intact. Thus, the waste package prevents rock fall from contacting and degrading the 
waste form. Hence, damage to the CSNF (waste form and cladding) located inside the waste 
package would not be expected. The CSNF (waste form and cladding) is located inside the 
waste package and no damage is expected. 
Mechanical failure of the cladding from external stresses originating inside the waste package, 
such as from the degradation of basket components and other waste package internals, has also 
been excluded from the TSPA-LA on the basis of low probability of occurrence. Volume 
increase from the waste package corrosion products is addressed in EBS Radionuclide Transport 
Abstraction (BSC 2004 [DIRS 169868]). Initially the waste package void fraction is 64% (4.7 
m3 free volume) but after the waste package internals corrode, the void fraction is 26% (1.9 m3 
free volume). Because of the high void fraction, even after corrosion, the waste package 
corrosion products do not mechanically load and fail the cladding and there is sufficient free 
volume for any helium that is released from the fuel (FEP 2.1.12.02.0A). 
Another approach is to consider dimensional changes in the fuel channel that could be produced 
from hematite (Fe2O3) formation. The inside width of fuel channel is 22.64 cm and the steel 
component (fuel basket tube) is 0.5 cm thick (BSC 2004 [DIRS 169868]). The iron (density = 
7.85 g/cm3) would change to hematite (density = 5.24 g/cm3 (BSC 2004 [DIRS 169868], 
Table 11). Considering the corrosion products can only form in the fuel channel, the inside 
dimension would decrease by 0.5 cm [(2  0.5  7.85/5.24)  2  0.5 = 0.5 cm]. The standard 
Westinghouse 17  17 assembly is 21.4 cm wide (DOE 1996 [DIRS 100320], p. 123). Although 
the free clearance would decrease from 1.2 cm to 0.7 cm after the formation of the hematite, fuel 
failure is not expected. In a fuel assembly, the rods do not contact each other and there is coolant 
spacing between the rods. Since no fuel failure is expected from the iron corrosion, this FEP is 
excluded in the TSPA-LA. 
Thus, mechanical failure of the cladding from external stresses has been excluded on the basis of 
low probability. 
Effects of longer term loading due to volume increase from waste form and cladding corrosion 
products (i.e., internal stresses) are addressed in FEP 2.1.09.03.0A, Volume Increase of 
Corrosion Products Impacts Cladding. Seismic-induced rockfall, drift degradation, and ground 
motion are also not treated within this FEP. A full discussion of seismic effects is contained in 
FEPs 1.2.03.02.0A, Seismic Ground Motion Damages EBS Components; 1.2.03.02.0B, Seismic 
Induced Rockfall Damages EBS Components; and 1.2.03.02.0C, Seismic Induced Drift Collapse 
Damages EBS Components. The effect of seismic events on fuel and cladding is addressed in 
Seismic Consequence Analysis (BSC 2004 [DIRS 167780]). 
TSPA Disposition: 
N/A 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-35 October 2004 
Supporting Reports: 
N/A 
6.2.15 Naval SNF Cladding (2.1.02.25.0B) 
FEP Description: 
DSNF to be disposed of in Yucca Mountain has a variety of fuel types that may not be similar to 
the CSNF to be disposed. Some of the fuel types may have initial cladding-degradation 
characteristics that are different from those for the CSNF. Therefore, the effectiveness of DSNF 
cladding as a barrier to radionuclide mobilization might be different from CSNF. This FEP 
addresses Naval SNF cladding only. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
Naval Nuclear Propulsion Program cladding and SNF performance is discussed in Naval Nuclear 
Propulsion Program Technical Support Document for the License Application, which is a 
classified document. Based upon the results from a sensitivity study (BSC 2001 [DIRS 152059]), 
waste packages containing naval fuel are conservatively modeled as containing CSNF. There are 
300 waste packages containing naval SNF compared to the commercial SNF inventory of 
approximately 7,500 waste packages. A comparison with an equivalent amount of Zircaloy-clad 
CSNF indicates that the total dose from the TSPA simulation, using the commercial-fuel 
equivalent, is more than four orders of magnitude higher than the total dose from the source-term 
simulation for naval SNF (BSC 2001 [DIRS 152059], p. 36 and Figure 6.1-2). Therefore, it is 
conservative to model naval SNF as CSNF. Details of the scientific analysis that describes the 
disposition in greater detail are presented in Clad Degradation  Summary and Abstraction for 
LA (BSC 2004 [DIRS 170023], Section 6). 
Supporting Reports: 
BSC 2004 [DIRS 170023] 
6.2.16 Diffusion-Controlled Cavity Growth in Cladding (2.1.02.26.0A) 
FEP Description: 
Diffusion-controlled cavity growth is a possible creep rupture mechanism that could occur under 
the temperature and pressure conditions that prevail during dry storage of spent fuel. It might 
also occur during disposal. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Diffusion-controlled cavity growth (DCCG) in cladding is excluded from the TSPA-LA on the 
basis of low consequence. 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-36 October 2004 
Applicants for dry storage licenses for CSNF were once required by the U.S. Nuclear Regulatory 
Commission (NRC) to model diffusion-controlled cavity growth to evaluate dry storage designs. 
That is, NUREG-1536 (NRC 1997 [DIRS 101903], Section 4.V.1) required the use of the DCCG 
method to calculate a maximum cladding temperature limit for a dry storage design. However, 
this design limit is overly restrictive and relatively inflexible. The literature does not support the 
use of this model for zirconium-based materials (Pescatore and Cowgill 1994 [DIRS 102066], 
pp. 83 to 85) since it has not been validated, and voids and cavities are rarely seen in irradiated 
Zircaloy. Pescatore and Cowgill (1994 [DIRS 102066], p. 85) recommend a methodology of 
calculating the amount of creep and comparing it to a creep failure criterion. The earlier NRC 
Interim Staff Guidance (ISG) Number 11 (NRC 2000 [DIRS 147797]) for transportation and 
storage recognizes the controversy with the DCCG conceptual model and permits license 
applicants to use other creep models in their license application. Cladding Considerations for 
the Transportation and Storage of Spent Fuel (NRC 2002 [DIRS 164593]) does not address 
specific creep models but concludes that creep failures are unlikely if peak temperatures are 
below 400C. The temperature profiles for dry storage (time at temperature; CRWMS M&O 
2001 [DIRS 151662], Figure 6) are similar to those in the early periods of repository closure 
when DCCG might occur. With peak cladding temperatures within the repository anticipated to 
be below 300C (Appendix A), creep failure (including DCCG failure) is not expected. 
Hayes et al. (1999 [DIRS 164598], Figures 2, 5, 6, 8, and 11) concluded that failure from DCCG 
is unlikely if the peak temperature is less than 330C to 400C depending on the specific DCCG 
model (LLNL versus PNNL), initial stress, and heat decay curve. None of their work suggests 
failure by DCCG at temperatures as low as 300C. In summary, omission of this FEP is justified 
on the basis that DCCG has a low consequence of occurrence because DCCG, as a mechanism to 
fail Zircaloy cladding, is not likely to significantly degrade the cladding at the temperatures 
(peak temperatures below 300C) predicted at the repository. 
The creep analysis described in FEP 2.1.02.19.0A is an alternative conceptual model for DCCG 
model. The creep analysis was selected because of the stronger experimental basis for this type 
of strain. 
Stainless steel cladding is modeled for TSPA-LA as failed; therefore, this FEP does not apply to 
stainless steel cladding. 
In conclusion, DCCG in cladding is excluded from the TSPA-LA on the basis of low 
consequence. The NRC requirements in 10 CFR 63.114 (e and f) [DIRS 156605] allow this 
omission because the impact of DCCG on cladding is likely to be insignificant and therefore will 
not significantly change the magnitude and time of the radiological exposures to the reasonably 
maximally exposed individual (RMEI), or radionuclide releases to the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-37 October 2004 
6.2.17 Localized (Fluoride Enhanced) Corrosion of Cladding (2.1.02.27.0A) 
FEP Description: 
Fluoride is present in Yucca Mountain groundwater, and zirconium has been observed to corrode 
in environments containing fluoride. Therefore, fluoride corrosion of cladding may occur in 
waste packages. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Fluoride enhanced corrosion of cladding is excluded from the TSPA-LA on the basis of low 
consequence. 
Hydrofluoric acid can contribute to an accelerated general corrosion rate with fluoride 
concentrations greater than 5 ppm and pH less than 3.18 (Pourbaix 1974 [DIRS 100817], p. 583). 
The in-package chemistry model (BSC 2004 [DIRS 167621], Attachment III) predicts Ph values 
greater than 4.5 and J13 well water containing only 2.2 ppm of fluoride. Even if pH dropped 
below 3, fluoride enhanced localized corrosion of the cladding would not be significant because 
the fluoride concentration would still be less than 5 ppm. 
As corroborating evidence, corrosion of zirconium by fluorides is addressed in Clad 
DegradationLocal Corrosion of Zirconium and Its Alloys Under Repository Conditions 
(CWRMS M&O 2000 [DIRS 136058], Sections 4.1, 6.1.5, 6.2.2.3, and III.4). Zirconium resists 
attack by most halides, including halogen acids. The major exceptions are hydrofluoric acid and 
ferric chloride (localized (pitting) corrosion of cladding, FEP 2.1.02.16.0A). As shown in 
Section 6.1.3 of Clad DegradationLocal Corrosion of Zirconium and Its Alloys Under 
Repository Conditions (CWRMS M&O 2000 [DIRS 136058]), zirconium is corrosion resistant 
to certain fluorides when the pH is sufficiently high. Low fluoride ion concentrations (F- ions), 
such as a few ppm, in city or ground water have little effect on zirconium's excellent corrosion 
resistance. However, a few ppm of hydrofluoric acid (HF molecule in solution) will noticeably 
increase zirconium's corrosion rate. Hydrofluoric acid only exists in solution at pH values 
below 3.18 (Pourbaix 1974 [DIRS 100817], p. 583). 
For accelerated corrosion to occur, the fluoride must be present as free ions (i.e., not complexed 
as compounds), and the pH must be low. A high insoluble fluoride concentration (in essence a 
low fluoride ion concentration) would not be expected to have much impact on the standard 
zirconium corrosion rate. Section 4.1.1 of Clad DegradationLocal Corrosion of Zirconium and 
Its Alloys Under Repository Conditions (CWRMS M&O 2000 [DIRS 136058], Test 12) shows 
that fluoride ion concentrations of less than 5 ppm, even at pH values as low as 1, produce 
similar corrosion rates to those with zero fluoride ion concentration. Thus, it is reasonable to 
conclude that low fluoride ion concentrations, as distinct from total fluoride content, will have 
limited impact on the uniform Zircaloy corrosion rate. 
Zirconium and its alloys generally exhibit low corrosion rates in fluoride solutions, including 
relatively high fluoride ion content solutions, if the temperature is sufficiently low and the pH is 
sufficiently high. This is illustrated with the results in Section 4.1.1 of Clad DegradationLocal 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-38 October 2004 
Corrosion of Zirconium and Its Alloys Under Repository Conditions (CWRMS M&O 2000 
[DIRS 136058], Test 13). However, if the metal is in contact with solutions containing HF, the 
corrosion rate can increase rapidly. From the Pourbaix (1974 [DIRS 100817], p. 583) diagram, 
HF can only exist when the pH is less than 3.18, although this does not necessarily mean that all 
fluoride ions are immediately converted to HF below this value. The data in Section 4.1.1 of 
Clad DegradationLocal Corrosion of Zirconium and Its Alloys Under Repository Conditions 
(CWRMS M&O 2000 [DIRS 136058], Tests 10 through 13) can be divided into fluoride ioncontaining 
solutions and HF-containing solutions using a pH of 3.18 as the demarcation point. 
The calcium fluoride, due to its low solubility, does not contribute to the fluoride ion 
concentration (about 2 ppm at 25C and less than 3 ppm at 90C). 
If the fluoride were to react with the cladding, the amount of corrosion would be limited by the 
amount of fluoride entering the waste package because the fluoride is consumed in the reaction 
Zr + 4F . Zr F4. For example, if 1 liter of J-13 well water enters a waste package each year, the 
total contained fluorine content is 2.2 mg (2.2 wt. ppm). If all the available fluorine reacts with 
zirconium to produce ZrF4, the maximum quantity of zirconium that could be corroded away is 
2.6 mg per year (4.0  10-4 cc/year). This volume represents a general corrosion mechanism and 
would be distributed over the entire wetted surface. If it were concentrated at one spot of 
cladding surface area, 0.008 square centimeters of fuel would be exposed (using nominal wall 
thickness of 500 m). This quantity of fuel exposure is not significant and, therefore, fluoride 
corrosion would be a low consequence degradation mechanism. Much larger quantities of water 
containing fluorides than were used in the example must enter the waste package to fail 
significant quantities of fuel. 
Repository conditions (as represented by J-13 well water) would not be expected to produce any 
significantly different corrosion rates in zirconium and its alloys than in general corrosion 
(FEP 2.1.02.13.0A). The well water analysis (Table 6-5) shows a neutral solution with impurity 
concentrations too low to be corrosive to zirconium and its alloys 
Table 6-5. Compositions of J-13 Well Water 
Concentration (mg/L) 
Ion J-13 
SO4 18.4 
Cl 7.14 
NO3 8.78 
F 2.18 
HCO3 128.9 
Na 45.8 
K 5.04 
Ca 13.0 
Mg 2.01 
SiO2 (aq) 61.0 
pH 7.4 
Source: BSC 2001 [DIRS 155640] Table 18. 
Stainless steel cladding is modeled for TSPA-LA as failed; therefore, this FEP does not apply to 
stainless steel cladding. 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-39 October 2004 
In conclusion, the magnitude and time of the resulting radiological exposures to the reasonably 
maximally exposed individual (RMEI), or radionuclide releases to the accessible environment, 
would not be significantly changed by the omission of this FEP (fluoride enhanced corrosion) 
from the performance assessment (TSPA-LA) model. The impact on cladding is not expected to 
be significant because the pH is too high (> 3.18) for the formation of hydrofluoric acid and the 
concentration of fluorine is too low (<5 ppm). Because conditions relevant to localized 
corrosion are not present, localized (fluoride-enhanced) corrosion of cladding has been excluded 
on the basis of low consequence. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.18 Rockfall (2.1.07.01.0A) 
FEP Description: 
Rockfalls may occur with blocks that are large enough to mechanically tear or rupture drip 
shields and/or waste packages. Seismic-induced rockfall is addressed in a separate FEP. 
Screening Decision: 
Excluded  Low Probability 
Screening Argument: 
As stated in FEPs Screening of Processes and Issues in Drip Shield and Waste Package 
Degradation (BSC 2004 [DIRS 169997], Section 6.2.23), nominal rockfall could damage the 
drip shield but will not deflect the drip shield sufficiently to contact the waste package. 
Therefore, during the 10,000-year regulatory period, no damage to the waste package is 
expected. The CSNF (waste form and cladding) is located inside the waste package and no 
damage is expected. While the fuel and cladding might experience some minor vibration from 
rockfall, no damage is expected. Fuel and cladding are exposed to vibration during 
transportation and damage is not observed (Debes 1999 [DIRS 161193]). 
Seismic induced rockfall and drift degradation are not considered within this FEP. A full 
discussion of seismic effects is contained in Engineered Barrier System Features, Events, and 
Processes [DIRS 169898], Sections 6.2.8 through 6.2.10) The effect of seismic events on fuel 
and cladding is addressed in Seismic Consequence Analysis (BSC 2004 [DIRS 167780]). 
Because the drip shield protects the waste package against rock fall, rock fall damage to the 
cladding has been excluded based on low probability. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-40 October 2004 
6.2.19 Volume Increase of Corrosion Products Impacts Cladding (2.1.09.03.0A) 
FEP Description: 
Corrosion products have a higher molar volume than the intact, uncorroded material. Increases in 
volume during waste form and cladding corrosion could change the stress state in the material 
being corroded and lead to cladding unzipping. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
The volume increase of corrosion products causes cladding axial splitting, or unzipping, and is 
included in the TSPA cladding degradation abstraction (BSC 2004 [DIRS 170023], 
Section 6.2.4). This FEP applies to failed cladding where water or moist air can interact with the 
fuel or cladding interior. The volume increase of corrosion products inside the cladding causes 
stress on the cladding and the cladding to tear open. This tearing is modeled to be instantaneous. 
All failed rods contain fuel pellet fragments for the full length of the fuel rod that are available 
for dissolution. Failed fuel rod unzipping (cladding axially splits down its length) is caused by 
the volume increase of corrosion products (fuel or cladding). It is based on experimental 
observations of two rods at ANL where both rods unzipped in less than two years. Unzipping 
leaves the fuel pellets exposed to the waste package internal environment. The scientific 
analysis that describes the disposition in greater detail is presented in Clad Degradation  
Summary and Abstraction for LA (BSC 2004 [DIRS 170023], Section 6.2.4). 
Unzipping by dry oxidation (oxidation of UO2 to U3O8) of the fuel requires low humidity and 
high temperature conditions. It is expected to occur in the repository if the waste package fails at 
closure and the fuel is exposed to the temperatures given in Appendix A. If dry oxidation should 
occur, it also would cause rapid unzipping and is well modeled with the instant unzipping model 
used in TSPA-LA. 
The effects of basket component degradation on external cladding integrity have been evaluated 
in FEP 2.1.02.24.0A, Mechanical Impact on the Cladding. 
Supporting Reports: 
BSC 2004 [DIRS 170023] 
6.2.20 Electrochemical Effects in EBS (2.1.09.09.0A) 
FEP Description: 
Electrochemical effects may establish an electric potential within the drift or between materials 
in the drift and more distant metallic materials. Migration of ions within such an electric field 
could affect corrosion of metals in the EBS and waste, and could also have a direct effect on the 
transport of radionuclides as charged ions. 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-41 October 2004 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Electrochemical effects (electrophoresis and galvanic coupling) are excluded from the TSPA-LA 
on the basis of low consequence. 
Cladding, with its thick oxide layer produced in reactor operation, is kinetically noble (Yau and 
Webster 1987 [DIRS 100494], Table 15, p. 718). Any unprotected cladding (due to mechanical 
removal of the protective oxide layer) oxidizes and forms a passive layer within seconds if 
exposed to water or humid air (Hansson 1984 [DIRS 101676], p. 6). Inside the waste package, 
the cladding would behave as a cathode and the carbon steel fuel box would be an anode if they 
were electrically coupled. The cladding cannot undergo hydrogen charging because the oxide 
layer prevents hydrogen absorption in the metal (FEP 2.1.12.03.0A). 
A report by the Center for Nuclear Waste Regulatory Analysis (Cragnolino 1999 
[DIRS 152354], p. 4-13) surveyed various corrosion mechanisms for zirconium cladding under 
repository conditions. They concluded: 
Zr is not susceptible to galvanic corrosion because the protective ZrO2 passive 
film leads to Ecorr values in the galvanic series in flowing seawater close to those 
of noble metals and graphite but slightly lower than that of Ag (Yau and Webster, 
1987). However, local corrosion promoted by galvanic coupling to a more noble 
metal may occur if the film is mechanically disrupted. Nevertheless, the 
repassivation rate of Zr and its alloys is sufficiently fast in many aqueous 
solutions that unless fretting is continuously occurring no substantial corrosion 
can be expected. 
Possible effects of galvanic coupling between cladding and the carbon steel fuel boxes have been 
considered. After reactor operation, fuel assemblies are coated with a thick oxide, which has a 
high electrical resistance and would minimize the galvanic effect. After the waste package is 
breached, the carbon steel fuel box would behave as an anode if coupled with the cladding and 
could corrode more quickly. An increase in corrosion of the carbon steel fuel box at assembly 
contact points would have a low consequence. Stainless steel and aluminum plates would 
maintain the geometry for some time. Galvanic corrosion of cladding or carbon steel fuel boxes 
is excluded from the TSPA because of low consequence. 
Galvanic coupling could theoretically affect the rate of redox reactions involved in waste form 
matrix degradation when the waste form matrix is an electronic conductor. For example, 
galvanic coupling between the CSNF matrix (a semiconductor) and cladding could influence the 
corrosion potential, and hence, the rate of oxidative dissolution of the CSNF matrix. However, 
data from tests performed on CSNF rod segments (which include the cladding) and on CSNF 
fragments in Zircaloy holders indicate that such hypothetical galvanic coupling has a negligible 
effect on the rate of corrosion of the fuel matrix and the associated radionuclide release rate. The 
effects of galvanic coupling on the corrosion and associated radionuclide release rate from DSNF 
are also negligible because the base case model for DSNF corrosion causes it to corrode 
completely in one TSPA-LA time step. Electrochemical effects on DHLW degradation are 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-42 October 2004 
negligible because the principal reactions involved are glass network hydrolysis reactions that 
are not influenced by electrochemical effects. 
Electrochemical effect on the waste form will also be minimal. The CSNF waste form is 
surrounded by the split zirconium cladding and would not be in direct contact with the waste 
package internals. If the UO2 did contact the steel fuel boxes and galvanic coupling did occur, 
the steel boxes would see the accelerated reaction and not the fuel pellets. As the UO2 corrodes, 
it coats itself with reaction products, which will minimize galvanic effects. Omission of 
electrochemical effects (electrophoresis and galvanic coupling) in the waste would not have a 
significant effect on the resulting radionuclide exposures to the RMEI because those effects are 
much smaller than the effects of modeling with a minimum flow rate through a failed container 
of 15 l/yr used in performing the equilibrium-model calculations (CRWMS M&O 2000 [DIRS 
136058]). This flow rate has a much greater effect than can be created by electrophoresis or 
electro-osmosis (Soderman and Jonsson 1996 [DIRS 149441]). 
Electrochemical effects of various solutions that might cause pitting of the cladding are 
addressed in Section 6.2.6. The pitting model (BSC 2004 [DIRS 170043]) is an electrochemical 
model and the in-package chemistry model (BSC 2004 [DIRS 167621]) addresses the effect of 
EBS and waste package corrosion on chemistry. Figure 6-2 shows pH variation with time, 
including the effects of radiolysis. The depression of pH in this figure is caused by waste 
package steel corrosion. The result from Sections 6.2.5 and 6.2.6 is that no failure is expected. 
Telluric currents can, under some conditions, affect corrosion rates in carbon steels, copper 
alloys and other less corrosion resistant alloy systems. This is especially true for cathodically 
protected buried carbon steel pipeline systems. However, in the case of the highly passive 
Alloy 22 waste package outer barrier material and the Titanium Grade 7 drip shield material, 
telluric current effects would not be expected to affect corrosion response since these materials 
remain highly passive over a very broad range of applied potentials. Thus small, varying 
changes in local surface potentials that could potentially result from telluric currents or other 
stray current sources would not measurably affect the very low corrosion rates of these materials. 
In conclusion, electrochemical effects (electrophoresis and galvanic coupling) are excluded from 
the TSPA-LA on the basis of low consequence. The NRC requirements in 10 CFR 63.114 (e 
and f) [DIRS 156605] allow this omission because electrochemical effects are expected to be 
small and will not significantly change the magnitude and time of the resulting radiological 
exposures to the reasonably maximally exposed individual (RMEI), or radionuclide releases to 
the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 6-43 October 2004 
6.2.21 Chemical Effects of Waste-Rock Contact (2.1.09.11.0A) 
FEP Description: 
Waste (CSNF, DSNF, and HLW) and rock may be placed in direct contact by mechanical failure 
of the drip shields and/or waste packages. Chemical effects on the waste (e.g. dissolution) may 
be enhanced or altered in a system where waste, rock minerals, and water are all in physical 
contact with one another, relative to a system where only waste and water are in physical contact. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
The chemical effects of wasterock contact are excluded from the TSPA-LA on the basis of low 
consequence. 
Wasterock contact will have no effect on CSNF, DSNF, and HLW dissolution. Water that 
contacts SNF will have previously been in contact (i.e., equilibrium with the host rock) and 
therefore, placement of the rock in physical contact with the SNF will not affect the 
wasterockwater system chemistry. Furthermore, In-Package Chemistry Abstraction (BSC 
2004 [DIRS 167621]) demonstrated that variations in the chemistry of the water contacting the 
waste package internal components, including the SNF, had an insignificant effect on the pH and 
total carbonate, the two key chemical parameters controlling the solubility of CSNF, DSNF, and 
HLW, of the in-package fluids. Waste form and cladding interaction with igneous intrusions are 
addressed in FEP 1.2.04.04.0A, Igneous intrusion interactions with EBS components. This FEP 
is addressed in Igneous Intrusion Impacts on Waste Package and Waste Forms (BSC 2004 
[DIRS 168960]). 
The chemical effect of wasterock contact has been excluded from TSPA-LA because of low 
consequence. Even if near-field chemistry contacts the cladding or waste form, accelerated 
corrosion is not expected. Tests involving the contact of zirconium with soils have not shown 
corrosion. Adler Flitton et al. (2002 [DIRS 161991], p. 4) buried various metal samples 
including zirconium in an arid vadose zone environment for three years. These samples were in 
intimate contact with the vadose (unsaturated) zone environment. They reported indications of 
pitting on some of the other metals, but observed no pitting on the zirconium samples when in 
contact with rock. Other corroborating information is also available. The waste form is 
protected from the rocks by both the drip shield and waste package during the 10,000-year 
regulatory period. Zirconium is impervious to most chemical solutions (CRWMS M&O 2000 
[DIRS 136058]) and only severe chemicals will pit cladding (BSC 2004 [DIRS 170043]). Yau 
(1984 [DIRS 102050]) performed corrosion tests with zirconium in geothermal fluids and reports 
no corrosion. Yau and Webster (1987 [DIRS 100494]) review the limited conditions where 
corrosion of zirconium is observed. This review suggests that contact with groundwater from 
YMP tuff would not corrode the cladding. 
The CSNF and DSNF waste forms will have little contact with the rock because the stainless 
steel inner barrier and Alloy 22 outer barrier waste package will prevent contact. For the CSNF 
and some forms of DSNF, the zirconium alloy cladding surrounding the UO2 pellets will prevent 
contact. Contact of the UO2 pellets with the rock would have little effect. The uranium in the 

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UO2 first oxidizes (U4+ to U6+), dissolves, and then precipitates immediately on the pellet surface 
as a U6+ mineral. The rock would have little contact with the UO2 pellets and little effect on the 
CSNF corrosion. DSNF and DHLW are also unlikely to contact the rock and direct effects of 
such contact, were it to occur, are also negligible. Indirect effects that could occur through the 
water chemistry (e.g., effects on dissolved silicon concentration) which would feed back into the 
rate of glass dissolution are small compared to the effects of the glass dissolution, as indicated by 
the results of tests conducted in J-13 water (i.e., J-13 well water preconditioned by reaction with 
tuff). 
In conclusion, the chemical effects of wasterock contact are excluded from the TSPA-LA on 
the basis of low consequence. NRC requirements in 10 CFR 63.114 (e and f) [DIRS 156605] 
allow this omission because contact is not expected. Even if it did occur, the resulting chemical 
effects would not significantly change the magnitude and time of the resulting radiological 
exposures to the reasonably maximally exposed individual (RMEI), or radionuclide releases to 
the accessible environment. If contact were to occur in the first 10,000 years following 
permanent closure, no chemical effect is expected. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.22 Thermal Expansion and Stress of In-Package EBS Components (2.1.11.05.0A) 
FEP Description: 
Thermally induced stresses could alter the performance of the waste or EBS. For example, 
thermal stresses could cause the waste form to develop cracks and create pathways for 
preferential fluid flow and, thereby, accelerate degradation of the waste. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Thermal expansion and stresses of in-package EBS components, including the waste form, are 
excluded from the TSPA-LA on the basis of low consequence. 
The waste package and its internals are designed for the thermal expansion from the 
temperatures discussed in Appendix A. The CSNF and DSNF are designed for the thermal 
cycles expected in reactors, which are more severe than the conditions anticipated to occur in the 
repository. As discussed in Section 6.2.1 of CSNF Waste Form Degradation: Summary 
Abstraction (BSC 2004 [DIRS 169987]), the in-reactor thermal cycles (principally that 
associated with the initial power escalation) result in extensive cracking of the fuel matrix. The 
effects of this cracking are included in the specific surface area parameter. Glass logs crack 
because of the cooldown during manufacturing. The cracking that results from this cooldown is 
included in the DHLW model surface area parameter. This cooldown (from molten glass, about 
950C, BSC 2004 [DIRS 169988], Section 7.5.3) is more severe than repository conditions. 

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Commercial nuclear fuel operates at higher temperatures than expected during the post-closure 
period at the repository. Under normal conditions, typical cladding operates at about 320C 
(Garde et al. 1991 [DIRS 101652], p. 582) with fuel centerline temperatures reaching 1,800C 
(Lanning et al. 1997 [DIRS 101704], V3, p. 3.2, Figure 3.1). Fuel is also designed to undergo 
anticipated operating occurrences (off normal transients that occur during the design life) without 
damage. These are more severe thermal cycles than normal reactor operation or repository 
closure. Every time a reactor shuts down and goes to cold shutdown, the fuel is cooled to below 
100C (coolant is less than boiling). These temperature transients are more severe than 
repository closure. DSNF is also exposed to reactor transients more severe than the post-closure 
cooldown. Because the temperature transients for spent fuel from normal in-reactor operations 
and for DHLW from normal manufacturing cooldown are more severe than the transient 
associated with repository closure, no further degradation (cracking) is expected from thermal 
expansion or secondary thermal stresses of in-package EBS components. 
In conclusion, thermal expansion and stress of in-package EBS components, including the waste 
form is excluded from the TSPA-LA on the basis of low consequence. The NRC requirements 
in 10 CFR 63.114 (e and f) [DIRS 156605] allow this omission because thermal expansion will 
not significantly change the magnitude and time of the resulting radiological exposures to the 
reasonably maximally exposed individual (RMEI), or radionuclide releases to the accessible 
environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.23 Gas Generation (He) From Waste Form Decay (2.1.12.02.0A) 
FEP Description: 
Helium (He) gas production may occur by alpha decay in the waste. Helium production might 
cause local pressure buildup in cracks in the fuel and in the void between fuel and cladding, 
leading to cladding and waste package failure. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Effects of helium gas generation from alpha decay on the cladding in spent fuel are excluded 
from the TSPA-LA on the basis of low consequence. 
Piron and Pelletier (2001 [DIRS 165318], Section 5.3) investigated pressurization of the fuel 
rods from alpha decay helium production and concluded that fuel with a burn-up 
of 47.5 MWd/kgU would produce 1,171 cm3 (STP) of helium in a rod after 10,000 years, too 
low a quantity to damage the fuel. The internal pressurization of less than 11 MPa (Section 
6.2.10) due to gas build up would have to be significantly higher for the cladding to fail from 
delayed hydride cracking. 

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For cladding, delayed hydride cracking, stress corrosion cracking, and strain failures are driven 
by the cladding stress, which may be caused by the internal gas (including initial fill gas, fission 
product gases, and helium gas from alpha decay) pressure buildup. The gas pressure will slowly 
increase over time by the production of helium as a result of alpha decay. For failed rods, the 
helium would be released into the waste package, possibly increasing the pressure there. Using 
complete helium release from the fuel is conservative at these temperatures. Helium is an inert 
gas and will not chemically react with in-package components (internals, cladding or UO2 
pellets). After waste package failure, it will also tend to diffuse out. The helium would have 
little effect on the corrosion of the UO2 pellets. UO2 corrosion is an oxidationdissolution 
phenomenon and the presence of traces of an inert gas will not affect this. The effects of 
microcracking due to pressure buildup in gas bubbles within the fuel matrix on the rate of matrix 
corrosion are expected to be negligible. As described in Sections 6.2.1 and 6.2.2 of CSNF Waste 
Form Degradation: Summary Abstraction (BSC 2004 [DIRS 169987]), evidence from CSNF 
testing indicates that the corrosion process is a general corrosion process occurring 
predominantly at the periphery of the fuel fragments. The effective specific surface area of the 
corroding fuel is comparable to the geometric surface area and is not sensitive to the internal 
grain boundary decohesion or microcracking that may result from helium buildup in the matrix. 
Likewise, because DHLW dissolution occurs at the periphery of glass fragments, the internal 
microstructural features (including helium gas bubbles) do not affect the dissolution rate. 
For an intact waste package, the free volume is 4.7 m3 (Section 6.2.14). If the waste package has 
failed, any helium that is released from the failed fuel can escape the waste package. It is 
expected that less than 1 percent of the fuel is failed (Section 6.2.2). If all of the rods were 
considered failed and all of the helium was released from the fuel matrix, then approximately 
(rounding up) 1,500 cm3 (STP) of gas (Section 6.2.10) would be released per rod. 
With 5,544 rods per waste package (BSC 2004 [DIRS 170023], Table 8-1), there would 
be 8.3  106 cm3 (STP) of helium available. When distributed into the free volume, 4.7 m3 
(Section 6.2.14), of the waste package at 73C (nominal peak temperature at 10,000 years; 
BSC 2004 [DIRS 169565] Figure 6.3-67), the pressure would increase by 0.22 MPa (2.2 atm). 
This pressure rise is negligible and, thus, would have no affect on the waste package. With most 
of the fuel rods intact, and most of the helium tied up in the fuel matrix, the actual pressure rise 
would be even smaller. 
In conclusion, effects of helium gas generation from alpha decay on the cladding are excluded 
from the TSPA-LA on the basis of low consequence. The NRC requirements in 10 CFR 63.114 
(e and f) [DIRS 156605] allow this omission because the pressures generated by helium 
generation are insufficient to damage the cladding and, therefore, will not significantly change 
the magnitude and time of the resulting radiological exposures to the reasonably maximally 
exposed individual (RMEI), or radionuclide releases to the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 

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6.2.24 Gas Generation (H2) From Waste Package Corrosion (2.1.12.03.0A) 
FEP Description: 
Gas generation can affect the mechanical behavior of the host rock and engineered barriers, 
chemical conditions, and fluid flow, and, as a result, the transport of radionuclides. Gas 
generation due to oxic corrosion of waste packages, cladding, and/or structural materials will 
occur at early times following closure of the repository. Anoxic corrosion may follow the oxic 
phase if all oxygen is depleted. The formation of a gas phase around the waste package may 
exclude oxygen from the iron, thus inhibiting further corrosion. 
Screening Decision: 
Excluded (low consequence) 
Screening Argument: 
Hydrogen gas (H2) generation from waste package corrosion, as a degradation mechanism for 
cladding, is excluded from the TSPA-LA on the basis of low consequence. The cladding absorbs 
little or no hydrogen from the outside environment, so waste package corrosion and hydride 
embrittlement will not occur. 
Many investigators have considered the hydriding of zirconium alloys (for example, the papers 
cited by Clayton (1989 [DIRS 149208], Tables 1 through 4)). Many of these investigations have 
been straightforward measurements of the rate of hydriding under various conditions. However, 
at least one set of experiments directly determined the origin of hydrogen in the metal. 
Waterside Corrosion of Zirconium Alloys in Nuclear Power Plants (IAEA 1998 [DIRS 150560], 
p. 92) discusses experiments in which zirconium-base alloys were oxidized in normal water 
(H2O) with dissolved tritium gas (T2). This experiment is extremely sensitive. If even 1 ppm of 
the hydrogen were from dissolved tritium gas, the radioactivity of T2 would result in thousands 
of decays per second per one square centimeter of surface. This level of activity would be readily 
detected. The IAEA (1998 [DIRS 150560], p. 92) gives the following discussion of the 
experiment: 
Oxidation studies using T2/H2O mixtures ... have shown that, during normal 
oxidation, no T2 enters the metal ... until the thermally-induced exchange reaction 
has progressed to the point where a measurable fraction of HTO has been formed. 
Thus, the hydrogen isotopes which enter the metal do so as an integral part of the 
reaction of the zirconium with water molecules, and not by reaction with any 
dissolved hydrogen in the water. Studies have shown that this situation persists ... 
until hydrogen over-pressures in the system of tens of MPa are present. 
Note that, according to this quotation, the hydrogen pressures required to cause hydriding are 
quite large. By comparison, the highest hydrogen pressure in a breached waste package is pure 
hydrogen gas at atmospheric pressure, or about 0.1 MPa. 
Additional corroborating information follows. The oxide film on the surface of the metal is 
important in preventing the uptake of hydrogen from the surrounding environment. Under 
specific experimental conditions, this oxide layer can be removed. At 200C, the solubility of 
oxygen in zirconium is greater than 1 percent by weight, and the solubility of oxygen increases 

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with increasing temperature (Baker 1992 [DIRS 149104], p. 2-326). As a result, the oxide film 
on the surface of a piece of zirconium is normally not thermodynamically stable. In his 
discussion of experimental procedure, Smith (1966 [DIRS 149107], p. 325) notes, Zirconium 
samples were first annealed at 700C under vacuum (~ 10-3 mm Hg) to remove any oxide film. 
The film dissolved into the samples, leaving them a bright metallic color. It is clear that the 
oxide film can be damaged or even destroyed by heat treatment in a suitable environment. 
However, the film can be maintained if there is a supply of oxygen. Water will serve as a source 
of oxygen, because the electrochemical domain of stability for zirconium metal lies well below 
that of water (Pourbaix 1974 [DIRS 100817], p. 226). Spent nuclear fuel has a robust oxide 
layer from reactor operation. Under the oxidizing and humid environment of a failed waste 
package, it is expected that the oxide layer will be preserved. 
Garzarolli et al. (1979 [DIRS 149256], p. 64) studied the effect of the oxygen supply and stated: 
The effect of the composition of the gas atmosphere on the electrical properties of 
ZrO2 corrosion films was measured. The results revealed a large decrease in the 
electric resistance when the atmosphere changed from oxidizing to non-oxidizing, 
indicating a drastic change of the morphology (passivity of the oxide film)  the 
obvious implication of all available results is that massive hydriding can start 
when the availability of oxygen to continuously repair the protective oxide film 
falls below a critical value. 
Hydriding of zirconium by absorption of gas has been observed in the laboratory. For example, 
Smith (1966 [DIRS 149107], Table 3) provides data on hydrogen absorption. However, the 
environment for these experiments was hot, extremely pure hydrogen. Smith (1966 
[DIRS 149107], Table 3) states that the temperatures for the hydrogen absorption experiments 
were 210C to 700C. Smith (1966 [DIRS 149107], p. 325) notes, hydrogen was purified by 
passing it through a Deoxo unit, a bed of platinized asbestos (300C), a tube of P2O5 and a liquid 
nitrogen trap. The evident intention is to react any oxygen impurities, and absorb or condense 
any water vapor that is formed. After this treatment, little oxygen would have been available to 
maintain the oxide film. Such an environment is not relevant for a repository at Yucca Mountain 
because air, water vapor, or liquid water will be present and will maintain the protective oxide 
film. 
Hydrogen would have no effect on in-package chemistry because it would immediately diffuse 
out of the package before reacting with anything. Also, hydrogen is a reduced gas and is, 
therefore, unlikely to form in an oxidizing environment. The uranium in the UO2 first oxidizes 
(U4+ to U6+), dissolves, and then precipitates immediately on the pellet surface as a schoepitetype 
mineral. Although the kinetics of hydrogen gas reactions are sluggish under the conditions 
expected in a breached waste package, the presence of hydrogen would reduce the concentration 
of oxidizing agents and hence the corrosion rate of the UO2; neglecting this effect is, therefore, 
conservative. The presence of some hydrogen gas would not affect the hydrolysis reactions that 
control the rate of DHLW dissolution. The hydrogen would be generated for the first 500 years 
while the waste package internals are corroding so this conservatism exists for a limited time. 
In conclusion, hydrogen gas (H2) generation from waste package corrosion, as a degradation 
mechanism for cladding, is excluded from the TSPA-LA on the basis of low consequence. The 

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cladding absorbs little or no hydrogen from the outside environment such as waste package 
corrosion and hydride embrittlement will not occur. The NRC requirements in 10 CFR 63.114 
(e and f) [DIRS 156605] allow this omission because H2 will not damage the cladding and 
therefore will not significantly change the magnitude and time of the resulting radiological 
exposures to the reasonably maximally exposed individual (RMEI), or radionuclide releases to 
the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 

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7. CONCLUSIONS 
This report summarizes the screening of 24 waste form and cladding degradation-related FEPs 
for possible inclusion in the TSPA-LA. The FEPs and the result of their screening are delineated 
in Table 7-1. This table provides the FEP number, name, screening decision 
(included/excluded), and basis for the exclusion (i.e., low probability or low consequence). Low 
consequence means that omitting the FEP will not significantly change the magnitude and time 
of the resulting radiological exposures to the reasonably maximally exposed individual (RMEI), 
or radionuclide releases to the accessible environment (10 CFR 63.114 e and f [DIRS 156605]). 
For the first nineteen FEPs in Table 7-1, the FEP screening applies to CSNF and naval spent 
fuel. The last five FEPs in Table 7-1 address cladding and all waste forms including 
DOE-owned spent nuclear fuel (DSNF) and defense high-level waste (HLW). Uncertainties 
pertaining to the implementation of the included FEPs are identified in the supporting reports 
listed at the end of each included FEP description. The results of this screening process are 
documented in the screening arguments and TSPA-LA disposition statements in Section 6.2. 
For the repository, cladding is classified as a barrier to release as part of the waste form. 
However, it is not designed nor controlled by the project. The behavior can be modeled but the 
design characteristics are not within the project controls. 
Any restrictions to the use of this document are inherently captured in the discussion of the 
included FEP and supporting reports. No other restrictions or limitations are identified. 
Table 7-1. Summary of Cladding Degradation and Waste Form FEPs 
FEP 
Number FEP Name 
Screening Decision and 
Basis Section 
2.1.02.11.0A Degradation of cladding from waterlogged rods Excluded (Low Consequence) 6.2.1 
2.1.02.12.0A Degradation of cladding prior to disposal Included 6.2.2 
2.1.02.13.0A General corrosion of cladding Excluded (Low Consequence) 6.2.3 
2.1.02.14.0A Microbially influenced corrosion (MIC) of cladding Excluded (Low Consequence) 6.2.4 
2.1.02.15.0A Localized (radiolysis enhanced) corrosion of cladding Excluded (Low Consequence) 6.2.5 
2.1.02.16.0A Localized (pitting) corrosion of cladding Excluded (Low Consequence) 6.2.6 
2.1.02.17.0A Localized (crevice) corrosion of cladding Excluded (Low Consequence) 6.2.7 
2.1.02.18.0A Enhanced corrosion of cladding from dissolved silica Excluded (Low Consequence) 6.2.8 
2.1.02.19.0A Creep rupture of cladding Excluded (Low Consequence) 6.2.9 
2.1.02.20.0A Internal pressurization of cladding Excluded (Low Consequence) 6.2.10 
2.1.02.21.0A Stress corrosion cracking (SCC) of cladding Excluded (Low Consequence) 6.2.11 
2.1.02.22.0A Hydride cracking of cladding Excluded (Low Consequence) 6.2.12 
2.1.02.23.0A Cladding unzipping Included 6.2.13 
2.1.02.24.0A Mechanical impact on cladding Excluded (Low Probability ) 6.2.14 
2.1.02.25.0B Naval SNF cladding Included 6.2.15 
2.1.02.26.0A Diffusion-controlled cavity growth in cladding Excluded (Low Consequence) 6.2.16 
2.1.02.27.0A Localized (fluoride enhanced) corrosion of cladding Excluded (Low Consequence) 6.2.17 
2.1.07.01.0A Rockfall Excluded (Low Probability) 6.2.18 
2.1.09.03.0A Volume increase of corrosion products impacts 
cladding 
Included 6.2.19 

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Table 7-1. Summary of Cladding Degradation and Waste Form FEPs (Continued) 
FEP 
Number FEP Name 
Screening Decision and 
Basis Section 
2.1.09.09.0A Electrochemical effects in EBS Excluded (Low Consequence) 6.2.20 
2.1.09.11.0A Chemical effects of waste-rock contact Excluded (Low Consequence) 6.2.21 
2.1.11.05.0A Thermal expansion/stress of in-package EBS 
components 
Excluded (Low Consequence) 6.2.22 
2.1.12.02.0A Gas generation (He) from waste form decay Excluded (Low Consequence) 6.2.23 
2.1.12.03.0A Gas generation (H2) from waste package corrosion Excluded (Low Consequence) 6.2.24 
The conclusions from this document (FEP screening decision, TSPA-LA disposition for included 
FEPs, or screening argument for excluded FEPs), along with any modifications 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, 
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 YMRP ACCEPTANCE CRITERIA 
The purpose of this report is to evaluate and document the inclusion or exclusion of cladding 
degradation 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 (AC) from Yucca Mountain Review Plan, Review Plan 
(NRC 2003 [DIRS 163274], Section 2.2.1.2.1.3) were addressed in this report: 
Acceptance Criterion 1The 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 large

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magnitude 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 cladding degradation-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 cladding degradation-related FEPs is consistent with 
the site characterization data. 
Acceptance Criterion 2Screening 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, 
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 cladding degradation-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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8. INPUTS AND REFERENCES 
8.1 DOCUMENTS CITED 
161991 Adler Flitton, M.K.; Mizia, R.E.; and Bishop, C.W. 2002. Underground Corrosion of 
Activated Metals in an Arid Vadose Zone Environment. Corrosion/2002, [57th 
Annual Conference & Exposition, April 7-11, 2002, Denver, Colorado]. Paper 
No. 02531. Houston, Texas: NACE International. TIC: 253838. 
103753 ASM International. 1987. Corrosion. Volume 13 of Metals Handbook. 9th Edition. 
Metals Park, Ohio: ASM International. TIC: 209807. 
149104 Baker, H. 1992. Binary Alloy Phase Diagrams. Volume 3 of ASM Handbook. 
Page 2.326. Materials Park, Ohio: ASM International. TIC: 247622. 
156122 Beckman, D.A. 2001. NRC Issue Resolution Status Report on Container Life and 
Source Term, Revision 3. Memorandum from D.A. Beckman (BSC) to Distribution, 
February 21, 2001, LV.LAP.AH.02/2001-236, with enclosure. 
ACC: MOL.20010418.0048. 
101564 Bradley, E.R.; Bailey, W.J.; Johnson, A.B., Jr.; and Lowry, L.M. 1981. Examination 
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Clad Degradation - FEPs Screening Arguments 
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152778 Cox, B. 1990. Environmentally-Induced Cracking of Zirconium Alloys  A 
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152354 Cragnolino, G.A.; Dunn, D.S.; Brossia, C.S.; Jain, V.; and Chan, K.S. 1999. 
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164598 Hayes, T.A.; Rosen, R.S.; and Kassner, M.E. 1999. Critical Analysis of Dry Storage 
Temperature Limits for Zircaloy-Clad Spent Nuclear Fuel Based on Diffusion 
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100455 Hillner, E.; Franklin, D.G.; and Smee, J.D. 1998. The Corrosion of Zircaloy-Clad 
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150560 IAEA (International Atomic Energy Agency) 1998. Waterside Corrosion of 
Zirconium Alloys in Nuclear Power Plants. IAEA-TECDOC-996. Vienna, 
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169906 Jones, D.A. 1992. Principles and Prevention of Corrosion. 1st Edition. New York, 
New York: Macmillan. TIC: 242631. 
123682 Knoll, R.W. and Gilbert, E.R. 1987. Evaluation of Cover Gas Impurities and Their 
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131519 Kohli, R. and Pasupathi, V. 1986. Investigation of Water-logged Spent Fuel Rods 
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100462 Kreyns, P.H.; Bourgeous, W.F.; White, C.J.; Charpentier, P.L.; Kammenzind, B.F.; 
and Franklin, D.G. 1996. Embrittlement of Reactor Core Materials. Zirconium in 
the Nuclear Industry, Eleventh International Symposium held in Garmisch- 
Partenkirchen, Germany, September 11-14, 1995. Bradley, E.R. and Sabol, G.P., eds. 
ASTM STP 1295. Pages 758-782. West Conshohocken, Pennsylvania: American 
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100051 Langmuir, D. 1997. Aqueous Environmental Geochemistry. Upper Saddle River, 
New Jersey: Prentice Hall. TIC: 237107. 
101704 Lanning, D.D.; Beyer, C.E.; and Painter, C.L. 1997. FRAPCON-3: Modifications to 
Fuel Rod Material Properties and Performance Models for High-Burnup Application. 
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160832 Lide, D.R., ed. 2002. CRC Handbook of Chemistry and Physics. 83rd Edition. 
Boca Raton, Florida: CRC Press. TIC: 253582. 
131533 Little B. and Wagner P. 1996. An Overview of Microbiologically Influenced 
Corrosion of Metals and Alloys Used in the Storage of Nuclear Wastes. Canadian 
Journal of Microbiology, 42, (4), 367-374. Ottawa, Canada: National Research 
Council of Canada. TIC: 246614. 

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101719 Manaktala, H.K. 1993. Characteristics of Spent Nuclear Fuel and Cladding Relevant 
to High-Level Waste Source Term. CNWRA 93-006. San Antonio, Texas: Center for 
Nuclear Waste Regulatory Analyses. TIC: 208034. 
131537 McNeil, M. and Odom, A. 1994. Thermodynamic Prediction of Microbiologically 
Influenced Corrosion (MIC) by Sulfate-Reducing Bacteria (SRB). Microbiologically 
Influenced Corrosion Testing. ASTM STP 1232. Pages 173-179. Philadelphia, 
Pennsylvania: American Society for Testing and Materials. TIC: 246989. 
100468 Merriam-Webster. 1993. Merriam-Webster's Collegiate Dictionary, 10th Edition. 
Springfield, Massachusetts: Merriam-Webster. TIC: 8883. 
101903 NRC (U.S. Nuclear Regulatory Commission) 1997. Standard Review Plan for Dry 
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Commission. ACC: MOL.20010724.0307. 
147797 NRC 2000. Interim Staff Guidance - 11. Storage of High Burnup Spent Fuel. ISG - 
11. Washington, D.C.: U.S. Nuclear Regulatory Commission. Accessed March 21, 
2000. TIC: 247227. http://www.nrc.gov/OPA/reports/isg11.htm 
164593 NRC 2002. Interim Staff Guidance - 11, Revision 2. Cladding Considerations for the 
Transportation and Storage of Spent Fuel. ISG-11, Rev. 2. Washington, D.C.: U.S. 
Nuclear Regulatory Commission. TIC: 254441. 
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Safety and Safeguards. TIC: 254568. 
102065 Peehs, M. and Fleisch, J. 1986. LWR Spent Fuel Storage Behaviour. Journal of 
Nuclear Materials, 137, (3), 190-202. Amsterdam, The Netherlands: North-Holland 
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Dry Storage of Spent Nuclear Fuel. EPRI TR-103949. Palo Alto, California: Electric 
Power Research Institute. TIC: 102933. 
165318 Piron, J.P. and Pelletier, M. 2001. State of the Art on the Helium Issues. Section 5.3 
of Synthesis on the Long Term Behavior of the Spent Nuclear Fuel. Poinssot, C., ed. 
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102067 Puls, M.P. 1988. The Influence of Hydride Size and Matrix Strength on Fracture 
Initiation at Hydrides in Zirconium Alloys. Metallurgical Transactions A, 19A, (6), 
1507-1522. Metals Park, Ohio: American Society for Metals. TIC: 237143. 

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161992 Ralph, R.; Purohit, A.; and Dille, E. 2002. Characterization of Oxide Formation on 
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121838 Reed-Hill, R.E. 1973. Physical Metallurgy Principles. Second Edition. 797-801. 
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149107 Smith, T. 1966. Kinetics and Mechanism of Hydrogen Permeation of Oxide Films 
on Zirconium. Journal of Nuclear Materials, 18, 323-336. Amsterdam, 
The Netherlands: Elsevier. TIC: 247624. 
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Geometries with Both Temporal and Spatial Resolution. Journal of Chemical 
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166303 tefanic I. and LaVerne, J.A. 2002. Temperature Dependence of the Hydrogen 
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102093 Tasooji, A.; Einziger, R.E.; and Miller, A.K. 1984. Modeling of Zircaloy Stress- 
Corrosion Cracking: Texture Effects and Dry Storage Spent Fuel Behavior. 
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Testing and Materials. TIC: 241417. 
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Edition. Pages 661-669. Materials Park, Ohio: ASM International. TIC: 241059. 
100097 Wolery, T.J. and Daveler, S.A. 1992. EQ6, A Computer Program for Reaction Path 
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Related Documentation (Version 7.0). UCRL-MA-110662 PT IV. Livermore, 
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165268 Wolfram, J.H.; Mizia, R.E.; Jex, R.; Nelson, L.; and Garcia, K.M. 1996. The Impact 
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102050 Yau, T. L. 1984. Zirconium Versus Corrosive Species in Geothermal Fluids. Paper 
140 of Corrosion 84. Houston, Texas: National Association of Corrosion Engineers. 
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110761 Yau, T.L. and Maguire, M. 1990. Control of Localized Corrosion of Zirconium in 
Oxidizing Chloride Media. Advances in Localized Corrosion, Proceedings of the 
Second International Conference on Localized Corrosion, June 1-5, 1987, Orlando, 
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Corrosion, Volume 13, Pages 707-721 of ASM Handbook. Formerly 9th Edition, 
Metals Handbook. [Materials Park, Ohio]: ASM International. TIC: 240704. 
149233 Yau, T-L. 1983. Corrosion Properties of Zirconium in Chloride Solutions. 
Corrosion 83, International Corrosion Forum, Anaheim, California, April 18-22, 
1983. Paper No. 26. Pages 26/1 - 26/13. Houston, Texas: National Association of 
Corrosion Engineers. TIC: 243849. 
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. 
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-2.27Q, Rev. 1, ICN 4. Planning for Science Activities. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC. 20040610.0006. 
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, DC: 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. 

Clad Degradation - FEPs Screening Arguments 
ANL-WIS-MD-000008 REV 02 8-9 October 2004 
AP-SV.1Q, Rev. 1, ICN 1. Control of the Electronic Management of Information. 
Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste 
Management. ACC: DOC.20040308.0001. 
LP-SI.11Q-BSC, Rev. 0, ICN 1. Software Management. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20041005.0008. 
105725 ASTM C 1174-97. 1998. Standard Practice for Prediction of the Long-Term 
Behavior of Materials, Including Waste Forms, Used in Engineered Barrier Systems 
(EBS) for Geological Disposal of High-Level Radioactive Waste. West 
Conshohocken, Pennsylvania: American Society for Testing and Materials. 
TIC: 246015. 
8.3 SOURCE DATA, LISTED BY DATA TRACKING NUMBER 
170760 MO0407SEPFEPLA.000. LA FEP List. Submittal date: 07/20/2004. 

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APPENDIX A 
CLADDING TEMPERATURE ESTIMATES FOR POSTCLOSURE 

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A. CLADDING TEMPERATURE ESTIMATES FOR POSTCLOSURE 
Many of the FEPs in this document depend on temperature. Peak temperatures are important for 
some FEPs such as delayed hydride cracking. Other FEPs such as general corrosion or creep 
depend on both the temperature levels and how long the temperatures stay elevated. 
The cladding temperatures are predicted by adding the interior temperature rise across the waste 
package to the waste package surface temperature. The waste package surface temperatures 
(BSC 2004 [DIRS 169565], Section 6.2) vary with waste package location in the repository, 
water infiltration flux, and time (heat from radioactive decay). The temperature rise across the 
waste package (BSC 2004 [DIRS 166692]) depends on the thermal loading (decay heat power) 
of the waste package and the loading pattern. The peak cladding temperature for a specific rod 
will also depend on the location of the rod within the waste package. The maximum cladding 
temperatures are in the center location and the coolest locations are along the outer row of 
locations. 
A.1 EVALUATION OF CLADDING TEMPERATURE 
The temperature histories from Multiscale Thermohydrologic Model (BSC 2004 [DIRS 169565], 
Figure 6.3-67) were developed based on all waste packages and accounting for the uncertainty of 
the host-rock thermal conductivity and percolation flux. The maximum waste package surface 
temperature of around 205C (BSC 2004 [DIRS 169565], Figure 6.3-67) occurs between 65 and 
70 years after emplacement (15 to 20 years after ventilation period). The waste package surface 
temperatures for this time period varies from around 100C to 205C due to the uncertainties. 
The internal waste package temperature difference (the temperature difference from the waste 
package surface to a specific location inside the waste package) decreases with time after 
emplacement (BSC 2004 [DIRS 166692]). Also, the decrease in the internal waste package 
temperature difference is less than the waste package surface temperature increase from the time 
after ventilation until the peak waste package temperature peaks between 15 to 20 years 
afterwards (BSC 2004 [DIRS 169565], Figure 6.3-67). Thus, maximum cladding temperature 
can be estimated at 65 years after emplacement, which results in a estimate of 296C (205C 
waste package surface temperature + 91.1C waste package internal temperature increase (BSC 
2004 [DIRS 166692]), using the internal temperature difference (T) from the surface of the 21- 
PWR waste package with 50 years of ventilation for 65 years in the center at the hottest 
location). After the peak at 65 years both the waste package surface and internal temperature 
increase will continue to decrease. 
This evaluation shows that with the predicted waste package surface temperatures, using the 
wide range of uncertainties, and the predicted waste package internal temperatures at the time 
when the peak temperatures are occurring, the temperature for the cladding remains 
below 300C. 
The above represents the TSPA nominal scenario repository environment where the drift remains 
open. Higher waste package surface and consequently cladding temperatures are possible in the 
case of a complete drift collapse early in the postclosure period. This could only occur as a 
result of a seismic event. The seismic event probabilities, consequences, and treatment for the 
TSPA-LA model are discussed in Seismic Consequence Abstraction (BSC 2004 [DIRS 169183]). 

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APPENDIX B 
QUALIFICATION OF DATA THAT SUPPORTS CLAD DEGRADATION - FEPS 
SCREENING ARGUMENTS (ANL-WIS-MD-000008) 

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B. QUALIFICATION OF DATA THAT SUPPORTS CLAD DEGRADATION - FEPS 
SCREENING ARGUMENTS (ANL-WIS-MD-000008) 
B.1 PURPOSE 
The Clad Degradation - FEPs (Features, Events and Processes) Screening Arguments 
(analysis) 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 qualifies those data that the cladding 
degradation analysis uses as direct input. The appendix was prepared according to AP-SIII.9Q, 
Scientific Analyses. The outside sources of data that are used as direct inputs to the screening of 
FEPs being qualified for use within this report are addressed in this appendix. 
B.2 METHODS AND CRITERIA 
The data obtained form outside sources that are not established facts are documented in this 
appendix for use within this technical report. The extent to which the data demonstrate the 
properties of interest must be addressed. 
One or more of the following factors from AP-SIII.9Q, Section 5.2.1 shall be used to qualify data 
obtained from an outside source: 
 Reliability of data source 
 Qualifications of personnel or organizations generating the information, 
 Prior uses of the data, and 
 Availability of corroborating data. 
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 twenty-one modes of 
cladding 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 factors used to qualify the information. The table 
designates the attributes or factors that are used to qualify the information. The data to be 
qualified is described in the following subsections. 

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Table B-1. Sources of Technical Information and Criteria Used to Qualify the Information 
Sources Factors 
Greene, C.A.; Brossia, C.S.; Dunn, D.S.; and Cragnolino, G.A. 
2000. Environmental and Electrochemical Factors on the Localized 
Corrosion of Zircaloy-4. Corrosion/2000, [55th Annual Conference 
& Exposition, March 26-31, 2000, Orlando, Florida]. Paper No. 
00210. Houston, Texas: NACE International. TIC: 246988. 
DIRS 145073 
Properties of Interest 
Availability of corroborating data. 
Hansson, C.M. 1984. The Corrosion of Zircaloy 2 in Anaerobic 
Synthetic Cement Pore Solution. SKB TR-84-13. Stockholm, 
Sweden: Svensk Krnbrnslefrsrjning A.B. TIC: 206293. 
DIRS 101676 
Properties of Interest 
Availability of corroborating data. 
Hayes, T.A.; Rosen, R.S.; and Kassner, M.E. 1999. Critical Analysis 
of Dry Storage Temperature Limits for Zircaloy-Clad Spent Nuclear 
Fuel Based on Diffusion Controlled Cavity Growth. UCRL-ID- 
131098. Livermore, California: Lawrence Livermore National 
Laboratory. TIC: 254551. 
DIRS 164598 
Properties of Interest 
Availability of corroborating information. 
Hillner, E.; Franklin, D.G.; and Smee, J.D. 1998. The Corrosion of 
Zircaloy-Clad Fuel Assemblies in a Geologic Repository 
Environment. WAPD-T-3173. West Mifflin, Pennsylvania: Bettis 
Atomic Power Laboratory. TIC: 237127. 
DIRS 100455 
Properties of Interest 
Availability of corroborating data. 
Piron, J.P. and Pelletier, M. 2001. State of the Art on the Helium 
Issues. Section 5.3 of Synthesis on the Long Term Behavior of the 
Spent Nuclear Fuel. Poinssot, C., ed. CEA-R-5958(E). Volume 1. 
[Paris], France: Commissariat  l'nergie Atomique. TIC: 253976. 
DIRS 165318 
Properties of Interest 
Availability of corroborating information 
Wolfram, J.H.; Mizia, R.E.; Jex, R.; Nelson, L.; and Garcia, K.M. 
1996. The Impact of Microbially Influenced Corrosion on Spent 
Nuclear Fuel and Storage Life. INEL-96/0335. Idaho Falls, Idaho: 
Idaho National Engineering Laboratory, Lockheed Martin Idaho 
Technologies Company. ACC: MOL.20030925.0039. 
DIRS 165268 
Properties of Interest 
Availability of corroborating data 
Both the discussions above and in the following sections fulfill requirements of AP-SIII.9Q to 
discuss the data for qualification, the method of qualification, and the evaluation criteria 
(AP-SIII.9Q, Section 5.2.1, paragraph l)). 
B.3.1 Qualification of Data from Greene, C.A.; Brossia, C.S.; Dunn, D.S.; and 
Cragnolino, G.A. 2000. [DIRS 145073] 
Properties of Interest This analysis (Section 6.7) evaluates the potential for localized (crevice) 
corrosion of the cladding. The conclusion of this source was that crevice corrosion was not 
observed. These data are integral to evaluation of the cladding degradation FEPs. 
Extent and Quality of Corroborating InformationTwo sources corroborate the above source. 
Yau and Webster (1987 [DIRS 100494], p. 717), an authoritative source (ASM International 
1987 [DIRS 103753], p. 717; an established fact), reported that zirconium resists crevice 
corrosion. In low-pH, chloride solutions zirconium is not subject to crevice attack. 
Brossia et al. (2002 [DIRS 161988]) performed pitting and crevice corrosion tests on Zircaloy-4. 
They report that no crevice corrosion is observed under the same environment and 
electrochemical conditions that promote pitting corrosion on exposed surfaces. 

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B.3.2 Qualification of Data from Hansson 1984 [DIRS 101676] 
Properties of Interest  This analysis assesses the extent to which silica could affect cladding 
corrosion. The analysis considers high dissolved silica content in water, the source considers the 
silica content in concrete, and an authoritative source (Yau and Webster 1987 [DIRS 100494]) 
considers the silica concentration in sea water. All three anticipate little cladding degradation 
that can be attributed to silica concentrations that far exceed those that the cladding would 
encounter in the repository. 
Extent and Quality of Corroborating Information An authoritative source, Yau and Webster 
(1987 [DIRS 100494]) corroborates this source's assessment of cladding's resistance to enhanced 
corrosion at the silica concentrations estimated for the repository. 
B.3.3 Qualification of Data from Hayes et al. 1999 [DIRS 164598] 
Properties of Interest This analysis assesses creep as a mechanism for failure of Zircaloy 
cladding. The analysis does not expect creep at peak repository temperatures (<300C), and 
others expect little creep damage under 400C. Evaluation of failure via this mechanism is 
essential to the screening of cladding degradation FEPs. 
Availability of Corroborating InformationAn authoritative source (NRC 2002 
[DIRS 164593]), corroborates this source's assessment of creep. Both state that cladding failure 
attributed to creep is unlikely for peak temperatures below 400C. 
B.3.4 Qualification of Data from Hillner et al. 1998 [DIRS 100455] 
Properties of InterestThis analysis evaluates the potential for microbes to cause cladding 
failure. The corrosion of Zircaloy-clad fuels via microbes under repository conditions and the 
organic acids produced by microbes are unlikely to significantly accelerate zirconium corrosion. 
Microbially induced corrosion is one of the FEPs that requires evaluation therefore, the data 
presented in this reference is essential. 
Availability of Corroborating InformationThree independent sources corroborate Hillner et al. 
(1998). Wolfram et al. (1996 [DIRS 165268]) found no evidence in the literature that zirconium 
or its alloys are susceptible to microbially induced corrosion. Yau and Webster (1987 
[DIRS 100494]), an authoritative source, states that zirconium resists organic acids that microbes 
produce. And the Yucca Mountain Project (BSC 2004 [DIRS 169991], Section 7.1) evaluated 
in-drift microbial communities and concluded that they would have little effect on the in-drift 
geochemistry. 
B.3.5 Qualification of Data from Piron and Pelletier 2001 [DIRS 165318] 
Properties of Interest  This analysis considers whether helium could pressurize the cladding and 
cause it to fail. They concluded that fuel with a burnup of 47.5 MWd/kgU would produce 
1,171 cm3 (STP) of helium in a rod after 10,000 years. Piron and Pelletiers (2001 [DIRS 
165318], Section 5.3) values are based on having all of the helium released. This provides a 
total overpressure of 90 bar or 9.10 MPa at 20C in 10,000 years after emplacement. This 
corresponds to 9.3 MPa at 30C. 

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Availability of Corroborating Information Manaktala (1993 [DIRS 101719]) and Rothman 
(1984 [DIRS 100417]) corroborate this sources assertion that helium production would not fail 
the cladding. Both found that cooling of the rods would offset helium pressure buildup. 
Consequently, the pressure at 10,000 years would be less than what this source predicted, and the 
likelihood of cladding failure would be even less likely. 
B.3.6 Qualification of Data from Wolfram et al. 1996 [DIRS 165268] 
Properties of Interest  This analysis does not expect microbes to cause cladding failure. It 
found no evidence in the literature that zirconium or its alloys are susceptible to microbialinduced 
corrosion. Because the evaluation of microbially induced corrosion is one of the FEPs 
being screened, these data are essential. 
Availability of Corroborating Information  Three independent sources corroborate 
Wolfram et al. (1996 [DIRS 165268]). Hillner et al. (1998 [DIRS 100455]) supports the concept 
that organic acids produced by microbes are unlikely to significantly accelerate zirconium 
corrosion. Yau and Webster, 1987 [DIRS 100494], an authoritative source, states that zirconium 
resists organic acids that microbes produce. And the Yucca Mountain Project (BSC 2004 
[DIRS 169991], Section 7.1) evaluated in-drift microbial communities and concluded that they 
would have little effect on the in-drift geochemistry. 
B.4 CONCLUSION 
In conclusion, the data are qualified for their intended use.