FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation REV 03, ICN 00

ANL-EBS-PA-000002

October 2004
 
1. PURPOSE 
The purpose of this report is to evaluate and document the inclusion or exclusion of features, 
events and processes (FEPs) with respect to drip shield and waste package modeling used to 
support the Total System Performance Assessment for License Application (TSPA-LA). 
Thirty-three FEPs associated with the waste package and drip shield performance have been 
identified (DTN: MO0407SEPFEPLA.000 [DIRS 170760]). A screening decision, either 
included or excluded, has been assigned to each FEP, with the technical bases for screening 
decisions, as required by the Nuclear Regulatory Commission (NRC) in 10 CFR 63.114 (d, e, 
and f) [DIRS 156605]. 
The FEPs analyses in this report address issues related to the degradation and potential failure of 
the drip shield and waste package over the post closure regulatory period of 10,000 years after 
permanent closure. For included FEPs, this report summarizes the disposition of the FEP in 
TSPA-LA. For excluded FEPs, this report provides the technical bases for the screening 
arguments for exclusion from TSPA-LA. 
The analyses are for the TSPA-LA base-case design (BSC 2004 [DIRS 168489]), where a drip 
shield is placed over the waste package without backfill over the drip shield (BSC 2004 
[DIRS 168489]). Each FEP includes one or more specific issues, collectively described by a 
FEP name and description. The FEP description encompasses a single feature, event, or process, 
or a few closely related or coupled processes, provided the entire FEP can be addressed by a 
single specific screening argument or TSPA-LA disposition. The FEPs were assigned to 
associated Project reports, so the screening decisions reside with the relevant subject-matter 
experts. 
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]). 
1.2 SCOPE 
The scope of this report is to identify the treatment of the FEPs affecting waste package and drip 
shield post closure performance over the 10,000-year regulatory period following permanent 
closure. Igneous and seismic effects are not treated within this report. A discussion of igneous 
and seismic FEPs can be found in Features, Events, and Processes: Disruptive Events 
(BSC 2004 [DIRS 170017]). 
The scope for this activity involves two tasks, namely: 
1. Identify the drip shield and waste package FEPs included in the TSPA-LA analysis and 
the reports in which they are addressed. 
2. Identify the drip shield and waste package FEPs excluded from analysis in the TSPA-LA 
and justify why they are excluded. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 1-2 October 2004 
This report describes and evaluates the technical bases for screening decisions associated with 
the drip shield and waste package FEPs. For included FEPs, this report summarizes how the 
FEP is addressed in TSPA-LA. For excluded FEPs, this report provides the technical basis for 
the decision to exclude (i.e., low probability, low consequence, or by regulation). In cases where 
a FEP encompasses multiple technical areas and is shared with other FEPs documents, this report 
provides a partial technical basis for the screening decision as it relates to drip shield and waste 
package. Collectively, the documents address the complete technical basis. This information is 
provided in Section 6.2 and subsequent sections. 
An overview of the Yucca Mountain Project 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), which describes the TSPA-LA FEP identification and 
screening process. As part of that process, the TSPA-LA FEP list 
(DTN: MO0407SEPFEPLA.000 [DIRS 170760]) was developed. This DTN was used as an 
initial input to the drip shield and waste package FEP analysis. The list of drip shield and waste 
package TSPA-LA FEPs, presented in Table 1-1, was derived from DTN: 
MO0407SEPFEPLA.000 [DIRS 170760] with subsequent modifications to the FEP list, FEP 
names, and/or FEP descriptions. These modifications are documented in the FEP History File 
in the FEP database (BSC 2004 [DIRS 168706], Table 6-1) and will be incorporated into a 
subsequent revision of the TSPA-LA FEP list (see Section 7). Table 1-1 also includes the 
designation of shared FEPs. 
Direct inputs supporting the screening decisions are listed in Section 4. Indirect inputs 
supporting the screening decisions are listed in Section 6.1. The individual FEP discussions 
providing identification (FEP number, name, and description) and screening information 
(screening decision and screening argument (if FEP is excluded) or TSPA disposition (if FEP is 
included)) are in Section 6.2. 
Table 1-1. Drip Shield and Waste Package FEPs for TSPA-LA 
FEP 
Number FEP Name 
Addressed 
in Section Sharing With 
1.1.03.01.0A Error in Waste Emplacement 6.2.1 EBS 
2.1.03.01.0A General Corrosion of Waste Packages 6.2.2 
2.1.03.01.0B General Corrosion of Drip Shields 6.2.3 
2.1.03.02.0A Stress Corrosion Cracking (SCC) of Waste Packages 6.2.4 
2.1.03.02.0B Stress Corrosion Cracking (SCC) of Drip Shields 6.2.5 
2.1.03.03.0A Localized Corrosion of Waste Packages 6.2.6 
2.1.03.03.0B Localized Corrosion of Drip Shields 6.2.7 
2.1.03.04.0A Hydride Cracking of Waste Packages 6.2.8 
2.1.03.04.0B Hydride Cracking of Drip Shields 6.2.9 
2.1.03.05.0A Microbially Influenced Corrosion (MIC) of Waste Package 6.2.10 
2.1.03.05.0B Microbially Influenced Corrosion (MIC) of Drip Shields 6.2.11 
2.1.03.06.0A Internal Corrosion of Waste Packages Prior to Breach 6.2.12 WF 
2.1.03.07.0A Mechanical Impact on Waste Package 6.2.13 
2.1.03.07.0B Mechanical Impact on Drip Shield 6.2.14 
2.1.03.08.0A Early Failure of Waste Packages 6.2.15 
2.1.03.08.0B Early Failure of Drip Shields 6.2.16 
2.1.03.09.0A Copper Corrosion in EBS 6.2.17 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 1-3 October 2004 
Table 1-1. Drip Shield and Waste Package FEPs for TSPA-LA (Continued) 
FEP 
Number FEP Name 
Addressed 
in Section Sharing With 
2.1.03.10.0A Advection of Liquids and Solids Through Cracks in the Waste 
Package 
6.2.18 EBS 
2.1.03.10.0B Advection of Liquids and Solids Through Cracks in the Drip Shield 6.2.19 EBS 
2.1.03.11.0A Physical Form of Waste Package and Drip Shield 6.2.20 
2.1.06.06.0B Oxygen Embrittlement of Drip Shields 6.2.21 
2.1.06.07.0B Mechanical Effects at EBS Component Interfaces 6.2.22 EBS 
2.1.07.01.0A Rockfall 6.2.23 EBS, CLAD 
2.1.07.05.0A Creep of Metallic Materials in the Waste Package 6.2.24 
2.1.07.05.0B Creep of Metallic Materials in the Drip Shield 6.2.25 
2.1.09.03.0B Volume Increase of Corrosion Products Impacts Waste Package 6.2.26 
2.1.09.09.0A Electrochemical Effects in EBS 6.2.27 CLAD 
2.1.11.06.0A Thermal Sensitization of Waste Packages 6.2.28 
2.1.11.06.0B Thermal Sensitization of Drip Shields 6.2.29 
2.1.11.07.0A Thermal Expansion/Stress of In-Drift EBS Components 6.2.30 EBS 
2.1.12.03.0A Gas Generation (H2) from Waste Package Corrosion 6.2.31 EBS, CLAD 
2.1.13.01.0A Radiolysis 6.2.32 EBS, WF 
2.1.13.02.0A Radiation Damage in EBS 6.2.33 EBS, WF 
NOTES: FEPs = features, events, and processes; EBS = engineered barrier system; WF = waste form; 
CLAD = cladding. 
1.3 SCIENTIFIC ANALYSIS LIMITATION AND USAGE 
This report, the documentation for inclusion or exclusion of drip shield and waste package FEPs, 
provides information for a database that promotes FEPs traceability and transparency. The 
following limitations apply to this report: 
 Because this report cites other documents as direct input, its limitations inherently 
include limitations or constraints described within those documents. 
 Where FEPs are shared, this report is limited to performance of the drip shield and waste 
package. 
 For screening purposes, this report generally uses mean values of probabilities and either 
amplitudes of events or of consequences (e.g., mean time to degradation) as a basis for 
reaching an include or exclude decision. Mean values are determined based on the 
range of possible values. 
 The screening results presented are specific to the Yucca Mountain Project repository 
design and processes. Changes in direct inputs listed in Section 4.1, in baseline 
conditions used for this evaluation or in other subsurface conditions, must be assessed to 
determine whether the changes are within the limits stated in the FEP evaluations. 
Engineering and design changes are reviewed to determine adverse impacts to nuclear 
safety, in accordance with 10 CFR 63.73, Subparts F and G [DIRS 156605], and 
10 CFR 63.44 [DIRS 156605] requirements. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 1-4 October 2004 
INTENTIONALLY LEFT BLANK 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 2-1 October 2004 
2. QUALITY ASSURANCE 
Development of this report is 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). 
This document contributes to the analysis and modeling used to support TSPA-LA. The FEPs 
documented herein involve the investigations of items or barriers on the Q-list and have the 
potential to affect the calculation of the performance of the natural barriers and various 
engineered barrier system (EBS) components included on the Q-list. However, the drip shield 
and waste package FEPs do not qualify as Q-list items. The evaluations and conclusions do 
not directly impact engineered features important to safety, as defined in AP-2.22Q, 
Classification Analyses and Maintenance of the Q-List. 
Approved quality assurance procedures identified in the technical work plan (BSC 2004 
[DIRS 171583], Section 4) have been used to conduct and document the activities described in 
this document. In particular, this work constitutes an analysis report, and the documentation has 
been prepared according to AP-SIII.9Q, Scientific Analyses, and in accordance with related 
procedures and guidance documents as outlined in the technical work plan (BSC 2004 
[DIRS 171583]). The technical work plan also identifies applicable controls for the electronic 
management of data (BSC 2004 [DIRS 171583], Section 8) during the analysis and 
documentation activities. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 2-2 October 2004 
INTENTIONALLY LEFT BLANK 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 3-1 October 2004 
3. USE OF SOFTWARE 
This report uses no computational software; therefore, it 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 for traceability and 
transparency purposes, but were not used in the analyses discussed herein. 
This report was developed using only commercial off-the-shelf software, Microsoft Word 2000, 
for word processing that is exempt from qualification requirements in accordance with LPSI.
11Q-BSC, Software Management. No additional applications (routines or macros) were 
developed using this commercial off-the-shelf software. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 3-2 October 2004 
INTENTIONALLY LEFT BLANK 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 4-1 October 2004 
4. INPUTS 
AP-3.15Q, Managing Technical Product Inputs, categorizes technical product input as 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 Section 4.1 and indirect inputs are 
addressed in Section 6.1.3. 
The direct inputs were obtained from controlled source documents and other appropriate sources 
in accordance with AP-3.15Q, Managing Technical Product Inputs. Section 4.2 identifies the 
FEP screening criteria described in 10 CFR Part 63 [DIRS 156605] and the YMRP. Section 4.3 
identifies applicable codes and standards. 
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 drip shield and waste package FEPs for screening in this report. 
The TSPA-LA FEP list identifies a FEP report or a set of sharing FEP 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 6-1). 
Other direct inputs used for the FEP screening analysis are listed in Table 4-1. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 4-2 October 2004 
Table 4-1. Direct Inputs 
Technical Product Input Specifically Used From Specifically Used In Input Description 
ASM International 1990 
[DIRS 144385] 
p. 626 Table 4-1, Section 6.2.25 Handbook of metal properties 
pp. 169, 557, 650-652 Table 4-1, Section 6.2.31 HIC of WPOB 
p.557 Table 4-1, Sections 6.2.8, 6.2.12, 
6.2.27 
Corrosion potential of Alloy 22, 316 Stainless Steel, 
and other drift materials 
pp. 169, 557, 650-652 Table 4-1, Section 6.2.8 HIC discussion. Used as input to screening argument. 
ASM International 1987 
[DIRS 103753] 
p. 681 Table 4-1, Section 6.2.21 Oxygen embrittlement of titanium 
Boyer and Gall 1984 
[DIRS 155318] 
Section 32 Table 4-1, Section 6.2.24 Creep temperature of nickel-based alloys and titanium 
BSC 2001 [DIRS 154441] Entire Table 4-1, Section 6.2.17 Copper in use in gantry rail system 
BSC 2001 [DIRS 156807] Section 6, Table 5-13 Table 4-1, Section 6.2.19 Crack plugging 
BSC 2001 [DIRS 152655] Section 5.3 Table 4-1, Section 6.2.30 Waste package outer barrier stress due to thermal 
expansion with various barrier gap sizes 
BSC 2004 [DIRS 171924] Section 8 Table 4-1, Section 6.2.28 Conclusions about aging and phase stability in the 
repository 
BSC 2004 [DIRS 170024] Sections 6.2.8, 6.3, 6.3.3, 6.4, 6.4.2, 
6.4.7, 6.4.8, 7 (Table 22) 
Table 4-1, Sections 6.2.1, 6.2.15, 
6.2.16 
Early failure summary for waste package and drip 
shield 
Table 23 Table 4-1, Section 6.2.22 Commercial SNF waste package design 
Table 1 Table 4-1, Section 6.2.1 WP spacing 
BSC 2004 [DIRS 169766] 
Entire Table 4-1, Section 6.2.20 LA waste package designs 
BSC 2004 [DIRS 169480] Entire Table 4-1, Section 6.2.30 Radial gap between the waste package inner vessel 
and the waste package outer barrier 
BSC 2004 [DIRS 170803] Sections 3, 4.2.2 Table 4-1, Section 6.2.12 Helium backfilling of waste packages 
BSC 2004 [DIRS 169593] Sections 5.1.4, 6.3, 6.4.1, 6.4.2, 
Tables 6.4-1, 6.4-3, Figure 6.4-1 
Table 4-1, Sections 6.2.32, 6.2.33 Gamma dose rates to waste package materials 
BSC 2004 [DIRS 166879] Sections 7.2.1, 7.2.2 Table 4-1, Sections 6.2.14, 6.2.17, 
6.2.23, 6.2.27 
Deflection of the drip shield due to rockfall 
BSC 2004 [DIRS 169868] Section 6.3.3 Table 4-1, Section 6.2.18 SCC crack opening 
Sections 6.4.1, 6.5.8, 6.6.3, 6.6.4, 
6.6.5, 8.2, 8.4, Tables 16 and 17 
Table 4-1, Sections 6.2.1, 6.2.3, 
6.2.7 
Corrosion of drip shield BSC 2004 [DIRS 169845] 
Sections 6.5.3, 6.6.3, 6.7.2 Table 4-1, Section 6.2.11, 6.2.29 Corrosion of drip shield 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 4-3 October 2004 
Table 4-1. Direct Inputs (Continued) 
Technical Product Input Specifically Used From Specifically Used In Input Description 
Section 6.4.2 Table 4-1, Sections 6.2.2, 6.2.13 Dry air oxidation 
Sections 6.4.5, 8.2 Table 4-1, Sections 6.2.2, 6.2.10 Effect of microbial activity on corrosion 
Section 6.4.3 Table 4-1, Section 6.2.2 Aqueous corrosion rates 
BSC 2004 [DIRS 169984] 
Sections 6.4.4, 7.2.4, 8.3, Table 7-4 Table 4-1, Section 6.2.6 Discussion of localized corrosion models 
BSC 2004 [DIRS 169847] Sections 6.1.2, 6.1.3, 6.1.4 6.1.5, 
6.1.6, 6.3.2, 8, 8.1 
Table 4-1, Sections 6.2.9, 6.2.17, 
6.2.31 
HIC of drip shield 
BSC 2004 [DIRS 169565] Figure 6.3-56 Table 4-1, Sections 6.2.8, 6.2.21, 
6.2.24, 6.2.25 
Temperature history of waste packages and drip 
shields 
Sections 6.3.7, 6.4.2, 6.5.2 Table 4-1, Sections 6.2.5, 6.2.18, 
6.2.19, 6.2.23, 6.2.26 
Stress corrosion cracking of the drip shield, the waste 
package outer barrier 
BSC 2004 [DIRS 169985] 
Sections 6, 8.3, Tables 8-1 thru 8-3 Table 4-1, Section 6.2.4 Discussion of stress corrosion cracking inputs to 
TSPA-LA 
Table 1 Table 4-1, Section 6.2.1 WP spacing BSC 2004 [DIRS 168489] 
Entire Table 4-1, Section 6.2.20 LA waste package designs 
BSC 2004 [DIRS 169996] Sections 6.3.2, 6.4.3 Table 4-1, Section 6.2.20 Nominal waste package configurations analyzed 
CRWMS M&O 2000 
[DIRS 150823] 
Section 6.3 Table 4-1, Section 6.2.13 Design analysis for the defense high-level waste 
disposal container 
CRWMS M&O 2000 
[DIRS 123881] 
p. II-10 Table 4-1, Section 6.2.12 Fuel drying 
Etherington 1958 [DIRS 164789] p. 10-107 Table 4-1, Section 6.2.33 Neutron fluence level thresholds for changes in 
mechanical properties 
Haynes International 1988 
[DIRS 101995] 
Entire Table 4-1, Section 6.2.24 Melting temperature 
LL021012712251.021 
[DIRS 163112] 
Entire Table 4-1, Sections 6.2.2, 6.2.3, 
6.2.7 
Long term corrosion testing results 
LL021105312251.023 
[DIRS 161253] 
Entire Table 4-1, Section 6.2.4 SCC of WPOB and DS 
MO0003RIB00073.000 
[DIRS 152926] 
Entire Table 4-1, Section 6.2.29 Physical characteristics of Titanium Grade 7 
MO0407SEPFEPLA.000 
[DIRS 170760] 
Entire Table 4-1, Sections 1, 1.2, 4.1, 
6.1.1 
FEPs database and description 
Revie et al. 2000 [DIRS 159370] Chapter 47 Table 4-1, Section 6.2.11 Titanium alloy properties 
Schutz and Thomas 1987 
[DIRS 144302] 
Entire Table 4-1, Sections 6.2.9, 6.2.31 Hydrogen absorption conditions in titanium alloys 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 4-4 October 2004 
The regulations in 10 CFR Part 63 [DIRS 156605] also provide direct inputs to the FEP 
screening process by specifying characteristics, concepts, and definitions for screening FEPs. 
The regulatory definitions and clarifications of concepts pertaining to the reference biosphere, 
geologic setting, reasonably maximally exposed individual (RMEI), and human intrusion are 
explained in detail in Section 4.1.3 of The Development of the TSPA-LA Features, Events, and 
Processes (BSC 2004 [DIRS 168706]). The regulations require that the standards shown in 
Table 4-2 are to be used in addressing these concepts. 
The definitions and concepts discussed in The Development of the Total System Performance 
Assessment-License Application Features, Events, and Processes (BSC 2004 [DIRS 168706], 
Section 4.1.3.3) indicate that the RMEI is located no closer than 18 km to the south in the 
direction of groundwater flow and over a contaminated groundwater plume (in accordance with 
10 CFR 63.312 [DIRS 156605]), and that the limit of the controlled area is no greater than 5 km 
from the repository in any direction other than to the south (as specified in 10 CFR 63.302 
[DIRS 156605]). The location of the RMEI and the associated distance from the repository is of 
primary interest in evaluating potential exposure risk due to potential releases at the repository. 
The location of the RMEI is also important for determining exposure and is part of the technical 
basis for included FEPs. 
As discussed in The Development of the Total System Performance Assessment-License 
Application Features, Events, and Processes (BSC 2004 [DIRS 168706], Sections 4.1.3.1), the 
reference biosphere must be consistent with present knowledge of conditions in the region. 
Changes in the biosphere (other than climate) from conditions at the time of license application 
should not be projected. 
As discussed in The Development of the Total System Performance Assessment-License 
Application Features, Events, and Processes (BSC 2004 [DIRS 168706], Sections 4.1.3.2), the 
geologic setting (geology, hydrology, and climate) may evolve based upon cautious, but 
reasonable assumptions, consistent with present knowledge of factors that could affect the 
system in the 10,000 years following permanent repository closure. 
By NRCs combining of the geologic and hydrologic factors within the section addressing 
required characteristics of the reference biosphere, it is inferred that the listed regulatory 
constraint of changes in the reference biosphere are also applicable to conditions at Yucca 
Mountain. This approach agrees with the statement from 10 CFR 63.102(i) [DIRS 156605] that: 
Characteristics of the reference biosphere and the reasonably maximally exposed 
individual are to be based on current human behavior and biospheric conditions in 
the region, as described in 63.305 and 63.312. 
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]). 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 4-5 October 2004 
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 (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 the 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 Yucca Mountain Review 
Plan, Final Report (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 the regulatory requirements and other project requirements. The regulatory 
requirements include criteria relevant to performance assessment activities, in general, and to 
FEP-related activities as they pertain to TSPA-LA, in particular. In Table 4-2, YMRP 
acceptance criteria are correlated to the corresponding regulations as they pertain to FEPs-related 
criteria. 
The basis of the NRC review of Scenario Analysis and Event Probability is described in Yucca 
Mountain Review Plan, Final Report (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 model report are based on meeting the requirements of 10 CFR 63.114 (a), (b), (d), (e), and 
(f) [DIRS 156605] from Section 2.2.1.2.1.3 of 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 YMRP (NRC 2003 [DIRS 163274]) 
pertaining to identification of events with a probability greater than 10-8 per year are not 
considered because this analysis does not develop probabilities for such events. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 4-6 October 2004 
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 
NOTE: FEPs = features, events, and processes, PRD = Project Requirements Document. 
4.2.2 FEPs Screening Criteria 
The NRC regulations and guidance specifically allow the exclusion of FEPs from the TSPA-LA 
if they can be shown to be of low probability or of low consequence. Additionally, FEPs 
can be excluded based on the constraints provided within 10 CFR Part 63 [DIRS 156605]. In 
this 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. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 4-7 October 2004 
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 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 4-8 October 2004 
concentrations, flow rates, transport times, or other measures as well as overall expected annual 
dose), there is no single quantitative test of significance. 
4.2.2.3 Exclusion by Regulation 
The provisions and constraints provided within 10 CFR Part 63 [DIRS 156605] pertaining to the 
reference biosphere, receptor, and performance assessment serve as the basis for exclusion of 
some FEPs. This process of screening out the FEPs that fall outside the parameters established 
by 10 CFR Part 63 [DIRS 156605] is described in Yucca Mountain Review Plan, Final Report 
(NRC 2003 [DIRS 163274]), Section 2.2.1.2.1.3 Acceptance Criterion 2) together with the 
screening criteria of low probability and low consequence: 
An acceptable justification for excluding features, events, and processes is that 
either the feature, event, and process is specifically excluded by regulation; 
probability of the feature, event, and process (generally an event) falls below the 
regulatory criterion; or omission of the feature, event, and process does not 
significantly change the magnitude and time of the resulting radiological 
exposures to the reasonably maximally exposed individual, or radionuclide 
releases to the accessible environment. 
Exclusion of FEPs by regulation involves consideration of those portions of 10 CFR Part 63 
[DIRS 156605] that define requirements and key concepts for performance assessment. In this 
context, portions of 10 CFR Part 63 [DIRS 156605] serve as criteria for screening related FEPs. 
4.3 CODES, STANDARDS, AND REGULATIONS 
This document was prepared to comply with NRC regulatory requirements presented in 
10 CFR Part 63 [DIRS 156605]. Subparts of this rule that are applicable to date include Subpart 
B, Section 15 (Site Characterization), Subpart E, Section 114 (Requirements for Performance 
Assessment), Subpart F (Performance Confirmation Program) and Subpart G (Quality 
Assurance). 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 5-1 October 2004 
5. ASSUMPTIONS 
This section addresses assumptions used in the FEP screening for the drip shield and waste 
package. 
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]), which 
indicates that assuming the behavior of the earth is generally Poissonian or random is the 
underlying assumption in all probabilistic hazard analyses. 
Although there may be cases where sufficient data and information exist to depart from this 
assumption, the Poissonian model is generally a reasonable 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 
period (a relatively short time 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 in Sections 6.2.4.8 (FEP 1.5.03.01.0A) and 
6.2.4.10 (FEP 1.5.01.01.0A). 
5.2 NATURALLY OCCURRING EVENTS 
Assumption: It is assumed that potential naturally occurring events, but perhaps of different 
magnitudes, 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 repository over the 10,000-year 
regulatory period. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 5-2 October 2004 
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 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. Because it is consistent with the regulations, no further confirmation 
is necessary. 
Use in the Analysis: This assumption is used throughout the document. It is particularly 
germane to FEPs related to processes or phenomena that, speculatively, could affect future states 
of the repository system, but for which the magnitude or coupling, or both, effect on the 
repository is not well defined, or for which consequences are presently known to be minor. 
These types of events are known to occur. However, the effects of the phenomenon or the 
effects associated with varying magnitudes of the event type and probabilities are not well 
documented (e.g., effects of a supernova); the form of the coupling process is not well defined 
(e.g., changes in the earth's magnetic field); or the phenomenon has been shown to have 
insignificant or no impact presently (e.g., earth tides). 
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. 
Deviations from design will be detected 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 the License 
Application and in accordance with NRC license requirements. This is inherent in performance 
evaluation of any engineering project, and design verification and performance confirmation are 
required as part of the construction and operation processes. Therefore, no further confirmation 
of the assumption is required 
Engineering and design changes are subject to evaluation to determine if there are any adverse 
manner impacts to nuclear safety as required in 10 CFR 63.73 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 requirements indicate that following the issuance of any license to construct, the DOE 
must submit periodic and special reports to the NRC regarding: 
1. Progress of construction 
2. Any data about the site, obtained during construction, not within the predicted limits 
on which the facility design was based 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 5-3 October 2004 
3. Any deficiencies in design and construction that, if uncorrected, could adversely affect 
safety at any future time 
4. Results of research and development programs being conducted to resolve safety 
questions. 
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. This assumption is used throughout 
this report. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 5-4 October 2004 
INTENTIONALLY LEFT BLANK 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-1 October 2004 
6. SCIENTIFIC ANALYSIS DISCUSSION 
The primary concern in this report is to address and document the screening decisions for 
the 33 waste package and drip shield degradation FEPs listed in Table 1-1. In some cases, where 
a FEP covers multiple technical areas shared with other FEP reports, this document provides 
only partial technical basis for the screening. The shared FEPs and the appropriate area of 
responsibility are identified in Table 1-1. The full technical basis for shared FEPs is addressed 
collectively by the sharing FEP reports. 
6.1 METHODS AND APPROACH 
The identification and screening of a comprehensive list of FEPs potentially relevant to the 
repository post closure performance is 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 a version of the LA FEP list 
(DTN: MO0407SEPFEPLA.000 [DIRS 170760]), used as initial input in this waste package and 
drip shield degradation 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 6-1). 
6.1.2 Feature, Event, and Processes Screening 
The second step of FEP analysis is screening, which addresses Acceptance Criterion 2 of Yucca 
Mountain Review Plan, Final Report (NRC 2003 [DIRS 163274], Section 2.2.1.2.1.3). The 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-2 October 2004 
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 4.2.2), 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 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 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 by 
regulation. 
A FEP need only satisfy one of the exclusion screening criteria to be excluded from analysis in 
TSPA-LA. A FEP that does not satisfy any of the exclusion screening criteria must be included 
(screened in) in the TSPA-LA model. For many of the FEPs addressed in this report, it was 
determined that the probability of a consequential event or condition occurring during the 
10,000-year period following permanent closure is extremely low. However, considering the 
limited availability of data, analyses of the potential impacts of such low probability events or 
conditions was performed in lieu of a quantitative probabilistic evaluation. These analyses 
resulted in a determination that even in the unlikely event that on a probability weighted basis the 
event or condition should occur, it would not have a significant consequential effect on the 
magnitude and time of the resulting radiological exposures to the reasonably maximally exposed 
individual (RMEI), or radionuclide release to the environment. As such, these FEPs have been 
excluded due to low consequences as permitted 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. 
This report documents the screening decisions for the drip shield and waste package FEPs. In 
cases where a FEP covers multiple technical areas and is shared with other FEP reports, this 
document provides only a partial technical basis for the screening decision as it relates to drip 
shield and waste package issues. 
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. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-3 October 2004 
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 two criteria (e.g., low probability 
and low consequence), but the final decision can never be listed as both included and 
excluded. 
Screening Argument: This field is used only for excluded FEPs. It provides the discussion for 
why a FEP has been excluded from analysis in TSPA-LA. For included FEPs, it is indicated as 
not applicable N/A. 
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 analysis, making reference to more 
detailed documentation in other supporting technical reports, as applicable. For excluded FEPs, 
it is indicated as not applicable N/A. 
Supporting Reports: This field is only used for included FEPs. It provides the list of 
supporting technical reports prepared by the Project (i.e., not all references are part of the 
supporting reports) that identify the FEP as an included FEP and contain information relevant to 
the implementation of the FEP within the TSPA-LA. This list of supporting technical reports 
provides traceability of the FEP through project documents. 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 input is 
used to provide additional information that is not used in the development of results or 
conclusions. Indirect inputs are listed in Table 6-1. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-4 October 2004 
Table 6-1. Indirect Inputs 
Technical Product Input 
Specifically Used 
From Specifically Used In Input Description 
10 CFR 63 [DIRS 156605] Entire Sections 1, 1.3, 4, 4.1, 4.2.1, 4.2.2, 4.3, 5.2, 
5.3, 6.1.2, 6.1.6, 7.1, Tables 4-2, 6-1 
NRC rule for Yucca Mountain Repository 
pp. 80-82 Table 6-1, Sections 6.2.2, 6.2.4 Evidence for existence of a relative humidity threshold. ASM International 1987 
[DIRS 103753] pp. 971-974 Table 6-1, Section 6.2.33 Radiation damage discussion. Used as input to screening 
argument. 
ASME 1995 
[DIRS 141257] 
Entire Table 6-1, Section 6.2.13 Allowable tensile strength 
Asphahani 1978 
[DIRS 160352] 
Entire Table 6-1, Section 6.2.8 Information on HIC of nickel alloys 
Beavers et al. 2002 
[DIRS 158781] 
Section 3.10 Table 6-1, Section 6.2.33 Possibility of radiation damage 
Brossia 2001 [DIRS 
159836] 
Section 4.1.3 Table 6-1, Section 6.2.11 Corrosion of titanium alloys 
BSC 2001 [DIRS 154441] Section 6.3.2 Table 6-1, Section 6.2.17 Copper in gantry rail system 
BSC 2001 [DIRS 152248] Section 6.3.3 Table 6-1, Section 6.2.19 Crack sizes 
BSC 2001 [DIRS 154004] Entire Table 6-1, Section 6.2.30 Effects of thermal expansion on WP 
BSC 2001 [DIRS 152655] Section 1, 5.3, 
Entire 
Table 6-1, Section 6.2.30 Thermal expansion effects on WPOB 
BSC 2003 [DIRS 161519] Entire Table 6-1, Section 6.2.30 Interlocking drip shield design 
BSC 2004 [DIRS 171924] Entire Table 6-1, Sections 6.2.1, 6.2.28 Aging and phase stability of Alloy 22 
BSC 2004 [DIRS 168030] Entire Section 5.1 Framework for seismicity and structural deformation at Yucca 
Mountain 
BSC 2004 [DIRS 170019] Entire Table 6-1, Section 6.2.26 FEPs shared with clad degradation 
BSC 2004 [DIRS 169766] Entire Table 6-1, Section 6.2.22 General reference 
Entire Sections 1, 6.2.1, 6.2.9, 6.2.26, 7, Table 6-1 LA Design BSC 2004 [DIRS 168489] 
Tables 5 and 6 Table 6-1, Sections 6.2.17, 6.2.27 Design information for Alloy 22 feet and pallet 
BSC 2004 [DIRS 169480] Entire Table 6-1, Sections 6.2.26, 6.2.30 Typical waste package components assembly 
BSC 2004 [DIRS 169472] Entire Table 6-1, Sections 6.2.12, 6.2.30 Typical waste package components assembly 
BSC 2004 [DIRS 166879] Section 7.2.2 Table 6-1, Sections 6.2.8, 6.2.17, 6.2.18, 
6.2.22, 6.2.27 
Function of emplacement pallet 
BSC 2004 [DIRS 169898] Entire Table 6-1, Sections 6.2.1, 6.2.4, 6.2.5, 6.2.13, 
6.2.14, 6.2.15, 6.2.16, 6.2.20, 6.2.23, 6.2.25, 
6.2.32 
FEPs shared with EBS 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-5 October 2004 
Table 6-1. Indirect Inputs (Continued) 
Technical Product Input 
Specifically Used 
From Specifically Used In Input Description 
BSC 2004 [DIRS 170017] Entire Section 1.2 General reference 
BSC 2004 [DIRS 169845] Entire Table 6-1, Sections 6.2.3, 6.2.7, 6.2.9, 6.2.20, 
6.2.22, 6.2.29, 6.1.31, 6.2.32 
Corrosion of drip shield 
Section 6.4.6 Table 6-1, Sections 6.2.1, 6.2.2, 6.2.28 Effect of aging and phase instability 
Entire Table 6-1, Sections 6.2.2, 6.2.6, 6.2.13, 
6.2.20, 6.2.22, 6.2.28 
Corrosion of waste package outer barrier 
Sections 6.4.3, 
6.4.5, Table 6-1, 
Entire 
Table 6-1, Sections 6.2.10, 6.2.11 MIC of waste package outer barrier 
BSC 2004 [DIRS 169984] 
Section 6 Table 6-1, Section 6.2.31 Corrosion of WPOB 
Entire Table 6-1, Section 6.2.17 HIC of drip shield 
Sections 6.1.2- 
6.1.5, 6.3.2 
Table 6-1, Section 6.2.31 HIC of drip shield 
BSC 2004 [DIRS 169847] 
Sections 6.1.3, 6.1.6 Table 6-1, Section 6.2.17 HIC of drip shield 
BSC 2004 [DIRS 169565] Entire Table 6-1, Section 6.2.13 Temperature table 
BSC 2004 [DIRS 168556] Entire Table 6-1, Section 6.2.1 FEPs that are shared or referenced from criticality 
Sections 6.3, 6.3.4 Table 6-1, Section 6.2.4 Stress corrosion cracking of the drip, slip dissolution-film rupture 
model shield, the waste package outer barrier 
Sections 6.2.1, 6.4.2 Table 6-1, Sections 6.2.4, 6.2.22 Stress corrosion cracking of the drip shield, the waste package 
outer barrier 
Section 8.3 Table 6-1, Section 6.2.22 Contact stresses 
BSC 2004 [DIRS 169985] 
Entire Table 6-1, Section 6.2.15 Manufacturing defects (weld flaws) on waste package 
BSC 2004 [DIRS 170265] Entire Table 6-1, Sections 6.2.13, 6.2.14, 6.2.23, 
6.2.25 
Subsurface facility description document 
Entire Sections 1.1, 2, Table 6-1 Technical work plan for regulatory integration modeling and 
analysis of the waste form and waste package 
BSC 2004 [DIRS 171583] 
Table 3-1 Section 4.2 Yucca Mountain Review Plan acceptance criteria 
BSC 2004 [DIRS 168706] Entire Sections 1.2, 4.1, 6.1, 6.1.1, 6.1.2, 6.1.6, 
Table 6-1 
The development of the total system performance assessment 
license application features, events, and processes 
Entire Table 6-1, Sections 6.2.2, 6.2.3, 6.2.4, 6.2.10, 
6.2.15, 6.2.20 
General reference to AMR BSC 2004 [DIRS 169996] 
Section 6.5 Table 6-1, Section 6.2.2 General reference to number of WPs and patches 
BSC 2004 [DIRS 170020] Entire Table 6-1, Sections 6.2.32, 6.2.33 FEPs shared with waste form 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-6 October 2004 
Table 6-1. Indirect Inputs (Continued) 
Technical Product Input 
Specifically Used 
From Specifically Used In Input Description 
PRD-013/T-036, 
PRD-013/T-045 
Table 6-1, Section 6.2.12 Project Requirements Document Canori and Leitner 2003 
[DIRS 166275] 
Entire Table 4-2, Sections 4.2, 4.2.1 Project Requirements Document 
Cowan and Weintritt 1976 
[DIRS 105212] 
pp. 1 to 39 and 376 
to 383 
Table 6-1, Section 6.2.19 Formation of deposits in seawater 
CRWMS M&O 2000 
[DIRS 150823] 
Table 5 Table 6-1, Section 6.2.12 Design analysis for the defense high-level waste disposal 
container 
CRWMS M&O 2001 
[DIRS 152016] 
Entire Table 6-1, Section 6.2.19 Water distribution and removal characteristics 
Dunn and Brossia 2002 
[DIRS 162213] 
Entire Table 6-1, Section 6.2.6 Discussion of localized corrosion 
Gdowski 1991 
[DIRS 100859] 
Sections 5, 5.1 Table 6-1, Section 6.2.8 Degradation modes for Ni-Cr-Mo alloys 
Gdowski 1997 
[DIRS 102789] 
p.1-8 Table 6-1, Section 6.2.29 Pd concentration in Titanium Grade 7 
Hua and Gordon 2003 
[DIRS 163111] 
Entire Table 6-1, Section 6.2.29 Corrosion of titanium in basic saturated water 
Hua et al. 2002 
[DIRS 160670] 
Entire Table 6-1, Section 6.2.29 Corrosion testing of container materials 
Kohli and Pasupathi 1986 
[DIRS 131519] 
Entire Table 6-1, Section 6.2.12 Investigation of water-logged spent fuel rods under dry storage 
conditions 
Little and Wagner 1996 
[DIRS 131533] 
Entire Table 6-1, Section 6.2.11 MIC of titanium alloy 
Lorenzo de Mele and 
Cortizo 2000 
[DIRS 159833] 
Entire Table 6-1, Section 6.2.3 Fluoride effects on titanium 
Merriam-Webster 1993 
[DIRS 100468] 
Entire Table 4-1, Section 5.1 Definitions 
NRC 2003 [DIRS 163274] Entire Tables 4-2, 6-1, Sections 4.2, 4.2.1, 4.2.2, 6.1, 
6.1.1, 6.1.2, 7.1.1 
Yucca Mountain Review Plan 
Pan et al. 2002 
[DIRS 165536] 
Entire Table 6-1, Section 6.2.4 Stress corrosion cracking and hydrogen embrittlement of container 
and drip shield materials 
Section 1 Table 6-1, Section 6.2.20 Waste package design Plinski 2001 
[DIRS 156800] Section 8.1.7 Table 6-1, Sections 6.2.4, 6.2.28 Annealing of outer cylinder 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-7 October 2004 
Table 6-1. Indirect Inputs (Continued) 
Technical Product Input 
Specifically Used 
From Specifically Used In Input Description 
Section 8.3.17 Table 6-1, Section 6.2.16 Drip shield assembly 
Section 8.1.8 Table 6-1, Sections 6.2.28, 6.2.30 Waste package design 
Pulvirenti et al. 2002 
[DIRS 165537] 
Entire Table 6-1, Section 6.2.4 Effects of lead, mercury, and reduced sulfur species on the 
corrosion of Alloy 22 in concentrated groundwaters as a function 
of pH and temperature 
Sakai et al. 1992 
[DIRS 154465] 
Entire Table 6-1, Section 6.2.4 Effect of lead water chemistry on oxide thin film of Alloy 600 
Shoesmith and Ikeda 1997 
[DIRS 151179] 
Sections 3, 6, Entire Table 6-1, Section 6.2.11 MIC of titanium alloys 
Shoesmith and King 1998 
[DIRS 112178] 
pp. 5, 29 to 30 Table 6-1, Section 6.2.32 Effects of gamma irradiation on waste package materials 
Shoesmith et al. 1995 
[DIRS 117892] 
Entire Table 6-1, Section 6.2.11 Initiation of biofilms under crevice corrosion 
Thomas 1994 
[DIRS 120498] 
Entire, p. 17:54 Table 6-1, Section 6.2.6 Discussion of localized corrosion 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-8 October 2004 
6.1.4 Qualification of Unqualified Direct Inputs 
N/A. No unqualified direct inputs are used. 
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 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 
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 nuclear safety as required in 
10 CFR 63.73 Subparts F and G [DIRS 156605]. See also the requirements in 10 CFR 63.44 and 
10 CFR 63.131 [DIRS 156605]. 
6.2 WASTE PACKAGE AND DRIP SHIELD FEATURE, EVENT, AND PROCESSES 
SCREENING AND ANALYSES 
Screening information for each of the 33 waste package and drip shield FEPs is presented in 
separate subsections. The FEPs are addressed in numeric order. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-9 October 2004 
6.2.1 Error in Waste Emplacement (1.1.03.01.0A) 
FEP Description: 
Deviations from the design and/or errors in waste emplacement could affect long-term 
performance of the repository. A specific example of such an error would be erroneously 
emplacing the waste packages in a saturated or wet zone of the repository. Errors of this 
type would impact repository performance by affecting waste package corrosion and 
radionuclide transport. 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
The features, events, or processes that could result in an error in waste emplacement that could 
lead to unanticipated operating conditions stem from administrative controls. The results of 
these errors could affect the waste package surface temperature and humidity history (thus 
impacting corrosion rates), result in placement in prohibited areas, or allow water to contact the 
waste package at times earlier than expected. 
Accidental misloading of waste packages with fuel assemblies having a lower burnup than 
required by the appropriate loading curve could cause the waste package surface temperature and 
relative humidity to be outside the expected ranges and impact the degradation characteristics of 
the waste package and drip shield. Analysis of Mechanisms for Early Waste Package Failure 
(BSC 2004 [DIRS 170024], Section 6.2.8) states that the temperature at the surface of the waste 
package is mainly governed by the temperature in the drift where the waste package is located, 
rather than the heat output within the waste package (a metallic container with a rather large heat 
transfer area). Therefore, the increase in heat output generated by a thermally overloaded waste 
package (which can not be expected to exceed a few kilowatts) would be quickly dissipated into 
the drift and is not expected to alter the waste package surface temperature to an extent 
significant enough to affect post closure performance. 
Similarly, emplacement of waste packages closer than the design specifies could cause the waste 
package surface temperature and relative humidity to be outside the expected ranges and impact 
the degradation characteristics of the waste package and drip shield. The current emplacement 
drift design specifies that the average waste package skirt-to-skirt spacing is 0.1 m or 10 cm 
(BSC 2004 [DIRS 168489], Table 1) indicating that any longitudinal misplacements would be 
small (on the order of 10 cm). For each locally created hot zone (resulting from placing a waste 
package closer than intended to one of its neighbors) there necessarily would be a corresponding 
cooler zone (resulting from the waste package being further than intended from its other 
neighbor). It is reasonable to conclude that, on a larger scale, the minute variations in local 
temperature and relative humidity caused by longitudinal misplacement of waste packages have 
low consequence to waste package or drip shield degradation. This conclusion is consistent with 
the discussion above regarding accidental misloadings (i.e., the temperature at the surface of the 
waste package is mainly governed by the temperature in the drift where the waste package is 
located, rather than the heat output within the waste package). 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-10 October 2004 
In addition, no effects of aging and phase instability on the corrosion behavior of the Alloy 22 
waste package outer barrier would be observed even if the waste package were maintained at less 
than 300C for a period of 500 years followed by temperatures less than 200C for a period of 
9,500 years (BSC 2004 [DIRS 171924], Section 8; BSC 2004 [DIRS 169984], Section 6.4.6). 
These thermal exposure conditions bound all repository-relevant thermal exposure conditions 
(BSC 2004 [DIRS 171924], Section 8). 
Another possible consequence of accidental misloading of waste packages with fuel assemblies 
having a lower burnup than required by the appropriate loading curve is the increased potential 
for criticality. Treatment of accidental misloading of lower burnup fuel assemblies is discussed 
in Screening Analysis of Criticality Features, Events, and Processes for License Application 
(BSC 2004 [DIRS 168556]). 
The areas that would be restricted from waste package emplacement would be those where the 
waste package would straddle a fault. An error in placement then would be defined as a waste 
package emplaced so that it straddles a fault. The evaluation of the event is discussed in 
FEP 1.2.02.03.0A in Engineered Barrier System Features, Events, and Processes (BSC 2004 
[DIRS 169898]). 
The current engineered barrier design includes a titanium drip shield that is placed over the waste 
packages (BSC 2004 [DIRS 168489]). The drip shield is installed in segments and, once 
installed, continues along the entire length of the drift. Each segment will slightly overlap the 
previously emplaced segment. Human error during placement of the drip shield resulting in a 
gap between the drip shield segments, could lead to water contact or rockfalls on the waste 
package at a time earlier than would be expected under nominal in-drift conditions. The gap 
between two adjacent drip shields improperly interlocked is expected to be small. It is not 
credible that the inspection performed by remotely operated cameras would not detect an 
incorrect emplacement leaving a gap exceeding the length of the connecting plates of the drip 
shields (BSC 2004 [DIRS 170024], Section 6.4.7). Therefore, dripping water and rockfalls are 
not expected to fall directly onto the underlying waste package, but will hit the connecting plate 
and not the waste package surface (BSC 2004 [DIRS 170024], Section 6.4.7). 
Similarly, erroneous emplacement of waste packages in a saturated or wet zone of the repository 
will not cause significant damage to the waste package outer barrier because the drip shields 
must fail for water to contact the underlying waste package. As discussed in FEP 2.1.03.02.0B, 
Stress Corrosion Cracking (SCC) of Drip Shields, SCC of drip shields does not compromise the 
water diversion function of the drip shield (i.e., SCC of the drip shield does not lead to water 
contacting the underlying waste package). As discussed in FEP 2.1.03.03.0B, Localized 
Corrosion of Drip Shields, localized corrosion of the drip shield will not occur under exposure 
conditions in the repository. Therefore, general corrosion is the only possible drip shield failure 
mode of any consequence to repository performance (note that seismic effects are not considered 
in this section). As shown in General Corrosion and Localized Corrosion of Drip Shield 
(BSC 2004 [DIRS 169845], Tables 16 and 17) the maximum general corrosion rate that can be 
applied to the underside of drip shield is about 1.13  10-4 mm/yr and the maximum general 
corrosion rate that can be applied to the top side of drip shield is about 3.20  10-4 mm/yr. 
Therefore, the earliest time at which the 15-mm-thick drip shield can fail as a result of general 
corrosion is approximately 35,000 years after final closure of the repository. With no water 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-11 October 2004 
expected to drip on the waste package, there is no possibility of any effect of erroneous 
emplacement of waste packages in a saturated or wet zone of the repository on waste package 
degradation. 
Based on the above discussion, this FEP is excluded based on low consequence to radiological 
exposures to the reasonable maximally exposed individual and radionuclide releases to the 
accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.2 General Corrosion of Waste Packages (2.1.03.01.0A) 
FEP Description: 
General corrosion may contribute to waste package failure. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
General corrosion (or passive corrosion) is the uniform thinning of the waste package outer 
barrier at its open-circuit corrosion potential. The presence of stable liquid water is required to 
initiate corrosion processes (including general corrosion) that are supported by electrochemical 
corrosion reactions. Typically, a threshold relative humidity is used to simulate such a corrosion 
initiation condition (e.g., ASM International 1987 [DIRS 103753], pp. 80-82). Surface 
cleanliness, corrosion product build-ups, and/or deliquescence of hygroscopic salts can cause 
water film formation on surfaces at lower relative humidities than would be required on smooth 
clean surfaces (ASM International 1987 [DIRS 103753], p. 80-82). Conservatively, in the waste 
package degradation analysis for TSPA-LA, no threshold relative humidity is used (BSC 2004 
[DIRS 169996], Section 5.1) (i.e., general corrosion can occur under any exposure conditions). 
Because general corrosion is likely to be operative for most of the repository operation period, it 
could lead to degradation and breach of waste packages in the repository. General corrosion due 
to dry-air oxidation and aqueous corrosion as well as the effects of microbially influenced 
corrosion and aging and phase instability on general corrosion of the waste package outer barrier 
are discussed in General Corrosion and Localized Corrosion of Waste Package Outer Barrier 
(BSC 2004 [DIRS 169984]). 
General Corrosion and Localized Corrosion of the Waste Package Outer Barrier (BSC 2004 
[DIRS 169984], Section 6.4.2) concluded that although dry air oxidation occurs, it results in a 
negligible amount of barrier thinning over repository time scales (only ~93 m even if the waste 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-12 October 2004 
package outer barrier were exposed for 10,000 years at 350C). Therefore, dry oxidation is not 
considered in TSPA-LA analyses. 
Penetration rates for general corrosion are provided in General Corrosion and Localized 
Corrosion of the Waste Package Outer Barrier (BSC 2004 [DIRS 169984] Section 6.4.3). 
General corrosion rates of the waste package outer barrier were estimated using the weight-loss 
of Alloy 22 crevice geometry specimens after a 5-year exposure in the LTCTF and the 
temperature dependence of corrosion rates measured using the polarization resistance technique 
(BSC 2004 [DIRS 169984], Section 6.4.3). The temperature dependence follows an Arrhenius 
relationship, as discussed in General Corrosion and Localized Corrosion of Waste Package 
Outer Barrier (BSC 2004 [DIRS 169984], Section 6.4.3.4). 
The patch size used to analyze general corrosion of the waste packages in WAPDEG Analysis of 
Waste Package and Drip Shield Degradation (BSC 2004 [DIRS 169996], Section 6.4.3) is four 
times the area of the crevice geometry specimen size used to evaluate the Alloy 22 general 
corrosion rates in General Corrosion and Localized Corrosion of the Waste Package Outer 
Barrier (BSC 2004 [DIRS 169984] Section 6.4.3). Therefore, the general corrosion rates are 
adjusted to account for the effects of this change of scale. Conceptually, the scaling method 
employed corresponds to using the highest of four sampled corrosion rates (from the crevice 
geometry specimen-based distribution) for the general corrosion rate applied to the waste 
package patch. The approach is conservative because it effectively uses the highest of the four 
corrosion rates sampled. The effect of this method is to shift the median general corrosion rate to 
higher values and to decrease the probability of sampling lower general corrosion rates. This 
general corrosion treatment applies to commercial spent nuclear fuel and codisposed waste 
packages. 
The effects of trace metals, including arsenic, calcium, and magnesium, are included in the 
general corrosion rate distributions based on samples exposed in the LTCTF. Samples of 
Alloy 22 were immersed in SAW for five years (DTN: LL021012712251.021 [DIRS 163112]). 
After five years, a sample of the acidified water, including any corrosion products, was analyzed 
for trace metals using inductively coupled plasma and mass spectroscopy methods. Low 
concentrations of potentially deleterious metals, including arsenic, calcium, and magnesium were 
detected. Given that the corrosion rates used for the Alloy 22 waste package outer barrier are 
based on samples exposed in these environments, the general corrosion analyses account for the 
presence of these trace metals both in the solutions and in the materials used for testing. On this 
basis, the effects of trace metals on general corrosion of the waste package outer barrier are 
included in TSPA-LA although the effects are not explicitly and separately modeled. 
The effect of microbial activity on the general corrosion process of the waste package outer 
barrier is represented in TSPA-LA analyses with a general corrosion rate-enhancement factor 
(BSC 2004 [DIRS 169984], Section 8.2). A more detailed discussion on the effect of microbial 
activity is provided in FEP 2.1.03.05.0A, Microbially Influenced Corrosion (MIC) of Waste 
Packages. 
Comparative analysis of the corrosion rates from the polarization resistance technique showed 
insignificant effects of welding and thermal aging of the waste package outer barrier on the 
general corrosion rates. It was determined that the aging of the base metal and welds of the 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-13 October 2004 
waste package outer barrier under the thermal conditions expected in the repository would not 
affect the corrosion performance of the waste package outer barrier and will not be specifically 
modeled in the TSPA-LA (BSC 2004 [DIRS 169984], Section 6.4.6). 
The general corrosion rate distributions along with the exposure condition parameters for the 
waste packages and drip shields are incorporated into the integrated waste package degradation 
analysis (BSC 2004 [DIRS 169996]). The output from the integrated waste package degradation 
analysis is a set of profiles (time histories) for the failure (i.e., initial breach) and subsequent 
number of penetration openings in the waste package and drip shield as a function of time. The 
integrated waste package degradation analysis is used directly in the TSPA-LA. 
The TSPA-LA waste package degradation analysis simulates the behavior of a few hundred 
waste packages (BSC 2004 [DIRS 169996], Section 6.5). Effects of spatial and temporal 
variations in the exposure conditions over the repository are modeled by explicitly incorporating 
relevant exposure condition histories into the analysis. The exposure condition parameters 
considered to vary over the repository are relative humidity and temperature at the waste package 
surface. Potentially variable corrosion processes for a single waste package are represented by 
dividing the waste package surface into subareas called patches and stochastically sampling 
the degradation parameter values for each patch. The use of patches explicitly represents the 
variability in degradation processes associated with a single waste package at a given time. 
In the TSPA-LA analysis, uncertainty in waste package degradation is analyzed with multiple 
realizations of the integrated waste package degradation analysis. For each realization, values 
are sampled for the uncertain degradation parameters and passed to the integrated waste package 
degradation analysis. Each realization is a complete integrated waste package degradation 
analysis simulation of a given number of waste packages, explicitly considering variability in the 
degradation processes. Accordingly, each of the integrated waste package degradation analysis 
outputs (i.e., the fraction of the total number of waste packages and drip shields failed versus 
time and of the average number of patch and crack penetrations per failed waste package (or drip 
shield)) are reported as a group of degradation profile curves (resulting from the multiple 
realizations) that represent the potential range of the output parameters. For example, the waste 
package breach time profiles are reported with a group of curves representing the cumulative 
probability of waste package breaches as a function of time. The outputs of the integrated waste 
package degradation analysis are used as input for waste form degradation analysis and 
radionuclide release analysis from failed waste packages conducted within the TSPA-LA model. 
The TSPA representation of general corrosion of waste package conservatively models the 
phenomenon studied and data generated in General Corrosion and Localized Corrosion of the 
Waste Package Outer Barrier (BSC 2004 [DIRS 169984]) and WAPDEG Analysis of Waste 
Package and Drip Shield Degradation (BSC 2004 [DIRS 169996]). 
Supporting Reports: 
BSC 2004 [DIRS 169984] 
BSC 2004 [DIRS 169996] 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-14 October 2004 
6.2.3 General Corrosion of Drip Shields (2.1.03.01.0B) 
FEP Description: 
General corrosion may contribute to drip shield failure. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
General corrosion is included in drip shield degradation analysis. This includes the effects of 
corrosive gases. General corrosion due to dry-air oxidation, humid-air and aqueous general 
corrosion, microbially influenced corrosion, and aging and phase instability of the Titanium 
Grade 7 drip shield are discussed in General Corrosion and Localized Corrosion of the Drip 
Shield (BSC 2004 [DIRS 169845]). 
It was concluded in General Corrosion and Localized Corrosion of the Drip Shield (BSC 2004 
[DIRS 169845], Section 8.2) that although dry air oxidation occurs, it results in a negligible 
amount of metal thinning over repository time scales (only ~1,320 nm even if the drip shield 
were exposed for 10,000 years at 200C). Therefore, dry oxidation does not need to be 
considered in the TSPA-LA analyses. 
Penetration rates for general corrosion are provided in General Corrosion and Localized 
Corrosion of the Drip Shield (BSC 2004 [DIRS 169845], Tables 16 and 17) and are used in 
TSPA-LA analyses. Humid-air and aqueous corrosion processes are considered part of general 
corrosion (BSC 2004 [DIRS 169845], Sections 6.1, 6.3.2, and 6.3.3). General corrosion rates of 
the drip shield were estimated using weight-loss data of Titanium Grade 16 samples after 
a 1-year exposure in the LTCTF and validated using 5-year Titanium Grade 16 and 2.5-year 
Titanium Grade 7 exposure test data (BSC 2004 [DIRS 169845], Section 7.4.2). Although 
analysis of this data shows that the corrosion rate of the titanium alloys decreases with time, a 
time-independent general corrosion rate (sampled from a distribution developed using the 1-year 
Titanium Grade 16 exposure data) is used in TSPA-LA (BSC 2004 [DIRS 169845], 
Sections 7.4.2 and 8.3). The maximum corrosion rates measured for the 2.5-year exposed 
Titanium Grade 7 specimens are lower than those for the 5-year exposed Titanium Grade 16 
specimens. This observation indicates that the general corrosion resistance of Titanium Grade 7 
is superior to that of Titanium Grade 16 and that the general corrosion rate distribution 
developed using the 1-year Titanium Grade 16 exposure data is a conservative representation of 
drip shield general corrosion rates. 
The drip shield outer surface may be exposed to a more complicated chemistry and geometry 
than the drip shield inner surface, since dust or mineral films (from evaporation of dripping 
water), or both, may form crevices on the drip shield outer surfaces. In contrast, the inner 
surfaces of the drip shield will not be exposed to dripping water or significant dust film 
formation (BSC 2004 [DIRS 169845], Sections 6.3, 6.5.2, and 6.5.3). Therefore, general 
corrosion of the inner and outer surfaces of the drip shield are modeled by using different sets of 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-15 October 2004 
corrosion data (BSC 2004 [169845], Sections 6.3, 6.5.2, and 6.5.3). Variations in the drip shield 
general corrosion rate distributions are considered to be entirely due to uncertainty (BSC 2004 
[DIRS 169845], Section 6.5.4.2). For each realization of the integrated waste package 
degradation analysis (BSC 2004 [DIRS 169996]), a single general corrosion rate is sampled from 
each general corrosion rate distribution (i.e., one general corrosion rate value is sampled from the 
general corrosion rate distribution applicable to the drip shield underside and another general 
corrosion rate value is sampled from the general corrosion rate distribution applicable to the drip 
shield outer surface). The two sampled values are then applied one each to the outer and inner 
surfaces of each drip shield simulated during the given realization. On each time step, general 
corrosion of the drip shield occurs. Using this conceptual model for drip shield general 
corrosion, all drip shields in the repository fail by general corrosion at the same time. The 
maximum general corrosion rate for the cumulative distribution function (CDF) applied to the 
under side of the drip shield is approximately 1.13  10-4 mm/yr and the maximum general 
corrosion rate for the CDF applied to the top side of the drip shield is 
approximately 3.2  10-4 mm/yr (BSC 2004 [DIRS 169845], Tables 16 and 17); therefore, drip 
shield failure is not anticipated to occur from general corrosion until at least 35,000 years after 
permanent closure. 
As discussed in General Corrosion and Localized Corrosion of the Drip Shield (BSC 2004 
[DIRS 169845], Section 6.5.7), the presence of dissolved fluoride can, under certain conditions, 
increase the general corrosion rate of titanium alloys including Titanium Grade 7. Direct 
incorporation of fluoride into the passive film leads to enhanced dissolution only under acidic 
conditions. A combination of a significant F- concentration (> 500 g/g) and a low pH (< 4.3) is 
necessary for loss of passivity (BSC 2004 [DIRS 169845], Section 6.5.7). Lorenzo de Mele and 
Cortizo (2000 [DIRS 159833]) conducted measurements in synthetic saliva (pH = 6.5, 
containing 0.2 mol/L of fluoride), which showed that if fluoride were added shortly after 
electrode immersion (while the oxide film was still growing and defective), the corrosion 
potential fell rapidly (indicating attack of the oxide). However, if the oxide were allowed to 
grow for four days (a sufficient period for the growth of a coherent oxide with a low defect 
concentration), then addition of the same amount of fluoride had no observable effect over the 
subsequent two days of exposure indicating that the susceptibility of titanium to corrosion in 
fluoride solutions is associated with defects and flaws in the oxide. 
In the case of the drip shield application in the repository, there is an early period of dry air 
exposure. During this time, the drip shield is subjected to a long period of thermal oxidation 
prior to aqueous exposure. Thermal treatments not only thicken the oxide film but also 
significantly decrease the defect density in the film compared to films grown in aqueous 
environments (BSC 2004 [DIRS 169845], Section 6.5.7). The increase in thickness and 
improvement in film properties would lead to a significant decrease in susceptibility to fluorideinduced 
film breakdown and an independence of the passive corrosion rate on fluoride 
concentration (BSC 2004 [DIRS 169845], Section 6.5.7). Therefore, fluoride will have a low 
consequence to the degradation rate of titanium in the repository. 
The effects of other trace metals, including arsenic, calcium, and magnesium are included in the 
general corrosion rate distributions based on samples exposed in the LTCTF. Samples of 
Titanium Grade 16, an excellent analogue for the more corrosion-resistant Titanium Grade 7, 
were immersed in simulated acidified water (DTN: LL021012712251.021 [DIRS 163112]) in 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-16 October 2004 
the LTCTF. Samples of the acidified water, including any corrosion products, were analyzed for 
trace metals using inductively coupled plasmamass spectrometry methods. Low concentrations 
of potentially deleterious metals, including arsenic, calcium, and magnesium, were detected in 
the sampled waters. As the corrosion rates used for the Titanium Grade 7 drip shield are based 
on samples exposed in these environments, the general corrosion analyses account for the 
presence of these trace metals (BSC 2004 [DIRS 169845], Section 6.5.1). On this basis, the 
effects of trace metals on general corrosion of the drip shield are included in TSPA-LA although 
the effects are not explicitly and separately modeled. 
The general corrosion rate distributions for the drip shields are incorporated into the integrated 
waste package degradation analysis (BSC 2004 [DIRS 169996]). The output from the integrated 
waste package degradation analysis is a set of profiles (time histories) for the failure (i.e., initial 
breach) and subsequent number of penetration openings in the waste package and drip shield as a 
function of time. The integrated waste package degradation analysis is used directly in the 
TSPA-LA. 
Supporting Reports: 
BSC 2004 [DIRS 169845] 
BSC 2004 [DIRS 169996] 
6.2.4 Stress Corrosion Cracking (SCC) of Waste Packages (2.1.03.02.0A) 
FEP Description: 
Waste packages may become wet at specific locations that are stressed leading to stress 
corrosion cracking (SCC). The possibility of SCC under dry conditions or due to thermal 
stresses are also addressed as part of this FEP. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
Stress corrosion cracking of the waste package outer barrier closure weld regions is included in 
TSPA-LA as part of waste package degradation analyses. Stress Corrosion Cracking of the Drip 
Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material (BSC 2004 
[DIRS 169985], Section 8.3, Tables 8.1 to 8.3) provides input to the TSPA-LA treatment of 
waste package degradation by stress corrosion cracking. 
All regions of the waste package (including the fabrication welds), except the waste package 
closure lid welds, are stress relief annealed before the waste packages are loaded with waste 
(Plinski 2001 [DIRS 156800], Section 8.1.7), and thus do not develop residual stress/stress 
intensity factors high enough for SCC to occur (BSC 2004 [DIRS 169985], Section 6.4.2). 
Welds are the most susceptible to SCC because (1) welding can produce high tensile residual 
stress in the weld; (2) preexisting flaws due to fabrication and welding have much higher 
concentration in the weld than in the base metal; and (3) welding could result in segregation and 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-17 October 2004 
nonequilibrium brittle phases, which could enhance material susceptibility to SCC (BSC 2004 
[DIRS 169985], Section 6.4.2). Plastic deformation resulting from seismic events also has the 
potential of leading to plastic upsets and resultant sustained residual stresses that may initiate 
cracks and drive them through the wall. Seismic effects are discussed 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 in Engineered Barrier System Features, Events, and Processes (BSC 2004 
[DIRS 169898]). 
The presence of stable liquid water is required to initiate corrosion processes (including SCC) 
that are supported by electrochemical corrosion reactions. Typically, a threshold relative 
humidity is used to simulate such a corrosion initiation condition (e.g., ASM International 1987 
[DIRS 103753], pp. 80 to 82). Conservatively, in the waste package degradation analysis for 
TSPA-LA, no threshold relative humidity is used (BSC 2004 [DIRS 169996], Section 5.1) 
(i.e., SCC is allowed to occur, if all stress and stress intensity criteria are satisfied, under any 
exposure conditions). 
As discussed in Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, 
and the Stainless Steel Structural Material (BSC 2004 [DIRS 169985], Section 6.3), the slip 
dissolutionfilm rupture model was used to assess the breach (or lack of it) of the waste package 
by SCC crack propagation due to the combined effects of stress and environment, or both. 
The SCC model for the waste package closure weld regions uses weld residual stress profiles and 
stress intensity factor profiles (BSC 2004 [DIRS 169985], Section 6). These input data were 
developed for the 25-mm outer lid closure weld regions (subjected to laser peening) and the 
as-welded 10-mm middle lid closure weld regions. The report (BSC 2004 [DIRS 169985], 
Section 6) also provides other input for the degradation of the waste package due to SCC 
including: a threshold stress for crack initiation; a threshold stress intensity factor for 
propagation; weld flaw size, density and orientation distributions; and an estimate of crackopening 
size. This SCC treatment applies to all waste packages. 
Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the 
Stainless Steel Structural Material (BSC 2004 [DIRS 169985], Section 6.2.1) describes the 
results of SCC crack-initiation measurements under constant load conditions while immersed in 
a basic saturated water solution. Alloy 22 (UNS N06022) exhibits excellent SCC resistance, as 
cracking was not observed for any of the 120 Alloy 22 (UNS N06022) specimens covering a 
variety of metallurgical conditions (including the as-welded condition). The applied stress ratios 
used in the experiments were up to about 2.1 times the yield strength of the as-received material 
and up to 2.0 times the yield strength of the welded material. This stress ratio corresponds to an 
applied stress of about 89 to 96 percent of the ultimate tensile strength. The high degree of 
SCC-initiation resistance for Alloy 22 (UNS N06022) is corroborated by results of high 
magnification visual examination of a number of Alloy 22 (UNS N06022) U-bend specimens 
exposed to a range of relevant environments at 60C and 90C in the LTCTF (BSC 2004 
[DIRS 169985] Section 6.2.1). No evidence of SCC initiation has been observed in these 
U-bend specimens after five years of exposure. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-18 October 2004 
The discussion in FEP 2.1.11.07.0A, Thermal Expansion/Stress of In-Drift EBS Components, 
indicates that thermal expansion is not a source of stress and, therefore, not a driving force for 
SCC in the repository. 
The SCC model is incorporated into the integrated waste package degradation analysis 
(BSC 2004 [DIRS 169996], Section 6.4). The output from the integrated waste package 
degradation analysis is a set of profiles (time histories) for the failure (i.e., initial breach) and 
subsequent number of penetration openings in the waste package and drip shield as a function of 
time. The integrated waste package degradation analysis is used directly in the TSPA-LA. The 
TSPA representation of stress corrosion cracking of waste package conservatively models the 
phenomenon studied and data generated in Stress Corrosion Cracking of the Drip Shield, the 
Waste Package Outer Barrier, and the Stainless Steel Structural Material (BSC 2004 
[DIRS 169985]) and WAPDEG Analysis of Waste Package and Drip Shield Degradation 
(BSC 2004 [DIRS 169996]). 
Lead (Pb) has been identified as a potential contributor to stress corrosion cracking in nickelbased 
alloys (Sakai et al. 1992 [DIRS 154465]; Pan et al. 2002 [DIRS 165536]; 
Pulvirenti et al. 2002 [DIRS 165537]). The results achieved by Pulvirenti et al. (2002 
[DIRS 165537]) have not been able to be reproduced by the same investigators or investigators 
from the Center for Nuclear Waste Regulatory Analyses (Pan et al. 2002 [DIRS 165536]). The 
effects of lead on SCC processes were investigated by the Project by evaluating the effect of 
significant additions of lead nitrate (PbNO3) to test solutions. Slow strain-rate stress corrosion 
cracking experiments were performed on specimens of Alloy 22 (UNS N06022) in 
lead-containing solutions, as discussed in Stress Corrosion Cracking of the Drip Shield, the 
Waste Package Outer Barrier, and the Stainless Steel Structural Material (BSC 2004 
[DIRS 169985], Section 6.3.4). SCC initiation test results were obtained using Slow Strain Rate 
Tests, at 76C to 95C in low pH brine solutions (pH = ~3) with and without 0.005% lead nitrate 
additions. These results also show no effect of Pb on SCC susceptibility. Thus, there appears to 
be no basis for concern that Pb will affect SCC susceptibility in these relevant concentrated brine 
environments over a broad range of pH values. Other experiments showed that even if SCC is 
forced to occur (by slow cyclic straining) before lead is added to the test solution, SCC 
crack-growth rates are not accelerated by subsequent lead additions 
(DTN: LL02110531105312251.023 [DIRS 161253]). This latter experiment used a basic 
saturated water solution with a pH ~ 12 at 110C. After 8,670 hours of exposure in the lead-free 
solution, 1,000 ppm Pb (as PbNO3) was added to the test solution and SCC crack growth rates 
were monitored for ~1,800 hours using extremely sensitive techniques (in situ reversing 
DC-potential drop technique). The presence of the lead in the test solution was concluded to 
have no measurable effect on the SCC growth rate (DTN: LL021105312251.023 [161253]). On 
this basis, any lead present in repository groundwaters is expected to have little consequence on 
Alloy 22 (UNS N06022) SCC processes. 
Supporting Reports: 
BSC 2004 [169985] 
BSC 2004 [169996] 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-19 October 2004 
6.2.5 Stress Corrosion Cracking (SCC) of Drip Shields (2.1.03.02.0B) 
FEP Description: 
Drip shields may become wet at specific locations that are stressed leading to stress 
corrosion cracking (SCC). The possibility of SCC under dry conditions or due to thermal 
stresses is also addressed as part of this FEP. 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
The sources of stress that could result in stress corrosion cracking in the Titanium Grade 7 drip 
shield are (1) weld-induced residual stress, (2) plasticity-induced residual stress caused by 
seismic events, and (3) residual stress produced by rock falls (BSC 2004 [DIRS 169985], 
Section 6.3.7). 
The weld induced residual stress in the drip shields will be mitigated by annealing before 
placement in the drifts. Therefore, drip shields are not subject to SCC upon emplacement 
(BSC 2004 [DIRS 169985], Section 6.3.7). However, the drip shields are subject to SCC under 
the action of seismic-induced loading and rockfall. Seismic effects on drip shield degradation 
are discussed in FEPs 2.1.07.01.0A, Rockfall, and 1.2.03.02.0A, Seismic Ground Motion 
Damages EBS Components, in Engineered Barrier System Features, Events, and Processes 
(BSC 2004 [DIRS 169898]). 
The consequence of stress corrosion cracking on drip shield performance is discussed in 
FEP 2.1.03.10.0B, Advection of Liquids and Solids Through Cracks in the Drip Shield. 
FEP 2.1.03.10.0B is excluded on the basis of low consequence because the amount of advective 
water flow through cracks in the drip shield will be severely limited based on the small aperture 
width (narrow opening and tight cracks), the presence of capillary forces, and the potential for 
plugging of the cracks due to mineral deposits. Thus, the effective water flow rate through 
cracks in the drip shield will be extremely low and will not contribute significantly to the overall 
radionuclide release rate from the repository. Therefore, since the primary role of the drip shield 
is to keep water from contacting the waste package, SCC of the drip shield does not compromise 
the design purpose of the drip shield. 
Based on the above rationale, this FEP is excluded for the drip shield due to low consequence to 
radiological exposures to the RMEI and radionuclide releases to the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-20 October 2004 
6.2.6 Localized Corrosion of Waste Packages (2.1.03.03.0A) 
FEP Description: 
Localized corrosion (pitting or crevice corrosion) could enhance degradation of the waste 
packages. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
Localized corrosion is a corrosion attack at discrete sites or in a nonuniform manner. Based on a 
conservative linear growth model the rate of localized corrosion penetration is generally higher 
than the rate of general corrosion penetration, and could lead to eventual breach of waste 
packages. 
In General Corrosion and Localized Corrosion of Waste Package Outer Barrier (BSC 2004 
[169984], Section 6.4.4) localized corrosion of the waste package outer barrier occurs when the 
corrosion potential (Ecorr), is equal to or greater than a threshold potential (Ecritical), that is, .E = 
(Ecritical  Ecorr) = 0. The magnitude of the .E is an index of the localized corrosion resistance 
(i.e., the larger the difference, the greater the localized corrosion resistance). The crevice 
corrosion initiation model components (i.e., Ecorr and Ecritical) are functions of the exposure 
conditions (temperature, pH, chloride ion concentration, and nitrate ion concentration). The 
model assumes that, once initiated, localized corrosion of the waste package outer barrier 
propagates at a (time independent) constant rate. As a conservative measure, the base-case 
localized corrosion model uses the crevice repassivation potential (Ercrev) as the critical potential 
for the localized corrosion initiation analysis (i.e., Ecritical = Ercrev). 
Nitrate ions have a strong inhibitive effect on localized corrosion of Alloy 22 in chloridecontaining 
solutions (BSC 2004 [DIRS 169984], Section 6.4.4.3.2; Dunn and Brossia 2002 
[DIRS 162213]). Anions containing nitrogen, phosphorus, and sulfur, which are abundant in the 
repository groundwater, exhibit varying degrees of inhibitive effects (Thomas 1994 
[DIRS 120498]). Inhibitive anions counteract the effects of aggressive anions (e.g., chloride 
ions), which tend to accelerate dissolution and breakdown of the oxide passive films formed on 
Alloy 22. The relationship between the inhibitive and aggressive anions corresponds to 
competitive adsorption or ion exchange at a fixed number of sites on the oxide surface. 
Inhibitive anions overcome the effects of aggressive anions through participation in reversible 
competitive adsorption such that the adsorbed inhibitive anions reduce the surface concentration 
of aggressive anions below a critical value (Thomas 1994 [DIRS 120498], p. 17:54). Because 
only the inhibitive effect of nitrate ions is accounted for in the model, results for solutions with 
significant amounts of other potentially inhibitive ions such as carbonate and sulfate (in addition 
to nitrate ions) are conservative (BSC 2004 [DIRS 169984], Section 8.3). 
The crevice repassivation potential model was constructed using data from pure chloride 
solutions and mixed chloride and nitrate solutions, including some data at high nitrate (up 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-21 October 2004 
to 18 m) and chloride (up to 36 m) concentrations and high temperatures (up to 160C). 
However, the long-term corrosion potential model was constructed using data from relatively 
dilute (in terms of chloride and nitrate ion content) mixed ionic solutions, pure chloride solutions 
(up to about 12 m), and mixed chloride (up to about 13 m) and nitrate (up to about 2.6 m) ion 
solutions. The maximum temperature for any data point used in construction of the corrosion 
potential model was 120C. The lack of measured long-term corrosion potential data in the high 
temperature (> 120C) regime indicates it would be prudent not to make use of the developed 
functional forms to predict localized corrosion initiation under these conditions. Localized 
corrosion (like general corrosion) requires the presence of a liquid water film on the waste 
package surface. To implement the waste package outer barrier localized corrosion initiation 
model, the following criteria are applied in a step-wise fashion (BSC 2004 [DIRS 169984], 
Section 6.4.4.6.7): 
1. If aqueous brine chemistry causes the initiation of localized corrosion, then localized 
corrosion continues to propagate regardless of changes in the bulk chemical exposure 
environment. This is a conservative modeling assumption made because no detailed 
model of the chemistry evolution of the crevice solution is available at this time. 
2. If the exposure temperature exceeds 160C and a water film is present on the waste 
package surface, then localized corrosion initiates. 
Localized corrosion initiated as a result of this criterion is reevaluated in accordance 
with Criterion 3 (below) when the exposure temperature drops below 160C. 
3. If the exposure temperature exceeds 120C but is less than or equal to 160C, then 
a) If the nitrate to chloride ion ratio is 0.5 (or greater), no localized corrosion will 
occur, or 
b) If the nitrate to chloride ion ratio is less than 0.5, then localized corrosion initiates 
and continues to propagate regardless of changes in the bulk chemical exposure 
environment (Criterion 1). 
4. If the exposure temperature is greater than or equal to 20C and less than or equal to 
120C, then the empirical correlations for the long-term corrosion potential (Ecorr) and 
crevice repassivation potential (Ercrev) are evaluated in accordance with the following 
implementation rules. If localized corrosion is determined to initiate, then localized 
corrosion continues to occur regardless of changes in the bulk chemical exposure 
environment (Criterion 1) 
a) If the nitrate to chloride ion ratio in the environment exceeds 0.5, 
then evaluate Ercrev and Ecorr at a nitrate to chloride ion ratio of 0.5. 
b) If the molality of chloride ion in the environment exceeds 36 molal, 
then evaluate Ercrev and Ecorr at a molality of chloride ion of 36 molal. 
If the molality of chloride ion is less than 0.001 molal, 
then evaluate Ercrev and Ecorr at a molality of chloride ion of 0.001 molal. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-22 October 2004 
c) If the molality of nitrate ion in the environment exceeds 18 molal, 
then evaluate Ercrev and Ecorr at a molality of nitrate ion of 18 molal. 
If the molality of nitrate ion is less than 0.001 molal, 
then Ercrev = o
rcrev E (i.e., the crevice repassivation potential in the absence 
of nitrate ions) and evaluate Ecorr at a molality of nitrate ion of 0.001 molal. 
d) If the pH in the environment exceeds 10.9, 
then evaluate Ercrev and Ecorr at a pH of 10.9. 
If the pH in the environment is less than 2.8, 
then initiate localized corrosion. 
In addition, Alloy 22 (UNS N06022) crevice samples were tested for over 5 years in three 
different solutions (simulated dilute water (SDW), simulated concentrated water (SCW) and 
simulated acidified water (SAW)) at 60C and 90C in the Projects LTCTF. None of the 
crevice samples has shown any indication of localized corrosion attack after being tested for over 
five years. The crevice corrosion initiation model also predicts no localized corrosion 
occurrence for these exposure conditions (BSC 2004 [DIRS 169984], Section 7.2.4, Table 7-4). 
A study for the welded Alloy 22 (UNS N06022) samples in 0.5 M NaCl solutions at 95C by the 
investigators at the Center for Nuclear Waste Regulatory Analysis demonstrated that nitrate ion 
is an effective inhibitor of localized corrosion of Alloy 22 (UNS N06022), when the nitrate to 
chloride concentration ratio is greater than 0.2 (Dunn and Brossia 2002 [DIRS 162213]). 
As discussed in FEP 2.1.03.01.0A, General Corrosion of Waste Packages, acidified waters that 
contain trace metals such as lead, arsenic, calcium, and magnesium were used to develop the 
Alloy 22 (UNS N06022) general corrosion model used in the TSPA-LA analyses. These same 
solutions were considered in the analysis of localized corrosion (BSC 2004 [DIRS 169984], 
Section 7.2.4). The small amounts of trace metals identified in the acidified waters were 
appropriately considered in analyzing localized corrosion of waste packages. 
Supporting Reports: 
BSC 2004 [DIRS 169984] 
6.2.7 Localized Corrosion of Drip Shields (2.1.03.03.0B) 
FEP Description: 
Localized corrosion (pitting or crevice corrosion) could enhance degradation of the drip 
shields. 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
In General Corrosion and Localized Corrosion of the Drip Shield (BSC 2004 [DIRS 169845], 
Section 6.6.1), localized corrosion of the titanium drip shield initiates when the corrosion 
potential (Ecorr) equals or exceeds the threshold potential for breakdown of the passive film 
(Ecritical). 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-23 October 2004 
A correlation between exposure parameters (temperature, chloride ion concentration, and pH) 
and the difference between the critical potential (Ecritical) and the corrosion potential (Ecorr) 
(i.e., (.E = Ecritical - Ecorr) was developed to show when localized corrosion could be initiated. 
Localized corrosion initiates when .E is less than or equal to zero (i.e., when Ecorr is greater than 
or equal to Ecritical). The results show that for the drip shield under repository conditions, .E is 
significantly greater than zero over all ranges of pH, chloride concentrations, and temperatures 
(BSC 2004 [DIRS 169845], Section 6.6.3) in the repository. Localized corrosion of Titanium 
Grade 7 will not initiate in repository-relevant environments even at pH values as high as 14 
(BSC 2004 [DIRS 169845], Section 8.4). 
The presence of crevices and concentrated calcium and magnesium chloride solutions and their 
influence on corrosion were also evaluated in General Corrosion and Localized Corrosion of the 
Drip Shield (BSC 2004 [DIRS 169845], Sections 6.5.8, 6.6.4, and 6.6.5). The results in 
Section 6.6.5 of General Corrosion and Localized Corrosion of the Drip Shield (BSC 2004 
[DIRS 169845]) show that the passive film on Titanium Grade 7 is stable in concentrated 
calcium and magnesium chloride solutions and general corrosion is not accelerated. Table 21 in 
General Corrosion and Localized Corrosion of the Drip Shield (BSC 2004 [DIRS 169845]) 
shows that the minimum .E (obtained in 9 molar CaCl2 at 150C) is 1.4V, which excludes the 
possibility of crevice corrosion in these environments. 
As discussed in FEP 2.1.03.01.0B, General Corrosion of Drip Shields, acidified waters 
containing trace metals such as arsenic, calcium, and magnesium were used to develop the 
Titanium Grade 7 general corrosion rates used in the TSPA-LA analyses. These same solutions 
were used in developing the localized corrosion initiation analyses (BSC 2004 [DIRS 169845], 
Section 6.6; DTN: LL021012712251.021 [DIRS 163112]). The small amounts of trace metals 
identified in the acidified waters were appropriately considered in analyzing localized corrosion 
of drip shields and are not considered in this discussion. 
Localized corrosion of Titanium Grade 7 will not initiate under repository exposure conditions 
and the passive film is stable even when crevices are present and the material is exposed to 
concentrated calcium and magnesium chloride solutions. This FEP is excluded based on low 
consequence because its omission does not have a significant effect on radiological exposures to 
the reasonable maximally exposed individual and radionuclide releases to the accessible 
environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.8 Hydride Cracking of Waste Packages (2.1.03.04.0A) 
FEP Description: 
The uptake of hydrogen and the formation of metal hydrides may mechanically weaken 
the waste packages and promote corrosion. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-24 October 2004 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
Hydrogen generated at cathodic sites in a corroding metal may be absorbed into the metal and 
potentially form hydride phases. Hydrogen incorporation could lead to degradation of the 
mechanical properties of the material and render it susceptible to cracking even in the absence of 
the formation of hydride phases. Hydrogen-induced cracking (HIC) is used to refer to the impact 
of hydrogen on waste package materials. HIC results from the combined action of hydrogen and 
residual or sustained applied tensile stresses. 
The Alloy 22 waste package when emplaced is protected by the Titanium Grade 7 drip shield. 
The drip shield is designed to protect the waste package from any fallen ground support 
contacting the waste package, thereby eliminating the chances of galvanic coupling. Likewise, 
the pallet will keep the waste package from contacting the invert (BSC 2004 [DIRS 166879], 
Section 7.2.2), thereby precluding any potential from galvanic coupling of the waste package 
contacting any material in the invert. 
HIC of the waste package outer barrier (Alloy 22 (UNS N06022)) is of low consequence under 
repository-relevant exposure conditions for the following reasons: 
 Handbook data (ASM International 1987 [DIRS 103753], pp. 650 to 652) states that 
fully annealed nickel-base alloys, such as Alloy 22, are essentially immune to hydrogeninduced 
cracking. Annealed alloys are extremely resistant to hydrogen-induced 
cracking, but may lack adequate strength for some structural applications. The structural 
requirements of the waste packages at the YMP are such that strengthening of the alloys 
is not required (Gdowski 1991 [DIRS 100859], Section 5). The structural requirements 
for the waste package are satisfied by the Stainless Steel Type 316 inner shell; 
 The extremely low corrosion rates exhibited by nickel alloys are not sufficient to 
generate enough hydrogen to cause hydrogen-induced cracking (ASM 
International 1987 [DIRS 103753], p. 652); 
 Ni-Cr-Mo alloys with compositions and properties similar to Alloy 22 (Alloys C-276, 
C-4, and 625) maintain their resistance to hydrogen-induced cracking even when heavily 
cold worked to yield strengths in excess of 1,240 MPa (180 ksi) (ASM 
International 1987 [DIRS 103753], p. 169); 
 Cold-worked Ni-Cr-Mo alloys are not susceptible to hydrogen-induced cracking unless 
they are galvanically coupled to a less noble material (or subjected to imposed cathodic 
currents) and strained beyond yield (Gdowski 1991 [DIRS 100859], Section 5.1; 
Asphahani 1978 [DIRS 160352]). 
Aging of Ni-Cr-Mo alloys at temperatures around 500C can lead to ordering, or grain-boundary 
segregation, of deleterious elements such as phosphorous and sulfur, which can increase 
susceptibility to hydrogen-induced cracking (ASM International 1987 [DIRS 103753], p. 169). 
However, since the waste package temperature does not exceed 300C even in the low

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-25 October 2004 
probability-seismic collapsed-drift scenario (BSC 2004 [DIRS 169565], Figure 6.3-56), 
significant ordering and grain-boundary segregation will not occur. Asphahani (1978 
[DIRS 160352]) tested cold worked (60 percent cold swaged) and aged (500C for 100 hours) 
samples of Alloy C-276 at applied stresses up to 92% of yield and cathodic current densities 
of 40 mA/cm2. No incidence of HIC was found. The conclusions reached concerning HIC of 
the waste package are applicable to all waste packages. 
As discussed in FEP 2.1.09.09.0A, Electrochemical Effects in EBS, contact between the 
Alloy 22 waste package outer barrier and Titanium Grade 7 drip shield is not expected. Based 
on the similarity of the two materials (ASM International 1987 [DIRS 103753], p. 557) in the 
electrochemical series, there is a very low probability of galvanic interaction and subsequent 
enhanced corrosion. 
As the Alloy 22 waste package outer barrier is fully annealed and Alloy 22 is highly resistant to 
hydrogen-induced cracking, this FEP is excluded based on low consequence to radiological 
exposures to the RMEI and radionuclide releases to the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.9 Hydride Cracking of Drip Shields (2.1.03.04.0B) 
FEP Description: 
The uptake of hydrogen and the formation of metal hydrides may mechanically weaken 
the drip shields and promote corrosion. 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
Hydrogen generated at cathodic sites in a corroding metal may migrate into the metal and 
potentially form hydride phases. Hydrogen incorporation could lead to degradation of the 
mechanical properties of the material and render it susceptible to cracking even in the absence of 
the formation of hydride phases. Hydrogen-induced cracking (HIC) is used to refer to the impact 
of hydrogen on waste package and drip shield materials. Hydrogen-induced cracking results 
from the combined action of hydrogen and residual or sustained applied tensile stresses. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-26 October 2004 
Hydrogen absorption in a-titanium alloys such as Titanium Grade 7 can occur when three 
general conditions are simultaneously met (Schutz and Thomas 1987 [DIRS 144302]; BSC 2004 
[DIRS 169847], Section 6.1.2): 
1. A mechanism exists for generating nascent (atomic) hydrogen on the surface 
2. Metal temperature is above approximately 80oC (175oF) where the diffusion rate of 
hydrogen into a-titanium is significant 
3. Solution pH is less than 3 or greater than 12, or impressed potentials are more negative 
than -0.7 V (Saturated Calomel Electrode). 
As discussed in General Corrosion and Localized Corrosion of Drip Shield (BSC 2004 
[DIRS 169845], Section 6.6) and FEP 2.1.03.03.0B, Localized Corrosion of Drip Shields, 
localized corrosion (crevice corrosion and pitting) will not occur under the exposure conditions 
anticipated in the repository. However, in the repository design that serves as the basis for the 
license application (BSC 2004 [DIRS 168489]), passive general corrosion and galvanic coupling 
of the drip shield to less noble materials are feasible processes in the repository that could lead to 
hydrogen generation on the surface of the drip shield. Some of the hydrogen produced can 
diffuse into the metal potentially forming hydrides. The direct absorption of radiolytically 
produced hydrogen is insignificant except at dose rates > 102 Gy/hr (Grays per hour 
where 1 Gray = 100 rads) and temperatures > 150oC (BSC 2004 [DIRS 169847], Section 6.1.2). 
Satisfaction of all three general conditions for hydrogen absorption is unattainable in the 
repository (BSC 2004 [DIRS 169847], Section 6.1.2). At certain repository locations, where 
temperatures are = 80C and concentrated groundwater is present, conditions (2) and (3) may be 
satisfied, however, only when all three conditions are met simultaneously can hydrogen 
absorption can be anticipated. 
For the passive noncreviced or inert crevice conditions expected to prevail, the corrosion of the 
titanium alloy will be sustained by reaction with water under neutral conditions (Ti + 2H2O . 
TiO2 + 2 H2) and will proceed at an extremely slow rate. This process will generate hydrogen, 
which must pass through the TiO2 film before absorption into the underlying titanium alloy 
(BSC 2004 [DIRS 169847], Section 6.1.2). 
As stated in Hydrogen Induced Cracking of Drip Shield (BSC 2004 [DIRS 169847], Section 8), 
a simple and conservative model was developed to evaluate the effects of hydrogen-induced 
cracking on the drip shield. The basic premise of the model is that failure will occur once the 
hydrogen content exceeds a certain limit or critical value, HC. Despite the potential occurrence 
of hydrogen absorption into the bulk structure, hydrogen induced cracking of Titanium Grade 7 
is not expected, due to the existence of the oxide film and the high critical value, HC, because of 
the addition of palladium (Pd) (BSC 2004 [DIRS 169847], Section 6.1.4). Recent analyses of 
published data for Titanium Grade 16, whose performance is similar to the Titanium Grade 7, 
suggest that the critical concentration may be well in excess of 1,000 g/g (BSC 2004 
[DIRS 169847], Section 6.1.3). Titanium Grade 7 has a higher Pd concentration than Titanium 
Grade 16; therefore, the HC value for Titanium Grade 7 must be at least 1,000 g/g and could be 
much higher (BSC 2004 [DIRS 169847], Section 6.1.3). The hydrogen concentration in the drip 
shield from passive corrosion 10,000 years after permanent closure is 510 g/g, resulting from a 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-27 October 2004 
conservative estimate (BSC 2004 [DIRS 169847], Section 8.1). This is below the threshold 
concentration and would not result in hydrogen induced cracking or any degradation of fracture 
toughness. 
In the repository design, hydrogen generation may be caused by the galvanic couple formed 
between the Titanium Grade 7 Drip Shield surface and less noble structural components (such as 
rock bolts, wire mesh, and steel liners used in the drift), which may fall onto the drip shield 
surface. If temperatures are = 80C and concentrated groundwaters are present, then the 
formation of locally hydrided hot spots are possible. However, the effect of these locally 
hydrided regions is negligible because (BSC 2004 [DIRS 169847], Section 6.3.2): 
1. The contact area is likely to be small, and the anode-to-cathode area ratio (area of steel 
and titanium, respectively) low leading to only limited amounts of hydrogen 
absorption. 
2. The intermittent nature of seepage at temperatures = 80C will lead to limited periods 
of the aqueous conditions required to sustain an active galvanic couple, thereby 
limiting hydrogen absorption while temperatures are high enough to drive hydrogen 
transport into the metal. The intermittent wetting and drying cycles anticipated on the 
drip shield will lead to the ready formation of calcareous and mineral deposits, which 
are well known to dramatically suppress galvanic currents, thereby stifling hydrogen 
absorption. 
3. Conditions in the repository will be oxidizing, making it less likely that the couple will 
sustain water reduction and, hence, hydrogen absorption. 
4. Titanium drip shield and the steel component surfaces will experience a considerable 
period of dry temperatures = 85C. This will leave the titanium and steel in the 
passive state (especially titanium) and avoid galvanic contact and hydrogen absorption 
by titanium. 
5. As discussed above, a-titanium alloys exhibit a protective oxide film and a relatively 
high critical value, HC, due to palladium addition, reducing the tendency for hydrogen 
induced cracking (BSC 2004 [DIRS 169847], Sections 6.1.3 and 6.1.5). 
The general corrosion rates of titanium alloys under the repository conditions are very low. 
Thus, hydrogen generation rates and, hence, hydrogen absorption rates, are expected to be very 
low (BSC 2004 [DIRS 169847], Section 6.1.2). 
In summary, hydrogen-influenced cracking of the Titanium Grade 7 is not expected to occur 
under the anticipated repository conditions. Locally hydrided regions, resulting from galvanic 
coupling, are possible, but their effects are negligible. Therefore, this FEP can be excluded 
based on low consequence to radiological exposures to the RMEI and radionuclide releases to 
the accessible environment. 
TSPA Disposition: 
N/A 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-28 October 2004 
Supporting Reports: 
N/A 
6.2.10 Microbially Influenced Corrosion (MIC) of Waste Packages (2.1.03.05.0A) 
FEP Description: 
Microbial activity may catalyze waste package corrosion by otherwise kinetically 
hindered oxidizing agents. The most likely process is microbial reduction of 
groundwater sulfates to sulfides and reaction of iron with dissolved sulfides. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
Microbially influenced corrosion (MIC) is the contribution to the corrosion of a metal or alloy 
due to the presence or activity, or both, of microorganisms. Microbially influenced corrosion 
most often occurs due to the increase in anodic or cathodic reactions due to the direct impact of 
microorganisms on the alloy or by indirect chemical effects on the surrounding solution 
(BSC 2004 [DIRS 169984], Section 6.4.5). Waste package microbially influenced corrosion is 
discussed in General Corrosion and Localized Corrosion of Waste Package Outer Barrier 
(BSC 2004 [DIRS 169984], Section 6.4.5). The effect of microbially influenced corrosion on 
waste package corrosion is analyzed with an enhancement factor approach (i.e., microbially 
influenced corrosion increases the general corrosion penetration rate). In this approach, the 
abiotic corrosion rate is multiplied by the enhancement factor when exposure conditions in the 
emplacement drift are conducive to significant microbial activity (BSC 2004 [DIRS 169984], 
Section 6.4.5). The microbially influenced corrosion enhancement factor, fMIC, is uniformly 
distributed between 1 and 2 (BSC 2004 [DIRS 169984], Section 6.4.5) and is applied to the 
waste package outer barrier general corrosion rate when the relative humidity at the waste 
package outer barrier surface is above 90 percent (BSC 2004 [DIRS 169984], Section 6.4.5). 
The general corrosion rate enhancement factor is applied to the entire waste package surface 
(BSC 2004 [DIRS 169984], Section 6.4.5) when the relative humidity threshold is satisfied. This 
treatment of microbially influenced corrosion applies to all waste packages. 
In the integrated waste package degradation analysis (BSC 2004 [DIRS 169996]), the general 
corrosion rate enhancement factor is sampled once per realization (i.e., the variation in the 
general corrosion rate microbially influenced corrosion enhancement factor, fMIC, is entirely due 
to uncertainty (BSC 2004 [DIRS 169984], Section 6.4.5)), and applied to the entire waste 
package surface. The output from the integrated waste package degradation analysis is a set of 
profiles (time histories) for the failure (i.e., initial breach) and subsequent number of penetration 
openings in the waste package and drip shield as a function of time. The integrated waste 
package degradation analysis is used directly in the TSPA-LA. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-29 October 2004 
Supporting Reports: 
BSC 2004 [DIRS 169984] 
BSC 2004 [DIRS 169996] 
6.2.11 Microbially Influenced Corrosion (MIC) of Drip Shields (2.1.03.05.0B) 
FEP Description: 
Microbial activity may catalyze drip shield corrosion by otherwise kinetically hindered 
oxidizing agents. The most likely process is microbial reduction of groundwater sulfates 
to sulfides and reaction of iron with dissolved sulfides. 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
Microbially influenced corrosion (MIC) of titanium is discussed in General Corrosion and 
Localized Corrosion of the Drip Shield (BSC 2004 [DIRS 169845], Section 6.7.2). Corrosion 
handbooks and literature reviews generally indicate that titanium alloys are immune to 
microbially influenced corrosion (Revie 2000 [DIRS 159370], Chapter 47; Little and 
Wagner 1996 [DIRS 131533]; and Brossia et al. 2001 [DIRS 159836], Section 4.1.3) under 
repository-like conditions. The remarkable stability of the TiO2 passive film formed on titanium 
alloys confers this immunity. While titanium is susceptible to biofouling in seawater solutions, 
the biofilm does not compromise the integrity of the passive film and, therefore, biofouled 
titanium maintains its resistance to localized corrosion processes (Revie 2000 [DIRS 159370], 
Chapter 47). It has been reported that production of nitrates, polythionates, thiosulfates, and 
oxygen associated with aerobic biologic activity does not significantly increase the corrosion rate 
of titanium alloys (Brossia et al. 2001 [DIRS 159836], Section 4.1.3). 
Steep gradients in O2 and pH can exist within biofilms; typically aerobic and near neutral in the 
outer layers becoming acidic and low in O2 close to the metal surface (Shoesmith and Ikeda 1997 
[DIRS 151179], Section 6). Hydrogen peroxide has been detected in biofilms at millimolar 
levels, the amount of which is thought to be controlled by bacteria enzymes during the aerobic 
respiration process (Shoesmith and Ikeda 1997 [DIRS 151179]). Hydrogen peroxide maintains a 
low pH (< 3) near the metal by oxidizing metal cations that then undergo hydrolysis. These 
chemical changes can lead to enoblement (a shift of the corrosion potential to more positive 
values) of titanium by up to 500 mV (Shoesmith and Ikeda 1997 [DIRS 151179], Section 6). As 
shown by Figures 19 and 20 of General Corrosion and Localized Corrosion of the Drip Shield 
(BSC 2004 [DIRS 169845], Section 6.6.3), .E far exceeds 500 mV at low pH values 
(i.e., localized corrosion will not initiate even if the corrosion potential is increased by 500 mV). 
Enoblement can also lead to several beneficial effects, including thickening of the passive film 
and a decrease in the number and density of film defects (Shoesmith and Ikeda 1997 
[DIRS 151179], Section 3 and 6). According to Shoesmith et al. (1995 [DIRS 117892]), 
initiation of crevice corrosion under biofilms is highly unlikely for titanium. Lastly, microbial 
growth in the repository will likely be limited by the availability of nutrients (BSC 2004 
[DIRS 169984], Section 6.4.5). 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-30 October 2004 
Microbially influenced corrosion has no significant effect on either general or localized corrosion 
processes of titanium alloys under the exposure conditions in the repository. Therefore, this FEP 
is excluded based on low consequence to radiological exposures to the RMEI and radionuclide 
releases to the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.12 Internal Corrosion of Waste Packages Prior to Breach (2.1.03.06.0A) 
FEP Description: 
Aggressive chemical conditions within the waste package could contribute to corrosion 
from the inside out. Effects of different waste forms, including CSNF and DSNF, are 
considered in this FEP. 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
Project Requirements Document (Canori and Leitner 2003 [DIRS 166275], PRD-013/T-036 and 
PRD-013/T-045) states that waste packages will be designed to preclude internal corrosion of the 
waste package and contained material. The waste package design will also preclude chemical, 
electrochemical, or other reactions (such as internal corrosion) of the waste package such that 
there will be no adverse effect on normal handling, transportation, storage, emplacement, 
containment, or isolation, or on abnormal occurrences such as a waste package drop accident and 
premature breach in the repository. This FEP pertains only to internal corrosion of the waste 
package. 
The CSNF assemblies will be dried prior to their insertion into the waste packages 
(CRWMS M&O 2000 [DIRS 123881], p. II-10). After being loaded with waste, the waste 
packages are evacuated, to remove air and ensure the waste form is dry, then backfilled with 
helium to severely limit oxidizing gases (e.g., oxygen and water vapor) and help preserve the 
chemical and physical stability of the waste form (BSC 2004 [DIRS 170803], Section 4.2.2). 
The inert gas environment is intended to ensure a negligible amount of corrosion degradation 
prior to the breach of the waste packages. The requirements are incorporated into DOE and 
Commercial Waste Package System Description Document (BSC 2004 [DIRS 170803], 
Section 3). The repository will be designed and constructed in accordance with an approved QA 
program thereby ensuring the requirements of DOE and Commercial Waste Package System 
Description Document are implemented correctly. 
Analyses performed by Kohli and Pasupathi (1986 [DIRS 131519]) suggest that the most likely 
cause of internal corrosion is the residual moisture remaining in the waste package at the time of 
emplacement. The primary source of this residual moisture is from waterlogged failed fuel rods. 
Analyses presented in the above reference indicate that the amount of moisture available to cause 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-31 October 2004 
internal corrosion is very limited, and even with very conservative assumptions, the potential for 
degradation of the waste package materials is very remote. Additionally, all waste package types 
have significant amounts of internals (e.g., basket materials composed of carbon steel) 
(BSC 2004 [DIRS 169472]) that will corrode in preference to the Alloy 22 (UNS N06022) waste 
package outer barrier and the Stainless Steel Type 316 inner vessel (ASM International 1987 
[DIRS 103753], p. 557). Thus, no significant corrosion damage to the waste package outer 
barrier and Stainless Steel Type 316 inner vessel internal surfaces will occur. 
DSNF waste packages containing N Reactor spent fuel may have significant quantities of 
residual free and chemically bound water at the time of sealing prior to placement in storage at 
its present location. However, the N Reactor spent fuel cladding is significantly damaged, thus 
exposing chemically reactive uranium metal surfaces, which could react with residual water 
producing uranium oxide and uranium hydride. Other forms of DSNF are less damaged, and 
will contain much lower quantities of residual water due to drying prior to sealing for storage at 
its present location. Damaged DSNF will be placed in waste packages (CRWMS M&O 2000 
[DIRS 150823], Table 5) that will not contain any residual water until breached. 
As discussed in FEP 2.1.03.04.0A, Hydride Cracking of Waste Packages, HIC of the waste 
package outer barrier (Alloy 22) is not considered to be a credible degradation mechanism under 
repository-relevant exposure conditions and will not enhance internal corrosion. 
In view of the above discussion, it can be concluded that insignificant corrosion damage of 
DSNF waste packages, DHLW glass waste packages, and CSNF waste packages will occur due 
to drying of the waste form before loading, backfilling of the waste packages with an inert gas, 
and the presence of significant amounts of waste package internals, which will corrode in 
preference to the Alloy 22 (UNS N06022) waste package outer barrier and the Stainless Steel 
Type 316 inner vessel. This FEP is excluded based on low consequence to radiological 
exposures to the RMEI and radionuclide releases to the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.13 Mechanical Impact on Waste Package (2.1.03.07.0A) 
FEP Description: 
Mechanical impact (dynamic loading) on the waste package may be caused by internal 
and external forces such as internal gas pressure, forces caused by swelling corrosion 
products, rockfall, and possible waste package or drip shield movement. Seismicinduced 
impacts are addressed in a separate FEP. 
Screening Decision: 
Excluded (Low Probability) (Rockfall) 
Excluded (Low Consequence) (Internal gas pressure and swelling of corrosion products) 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-32 October 2004 
Screening Argument: 
The arguments put forth and the conclusions reached in this section are equally applicable to all 
waste packages. 
Internal gas pressure: A calculation of the maximum stresses developed in the waste package 
due to internal pressurization as a result of fuel rod rupture at 400C is less than 1995 ASME 
Boiler and Pressure Vessel Code (ASME 1995 [DIRS 141257]) requirements for the allowable 
tensile strength of the materials used for the waste package (CRWMS M&O 2000 
[DIRS 150823], Section 6.3). In the repository, the maximum waste package surface 
temperature will be less than 300C even in the low-probability-seismic collapsed-drift scenario 
(BSC 2004 [DIRS 169565], Figure 6.3-56). Therefore, with the current robust waste package 
design, the pressurization of the internal gas under the expected repository condition would not 
cause mechanical damage to the waste package and can be excluded based on low consequence 
to radiological exposures to the RMEI and radionuclide releases to the accessible environment. 
Swelling of corrosion products: Mechanical damage to the waste package from swelling 
corrosion products is discussed in greater detail under FEP 2.1.09.03.0B, Volume Increase of 
Corrosion Products Impacts Waste Package. Analyses cited in General Corrosion and Localized 
Corrosion of Waste Package Outer Barrier (BSC 2004 [DIRS 169984], Section 6.4.2), indicate 
that, even under very conservative assumptions, the growth of the corrosion product (primarily 
Cr2O3 as stated in BSC 2004 [DIRS 169984], Section 6.4.2) oxide layer (93 m in 10,000 years) 
is not thick enough to produce enough pressure to cause mechanical damage to the Stainless 
Steel Type 316 inner shell or the Alloy 22 (UNS N06022) outer barrier. Alloy 22 
(UNS N06022) contains 20.0 to 22.5 weight % chromium (BSC 2004 [DIRS 169984], Section 
1.1). Therefore, waste package damage from swelling corrosion products is excluded based on 
low probability of occurrence under the exposure conditions in the repository. 
Rockfall: Mechanical damage of the waste package by rockfall is discussed in greater detail 
under FEP 2.1.07.01.0A, Rockfall. As indicated in Subsurface Facility Description Document 
(BSC 2004 [DIRS 170265], Section 3.1.1.4.14.1) the drip shield will be designed to protect the 
waste package from rockfalls during postclosure. Because the drip shield provides adequate 
protection to the waste packages from rockfall (FEP 2.1.03.07.0B, Mechanical Impact on Drip 
Shield), the effects of rockfall on the waste package are excluded from consideration due to low 
probability of occurrence. 
Seismic: Mechanical damage of the waste packages and drip shields by ground motion and 
rockfalls during seismic events is discussed 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 in 
Engineered Barrier System Features, Events, and Processes (BSC 2004 [DIRS 169898]). 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-33 October 2004 
6.2.14 Mechanical Impact on Drip Shield (2.1.03.07.0B) 
FEP Description: 
Mechanical impact (dynamic loading) on the drip shield may be caused by forces such as 
rockfall and possible waste package or drip shield movement. Seismic-induced impacts 
are addressed in a separate FEP. 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
Mechanical damage of the drip shield by rockfall is discussed in greater detail under 
FEP 2.1.07.01.0A, Rockfall. That FEP discussion provides relevant references discussing the 
issue in greater detail. In addition, the emplacement drift system design criteria require the drip 
shield to be designed to protect the waste package from rockfalls during postclosure (BSC 2004 
[DIRS 170265], Section 3.1.1.4.14.1). Drip Shield and Waste Package Emplacement Pallet 
Design Report (BSC 2004 [DIRS 166879], Section 7.2.1) shows that the deflection of the drip 
shield due to an 11.5 metric ton rockfall (which produces the maximum vertical displacement in 
the drip shield components) is not large enough to cause the drip shield to contact the waste 
package. The maximum displacement from the 11.5 metric ton rockfall event is 254 mm 
(BSC 2004 [DIRS 166879], Section 7.2.1). The minimum gap upon emplacement between the 
drip shield and waste package outer barrier is about 367 mm (BSC 2004 [DIRS 166879], 
Section 7.2.1). 
Thus, the drip shield provides adequate protection to the waste package from rockfall. In view of 
the above rationale, this FEP is excluded as low consequence to radiological exposures to the 
RMEI and radionuclide releases to the accessible environment. 
Mechanical damage of the waste package and drip shield by ground motion during seismic 
events is discussed in greater detail under 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, in Engineered Barrier 
System Features, Events, and Processes (BSC 2004 [DIRS 169898]). 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.15 Early Failure of Waste Packages (2.1.03.08.0A) 
FEP Description: 
Waste packages may fail prematurely because of manufacturing defects, improper 
sealing, or other factors related to quality control during manufacture and emplacement. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-34 October 2004 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
Analysis of Mechanisms for Early Waste Package/Drip Shield Failure (BSC 2004 
[DIRS 170024]) evaluates several mechanisms for early breach of the waste package. Of the 
mechanisms evaluated, weld flaws, improper heat treatment, improper laser peening, and 
improper handling of waste packages were determined to be included in the TSPA-LA model 
(BSC 2004 [DIRS 170024], Section 7). 
As discussed in FEP2.1.03.02.0A, Stress Corrosion Cracking of Waste Packages, weld flaws act 
as sites for initiation of SCC in the waste package closure weld regions (BSC 2004 
[DIRS 169985]; BSC 2004 [DIRS 169996]). Weld flaws are included in TSPA-LA analysis 
through their effect on SCC of waste packages. 
Improper heat treatment results primarily from improper stress-relief annealing. The 
consequence of improper laser peening is the introduction of unacceptable amounts of cold work 
in the material and increased susceptibility to stress corrosion cracking. Improper handling of 
the waste packages may lead to gouges in the waste package outer surface and provide sites for 
stress corrosion cracks. A discussion of these causes and consequences is provided in Analysis 
of Mechanisms for Early Waste Package/Drip Shield Failure (BSC 2004 [DIRS 170024], 
Section 6.4.8). 
The number of early failed waste packages per realization is represented by a Poisson 
distribution with an uncertain intensity. The Poisson intensity is sampled from a log normal 
distribution with a median of 7.2  10-6 and an error factor of 15 (BSC 2004 [DIRS 170024], 
Section 7, Table 22). Improperly heat-treated waste packages might be susceptible to stress 
corrosion cracking or lead to the formation of grain-boundary precipitates. In lieu of trying to 
identify a single specific mechanism of degradation, a very conservative approach is used based 
on recommendations in Analysis of Mechanisms for Early Waste Package/Drip Shield Failure 
(BSC 2004 [DIRS 170024], Section 6.4.8) for evaluating waste package early breach: 
 A failure of the waste package outer barrier shell and outer and inner closure lids should 
be assumed, as well as the failure of the stainless steel structural inner vessel and closure 
lid. 
 The affected waste packages should be assumed to fail immediately upon initiation of 
degradation processes. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-35 October 2004 
 The entire waste package surface area should be considered affected by waste package 
early failure. 
 The materials of the entire affected area should be assumed lost upon failure of the waste 
package because the affected area will be subjected to stress corrosion cracking and 
highly enhanced localized and general corrosion. 
This early breach treatment applies to all waste packages. 
Waste package emplacement errors, including design deviations, improper quality control, 
surface contamination, and administrative errors leading to unanticipated conditions, are 
addressed in FEP 1.1.03.01.0A, Error in Waste Emplacement, which is shared between this 
report and Engineered Barrier System Features, Events, and Processes (BSC 2004 
[DIRS 169898]). 
Supporting Reports: 
BSC 2004 [DIRS 169996] 
BSC 2004 [DIRS 170024] 
6.2.16 Early Failure of Drip Shields (2.1.03.08.0B) 
FEP Description: 
Drip shields may fail prematurely because of manufacturing defects, improper sealing, or 
other factors related to quality control during manufacture and emplacement. 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
Evaluation of several mechanisms for early failure of the drip shield including weld flaws, base 
metal flaws, improper weld material or base metal, improper heat treatment, contamination, 
improper handling, and drip shield emplacement error was performed in Analysis of Mechanisms 
for Early Waste Package/Drip Shield Failure (BSC 2004 [DIRS 170024], Section 6.3). 
The consequence of weld flaws, base metal flaws, improper heat treatment, and damage by 
mishandling is an increased susceptibility to stress corrosion cracking (BSC 2004 
[DIRS 170024], Section 6.4). 
For the drip shields, fabrication welds will be fully stress-relief annealed before placement in the 
drifts (Plinski 2001 [DIRS 156800], Section 8.3.17). Therefore, drip shields are not subject to 
SCC upon emplacement (BSC 2004 [DIRS 169985], Section 6.3.7). However, the drip shields 
are subject to SCC as a result of seismic-induced loading and rockfalls. Seismic effects on drip 
shield degradation are discussed 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 (BSC 2004 
[DIRS 169898]). The consequence of stress corrosion cracking (see also FEP 2.1.03.02.0B, 
Stress Corrosion Cracking (SCC) of Drip Shields) on drip shield performance is discussed in 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-36 October 2004 
FEP 2.1.03.10.0B, Advection of Liquids and Solids Through Cracks in the Drip Shield. 
FEP 2.1.03.10.0B is excluded on the basis of low consequence because the amount of advective 
water flow through cracks in the drip shield will be severely limited based on the small aperture 
width (narrow opening and tight cracks), the presence of capillary forces, and the potential for 
plugging of the cracks due to mineral deposits. Thus, the effective water flow rate through 
cracks in the drip shield will be extremely low and will not contribute significantly to the overall 
radionuclide release rate from the repository. Therefore, since the primary role of the drip shield 
is to keep water from contacting the waste package, SCC of the drip shield (which is the 
consequence of weld flaws, base metal flaws, improper heat treatment, and damage by 
mishandling) does not compromise the design purpose of the drip shield. 
The use of improper weld or base metal material is possible in the repository (BSC 2004 
[DIRS 170024], Section 6.3.3); however, due to the strict controls that will govern the 
fabrication of the drip shield, it is expected that the material composition of the improper weld or 
base metal material will differ only slightly from the intended composition (BSC 2004 
[DIRS 170024], Section 6.4.2). In view of the high corrosion resistance of the materials in 
question, the consequences of improper weld or base metal will be insignificant (BSC 2004 
[DIRS 170024], Section 6.4.2). 
The probability of drip shield surface contamination is also evaluated in Analysis of Mechanisms 
for Early Waste Package/Drip Shield Failure (BSC 2004 [DIRS 170024], Section 6.3.5). It is 
found that the consequence of drip shield surface contamination is not significant from a 
corrosion standpoint (BSC 2004 [DIRS 170024], Section 6.4.5). On this basis, drip shield 
surface contamination is considered to be insignificant. 
As discussed in FEP 1.1.03.01.0A, Error in Waste Emplacement, errors in drip shield 
emplacement large enough to allow dripping water to contact underlying waste package would 
be detected and remedied. FEP 1.1.03.01.0A is excluded based on low consequence to 
radiological exposures to the reasonable maximally exposed individual and radionuclide releases 
to the accessible environment. 
Thus, manufacturing defects in the drip shield and early failure mechanisms (not including drip 
shield emplacement error (FEP 1.1.03.01.0A)) of the drip shield can be excluded based on low 
consequence to radiological exposures to the RMEI and radionuclide releases to the accessible 
environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.17 Copper Corrosion in EBS (2.1.03.09.0A) 
FEP Description: 
Chemical reactions involving copper corrosion have been identified as being of potential 
interest for repository programs considering the use of copper containers. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-37 October 2004 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
The repository does not use copper containers and thus this FEP description does not apply as 
written. However, consideration of copper effects is of merit since a small amount of copper 
may be present in the emplacement drifts as part of the gantry rail system (BSC 2001 
[DIRS 154441]). The copper would have no adverse effects on the performance of the Alloy 22 
(UNS N06022) waste package outer barrier or Titanium Grade 7 drip shield material as there is 
no potential for the waste package or the drip shield to come in contact with copper. The waste 
package is designed to rest on a pallet, which is constructed of Alloy 22 (BSC 2004 
[DIRS 168489]) and is designed to keep the waste package from contacting other dissimilar 
metals (BSC 2004 [DIRS 166879], Section 7.2.2). Similarly, the drip shields are designed to 
contact no other material except Alloy 22 (UNS N06022) feet, which are attached to the bottom 
of the drip shields (BSC 2004 [DIRS 168489]). Therefore, the effect of copper corrosion is of 
low consequence to radiological exposures to the RMEI and radionuclide releases to the 
accessible environment. 
If, however, the drip shield were to come in contact with copper due to the failure of gantry 
system, there is a potential for galvanic interaction between the titanium and copper and 
subsequent hydrogen absorption. The potential for hydrogen absorption, or hydrogen induced 
cracking, is considered low since Titanium Grade 7 contains Pd and because a protective oxide 
film forms on the surface of the drip shield, which hinders hydrogen absorption (BSC 2004 
[DIRS 169847], Section 6.1.3). When Titanium Grade 7 is cathodically polarized at -1.2 V 
(versus saturated calomel electrode), a more severe condition than galvanically coupling to 
copper, the hydrogen absorption efficiency was found to be 0.015 and the resulting hydrogen 
absorption insignificant (BSC 2004 [DIRS 169847], Section 6.1.6). Thus, corrosion due to 
copper in the gantry rail system may be excluded based on low consequence to radiological 
exposures to the RMEI and radionuclide releases to the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-38 October 2004 
6.2.18 Advection of Liquids and Solids Through Cracks in the Waste Package 
(2.1.03.10.0A) 
FEP Description: 
The presence of one or more cracks or other small openings of sufficient size in a waste 
package may provide a pathway for the advective flow of water (e.g., thin films or 
droplets) or solid material into the waste package. The resulting presence of sufficient 
water or solid material in the waste package may affect in-package chemistry and/or 
criticality. Partial or full plugging of the waste package cracks by chemical or physical 
reactions after their formation (i.e., healing) could also affect water flow and radionuclide 
transport through the waste package. Passivation by corrosion products is a potential 
mechanism for waste package healing. 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
Formation of stress corrosion cracks in the waste packages are modeled in Stress Corrosion 
Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural 
Material (BSC 2004 [DIRS 169985]). This document states that the through wall cracks will be 
located in the closure lid weld region and oriented in the radial direction. Circumferential cracks 
in the closure lid weld region may occur but they are not expected to penetrate through thickness 
of the lid because of the stress profile through the thickness of the lid (BSC 2004 
[DIRS 169985]). 
Typically the stress corrosion cracks are very tight and tortuous and often with many branches. 
EBS Radionuclide Transport Abstraction (BSC 2004 [DIRS 169868], Section 6.3.3) (RTA 
model) provides a prediction of crack aperture size for the cracks in the waste package. A range 
of crack aperture values are calculated using the estimated residual stress (from finite element 
simulations) and the crack length (approximately twice the lid thickness). EBS Radionuclide 
Transport Abstraction (BSC 2004 [DIRS 169868], Section 6.3.3) calculates the following crack 
widths at the inner and outer surfaces (118 m and 196 m, respectively) of the through wall 
radial cracks in the waste package lid using the method outlined in Stress Corrosion Cracking of 
the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material 
(BSC 2004 [DIRS 169985], Section 6.5.2). 
Water vapor diffusion is treated in the RTA model as the only probable mechanism for water to 
enter the waste package through the SCCs. Advective flow through the cracks is of low 
consequence based on the following argument. First, a film that completely spans the opening of 
a stress corrosion crack creates a differential in the capillary forces that will prevent any further 
ingress of flowing water into the waste package. Second, the presence of corrosion products in 
the tight stress corrosion crack may provide capillary barrier for advective flow into the waste 
package. Third, the corrosion products in the cracks will provide resistance to flow requiring a 
substantial pressure gradient that is unlikely to exist. Additionally, the RTA model follows the 
logic that dripping water is capable of contacting a stress corrosion crack in the lid only if the 
waste package is tilted upward. This will not occur as the ends of the pallet designed to support 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-39 October 2004 
the waste package are made of Alloy 22 and are expected to last beyond the regulatory period 
(BSC 2004 [DIRS 166879], Section 7.2.2). 
Based on the above discussion, the amount of advective water flow through cracks in waste 
packages will be severely limited based on the low probability of SCC formation, small aperture 
width (narrow opening and tight cracks), and the presence of capillary forces. On these bases, 
advective flux of liquids and solids through cracks in the waste package is excluded based on 
low consequence to radiological exposures to the RMEI and radionuclide releases to the 
accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.19 Advection of Liquids and Solids Through Cracks in the Drip Shield (2.1.03.10.0B) 
FEP Description: 
The presence of one or more cracks or other small openings of sufficient size in a drip 
shield may provide a pathway for the advective flow of water (e.g., thin films or droplets) 
or solid material through the drip shield. The resulting flux may affect drip shield 
performance and/or subsequent dripping onto the waste packages. Partial or full 
plugging of the drip shield cracks by chemical or physical reactions after their formation 
(i.e., healing) could also affect water flow through the drip shield. 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
As discussed in FEP 2.1.03.03.0B, Localized Corrosion of Drip Shields, no localized corrosion 
of the drip shield is expected under repository exposure conditions. However, drip shields are 
subject to stress corrosion cracking (SCC), under seismic induced loading and rockfalls due to 
drift collapse. If cracks were to be produced in the drip shield due to rockfalls, they tend to be 
very tight in passive alloys such as Titanium Grade 7 (BSC 2004 [DIRS 169985], Section 6.3.7). 
Crack sizes are estimated in Flow of Water and Pooling in a Waste Package to be between 130 
to 208 m (BSC 2001 [DIRS 152248], Section 6.3.3). Another estimate of the crack width was 
made in Plugging of Stress Corrosion Cracks by Precipitation (BSC 2001 [DIRS 156807], 
Table 5-13) and suggests a value of 157 m which is within the range provided above. 
Flow of water through the stress corrosion cracks in the drip shields is addressed in Water 
Distribution and Removal Model (CRWMS M&O 2001 [DIRS 152016]). An analysis of crevice 
water holding capacity in that report shows that for a drip shield with a wall thickness of 15 mm, 
the maximum crevice width that can hold a full column of water (fully saturated crack) 
is 0.916 mm (CRWMS M&O 2001 [DIRS 152016], Section 6.1.4), based on the contact angle of 
the pendant drop on the surface. Cracks capable of supporting pendant drop contact angles 
of 180 or larger will not be fully saturated. This analysis implies that crevices smaller 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-40 October 2004 
than 0.916 mm when fully saturated will be capable of supporting surface water film, and 
thereby not allow water to pass through the crevice. Cracks larger than 0.916 mm can be 
partially saturated and may allow water to pass through the crack. However, as discussed above 
the width of cracks in the drip shields are significantly below the aperture width (0.916 mm) 
needed for supporting flow through the crack. 
Even if some flow through the cracks is possible, the small heat flux across the thickness of the 
drip shield will result in the evaporation of the slow flowing water and deposition of mineral 
scale deposit (principally calcium carbonate or calcite) in the crack and at the surface. This 
formation of calciferous deposits is well documented in seawater environments and in heat 
exchangers through which natural brines are forced to flow, such as in desalination plants and in 
potash plants (Cowan et al. 1976 [DIRS 105212], pp. 1 to 39 and 376 to 383). In both these 
cases, titanium surfaces are heat sources operating at about 100C. Such deposits form rapidly 
under flowing conditions, and have to be removed regularly to avoid heat exchanger efficiency. 
A detailed calculation of the expected rate of SCC plugging due to calcite precipitation has been 
performed and is documented in Plugging of Stress Corrosion Cracks by Precipitation 
(BSC 2001 [DIRS 156807], Section 6). The calculation conservatively assumes that the 
corrosion products generated on the crack faces, as well as colloids, particulates, and any 
precipitated minerals, do not help in resisting the water flow and that there is a uniform water 
seepage flow in space and time. It is concluded that under these assumptions, the cracks will be 
sealed in a few hundred years at most when water is allowed flow at a low film flow rates. 
Based on the above discussion, the amount of advective water flow through cracks in drip shields 
will be severely limited based on the small aperture width (narrow opening and tight cracks), the 
presence of capillary forces, and the potential for plugging of the cracks due to mineral deposits. 
On these bases, advective flux of liquids and solids through cracks in the drip shields is excluded 
based on low consequence to radiological exposures to the RMEI and radionuclide releases to 
the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-41 October 2004 
6.2.20 Physical Form of Waste Package and Drip Shield (2.1.03.11.0A) 
FEP Description: 
The specific forms of the various drip shields, waste packages, and internal waste 
containers that are proposed for the Yucca Mountain repository can affect long-term 
performance. Waste package form may affect container strength through the shape and 
dimensions of the waste package and affect heat dissipation through waste package 
volume and surface area. Waste package and drip shield materials may affect physical 
and chemical behavior of the disposal area environment. Waste package and drip shield 
integrity will affect the releases of radionuclides from the disposal system. Waste 
packages may have both local effects and repository-scale effects. All types of waste 
packages and containers, including CSNF, DSNF, and DHLW, should be considered. 
Screening Decision: 
Included 
Screening Argument: 
N/A 
TSPA Disposition: 
The waste package, drip shield, and repository design configurations for the Yucca Mountain 
Project are shown in D&E/PA/C IED Emplacement Drift Configuration and Environment 
(BSC 2004 [DIRS 168489]). Only one drip shield configuration is used in the repository. While 
more than one waste package configuration will be used in the repository, they are all similar in 
their general design, fabrication methodology, and dimensions (Plinski 2001 [DIRS 156800], 
Section 1). Therefore, there will be little variation in strength, dimensions, and shape of the 
waste packages used in the repository. Waste package and drip shield degradation modes are 
modeled in General Corrosion and Localized Corrosion of Waste Package Outer Barrier 
(BSC 2004 [DIRS 169984]), General Corrosion and Localized Corrosion of the Drip Shield 
(BSC 2004 [DIRS 169845]), and analyzed in WAPDEG Analysis of Waste Package and Drip 
Shield Degradation (BSC 2004 [DIRS 169996]). 
WAPDEG Analysis of Waste Package and Drip Shield Degradation (BSC 2004 [DIRS 169996], 
Section 6.3.2) analyzes two nominal waste package configurations. The first waste package 
configuration is referred to as the commercial spent nuclear fuel (CSNF) waste package 
configuration for which the 21-PWR waste package configuration parameters are used because 
the 21-PWR waste package configuration is the most common waste package configuration in 
the repository. The second waste package configuration analyzed is the codisposal (CDSP) 
waste package configuration whose length is considered to be the average length of the 5 HLW/1 
DOE SNF long and short waste package configurations (which have roughly equal populations 
in the inventory). Because the integrated waste package degradation analysis documented in 
WAPDEG Analysis of Waste Package and Drip Shield Degradation (BSC 2004 [DIRS 169996]) 
is used directly in the TSPA-LA, the effects of differences in the physical form of waste 
packages on corrosion processes are included in TSPA-LA. 
The physical effects of degraded waste packages and drip shields on flow and transport of 
radionuclides are indirectly included in the selection of the EBS flow pathways, but they do not 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-42 October 2004 
have an explicit effect because the flow pathways are modeled without regard to the detailed 
mechanisms of flow (FEP 2.1.08.07.0A, Unsaturated Flow in the EBS) (BSC 2004 
[DIRS 169898]). Chemical effects of the waste package and drip shield materials are discussed 
in FEPs 2.1.09.01.0A, Chemical Characteristics of Water in Drifts; and 2.1.09.02.0A, Chemical 
Interactions with Corrosion Products, in Engineered Barrier System Features, Events and 
Processes (BSC 2004 [DIRS 169898]). 
Supporting Reports: 
BSC 2004 [DIRS 169984] 
BSC 2004 [DIRS 169845] 
BSC 2004 [DIRS 169996] 
6.2.21 Oxygen Embrittlement of Drip Shields (2.1.06.06.0B) 
FEP Description: 
A potential failure mechanism for drip shields is oxygen embrittlement, resulting from 
the diffusion of interstitial oxygen in the titanium at high temperatures. 
Screening Decision: 
Excluded (Low Probability) 
Screening Argument: 
Oxygen embrittlement of titanium results from diffusion of interstitial oxygen into the metal at 
temperatures > 340C (ASM International 1987 [DIRS 103753], p. 681). The time to failure 
depends on the alloy composition, material thickness, and stress state. In the repository, the 
maximum waste package surface temperature (because the waste package is the heat source, the 
waste package surface temperature would be greater than the drip shield surface temperature) is 
less than 300C even in the low-probability-seismic collapsed-drift scenario (BSC 2004 
[DIRS 169565], Figure 6.3-56). Because this temperature is less than the threshold temperature 
for oxygen embrittlement of 340C, oxygen embrittlement of the titanium drip shields is 
excluded on the basis of low probability of occurrence under the exposure conditions in the 
repository. 
Hydrogen embrittlement is discussed in FEP 2.1.03.04B, Hydride Cracking of Drip Shields. The 
effects of potential acidification in creviced regions such as cracks is discussed in 
FEP 2.1.03.03.0B, Localized Corrosion of Drip Shields. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-43 October 2004 
6.2.22 Mechanical Effects at EBS Component Interfaces (2.1.06.07.0B) 
FEP Description: 
Physical effects of steady-state contact (static loading) that occur at the interfaces 
between materials in the drift may affect the performance of the system. 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
The waste package and the drip shield, as designed and emplaced, come in contact with very few 
other EBS components. For example, the waste package is designed to rest on a pallet, which is 
constructed of Alloy 22 (UNS N06022) and is designed to keep the waste package from 
contacting other dissimilar metals (BSC 2004 [DIRS 168489]). The pallet is also designed to 
keep the waste package supported in a horizontal position, and away from the invert and ground 
support under nonseismic scenarios (BSC 2004 [DIRS 166879], Section 7.2.2). Similarly the 
drip shields are designed to contact no other material except the Alloy 22 feet, which are 
attached to the bottom of the drip shields. These feet are in contact with the invert, which is 
covered by crushed tuff as ballast. 
The possibility of the drip shield contacting the waste package as a result of mechanical damage 
caused by rockfall was considered. This possibility was, however, addressed and excluded as 
discussed in FEP 2.1.03.07.0B, Mechanical Impact on Drip Shield. 
The mechanical loading at the interfaces between the waste package and (degraded) pallet has 
been analyzed (BSC 2004 [DIRS 169766]). The contact stresses are shown to be much less 
(maximum stress intensity ~116 MPa; (BSC 2004 [DIRS 169766], Table 23)) than the stress 
threshold for initiation of stress corrosion cracking (~285 MPa) (BSC 2004 [DIRS 169985], 
Section 6.2.1). On this basis, no enhanced degradation due to mechanical loading at the waste 
packagepallet interfaces is expected. The waste package and drip shield corrosion degradation 
analyses include the effects of material interfaces in the repository on thermal-hydrologicgeochemical 
analyses (e.g., FEP 2.1.09.09.0A, Electrochemical Effects in EBS). These include 
the effects of materials present in the emplacement drift, including the waste package and the 
drip shield, which are described in General Corrosion and Localized Corrosion of Waste 
Package Outer Barrier (BSC 2004 [DIRS 169984]) and General Corrosion and Localized 
Corrosion of the Drip Shield (BSC 2004 [DIRS 169845]). 
This treatment of mechanical effects at EBS component interfaces applies to all waste packages. 
This FEP is excluded based on low consequence to radiological exposures to the RMEI and 
radionuclide releases to the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-44 October 2004 
6.2.23 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 Consequence) (drip shield) 
Excluded (Low Probability) (waste package) 
Screening Argument: 
Seismic-induced rockfall and drift degradation are not treated within this FEP discussion. 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, 
treated in Engineered Barrier System Features, Events, and Processes (BSC 2004 
[DIRS 169898]). 
The emplacement drift system design criteria require the drip shield to protect the waste package 
from rockfalls during postclosure (BSC 2004 [DIRS 170265], Section 3.1.1.4.14.1). Drip Shield 
and Waste Package Emplacement Pallet Design Report (BSC 2004 [DIRS 166879], 
Section 7.2.1), shows that the deflection of the drip shield due to rockfall is not large enough to 
contact the waste package. The drip shield will withstand an 11.5 metric ton rockfall without 
contacting the waste package. The maximum displacement from the 11.5 metric ton rockfall 
event is 254 mm (BSC 2004 [DIRS 166879], Section 7.2.1) and the minimum gap between the 
drip shield and waste package outer barrier is about 367 mm (see Section 6.2.14, 
FEP 2.1.03.07.0B, Mechanical Impact on Drip Shield). Thus, the drip shield provides adequate 
protection to the waste package from rockfall. 
The effects of rockfall on crack initiation in the drip shield are discussed in Stress Corrosion 
Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural 
Material (BSC 2004 [DIRS 169985], Section 6.3.7). These cracks are extremely tight and, with 
time, become plugged with corrosion products and other mineral precipitates (BSC 2004 [DIRS 
169985], Section 6.3.7; FEP 2.1.03.02.0B, Stress Corrosion Cracking (SCC) of Drip Shields). 
This plugging process limits water transport through the drip shield to negligible amounts, and 
maintains the functionality of the drip shield. Therefore, rockfall on drip shield is of low 
consequence to radiological exposures to the RMEI and radionuclide releases to the accessible 
environment. 
This document does not address multiple rockfalls. Bounding effects of multiple rockfalls and 
drift degradation are addressed as part of the seismic consequences in Engineered Barrier System 
Features, Events, and Processes (BSC 2004 [DIRS 169898]). 
Stress fractures in the host rocks and rockfall events may increase the flow of water into the 
repository. Given that the drip shield is capable of withstanding a rockfall event the drip shield 
continues to keep water from falling on the waste package until the drip shield fails. Therefore, 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-45 October 2004 
increased inflow of water related to rockfall is excluded from the TSPA-LA analysis based on 
low consequence to radiological exposures to the RMEI and radionuclide releases to the 
accessible environment. 
Based on the above discussion, the effects of rockfall on the waste package are excluded from 
consideration based on low probability of occurrence (i.e., the probability of rockfall impacting 
the waste package is less than 1 in 10,000 during the first 10,000 years after permanent closure). 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.24 Creep of Metallic Materials in the Waste Package (2.1.07.05.0A) 
FEP Description: 
Metals used in the waste package may deform by creep processes in response to 
deviatoric stress or internal void space. 
Screening Decision: 
Excluded (Low Probability) 
Screening Argument: 
In the repository, the maximum waste package surface temperature will be less than 300C even 
in the low-probability-seismic collapsed-drift scenario (BSC 2004 [DIRS 169565], Figure 6.3- 
56). Adverse elevated-temperature behavior (i.e., creep deformation or creep fracture) of nickelbased 
alloys is not expected at temperatures under 650C (Boyer and Gall 1984 [DIRS 155318], 
Section 32). No directly relevant data exist for Alloy 22 (UNS N06022) in this temperature 
regime; however, the melting temperature of Alloy 22 (UNS N06022) is approximately 1,370C 
(1,643 K) (Haynes International 1988 [DIRS 101995]) and the maximum surface temperature 
(less than 300C or 573 K) is only about 35% of the melting temperature. Creep of Alloy 22 at 
such low temperatures is not expected. Therefore, high temperature creep has a low probability 
of occurrence. This treatment of creep of metallic materials in the waste package applies to all 
waste packages. 
External stress, by rock displacements or ground motion for example, may lead to plastic 
deformations and mechanical damage of the waste package and subsequent leakage of 
radionuclides. The drip shield is designed to protect the waste package during rockfall and 
ground motion events (FEP 2.1.07.05.0B, Creep of Metallic Materials in the Drip Shield). Even 
if mechanical damage were to occur, creep of metallic materials in the waste package will not 
occur unless an external factor raises the temperature of the waste package surface above 650C 
(Boyer and Gall [DIRS 155318], Section 32). In view of the above rationale, this FEP is 
excluded based on low probability of occurrence. 
TSPA Disposition: 
N/A 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-46 October 2004 
Supporting Reports: 
N/A 
6.2.25 Creep of Metallic Materials in the Drip Shield (2.1.07.05.0B) 
FEP Description: 
Metals used in the drip shield may deform by creep processes in response to deviatoric 
stress. 
Screening Decision: 
Excluded (Low Probability) 
Screening Argument: 
In the repository, the maximum waste package surface temperature (because the waste package 
is the heat source, the waste package surface temperature would be greater than the drip shield 
surface temperature) will be less than 300C even in the low-probability-seismic collapsed-drift 
scenario (BSC 2004 [DIRS 169565], Figure 6.3-56). Literature indicates that between 200C 
and 315C (400F and 600F), the deformation of many titanium alloys loaded to yield point 
does not increase with time (ASM International 1990 [DIRS 144385], p. 626). Given that creep 
rates decrease at lower temperatures, creep deformation will not occur to any appreciable extent 
under repository exposure conditions. 
External stress, by rock displacements or ground motion for example, may lead to plastic 
deformations and mechanical damage of the drip shield. Mechanical damage of the drip shield 
by rockfall is discussed in greater detail under FEP 2.1.07.01.0A, Rockfall. This FEP discussion 
also provides relevant references discussing the issue in greater detail. In addition, the 
Emplacement Drift System design criteria require that the drip shield protect the waste package 
from rockfalls during post closure (BSC 2004 [DIRS 170265], Section 3.1.1.4.14.1). 
Mechanical damage of the drip shield during seismic events is discussed 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 in Engineered Barrier System Features, Events, and Processes (BSC 2004 
[DIRS 169898]). 
In view of the above rationale, this FEP is excluded based on low probability of occurrence. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-47 October 2004 
6.2.26 Volume Increase of Corrosion Products Impacts Waste Package (2.1.09.03.0B) 
FEP Description: 
Corrosion products have a higher molar volume than the intact, uncorroded material. 
Increases in volume during waste form, cladding, and waste package corrosion could 
change the stress state in the material being corroded and lead to waste package damage. 
Screening Decision: 
Excluded (Low Probability) 
Screening Argument: 
In general, corrosion products have greater volume than the metal. When the corrosion products 
form in a tightly confined space, their increase in volume generates swelling pressures that could 
lead to mechanical damage of the surrounding material. Since the current design precludes the 
use of shrink fitting the outer and inner cylinder components, mechanical damage to the Alloy 22 
(UNS N06022) waste package outer barrier due to the pressure exerted by the corrosion product 
(Cr2O3) of the inner shell (Stainless Steel Type 316) will not occur. Current waste package 
designs require the radial gap between the inner vessel and the outer barrier to be at least 1 mm 
and could be a maximum of 5 mm (BSC 2004 [DIRS 169480]). Analyses cited in General 
Corrosion and Localized Corrosion of Waste Package Outer Barrier (BSC 2004 
[DIRS 169985], Section 6.4.2), indicate that for chromia (Cr2O3) scale-forming alloys 
(e.g., Alloy 22 and Stainless Steel Type 316), even under very conservative assumptions, the 
growth of corrosion product will not exceed 93 m after 10,000 years. This oxide layer is not 
thick enough to produce enough pressure to cause mechanical damage to the waste package outer 
barrier or inner vessel. In the current design of waste package and engineered barrier system in 
the emplacement drift (BSC 2004 [DIRS 168489]), there is no possibility of forming such a 
tightly confined space such that the swelling corrosion products could cause mechanical damage 
to the Alloy 22 (UNS N06022) outer barrier. Therefore, waste package damage from swelling 
corrosion products is excluded based on low probability of occurrence under the exposure 
conditions in the repository. 
A related FEP (2.1.09.03.0A, Volume increase of corrosion products impacts cladding) is 
discussed in Clad Degradation - FEPs Screening Arguments (BSC 2004 [DIRS 170019]). 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.27 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. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-48 October 2004 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
Due to the large distances involved, it is reasonable to consider electrochemical effects between 
materials in the drift and more distant metallic materials in the natural system to be less 
important to waste package materials degradation than the electrochemical interactions between 
materials or components within the drift. Such long-range interactions are more appropriate for 
consideration in modeling processes, such as radionuclide transport away from the repository, 
rather than in consideration of relatively local phenomena, such as waste package or drip shield 
degradation. 
The waste package and the drip shield, as designed and emplaced, come in contact with very few 
other EBS components. For example, the waste package is designed to rest on a pallet 
constructed of Alloy 22 (UNS N06022), and is designed to keep the waste package from 
contacting other dissimilar metals (BSC 2004 [DIRS 168489]). The pallet is also designed to 
keep the waste package supported in a horizontal position away from the invert and ground 
support under nonseismic scenarios (BSC 2004 [DIRS 166879], Section 7.2.2). Similarly, the 
drip shields are designed to contact no other material except the Alloy 22 feet, which are 
attached to the bottom of the drip shields. These feet are in contact with the invert, which is 
covered by crushed tuff as ballast. 
The TSPA-LA waste package design (BSC 2004 [DIRS 168489]) includes an outer barrier of 
Alloy 22 (UNS N06022) over a Stainless Steel Type 316 inner vessel. In addition, a titanium 
drip shield is placed over the waste packages. Although the Stainless Steel Type 316 inner 
vessel provides structural stability to the Alloy 22 (UNS N06022) outer barrier, no other 
performance credit is taken for the waste package inner vessel. The corrosion potentials of 
Alloy 22 (UNS N06022) and Stainless Steel Type 316 are very close to each other under similar 
exposure conditions (e.g., ASM International [DIRS 103753], p. 557, Figure 10) with Alloy 22 
(UNS N06022) slightly more noble than Stainless Steel Type 316. After breach of the Alloy 22 
(UNS N06022) waste package outer barrier, electrochemical coupling of the Alloy 22 
(UNS N06022) waste package outer barrier with the Stainless Steel Type 316 waste package 
inner vessel could occur. Due to the similarity in corrosion potential of Alloy 22 (UNS N06022) 
and Stainless Steel Type 316, any enhanced degradation of either material due to galvanic 
interaction would be negligible. Therefore, electrochemical coupling of the Alloy 22 
(UNS N06022) waste package outer barrier and the Stainless Steel Type 316 waste package 
inner vessel is of low consequence to radiological exposures to the RMEI and radionuclide 
releases to the accessible environment. 
The possibility of the drip shield contacting the waste package as a result of mechanical damage 
caused by rockfall was considered. This possibility was, however, addressed and excluded as 
discussed in FEP 2.1.03.07.0B, Mechanical Impact on Drip Shield. The only contact between 
Titanium Grade 7 and Alloy 22 (UNS N06022) occurs at the bottom of the drip shields where 
Alloy 22 (UNS N06022) feet are attached to prevent contact between titanium and the invert. 
The choice of Alloy 22 (UNS N06022) for the feet was based on similarity of the two materials 
in the electrochemical series (ASM International 1987 [DIRS 103753], p. 557, Figure 10), which 
indicates any enhanced degradation of either material due to galvanic interaction would be 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-49 October 2004 
negligible. Therefore, electrochemical coupling of the Alloy 22 (UNS N06022) waste package 
outer barrier and the Titanium Grade 7 drip shield is of low consequence to radiological 
exposures to the RMEI and radionuclide releases to the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.28 Thermal Sensitization of Waste Packages (2.1.11.06.0A) 
FEP Description: 
Phase changes in waste package materials can result from long-term storage at 
moderately hot temperatures in the repository. Stress corrosion cracking (SCC), 
intergranular corrosion, or mechanical degradation may ensue. 
Screening Decision: 
Excluded (Low Probability) 
Screening Argument: 
Alloy 22 (UNS N06022) could to be subject to aging and phase instability when exposed to 
elevated temperatures that are well above the anticipated maximums expected in the repository 
(BSC 2004 [DIRS 171924], Section 6.3). The processes involve precipitation of different 
secondary phases and restructuring of the microstructure. The affected material exhibits 
increased brittleness and decreased resistance to corrosion, especially to localized corrosion and 
SCC (BSC 2004 [DIRS 171924], Section 1.1). 
Before waste loading, the waste packages (base-metal and fabrication welds) are fully solution 
annealed (Plinski 2001 [DIRS 156800], Section 8.1.7). After waste loading, the closure lids are 
welded onto the waste package (Plinski 2001 [DIRS 156800], Section 8.1.8). No effects of 
aging and phase instability on the Alloy 22 waste package outer barrier would be observed even 
if the waste package were maintained at less than 300C for a period of 500 years followed by 
temperatures less than 200C for a period of 9,500 years (BSC 2004 [DIRS 171924], Section 8; 
BSC 2004 [DIRS 169984], Section 6.4.6). These thermal exposure conditions bound all 
repository-relevant thermal exposure conditions (BSC 2004 [DIRS 171924], Section 8). 
The closure lid welds cannot be solution annealed without risking damage to the waste form. 
Therefore, the closure welds of the waste package outer barrier could be more prone to thermal 
aging and phase instability than the base metal under long-term thermal exposure in the 
repository (BSC 2004 [DIRS 169984], Section 6.4.6). Analyses conducted in General Corrosion 
and Localized Corrosion of Waste Package Outer Barrier (BSC 2004 [DIRS 169984], 
Section 6.4.6) studied the effect of thermal aging on corrosion of Alloy 22. Three metallurgical 
conditions of Alloy 22 (UNS N06022) were studied using the multiple crevice assembly 
samples: mill annealed, as-welded, and as-welded plus thermally aged (at 700C for 173 hours). 
The samples were tested in 5 M CaCl2 solutions with and without Ca(NO3)2 additions (up 
to 0.5 m) at test temperatures varying from 45C to 120C. Comparison of the calculated 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-50 October 2004 
corrosion rates of the mill annealed, as-welded, and as-welded plus thermally aged samples 
showed no apparent enhancement of the corrosion rate due to the presence of welds or thermal 
aging of the welded samples for the tested conditions. 
Based on this analysis, insignificant aging and phase stability will occur under the thermal 
conditions expected in the repository (BSC 2004 [DIRS 169984], Section 6.4.6) and the 
corrosion performance of the waste package outer barrier material is not expected to be affected 
by aging and phase stability in the repository. This treatment of thermal sensitization of waste 
packages applies to all waste packages. Thermal sensitization of waste packages is excluded on 
the basis of low probability of occurrence. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.29 Thermal Sensitization of Drip Shields (2.1.11.06.0B) 
FEP Description: 
Phase changes in drip shield materials can result from long-term storage at moderately 
hot temperatures in the repository. Stress corrosion cracking (SCC), intergranular 
corrosion, or mechanical degradation may ensue. 
Screening Decision: 
Excluded (Low Probability) 
Screening Argument: 
Aging and phase stability of the drip shield is considered in Section 6.5.3 of General Corrosion 
and Localized Corrosion of the Drip Shield (BSC 2004 [DIRS 169845]) that states 
Titanium Grade 7 is a stabilized alpha (a) phase alloy and possesses outstanding phase stability. 
While Titanium Grade 7 contains small amounts of alloying elements 
(DTN: MO0003RIB00073.000 [DIRS 152926]), most notably palladium, it is essentially a pure 
titanium alloy that has no capability of forming intermetallic compounds under the thermal 
exposure conditions in the repository. 
The solubility of palladium in Titanium Grade 7 is about 1 weight percent at 400C. The 
nominal concentration of palladium in Titanium Grade 7 is well below the solubility limit at this 
temperature (Gdowski 1997 [DIRS 102789], pp. 1 to 8). Titaniumpalladium intermetallic 
compounds capable of being formed in the Ti-Pd system have not been reported to occur in 
Titanium Grade 7 with normal heat treatments. Hua et al. (2002 [DIRS 160670]) tested both the 
base metal and welded metal of Titanium Grade 7 in a highly concentrated basic environment at 
60C, 70C, 80C, 90C, 100C, and 105C for up to eight weeks (Hua et al. 2002 
[DIRS 160670]; Hua and Gordon 2003 [DIRS 163111]). No difference in weight loss and, 
therefore, in corrosion rate was observed between the base metal and welds. The boundaries 
between the welds and heat-affected zone and between the heat-affected zone and base metal 
were not visibly attacked. Therefore, based on the above experimental evidence, thermal 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-51 October 2004 
sensitization of the drip shield can be excluded based on low probability of occurrence. Also, 
because thermal sensitization of the drip shield will not occur, stress corrosion cracking (SCC), 
intergranular corrosion, or mechanical degradation will not ensue due to thermal sensitization of 
the drip shield. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.30 Thermal Expansion/Stress of In-Drift EBS Components (2.1.11.07.0A) 
FEP Description: 
Repository heat at Yucca Mountain could result in thermally induced stress changes that 
would affect the mechanical and chemical evolution of the repository. These stress 
changes could affect the EBS components, thus causing the formation of pathways for 
groundwater flow through the EBS or altering and/or enhancing existing pathways. 
Relevant processes include changes in physical properties of the drip shields, waste 
packages, pallet, and invert. 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
The coefficient of thermal expansion for Stainless Steel Type 316L (an analogue for the Stainless 
Steel Type 316 used for the waste package inner shell) is larger than the coefficient of thermal 
expansion for Alloy 22 (UNS N06022). Thus, changes in temperature could lead to contact 
stresses between the waste package barriers. In Waste Package Outer Barrier Stresses Due to 
Thermal Expansion with Various Barrier Gap Sizes (BSC 2001 [DIRS 152655]), the maximum 
tangential stress at the waste package outer barrier inner and outer surfaces were evaluated for 
several waste package types (21-PWR, 44-BWR, 12-PWR Long, 5 DHLW/DOE SNF  Short, 
2-MCO/2-DHLW, and naval SNF Long) as a function of temperature and barrier gap size 
(difference in radius of the two barriers evaluated at room temperature (BSC 2001 
[DIRS 152655], Section 5.3). An earlier calculation (BSC 2001 [DIRS 154004]) using a barrier 
gap size of zero, showed that under thermal expansion loading, tangential stresses are 
significantly higher than radial stresses (BSC 2001 [DIRS 152655], Section 1.0). The conclusion 
of these studies was that a barrier gap size of at least 1 mm would result in no tangential stresses 
due to thermal expansion. Current waste package designs require the radial gap between the 
waste package inner vessel and the waste package outer barrier to be at least 1 mm and could be 
a maximum of 5 mm (BSC 2004 [DIRS 169480]). 
Waste Package Operation Fabrication Process Report (Plinski 2001 [DIRS 156800], 
Section 8.1.8) requires a loose fit between the outer barrier (Alloy 22 [UNS N06022]) and the 
inner vessel (Stainless Steel Type 316) to accommodate the differing thermal expansion 
coefficients. Typical waste package designs also require large longitudinal barrier gaps 
(~30 mm) (BSC 2004 [DIRS 169472], Table 1A). Therefore, although thermal expansion of 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-52 October 2004 
waste package components occurs, no significant stresses due to differing thermal expansion 
between the barriers develop. This FEP is excluded for the waste packages based on low 
consequence to radiological exposures to the RMEI and radionuclide releases to the accessible 
environment. 
In the LA drip shield design (BSC 2003 [DIRS 161519]), the drip shield connectors are designed 
to allow for thermal expansion with no effect on drip shield performance. The drip shield 
segments are interlocked with a significant amount of freedom to expand and still maintain their 
intended purpose. The space between the drip shield and waste package is large enough to 
accommodate deflection due to rockfall (see FEP 2.1.03.07.0B, Mechanical Impact on Drip 
Shield). The space required for thermal expansion is very small by comparison. Therefore, this 
FEP can be excluded for the drip shields based on low consequence to radiological exposures to 
the RMEI and radionuclide releases to the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.31 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: 
The materials selected for waste package outer barrier and the drip shield are highly corrosion 
resistant materials (BSC 2004 [DIRS 169845], Section 6; BSC 2004 [DIRS 169984], Section 6). 
These form a thin, highly protective oxide layer (passive film) that protects the materials from 
further corrosion. As a result, limited corrosion of these materials leads to negligible gas 
generation. In addition, a drift in the unsaturated zone of the repository is expected to be 
connected to the atmosphere and to be operating under oxidizing conditions. Therefore, any 
gases generated by metal corrosion would escape from the drifts. 
Hydrogen generated at cathodic sites in a corroding metal may migrate into the metal and 
potentially form hydride phases. Hydrogen incorporation could lead to degradation of the 
mechanical properties of the material and render it susceptible to cracking even in the absence of 
the formation of hydride phases. Hydrogen-induced cracking (HIC) is one of the results that 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-53 October 2004 
may happen from the combined action of hydrogen and residual or sustained applied tensile 
stresses. 
As discussed in FEP 2.1.03.04.0A, Hydride Cracking of Waste Packages, HIC of the waste 
package outer barrier (Alloy 22 (UNS N06022)) is not considered to be an effective degradation 
mechanism under repository-relevant exposure conditions based on analyses of handbook data 
(ASM International. 1987 [DIRS 103753], pp. 169, 557, 650-652). As discussed in 
FEP 2.1.03.04.0B, Hydride Cracking of Drip Shields, HIC of the drip shields is not considered to 
be an effective degradation mechanism under repository-relevant exposure conditions based on 
analyses of handbook data (Schutz and Thomas 1987 [DIRS 144302]) and Project analyses 
(BSC 2004 [DIRS 169847], Sections 6.1.2, 6.1.3, 6.1.4, 6.1.5, 6.3.2). This treatment of gas 
generation from waste package corrosion applies to all waste packages. Overall, even though 
gas generation is possible, it is of low consequence to radiological exposures to the RMEI and 
radionuclide releases to the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 
6.2.32 Radiolysis (2.1.13.01.0A) 
FEP Description: 
Alpha, beta, gamma, and neutron irradiation of water can cause disassociation of 
molecules, leading to gas production and changes in chemical conditions (potential, pH, 
and concentration of reactive radicals). 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
Alpha, beta, gamma, and neutron irradiation of water leads to formation of highly reactive 
excited and ionized species, which in turn can undergo various reactions. In pure water, the final 
products are hydrogen and oxidants. This section addresses only the effects of radiolysis of 
water on degradation of the waste package outer barrier. Radiolysis effects on waste form are 
addressed in Waste-Form Features, Events, and Processes (BSC 2004 [DIRS 170020]). 
Chemical effects from radiolysis are addressed in Engineered Barrier System Features, Events, 
and Processes (BSC 2004 [DIRS 169898]). Gamma radiation is the dominant contributor to 
dose rate at the waste package surface because the neutron dose rate is relatively small by 
comparison (BSC 2004 [DIRS 169593], Section 6.3) and alpha and beta radiation will not 
penetrate the container. Analogous to the discussion in FEP 2.1.12.03.0A, Gas Generation (H2) 
from Waste Package Corrosion, a drift in the unsaturated zone of the Yucca Mountain repository 
is expected to be operating under oxidizing conditions and will be connected to the atmosphere 
allowing any gases generated by radiolysis to escape from the drifts (Shoesmith and King 1998 
[DIRS 112178], p. 5). The effects of radiation on the corrosion of the waste package materials 
differ depending on the amount of liquid present on their surfaces (i.e., humid air or aqueous 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-54 October 2004 
conditions). Under humid air conditions, a thin film of liquid forms possibly containing trace 
constituents (e.g., dissolved gases). Irradiation of these films could lead to acidic conditions and 
to enhanced corrosion rates. Under aqueous conditions (bulk solutions), anodic shifts in the open 
circuit potential of stainless steel in gamma-irradiated solutions have been experimentally 
observed (BSC 2004 [DIRS 169845], Section 6.7.1). These shifts in potential have been shown 
to be due to the formation of hydrogen peroxide (BSC 2004 [DIRS 169845], Section 6.7.1). 
Calculations of the expected radiation levels at the surface of the waste package have been 
performed in (BSC 2004 [DIRS 169593], Section 6.4.2). For a bounding-case waste package 
containing 21-PWR spent fuel assemblies (80 GWd/MTU burnup, and 5-year decay), the 
maximum surface radiation level was calculated to be about 1,170 rad/hr at the waste package 
outer barrier surface and 1,650 rad/hr at the outer lid of the waste package (BSC 2004 
[DIRS 169593], Table 6.4-3). During the 50-year ventilation period, an aqueous or humid air 
environment will not exist and, as a result, radiolysis is not anticipated. After 50 years, the 
maximum surface radiation level decreases to about 100 rad/hr for the waste package outer 
barrier surface and 80 rad/hr for the outer lid (BSC 2004 [DIRS 169593], Figure 6.4-1). One 
hundred years after emplacement, the maximum calculated levels reduce to about 35 rad/hr for 
the waste package outer barrier surface and 20 rad/hr for the outer lid region (BSC 2004 
[DIRS 169593], Figure 6.4-1). The bounding radiation levels for the highest burnup spent fuel 
are well below the levels at which the effect of radiation has been observed. 
Although there is little information available in the literature on the effects of radiation on 
Alloy 22 (UNS N06022), some data are available on the corrosion of Alloy C-4, which is 
compositionally similar to Alloy 22. Gamma irradiation in aggressive MgCl2 brines showed that 
below ~100 rad/hour (Shoesmith and King 1998 [DIRS 112178], p. 29) irradiation has no 
observable influence on the corrosion behavior of Alloy C-4. In this same environment, it was 
found, even at dose rates above 1,000 rad/hr, only a minor enhancement of film growth rates on 
Titanium Grade 7 was observed and passivity was not threatened (Shoesmith and King 1998 
[DIRS 112178], p. 30). Based on this data it is concluded that, even in aggressive MgCl2 brines, 
radiation levels in the repository are not high enough to result in an enhancement of corrosion 
processes on Alloy 22 (UNS N06022) or Titanium Grade 7. On this basis, radiolysis is excluded 
based on low consequence to radiological exposures to the RMEI and radionuclide releases to 
the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-55 October 2004 
6.2.33 Radiation Damage in EBS (2.1.13.02.0A) 
FEP Description: 
Radiolysis due to the alpha, beta, gamma-ray, and neutron irradiation of water could 
result in enhancement of the radionuclide migration from the surface of a degraded waste 
form into groundwater. When radionuclides decay, the emitted high-energy particle 
could result in the production of radicals in the water or air surrounding the spent nuclear 
fuel. If these radicals migrate (diffuse) to the surface of the fuel, they may then enhance 
the degradation/corrosion rate of the fuel (UO2). This effect would increase the 
dissolution rate for radionuclides from the fuel material (fuel matrix) into the 
groundwater. Strong radiation fields could lead to radiation damage to the waste forms 
(CSNF, DSNF, DHLW), waste packages, drip shields, seals, and surrounding rock. 
Screening Decision: 
Excluded (Low Consequence) 
Screening Argument: 
The enhanced degradation or corrosion of the fuel is covered in Waste-Form Features, Events, 
and Processes (BSC 2004 [DIRS 170020]). The cumulative neutron fluence at the waste 
package emplacement pallet as a function of years after emplacement is determined in Dose Rate 
Calculation for 21-PWR Waste Package (BSC 2004 [DIRS 169593], Section 6.4.1). The 
calculation uses the design-basis source term (60 GWd/MTU burnup, 10-year decay time) 
(BSC 2004 [DIRS 169593], Section 5.1.4). At 290 years and 340 years after emplacement (the 
two longest times reported), the cumulative neutron fluence at the emplacement pallet location 
was determined to be 3.78  1014 n/cm2 and 3.90  1014 n/cm2, respectively (where n stands for 
neutrons) (BSC 2004 [DIRS 169593], Table 6.4-1). The cumulative neutron fluence at the 
emplacement pallet location can be approximated to be increasing at a rate of 
(1.2  1013)/50 n/cm2/yr. Using this approximation, the cumulative neutron fluence at the 
emplacement pallet location after 10,000 years would be about 3  1015 n/cm2. Due to the close 
proximity of the waste package and the waste package emplacement pallet, it is reasonable to 
equate the cumulative neutron fluence at the emplacement pallet location to the cumulative 
neutron fluence at the waste package location. Nuclear Engineering Handbook 
(Etherington 1958 [DIRS 164789], p. 10-107) identifies neutron fluence levels below which 
there is no change in the mechanical properties of Stainless Steel Type 316 (5  1019 n/cm2), 
nickel (1020 n/cm2), and molybdenum (1020 n/cm2). Because Alloy 22 is a nickel-basedmolybdenum-
containing alloy, there is no reason that the neutron fluence threshold for Alloy 22 
should be more than four orders of magnitude lower than for these materials (as it would have to 
be in order for the estimated neutron fluence to exceed the neutron fluence threshold for changes 
in mechanical properties). Therefore, while it is likely that some radiation damage occurs, 
because the estimated neutron fluence is below the neutron fluence threshold by a considerable 
margin, this FEP is excluded based on low consequence to radiological exposures to the RMEI 
and radionuclide releases to the accessible environment. 
Furthermore, the Waste Package Materials Performance Peer Review Panel addressed the 
possibility of radiation damage in the repository in their final report (Beavers et al. 2002 
[DIRS 158781], Section 3.10). They stated that the waste package will be subjected to a flux of 
neutrons and gamma rays from the stored radioactive waste (Beavers et al. 2002 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 6-56 October 2004 
[DIRS 158781]). These fluxes could cause the following: 1) neutrons could produce atomic 
displacement damage in the metal, 2) neutrons could produce atomic displacement damage and 
gamma rays will cause electron-hole pairs in the passive film, and 3) gamma rays could cause 
radiolysis of the surrounding environment (Beavers et al. 2002 [DIRS 158781], Section 3.10). 
The peak neutron flux has been calculated to be about 5  104 n/cm2sec) in the repository 
environment (Beavers et al. 2002 [DIRS 158781], Section 3.10). The total neutron fluence, 
taking the most conservative estimate with no nuclear decay of the waste, will be 
about 1.5  1016 n/cm2 in 10,000 years in the repository environment (Beavers et al. 2002 
[DIRS 158781], Section 3.10). The report concluded that there is no evidence to suggest that 
radiation damage to the waste package will alter its mechanical properties (Beavers et al. 2002 
[DIRS 158781], Section 3.10). The drip shield is located farther away from the source of 
radiation (the waste form) than is the waste package barrier. On this basis, the drip shield 
material will be subject to less radiation damage than the waste package (i.e., radiation damage is 
of even less consequence to drip shield performance than it is to waste package performance). 
Therefore, the only potential effect of radiation will be the change in external environment due to 
groundwater radiolysis (e.g., ASM International 1987 [DIRS 103753], pp. 971 to 974), which is 
addressed in FEP 2.1.13.01.0A, Radiolysis. 
Based on the above rationale, this FEP is excluded based on low consequence to radiological 
exposures to the RMEI and radionuclide releases to the accessible environment. 
TSPA Disposition: 
N/A 
Supporting Reports: 
N/A 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 7-1 October 2004 
7. CONCLUSIONS 
The analyses documented in this report are for the TSPA-LA design in which a drip shield is 
placed over the waste package (BSC 2004 [DIRS 168489]). 
Thirty-three FEPs relevant to waste package and drip shield degradation processes were screened 
against criteria presented in Section 4.2.2. The results of this screening process are documented 
in the screening arguments and the TSPA disposition statements in Section 6.2. The screening 
basis for each FEP is summarized in Table 7-1. This table shows the FEP number, name, 
screening decision (included or excluded in TSPA-LA), and basis for the screening decision 
(i.e., low probability of occurrence or low consequence to radiological exposures to the RMEI 
and radionuclide releases to the accessible environment). 
Table 7-1. Summary of Waste Package FEPs 
FEP Number FEP Name Screening Decision and Basis Addressed in 
Section 
1.1.03.01.0A Error in Waste Emplacement Excluded (Low Consequence) 6.2.1 
2.1.03.01.0A General Corrosion of Waste Packages Included 6.2.2 
2.1.03.01.0B General Corrosion of Drip Shields Included 6.2.3 
2.1.03.02.0A Stress Corrosion Cracking (SCC) of Waste 
Packages Included 6.2.4 
2.1.03.02.0B Stress Corrosion Cracking (SCC) of Drip 
Shields Excluded (Low Consequence) 6.2.5 
2.1.03.03.0A Localized Corrosion of Waste Packages Included 6.2.6 
2.1.03.03.0B Localized Corrosion of Drip Shields Excluded (Low Consequence) 6.2.7 
2.1.03.04.0A Hydride Cracking of Waste Packages Excluded (Low Consequence) 6.2.8 
2.1.03.04.0B Hydride Cracking of Drip Shields Excluded (Low Consequence) 6.2.9 
2.1.03.05.0A Microbially Influenced Corrosion (MIC) of 
Waste Package Included 6.2.10 
2.1.03.05.0B Microbially Influenced Corrosion (MIC) of Drip 
Shields Excluded (Low Consequence) 6.2.11 
2.1.03.06.0A Internal Corrosion of Waste Packages prior to 
breach Excluded (Low Consequence) 6.2.12 
2.1.03.07.0A Mechanical Impact on Waste Package Excluded (Low Consequence 
and Low Probability) 
6.2.13 
2.1.03.07.0B Mechanical Impact on Drip Shield Excluded (Low Consequence) 6.2.14 
2.1.03.08.0A Early failure of Waste Packages Included 6.2.15 
2.1.03.08.0B Early failure of Drip Shields Excluded (Low Consequence) 6.2.16 
2.1.03.09.0A Copper corrosion in EBS Excluded (Low Consequence) 6.2.17 
2.1.03.10.0A Advection of liquids and solids through cracks 
in the waste package Excluded (Low Consequence) 6.2.18 
2.1.03.10.0B Advection of liquids and solids through cracks 
in the drip shield Excluded (Low Consequence) 6.2.19 
2.1.03.11.0A Physical form of Waste Package and Drip 
Shield Included 6.2.20 
2.1.06.06.0B Oxygen embrittlement of Drip Shields Excluded (Low Probability) 6.2.21 
2.1.06.07.0B Mechanical effects at EBS component 
interfaces Excluded (Low Consequence) 6.2.22 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 7-2 October 2004 
Table 7-1. Summary of Waste Package FEPs (Continued) 
FEP Number FEP Name Screening Decision and Basis Addressed in 
Section 
2.1.07.01.0A Rockfall Excluded (Low Consequence) 
and Excluded (Low Probability) 6.2.23 
2.1.07.05.0A Creep of metallic materials in the Waste 
Package Excluded (Low Probability) 6.2.24 
2.1.07.05.0B Creep of metallic materials in the Drip Shield Excluded (Low Probability) 6.2.25 
2.1.09.03.0B Volume increase of Corrosion products 
impacts Waste Package Excluded (Low Probability) 6.2.26 
2.1.09.09.0A Electrochemical effects in EBS Excluded (Low Consequence) 6.2.27 
2.1.11.06.0A Thermal sensitization of Waste Packages Excluded (Low Probability) 6.2.28 
2.1.11.06.0B Thermal sensitization of Drip Shields Excluded (Low Probability) 6.2.29 
2.1.11.07.0A Thermal expansion/stress of In-Drift EBS 
components Excluded (Low Consequence) 6.2.30 
2.1.12.03.0A Gas generation (H2) from Waste Package 
corrosion Excluded (Low Consequence) 6.2.31 
2.1.13.01.0A Radiolysis Excluded (Low Consequence) 6.2.32 
2.1.13.02.0A Radiation damage in EBS Excluded (Low Consequence) 6.2.33 
The conclusions from this document (FEP screening decision, TSPA disposition for included 
FEPs, or screening argument for excluded FEPs), along with any modifications to the FEP list, 
FEP names, and/or FEP 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 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 waste 
package and drip shield 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 is 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 (NRC 2003 
[DIRS 163274], Section 2.2.1.2.1.3) were addressed in this report: 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 7-3 October 2004 
Acceptance Criterion 1 The Identification of a List of Features, Events, and Processes 
Is Adequate. 
(1) The Safety Analysis Report contains a complete list of features, events, and 
processes, related to the geologic setting or the degradation, deterioration, or 
alteration of engineered barriers (including those processes that would affect the 
performance of natural barriers), that have the potential to influence repository 
performance. The list is consistent with the site characterization data. Moreover, 
the comprehensive features, events, and processes list includes, but is not limited 
to, potentially disruptive events related to igneous activity (extrusive and 
intrusive); seismic shaking (high-frequency-low magnitude, and rare largemagnitude 
events); tectonic evolution (slip on existing faults and formation of new 
faults); climatic change (change to pluvial conditions); and criticality. 
Documentation of the origin of the FEPs list is provided in Section 6.1.1; FEP descriptions are 
provided in Section 6.2. This analysis contains a list of waste package and drip shield-related 
FEPs (Table 1-1). This list of FEPs includes those related to the degradation, deterioration, or 
alteration of engineered barriers. The list of waste package and drip shield-related FEPs is 
consistent with the site characterization data. 
Acceptance Criterion 2 Screening of the List of Features, Events, and Processes Is Appropriate. 
(1) The U.S. Department of Energy has identified all features, events, and 
processes related to either the geologic setting or to the degradation, 
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. Table 7-1 provides a 
list of excluded waste package and drip shield-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 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 7-4 October 2004 
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. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 8-1 October 2004 
8. INPUTS AND REFERENCES 
8.1 DOCUMENTS CITED 
144385 ASM International 1990. Properties and Selection: Nonferrous Alloys and Special- 
Purpose Materials, Specific Metals and Alloys. Volume 2 of Metals Handbook. 10th 
Edition. Page 666. Materials Park, Ohio: American Society for Metals. TIC: 239807. 
103753 ASM International. 1987. Corrosion. Volume 13 of Metals Handbook. 9th Edition. 
Metals Park, Ohio: ASM International. TIC: 209807. 
160352 Asphahani, A.I. 1978. Hydrogen Cracking of Nickel-Base Alloys. Hydrogen in 
Metals, Proceedings of the 2nd International Congress, Paris, France, 6-10 June, 
1977. 1, Paper 3C2. New York, New York: Pergamon. ACC: NNA.19910419.0006. 
158781 Beavers, J.A.; Devine, T.M., Jr.; Frankel, G.S.; Jones, R.H.; Kelly, R.G.; Latanision, 
R.M.; and Payer, J.H. 2002. Final Report, Waste Package Materials Performance 
Peer Review Panel, February 28, 2002. Las Vegas, Nevada: Waste Package 
Materials Performance Peer Review Panel. ACC: MOL.20020614.0035. 
155318 Boyer, H.E. and Gall, T.L., eds. 1984. Metals Handbook. Desk Edition. 10th Printing 
1997. Metals Park, Ohio: American Society for Metals. TIC: 250192. 
159836 Brossia, C.S.; Browning, L.; Dunn, D.S.; Moghissi, O.C.; Pensado, O.; and Yang, L. 
2001. Effect of Environment on the Corrosion of Waste Package and Drip Shield 
Materials. CNWRA 2001-003. San Antonio, Texas: Center for Nuclear Waste 
Regulatory Analyses. TIC: 252324. 
154441 BSC (Bechtel SAIC Company) 2001. Committed Materials in Repository Drifts. 
CAL-GCS-GE-000002 REV 00. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: MOL.20010412.0157. 
152248 BSC 2001. Flow of Water and Pooling in a Waste Package. ANL-EBS-MD-000055 
REV 00. Las Vegas, Nevada: Bechtel SAIC Company. ACC: MOL.20010604.0218. 
156807 BSC 2001. Plugging of Stress Corrosion Cracks by Precipitates. CAL-EBS-MD- 
000017 REV 00. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: MOL.20011010.0168. 
154004 BSC 2001. Waste Package Barrier Stresses Due to Thermal Expansion. CAL-EBSME-
000008 REV 00. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: MOL.20011211.0149. 
152655 BSC 2001. Waste Package Outer Barrier Stress Due to Thermal Expansion with 
Various Barrier Gap Sizes. CAL-EBS-ME-000011 REV 00. Las Vegas, 
Nevada: Bechtel SAIC Company. ACC: MOL.20011212.0222. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 8-2 October 2004 
161519 BSC 2003. Interlocking Drip Shield. 000-MW0-TED0-00101-000-00A. Las Vegas, 
Nevada: Bechtel SAIC Company. ACC: ENG.20030205.0001. 
171924 BSC 2004. Aging and Phase Stability of Waste Package Outer Barrier. ANL-EBSMD-
000002 REV 02. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: DOC.20041005.0003. 
170024 BSC 2004. Analysis of Mechanisms for Early Waste Package/Drip Shield Failure. 
CAL-EBS-MD-000030 REV 00C. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: DOC.20040913.0006. 
168030 BSC 2004. Characterize Framework for Seismicity and Structural Deformation at 
Yucca Mountain, Nevada. ANL-CRW-GS-000003 REV 00, with 1 errata. Las Vegas, 
Nevada: Bechtel SAIC Company. ACC: MOL.20000510.0175; 
DOC.20040223.0007. 
170019 BSC 2004. Clad Degradation  FEPs Screening Arguments. ANL-WIS-MD-000008, 
Rev. 02. Las Vegas, Nevada: Bechtel SAIC Company. 
169766 BSC 2004. Commercial SNF Waste Package Design Report. 000-00C-DSU0-02800- 
000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040709.0001. 
168489 BSC 2004. D&E / PA/C IED Emplacement Drift Configuration and Environment. 
800-IED-MGR0-00201-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: ENG.20040326.0001. 
169480 BSC 2004. D&E / PA/C IED Typical Waste Package Components Assembly. 800- 
IED-WIS0-00201-000-00E. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: ENG.20040517.0007. 
169472 BSC 2004. D&E/PA/C IED Typical Waste Package Components Assembly. 800-IEDWIS0-
00202-000-00C. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: ENG.20040517.0008. 
170803 BSC 2004. DOE and Commercial Waste Package System Description Document. 
000-3YD-DS00-00100-000-004. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: ENG.20040720.0009. 
169593 BSC 2004. Dose Rate Calculation for 21-PWR Waste Package. 000-00C-DSU0- 
01800-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: ENG.20040607.0001. 
166879 BSC 2004. Drip Shield and Waste Package Emplacement Pallet Design Report. 000- 
00C-SSE0-00100-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: ENG.20040225.0016. 
169868 BSC 2004. EBS Radionuclide Transport Abstraction. ANL-WIS-PA-000001, Rev. 
01. Las Vegas, Nevada: Bechtel SAIC Company. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 8-3 October 2004 
169898 BSC 2004. Engineered Barrier System Features, Events, and Processes. ANL-WISPA-
000002, Rev. 03. Las Vegas, Nevada: Bechtel SAIC Company. 
170017 BSC 2004. Features, Events, and Processes: Disruptive Events. ANL-WIS-MD- 
000005, Rev. 02. Las Vegas, Nevada: Bechtel SAIC Company. 
169845 BSC 2004. General Corrosion and Localized Corrosion of the Drip Shield. ANLEBS-
MD-000004 REV 02. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: DOC.20040921.0002. 
169984 BSC 2004. General Corrosion and Localized Corrosion of Waste Package Outer 
Barrier. ANL-EBS-MD-000003 REV 02. Las Vegas, Nevada: Bechtel SAIC 
Company. ACC: DOC.20041004.0001. 
169847 BSC 2004. Hydrogen-Induced Cracking of the Drip Shield. ANL-EBS-MD-000006 
REV 02. Las Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20040909.0004. 
169565 BSC 2004. Multiscale Thermohydrologic Model. ANL-EBS-MD-000049, Rev. 02. 
Las Vegas, Nevada: Bechtel SAIC Company. 
168556 BSC 2004. Screening Analysis for Criticality Features, Events, and Processes for 
License Application. ANL-EBS-NU-000008, Rev. 01. Las Vegas, Nevada: Bechtel 
SAIC Company. 
169985 BSC 2004. Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer 
Barrier, and the Stainless Steel Structural Material. ANL-EBS-MD-000005, Rev. 02. 
Las Vegas, Nevada: Bechtel SAIC Company. 
170265 BSC 2004. Subsurface Facility Description Document. 800-3YD-MGR0-00100-000- 
004. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040630.0006. 
171583 BSC 2004. Technical Work Plan For: Regulatory Integration Modeling and Analysis 
of the Waste Form and Waste Package. TWP-WIS-MD-000009 REV 00 ICN 01. Las 
Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20040910.0001. 
168706 BSC 2004. The Development of the Total System Performance Assessment-License 
Application Features, Events, and Processes. TDR-WIS-MD-000003, Rev. 01. Las 
Vegas, Nevada: Bechtel SAIC Company. 
169996 BSC 2004. WAPDEG Analysis of Waste Package and Drip Shield Degradation. 
ANL-EBS-PA-000001 REV 02. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: DOC.20041004.0005. 
170020 BSC 2004. Waste-Form Features, Events, and Processes. ANL-WIS-MD-000009, 
Rev. 02. Las Vegas, Nevada: Bechtel SAIC Company. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 8-4 October 2004 
166275 Canori, G.F. and Leitner, M.M. 2003. Project Requirements Document. TER-MGRMD-
000001 REV 02. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: DOC.20031222.0006. 
105212 Cowan, J.C. and Weintritt, D.J. 1976. Water-Formed Scale Deposits. Houston, 
Texas: Gulf Publishing Company. TIC: 245620. 
150823 CRWMS M&O 2000. Design Analysis for the Defense High-Level Waste Disposal 
Container. ANL-DDC-ME-000001 REV 00. Las Vegas, Nevada: CRWMS M&O. 
ACC: MOL.20000627.0254. 
123881 CRWMS M&O 2000. Engineering Files for Site Recommendation. TDR-WHS-MD- 
000001 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20000607.0232. 
152016 CRWMS M&O 2001. Water Distribution and Removal Model. ANL-EBS-MD- 
000032 REV 01. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20010214.0031. 
162213 Dunn, D.S. and Brossia, C.S. 2002. Assessment of Passive and Localized Corrosion 
Processes for Alloy 22 as a High-Level Nuclear Waste Container Material. 
Corrosion/2002, 57th Annual Conference & Exposition, April 7-11, 2002, Denver, 
Colorado. Paper No. 02548. Houston, Texas: NACE International. TIC: 254579. 
164789 Etherington, H., ed. 1958. Nuclear Engineering Handbook. 1st Edition. New York, 
New York: McGraw-Hill. TIC: 231758. 
100859 Gdowski, G.E. 1991. Survey of Degradation Modes of Four Nickel-Chromium- 
Molybdenum Alloys. UCRL-ID-108330. Livermore, California: Lawrence Livermore 
National Laboratory. ACC: NNA.19910521.0010. 
102789 Gdowski, G.E. 1997. Degradation Mode Survey Candidate Titanium - Base Alloys 
for Yucca Mountain Project Waste Package Materials. UCRL-ID-121191, Rev. 1. 
Livermore, California: Lawrence Livermore National Laboratory. 
ACC: MOL.19980120.0053. 
101995 Haynes International 1988. Hastelloy Alloy C-22. Kokomo, Indiana: Haynes 
International. TIC: 239938. 
163111 Hua, F. and Gordon, G. 2003. On Apparent Bi-Linear Corrosion Rate Behavior of Ti 
Grade 7 in Basic Saturated Water (BSW-12) Below and Above 80C. 
Corrosion/2003, 58th Annual Conference & Exposition, March 16-20, 2003, San 
Diego, California. Paper No. 03687. Houston, Texas: NACE International. 
TIC: 254248. 
160670 Hua, F.; Sarver, J.; Jevec, J.; and Gordon, G. 2002. General Corrosion Studies of 
Candidate Container Materials in Environments Relevant to Nuclear Waste 
Repository. Corrosion/2002, 57th Annual Conference & Exposition, April 7-11, 
2002, Denver, Colorado. Paper No. 02530. Houston, Texas: NACE International. 
TIC: 252067. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 8-5 October 2004 
131519 Kohli, R. and Pasupathi, V. 1986. Investigation of Water-logged Spent Fuel Rods 
Under Dry Storage Conditions. PNL-5987. Richland, Washington: Pacific Northwest 
Laboratory. TIC: 246472. 
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. 
159833 Lorenzo de Mele, M.F. and Cortizo, M.C. 2000. Electrochemical Behaviour of 
Titanium in Fluoride-Containing Saliva. Journal of Applied Electrochemistry, 30, 
(1), 95-100. Dordrecht, The Netherlands: Kluwer Academic Publishers. 
TIC: 253126. 
100468 Merriam-Webster. 1993. Merriam-Webster's Collegiate Dictionary, 10th Edition. 
Springfield, Massachusetts: Merriam-Webster. TIC: 8883. 
163274 NRC (U.S. Nuclear Regulatory Commission) 2003. Yucca Mountain Review Plan, 
Final Report. NUREG-1804, Rev. 2. Washington, D.C.: U.S. Nuclear Regulatory 
Commission, Office of Nuclear Material Safety and Safeguards. TIC: 254568. 
165536 Pan, Y.-M.; Brossia, C.S.; Cragnolino, G.A.; Dunn, D.S.; Gute, G.D.; and Yang, L. 
2002. Stress Corrosion Cracking and Hydrogen Embrittlement of Container and Drip 
Shield Materials. CNWRA 2003-02. San Antonio, Texas: Center for Nuclear Waste 
Regulatory Analyses. TIC: 254055. 
156800 Plinski, M.J. 2001. Waste Package Operations Fabrication Process Report. TDREBS-
ND-000003 REV 02. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: MOL.20011003.0025. 
165537 Pulvirenti, A.L.; Needham, K.M.; Adel-Hadadi, M.A.; Marks, C.R.; Gorman, J.A.; 
and Barkatt, A. 2002. Effects of Lead, Mercury, and Reduced Sulfur Species on the 
Corrosion of Alloy 22 in Concentrated Groundwaters as a Function of pH and 
Temperature. Scientific Basis for Nuclear Waste Management XXV, Symposium held 
November 26-29, 2001, Boston, Massachusetts. McGrail, B.P. and Cragnolino, G.A., 
eds. 713, 89-95. Warrendale, Pennsylvania: Materials Research Society. 
TIC: 248663. 
159370 Revie, R.W., ed. 2000. Uhlig's Corrosion Handbook. 2nd Edition. New York, New 
York: John Wiley & Sons. TIC: 248360. 
154465 Sakai, T.; Aoki, K.; Shigemitsu, T.; and Kishi, Y. 1992. Effect of Lead Water 
Chemistry on Oxide Thin Film of Alloy 600. Corrosion, 48, (9), 745-750. Houston, 
Texas: National Association of Corrosion Engineers. TIC: 249773. 
144302 Schutz, R.W. and Thomas, D.E. 1987. Corrosion of Titanium and Titanium Alloys. 
In Corrosion, Volume 13, Pages 669-706 of Metals Handbook. 9th Edition. Metals 
Park, Ohio: ASM International. TIC: 209807. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 8-6 October 2004 
151179 Shoesmith, D.W. and Ikeda, B.M. 1997. The Resistance of Titanium to Pitting, 
Microbially Induced Corrosion and Corrosion in Unsaturated Conditions. AECL- 
11709. Pinawa, Manitoba, Canada: Whiteshell Laboratories. TIC: 236226. 
112178 Shoesmith, D.W. and King, F. 1998. The Effects of Gamma Radiation on the 
Corrosion of Candidate Materials for the Fabrication of Nuclear Waste Packages. 
AECL-11999. Pinawa, Manitoba, Canada: Atomic Energy of Canada Limited. 
ACC: MOL.19990311.0212. 
117892 Shoesmith, D.W.; Ikeda, B.M.; Bailey, M.G.; Quinn, M.J.; and LeNeveu, D.M. 1995. 
A Model for Predicting the Lifetimes of Grade-2 Titanium Nuclear Waste Containers. 
AECL-10973. Pinawa, Manitoba,Canada: Atomic Energy of Canada Limited. 
TIC: 226419. 
120498 Thomas, J.G.N. 1994. The Mechanism of Corrosion Prevention by Inhibitors. In 
Corrosion Control, Volume 2, Chapter 17.3 of Corrosion. 3rd Edition. Reprinted 
1998. Sheir, L.L.; Jarman, R.A.; and Burstein. Woburn, Massachusetts: Butterworth- 
Heinemann. TIC: 244694. 
8.2 CODES, STANDARDS, REGULATIONS, AND PROCEDURES 
156605 10 CFR 63. Energy: Disposal of High-Level Radioactive Wastes in a Geologic 
Repository at Yucca Mountain, Nevada. Readily available. 
AP-2.22Q, Rev. 1, ICN 1. Classification Analyses and Maintenance of the Q-List. 
Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste 
Management. ACC: DOC.20040714.0002. 
AP-3.15Q, Rev. 4, ICN 5. Managing Technical Product Inputs. Washington, D.C.: 
U.S. Department of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20040812.0004. 
AP-SIII.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. 
141257 ASME (American Society of Mechanical Engineers) 1995. Rules for Construction 
of Nuclear Power Plant Components, Division 1, Subsection NB, Class 1 
Components. Section III of 1995 ASME Boiler and Pressure Vessel Code. New 
York, New York: American Society of Mechanical Engineers. TIC: 245287. 
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. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 8-7 October 2004 
8.3 SOURCE DATA, LISTED BY DATA TRACKING 
163112 LL021012712251.021. Chemical Analysis of Waters from Select Vessels in the 
Long-Term Corrosion Test Facility at Lawrence Livermore National Laboratory. 
Submittal date: 01/20/2003. 
161253 LL021105312251.023. Stress Corrosion Crack Growth and Initiation Measurements 
for C-22 and Ti-7, General Electric Global Research Center (GEGRC) 121202. 
Submittal date: 01/08/2003. 
152926 MO0003RIB00073.000. Physical and Chemical Characteristics of Ti Grades 7 and 
16. Submittal date: 03/13/2000. 
170760 MO0407SEPFEPLA.000. LA FEP List. Submittal date: 07/20/2004. 

FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 03 8-8 October 2004 
INTENTIONALLY LEFT BLANK