FEPs Screening of Processes and Issues and 
Drip Shield and Waste Package Degradation 
ANL-EBS-PA-000002 REV 002 
March 2004 





1. PURPOSE 
As directed by a written development plan (BSC 2002 [DIRS 161132]), the primary purpose of 
this scientific analysis is to identify and document the analyses and resolution of the features, 
events, and processes (FEPs) affecting the waste package and drip shield performance in the 
repository. Thirty-three FEPs were identified that are associated with the waste package and drip 
shield performance. This scientific analysis has been prepared to document the screening 
methodology used in the process of FEP inclusion and exclusion. 
The scope of this scientific analysis is to identify the treatment of the FEPs affecting postclosure 
waste package and drip shield performance. It should be noted that seismic effects are not 
treated within this report. A full discussion of seismic effects is contained in the Engineered 
Barrier System Features, Events, and Processes report (BSC 2004 [DIRS 167253]). The FEPs 
that are deemed potentially important to repository postclosure performance are evaluated, either 
as components of the total system performance assessment (TSPA) or as a separate discussion in 
a scientific analysis report. The scope for this activity involves two tasks, namely: 
Task 1: Identify which FEPs are to be considered explicitly in the TSPA (called 
included FEPs) and in which scientific analyses these FEPs are addressed. 
Task 2: Identify FEPs not to be included in the TSPA (called excluded FEPs) and 
provide justification for why these FEPs do not need to be a part of the TSPA 
model. 
The analyses documented in this scientific analysis are for the license application (LA) base case 
design (BSC 2004 [DIRS 167040]). In this design, a drip shield is placed over the waste 
package and no backfill is placed over the drip shield (BSC 2004 [DIRS 167040]). Each FEP 
may include one or more specific issues that are collectively described by a FEP name, a FEP 
description, and descriptor phrases. The FEP Description may encompass a single feature, 
process or event, or a few closely related or coupled processes provided that the entire FEP can 
be addressed by a single specific screening argument or TSPA disposition. Descriptor phrases 
provide additional detail about the subject and content of the FEP beyond the FEP name and 
description. The FEPs have been assigned to associated scientific analyses. The assignments 
were based on the nature of the FEPs so that the analysis and resolution for screening decisions 
reside with the subject-matter experts in the relevant disciplines. This scientific analysis 
addresses the screening decisions associated with the FEPs for the Performance Assessment 
Project Waste Package Degradation Modeling group. 
1.1 OVERVIEW OF FEATURES, EVENTS PROCESSES ANALYSIS AND 
DEVELOPMENT 
The overall FEPs identification and selection processes are documented in The Enhanced Plan 
for Features, Events, and Processes (FEPs) at Yucca Mountain (BSC 2002 [DIRS 158966]) and 
supplemented by the Key Technical Issue Letter Report Response to Additional Information 
Needs on TSPAI 2.05 and TSPAI 2.06 (Freeze 2003 [DIRS 165394]). The initial set of FEPs was 
created for the Yucca Mountain Project (YMP) TSPA by combining lists of FEPs identified as 
relevant to the YMP. This list consists of FEPs from the Nuclear Energy Agency working group, 

FEPs from YMP literature and site studies, and FEPs identified during YMP project staff 
workshops. The FEPs were identified by a variety of methods, including expert judgment, 
informal elicitation, event tree analysis, stakeholder review, and regulatory stipulation. All 
potentially relevant FEPs were evaluated, regardless of origin. This approach led to considerable 
redundancy in the FEP list because the same FEPs are frequently identified by multiple sources, 
but it also ensures that a comprehensive review of all of the FEPs will be performed. 
To eliminate the redundancy in the FEP list and to create a more efficient aggregation of FEPs to 
carry forward into the TSPA-LA screening process, the FEPs were classified into a 
two-dimensional FEP matrix having a physical (subsystem elements and features) hierarchy 
along one axis (the rows) and a process/event hierarchy along the other axis (the columns) 
(Table 1-1). Each of the specific FEPs was assigned to one or more of the matrix 
intersections, generally representative of a process or event acting upon a feature or element. 
In the case of coupled processes (e.g., thermal-hydrologic-chemical effects on the waste 
package), several of the intersections along a physical row (e.g., the Waste Package row) may 
include the same coupled-process FEP. Similarly, in the case of a high-level process or event 
(e.g., repository dry-out due to waste heat), several intersections along a process column may 
include the same FEP. 
Table 1-1. Example FEP Matrix 
Repository Processes and Events 
Flow 
Transport 
Chemical 
Mechanical 
Thermal 
Microbiological 
Radiological 
Repository Subsystem 
Physical Elements and 
Features 
Ground Support 
Backfill 
Drip Shield 
Waste Package 
Cladding 
Waste Form 
Invert 
Note: Only a few of the processes/events and subsystem elements/features are included in 
this example table. 
The matrix classification process resulted in 367 unique FEPs (DTN: MO0307SEPFEPS4.000 
[DIRS 164527]), each of which encompassed a single process or event, or a few closely related 
or coupled processes or events that could be addressed by a specific screening discussion. A few 
of the 33 FEPs are broad ranging and cover multiple technical areas in addition to waste package 
and drip shield. These FEPs are shared with other FEP reports, and this FEP report may provide 
only a partial technical basis for the screening of the FEP. The full technical basis for these 
shared FEPs is addressed collectively by all of the sharing FEP reports. 

The approach used for the evaluation of the 33 waste package and drip shield FEPs is a 
combination of qualitative and quantitative analyses. The analyses are based on the U.S. Nuclear 
Regulatory Commission (NRC) regulatory requirements provided in 10 CFR Part 63 
[DIRS 159535]) to determine whether or not each FEP should be included in the TSPA. For 
FEPs that are excluded from the TSPA based on the NRC regulatory requirements, the screening 
argument includes a summary of the basis and results that indicate either low probability of 
occurrence or low consequence to radiological exposures to the reasonably maximally exposed 
individual (RMEI) and radionuclide releases to the accessible environment. As appropriate, 
screening arguments may cite work done outside the YMP, such as in other scientific analyses. 
For FEPs that are included in the TSPA, the TSPA disposition includes a short summary of how 
the FEP has been incorporated in the process models and the TSPA models and, where 
necessary, a reference to the scientific analysis that describes the disposition in greater detail. 

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2. QUALITY ASSURANCE 
The quality assurance (QA) program applies to this analysis. The technical work plan entitled 
Waste Package Materials Data Analyses and Modeling (BSC 2002 [DIRS 161132]) determined 
that this activity is subject to the Quality Assurance Requirements and Description (DOE 2003 
[DIRS 162903]) requirements. All waste package configurations have been determined to be 
important to waste isolation in accordance with AP-2.22Q, Classification Analyses and 
Maintenance of the Q-List, and therefore are classified as safety category (SC) on the Q-List 
(BSC 2003 [DIRS 165179], Appendix A; BSC 2003 [DIRS 164554], Section 6.4.2). The drip 
shields have been determined to be important to waste isolation in accordance with AP-2.22Q 
and therefore are classified as SC on the Q-List (BSC 2003 [DIRS 165179], Appendix A; 
BSC 2003 [DIRS 164554], Section 6.4.2). 
This document was prepared in accordance with AP-SIII.9Q, Scientific Analyses, and reviewed 
in accordance with AP-2.14Q, Document Review. 
The electronic management of data is not applicable to this analysis report since data were not 
developed as part of this analysis. 

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3. USE OF SOFTWARE 
This analysis report uses no computational software; therefore, this analysis is not subject to 
software controls. The analyses and arguments presented herein are based on guidance and 
regulatory requirements, results of analyses presented and documented in other analysis reports, 
or on other technical literature. Software and models used in the supporting documents are cited 
in this analysis report for traceability and transparency purposes, but were not used in its 
development. This includes the FEPS Database Software Program, V.2.PC, tracking number 
10418-.2-00 (BSC 2002 [DIRS 159684]). This database program was utilized outside of this 
analysis to establish and organize the original list of TSPA-Site Recommendation (TSPA-SR) 
FEPs to be addressed for the repository (refer to Section 4.1.3). 
This analysis report was developed using only commercial-off-the-shelf software, Microsoft 
Word 2000, for word processing which is exempt from qualification requirements in accordance 
with AP-SI.1Q, Software Management. No additional applications (Routines or Macros) were 
developed using this commercial off-the-shelf software. 

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4. INPUTS 
4.1 DIRECT INPUTS 
4.1.1 Data 
The technical information used in this scientific analysis as input has been obtained, where 
possible, from controlled source documents and references using the appropriate document 
identifiers or records system accession numbers. These inputs were documented according to the 
procedure outlined in AP-3.15Q, Managing Technical Product Inputs. Inputs that are relied 
upon in screening arguments within Section 6.2 of this AR are summarized in Table 4-1. 
Table 4-1. Direct Inputs 
Reference 
[DIRs Identifier] 
Description of Input 
Input Status 
BSC 2003 [DIRS 168489] 
Displacement of DS based on 11.5 MT rockfall 
Product Output 
BSC 2003 [DIRS 168489] 
Mechanical loading of waste package interfaces 
Product Output 
BSC 2004 [DIRS 167040] 
Configuration of WP and DS design; Distance from top of 
invert to centerline of WP (5 DHLW) 
Product Output 
BSC 2003 [DIRS 165269], 
Tables 59 and 60 
Radiation dose rates at waste package surface 
Product Output 
BSC 2003 [DIRS 16168489] 
Analysis of multiple rockfalls 
Product Output 
LL021012712251.021 [DIRS 
163112] 
Results of long term corrosion testing 
Data  Qualified 
BSC 2003 [DIRS 168489], 
Table 1 
Interior height of drip shield 
Product Output 
BSC 2004 [DIRS 167207], 
Table 1 
Nominal diameter of 5 DHLW/DOE SNF short waste 
package 
Product Output 
BSC 2004 [DIRS 166694] 
Gap between WP barriers 
Product Output 
BSC 2003 [DIRS 163855] 
Mass of WP internals 
Product Output 
MT = metric ton, WP = waste package, DS = drip shield, DHLW = defense high-level radioactive waste, 
DOE SNF = U.S. Department of Energy spent nuclear fuel. 
4.1.2 Parameters 
The analyses and arguments presented herein are based on guidance and regulatory 
requirements, results of analyses presented and documented in other analysis reports, model 
reports, or other technical literature. There were no parameters used and this subsection is not 
applicable. 
4.1.3 Other Model/Analysis Inputs and Technical Information 
Table 4-2 summarizes inputs used in this analysis that originated from other models or analyses 
and from other sources of technical information. 

Table 4-2. Other Model/Analysis Inputs and Technical Information 
Reference 
[DIRS Identifier] 
Description of Input 
Input Status 
ASM International 1987 
[DIRS 103753] 
Effects of radiation on Alloy 22; oxygen embrittlement of 
titanium; effects of hydrogen induced cracking on Alloy 22 
and titanium 
Technical Information 
ASM International 1990 
[DIRS 144385], p. 626 
Deformation characteristics of titanium and nickel-based 
alloys 
Technical Information 
Boyer and Gall [DIRS 
155318], Section 32 
Creep temperatures for nickel-based alloys and titanium 
Technical Information 
Haynes International 1988 
[DIRS 101995] 
Melting temperature of Alloy 22 
Technical Information 
Revie, R.W. ed. 2000 
[DIRS 159370], Chapter 47 
Titanium alloy properties 
Technical Information 
Schutz, R.W. and Thomas, 
D.E. 1987. [DIRS 144302], 
p. 669-706 
Hydrogen absorption conditions in titanium alloys 
Technical Information 
The TSPA-LA FEPs list for this study originated from the TSPA-SR FEP list (BSC 2002 
[DIRS 159684]) that is described in BSC 2001 [DIRS 154365]. The development of a 
comprehensive list of FEPs potentially relevant to post-closure performance of the Yucca 
Mountain repository is an ongoing, iterative process based on site-specific information, design, 
and regulations. The approach described in Section 0 for developing an initial list of FEPs, in 
support of TSPA-SR (CRWMS M&O 2000 [DIRS 153246]), was documented in BSC 2001 
[DIRS 154365]. The initial FEP list contained 328 FEPs, of which 176 were included in 
TSPA-SR models (CRWMS M&O 2000 [DIRS 153246], Appendix B, Tables B-9 through 
B-17). To support TSPA-LA, the FEP list was re-evaluated in accordance with The Enhanced 
Plan for Features, Events, and Processes (FEPs) at Yucca Mountain (BSC 2002 
[DIRS 158966]) and Key Technical Issue Letter Report Response to Additional Information 
Needs on TSPAI 2.05 and TSPAI 2.06 (Freeze 2003 [DIRS 165394]). The resulting TSPA-LA 
list contains 367 FEPs (DTN: MO0307SEPFEPS4.000 [DIRS 164527]). Significant changes to 
the waste package and drip shield FEPs from TSPA-SR to TSPA-LA are summarized in a 
general fashion in Table 4-3. In addition, descriptions for five of the 33 waste package and drip 
shield FEPs addressed in this report differ from the descriptions in the TSPA-LA list 
(DTN: MO0307SEPFEPS4.000 [DIRS 164527]) due to recommended changes that were made 
during the review process. The five FEPs are FEP 2.1.03.07.0A, Mechanical Impact on Waste 
Package, FEP 2.1.03.07.0B, Mechanical Impact on Drip Shield, FEP 2.1.07.01.0A, Rockfall, 
FEP 2.1.11.07.0A, Thermal Expansion/Stress of In-Drift EBS Components, and 
FEP 2.1.13.02.0A, Radiation Damage in EBS. 

Table 4-3. Significant Changes to the Waste Package FEPs from TSPA-SR to TSPA-LA 
TSPASR 
FEP Number and Name 
TSPA-LA 
FEP Number and Name 
Significant Changes from TSPA-SR to TSPA-LA 
1.1.03.01.00 
Error in waste or backfill 
emplacement 
1.1.03.01.0A 
Error in waste 
emplacement 
Discussion of waste package and drip shield 
emplacement error was added. 
2.1.03.01.00 
Corrosion of waste 
containers 
2.1.03.01.0A 
General corrosion of waste 
packages 
FEP was split meaning that the FEP only addresses 
general corrosion of waste packages for LA, but 
addressed general corrosion of waste packages and drip 
shields for SR. The FEP also discussed other forms of 
corrosion for SR, but for LA, these other forms of 
corrosion are completely addressed in other LA FEPs. 
2.1.03.01.00 
Corrosion of waste 
containers 
2.1.03.01.0B 
General corrosion of drip 
shields 
This is a new FEP for LA (given a new number) although 
the FEP was discussed in SR FEP 2.1.03.01.00. 
2.1.03.02.00 
SSC of waste containers 
and DSs 
2.1.03.02.0A 
SCC of waste packages 
FEP was split meaning that the FEP only addresses SCC 
of waste packages for LA but addressed SCC of waste 
packages and drip shields for SR. 
2.1.03.02.00 
SSC of waste containers 
and drip shields 
2.1.03.02.0B 
SCC of drip shields 
This is a new FEP for LA (given a new number) although 
the FEP was discussed in SR FEP 2.1.03.02.00. 
2.1.03.03.00 
Pitting of waste 
containers and DSs 
2.1.03.03.0A 
Localized corrosion of 
waste packages 
FEP was split meaning that the FEP only addresses 
localized corrosion of waste packages for LA but 
addressed localized corrosion of waste packages and drip 
shields for SR. 
2.1.03.03.00 
Pitting of waste 
containers and DSs 
2.1.03.03.0B 
Localized corrosion of drip 
shields 
This is a new FEP for LA (given a new number) although 
the FEP was discussed in SR FEP 2.1.03.03.00. 
2.1.03.04.00 
Hydride cracking of 
waste containers and 
DSs 
2.1.03.04.0A 
Hydride cracking of waste 
packages 
FEP was split meaning that the FEP only addresses 
hydride cracking of WPs for LA but addressed hydride 
cracking of WPs and DSs for SR. 
2.1.03.04.00 
Hydride cracking of 
waste containers and 
DSs 
2.1.03.04.0B 
Hydride cracking of drip 
shields 
This is a new FEP for LA (given a new number) although 
the FEP was discussed in SR FEP 2.1.03.04.00. 
2.1.03.05.00 
Microbially-mediated 
corrosion of waste 
container andDSs 
2.1.03.05.0A 
MIC of waste packages 
FEP was split meaning that the FEP only addresses MIC 
of WPs for LA but addressed MIC of WPs and DSs for 
SR. 
2.1.03.05.00 
Microbially-mediated 
corrosion of waste 
container and DS 
2.1.03.05.0B 
MIC of DSs 
This is a new FEP for LA (given a new number) although 
the FEP was discussed in SR FEP 2.1.03.05.00. 
2.1.03.06.00 
Internal corrosion of 
waste container 
2.1.03.06.0A 
Internal corrosion of waste 
packages prior to breach 
None 
2.1.03.07.00 
Mechanical impact on 
waste container and DS 
2.1.03.07.0A 
Mechanical impact on WP 
FEP was split meaning that the FEP only addresses 
mechanical impact on WPs for LA, but addressed 
mechanical impacts on WPs and DSs for SR. The FEP 
does not address static loading or seismic impacts for LA 
(they are addressed in other LA FEPs) but addressed all 
types of mechanical impacts for SR. 

Table 4. Significant Changes to the WP FEPs from TSPA-SR to TSPA-LA (Continued) 
TSPASR 
FEP Number and Name 
TSPA-LA 
FEP Number and Name 
Significant Changes from TSPA-SR to TSPA-LA 
2.1.03.07.00 
Mechanical impact on 
waste container and DS 
2.1.03.07.0B 
Mechanical impact on DS 
This is a new FEP for LA (given a new number) although 
the FEP was discussed in SR FEP 2.1.03.07.00. 
2.1.03.08.00 
Juvenile and early failure 
of waste containers and 
DSs 
2.1.03.08.0A 
Early failure of DSs 
FEP was split meaning that the FEP only addresses early 
failure of waste packages for LA but addressed early 
failure of waste packages and drip shields for SR. 
2.1.03.08.00 
Juvenile and early failure 
of waste containers and 
DSs 
2.1.03.08.0B 
Early failure of DSs 
This is a new FEP for LA (given a new number) although 
the FEP was discussed in SR FEP 2.1.03.08.00. 
2.1.03.09.00 
Copper corrosion 
2.1.03.09.0A 
Copper corrosion in EBS 
None 
2.1.03.10.00 
Container healing 
2.1.03.10.0A 
Healing of WPs 
FEP was split meaning that the FEP only addresses 
healing of WPs for LA but addressed healing of WPs and 
DSs for SR. 
2.1.03.10.00 
Container healing 
2.1.03.10.0B 
Healing of DSs 
This is a new FEP for LA (given a new number) although 
the FEP was discussed in SR FEP 2.1.03.10.00. 
2.1.03.11.00 
Container form 
2.1.03.11.0A 
Physical form of WP and 
DS 
None 
2.1.06.06.00 
Effects and degradation 
of DS 
2.1.06.06.0B 
Oxygen embrittlement of 
DSs 
This is a new FEP for LA (given a new number) although 
the FEP was discussed in SR FEP 2.1.06.06.00. 
2.1.06.07.00 
Effects of material 
interfaces 
2.1.06.07.0B 
Mechanical effects at EBS 
component interfaces 
FEP was split meaning that the FEP only addresses 
mechanical effects at EBS component interfaces for LA 
but addressed chemical and mechanical effects at EBS 
component interfaces for SR. Chemical effects are 
discussed under LA FEP 2.1.06.07.0A in the EBS FEPs 
AR. 
2.1.07.01.00 
Rockfall (Large Block) 
2.1.07.01.0A 
Rockfall 
FEP was modified for LA to only address rockfall under 
nominal conditions. Seismic-induced rockfall is 
addressed under LA FEP 1.2.03.02.0B in the EBS and DE 
FEPs ARs. 
2.1.07.05.00 
Creeping of metallic 
materials in the EBS 
2.1.07.05.0A 
Creep of metallic materials 
in the WP 
FEP was split meaning that the FEP only addresses creep 
of metallic materials in the WP for LA but addressed creep 
of metallic materials in the WP and DSs for SR. 
2.1.07.05.00 
Creeping of metallic 
materials in the EBS 
2.1.07.05.0B 
Creep of metallic materials 
in the DS 
This is a new FEP for LA (given a new number) although 
the FEP was discussed in SR FEP 2.1.07.05.00. 
2.1.09.03.00 
Volume increase of 
corrosion products 
2.1.09.03.0B 
Volume increase of 
corrosion products impacts 
WP 
FEP was split meaning that the FEP only addresses 
volume increase of corrosion products impacts WP for LA 
but addressed volume increase of corrosion products 
impacts on WP, cladding, and other EBS components for 
SR. 

Table 5. Significant Changes to the WP FEPs from TSPA-SR to TSPA-LA (Continued) 
TSPASR 
FEP Number and Name 
TSPA-LA 
FEP Number and Name 
Significant Changes from TSPA-SR to TSPA-LA 
2.1.09.09.00 
Electrochemical effects 
in waste and EBS 
2.1.09.09.0A 
Electrochemical effects in 
EBS 
Discussion of galvanic coupling added. 
2.1.11.06.00 
Thermal sensitization of 
waste containers and DS 
increases their fragility 
2.1.11.06.0A 
Thermal sensitization of 
WPs 
FEP was split meaning that the FEP only addresses 
thermal sensitization of WPs for LA but addressed thermal 
sensitization of WPs and DSs for SR. 
2.1.11.06.00 
Thermal sensitization of 
waste containers and 
DSs increases their 
fragility 
2.1.11.06.0B 
Thermal sensitization of 
DSs 
This is a new FEP for LA (given a new number) although 
the FEP was discussed in SR FEP 2.1.11.06.00. 
2.1.11.07.00 
Thermally-induced stress 
changes in waste and 
EBS 
2.1.11.05.00 
Differing Thermal 
Expansion of Repository 
Components 
2.1.11.07.0A 
Thermal expansion/stress 
of in-drift EBS components 
This FEP was re-scoped for LA. SR FEPs 2.1.11.05.00 
and 2.1.11.07.00, which collectively addressed thermal 
effects in the repository, were was combined for LA and 
then split into 2 LA FEPs. This FEP addresses in-drift 
thermal effects. 
2.1.12.03.00 
Gas generation (H2) from 
metal corrosion 
2.1.12.03.0A 
Gas generation (H2) from 
WP corrosion 
None 
2.1.13.01.00 
Radiolysis 
2.1.13.01.0A 
Radiolysis 
None 
2.1.13.02.00 
Radiation damage in 
waste and EBS 
2.1.13.02.0A 
Radiation damage in EBS 
None 
DS = drip shield, WP = waste page(s), LA = license application, SR = site recommendation, FEP = feature, event, 
and process, SSC = stress corrosion cracking, MIC = microbially influenced corrosion, EBS = engineered barrier 
system, DE = disruptive events 
4.2 CRITERIA 
4.2.1 NRC Regulatory Requirements 
As discussed in The Enhanced Plan for Features, Events, and Processes (FEPs) at Yucca 
Mountain (BSC 2002 [DIRS 158966]), FEPs will be screened and arguments documented in 
accordance with the NRC technical screening regulatory requirements provided in 10 CFR 63 
[DIRS 158535]. 
These regulatory requirements are summarized as follows: 
10 CFR 63.21(c)(1)  The Safety Analysis Report must include a description of the Yucca 
Mountain site, with appropriate attention to those features, events, and processes of the site 
that might affect design of the geologic repository operations area and performance of the 
geologic repository. The description of the site must include information regarding features, 

events, and processes outside of the site to the extent the information is relevant and material 
to safety or performance of the geologic repository. The information referred to in this 
paragraph must include: 
(i) The location of the geologic repository operations area with respect to the 
boundary of the site 
(ii) Information regarding the geology, hydrology, and geochemistry of the site, 
including geomechanical properties and conditions of the host rock 
(iii) Information regarding surface water hydrology, climatology, and meteorology 
of the site 
(iv) Information regarding the location of the reasonably maximally exposed 
individual, and regarding local human behaviors and characteristics, as needed 
to support selection of conceptual models and parameters used for the reference 
biosphere and reasonably maximally exposed individual. 
10 CFR 63.21(c)(9)  The Safety Analysis Report must include an assessment to determine 
the degree to which those features, events, and processes of the site that are expected to 
materially affect compliance with 10 CFR 63.113  whether beneficial or potentially adverse 
to the performance of the geologic repository  have been characterized, and the extent to 
which they affect waste isolation. 
10 CFR 63.102(j) - . . . Those features, events, and processes expected to materially affect 
compliance with 10 CFR 63.113(b) or be potentially adverse to performance are included . . . 
10 CFR 63.114(d)  Consider only events that have at least one chance in 10,000 of 
occurring over 10,000 years. 
10 CFR 63.114(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. 
10 CFR 63.114(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. 
4.2.2 Yucca Mountain Review Plan 
The Waste Package TWP (BSC 2002 [161132], Attachment C, Table C5) identified acceptance 
criteria based on the requirements mentioned in the Project Requirements Document (PRD) 

(Canori and Leitner 2003 [DIRS 166275]) and the Yucca Mountain Review Plan (NRC 2003 
[DIRS 163274]). The following acceptance criteria from the PRD and Yucca Mountain Review 
Plan are updates to those mentioned in the Waste Package TWP (BSC 2002 [DIRS 161132], 
Attachment C, Table C5): 
System Description and Demonstration of Multiple Barriers (NRC 2003 [DIRS 163274], 
Section 2.2.1.1; Canori and Leitner 2003 [DIRS 166275], PRD-002/T-014, PRD-002/T-016) 
 Specific requirements involve identification of multiple barriers (natural and engineered), 
describing the capabilities of these barriers to isolate waste, and providing technical bases for 
capabilities descriptions consistent with the postclosure performance objectives as described 
below: 
1. Identification of barriers relied on for postclosure performance; (including at least one 
barrier from the engineered system and one from the natural system) 
2. Description of the capability of identified barriers to prevent or substantially reduce 
the rate of movement of water or radionuclides from the Yucca Mountain repository to 
the accessible environment, or prevent the release or substantially reduce the release of 
radionuclides from the waste including the uncertainty associated with this capacity 
and the consistency with approaches used in the total system performance assessment 
3. Discussion of the technical bases for assertions of barrier capability commensurate 
with the importance of a particular barrier in the performance assessment and with the 
associated uncertainties. 
To comply with these requirements, the following acceptance criteria are identified in the Waste 
Package TWP for this report (BSC 2002 [DIRS 161132], Attachment C, Table C5): 
AC1: Identification of Barriers is Adequate: Barriers relied on to achieve compliance 
with 10 CFR 63.113(b), as demonstrated in the TSPA, are adequately identified, and are 
clearly linked to their capability. The barriers identified include at least one from the 
engineered system and one from the natural system. 
AC2: Description of the Capability to Isolate Waste is Acceptable: The capability of the 
identified barriers to prevent or substantially reduce the rate of movement of water or 
radionuclides from the Yucca Mountain repository to the accessible environment, or prevent 
the release or substantially reduce the release rate of radionuclides from the waste is 
adequately identified and described: 
1. The information on the time period over which each barrier performs its intended 
function, including any changes during the compliance period, is provided 
2. The uncertainty associated with barrier capabilities is adequately described 
3. The described capabilities are consistent with the results from the total system 
performance assessment 

4. The described capabilities are consistent with the definition of a barrier at 
10 CFR 63.2. 
Scenario Analysis and Event Probability (NRC 2003 [DIRS 163274], Section 2.2.1.2; Canori 
and Leitner 2003 [DIRS 166275], PRD-002/T-015)  Specific requirements include providing 
technical bases for inclusion or exclusion of specific FEPs. In order to meet these requirements, 
the following acceptance criteria are identified in the Waste Package TWP for this report 
(BSC 2002 [DIRS 161132], Attachment C, Table C5): 
AC1: The Identification and Initial List of Features, Events, and Processes is 
Adequate: The Safety Analysis Report contains a complete list of FEPs 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 FEPs list includes, but is not limited to, 
potentially disruptive events related to igneous activity (extrusive and intrusive); seismic 
shaking (high-frequency-low magnitude, and rare large-magnitude events); tectonic 
evolution (slip on existing faults and formation of new faults); climatic change (change to 
pluvial conditions); and criticality. 
AC2: Screening of the Initial List of Features, Events, and Processes is Appropriate: 
1. The U.S. Department of Energy (DOE) has identified all FEPs 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; 
2. The DOE has provided justification for those FEPs that have been excluded. An 
acceptable justification for excluding FEPs is that either the FEP is specifically 
excluded by regulation; probability of the FEP (generally an event) falls below the 
regulatory criterion; or omission of the FEP 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 
3. The DOE has provided an adequate technical basis for each FEP, excluded from the 
performance assessment, to support the conclusion that either the FEP is specifically 
excluded by regulation; the probability of the FEP falls below the regulatory criterion; 
or omission of the FEP 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. 
AC3: Formation of Scenario Classes Using the Reduced Set of Events is Adequate: 
1. Scenario classes are mutually exclusive and complete, clearly documented, and 
technically acceptable. 

AC4: Screening of Scenario Classes is Appropriate: 
1. Screening of scenario classes is comprehensive, clearly documented, and technically 
acceptable; 
2. The DOE has adequately considered coupling of processes in estimates of 
consequences used to screen scenario classes. Scenario classes were not prematurely 
excluded by a narrow definition; 
3. Scenario classes that are screened from the performance assessment, on the basis that 
they are specifically ruled out by regulation or are contrary to stated regulatory 
assumptions are identified, and sufficient justifications are provided; 
4. Scenario classes that are screened from the performance assessment, on the basis that 
their probabilities fall below the regulatory criterion, are identified, and sufficient 
justifications are provided 
5. Scenario classes that are screened from the performance assessment, on the basis that 
their omission would not significantly change the magnitude and time of the resulting 
radiological exposure to the reasonably maximally exposed individual, or radionuclide 
releases to the accessible environment, are identified, and sufficient justifications are 
provided. 
Degradation of Engineered Barriers (NRC 2003 [DIRS 163274], Section 2.2.1.3.1; Canori 
and Leitner 2003 [DIRS 166275]; PRD-002/T-015)  Specific requirements include 
describing deterioration or degradation of engineered barriers and modeling degradation 
processes using data for performance assessment, including TSPA. Consideration of 
uncertainties and variabilities in model parameters and alternative conceptual models are also 
required. Specific wording from the Yucca Mountain Review Plan (NRC 2003 
[DIRS 163274], Section 2.2.1.3.1) is as follows: 
To review this model abstraction, evaluate the adequacy of the DOE LA, relative to the 
degree to which degradation of engineered barriers affects the DOE LA. Review this model 
abstraction, considering the risk information evaluated in the Multiple Barriers 
Section 2.2.1.1. For example, if the DOE relies on the engineered barriers to provide 
significant delay in the transport of radionuclides to the reasonably maximally exposed 
individual, then perform a detailed review of this abstraction. If, on the other hand, the DOE 
demonstrates this abstraction to have a minor impact on the dose to the RMEI, then conduct a 
simplified review focusing on the bounding assumptions. The review methods and 
acceptance criteria provided here are for a detailed review. Some of the review methods and 
acceptance criteria may not be necessary in a simplified review for those abstractions that 
have a minor impact on performance. The demonstration of compliance with the 
performances objective is evaluated using Section 2.2.1.4 of the Yucca Mountain Review 
Plan. 

To fulfill these requirements, the following acceptance criteria are identified in the TWP for 
this report (BSC 2002 [DIRS 161132], Attachment C, Table C5): 
AC1: System Description and Model Integration are Adequate: 
1. The TSPA adequately incorporates important design features, physical phenomena, 
and couplings, and uses consistent and appropriate assumptions throughout the 
degradation of engineered barriers abstraction process. 
2. Assessment abstraction of the degradation of engineered barriers uses assumptions, 
technical bases, data, and models that are appropriate and consistent with other related 
DOE abstractions. For example, the assumptions used for degradation of engineered 
barriers should be consistent with the abstractions of the quantity and chemistry of 
water contacting waste packages and waste forms (Section 2.2.1.3.3); climate and 
infiltration (Section 2.2.1.3.5); and mechanical disruption of waste packages 
(Section 2.2.1.3.2). The descriptions and technical bases provide transparent and 
traceable support for the abstraction of the degradation of engineered barriers. 
3. The descriptions of engineered barriers, design features, degradation processes, 
physical phenomena, and couplings that may affect the degradation of the engineered 
barriers are adequate. For example, materials and methods used to construct the 
engineered barriers are included, and degradation processes, such as uniform 
corrosion, pitting corrosion, crevice corrosion, stress corrosion cracking, intergranular 
corrosion, microbially influenced corrosion, dry-air oxidation, hydrogen 
embrittlement, and the effects of wet and dry cycles, material aging and phase 
instability, welding, and initial defects on the degradation modes for the engineered 
barriers are considered. 
4. Boundary and initial conditions used in the TSPA abstractions are propagated 
consistently throughout the abstraction approaches. For example, the conditions and 
assumptions used in the degradation of engineered barriers abstraction are consistent 
with those used to model the quantity and chemistry of water contacting waste 
packages and waste forms (Section 2.2.1.3.3); climate and infiltration 
(Section 2.2.1.3.5); and mechanical disruption of waste packages (Section 2.2.1.3.2). 
5. Sufficient technical bases for the inclusion of FEPs related to degradation of 
engineered barriers in the TSPA abstractions are provided. 
6. Adequate technical bases are provided, for selecting the design criteria, that mitigate 
any potential impact of in-package criticality on repository performance, including 
considering all FEPs that may increase the reactivity of the system inside the waste 
package. For example, the technical bases for the abstraction of the degradation of 
engineered barriers include configuration classes and configurations that have 
potential for nuclear criticality, changes in radionuclide inventory, and changes in 
thermal conditions. 

7. Guidance in NUREG1297 (Altman et al. 1988 [103597]) and NUREG1298 
(Altman et al. 1988 [103750]), or other acceptable approaches, is followed. 
AC2: Data are Sufficient for Model Justification: 
1. Parameters used to evaluate the degradation of engineered barriers in the LA are 
adequately justified (e.g., laboratory corrosion tests, site-specific data such as data 
from drift-scale tests, in-service experience in pertinent industrial applications, and test 
results not specifically performed for the Yucca Mountain site, etc.). The DOE 
describes how the data were used, interpreted, and appropriately synthesized into the 
parameters 
2. Sufficient data have been collected on the characteristics of the engineered 
components, design features, and the natural system to establish initial and boundary 
conditions for abstraction of degradation of engineered barriers 
3. Data on the degradation of the engineered barriers (e.g., general and localized 
corrosion, microbially influenced corrosion, galvanic interactions, hydrogen 
embrittlement, and phase stability), used in the abstraction, are based on laboratory 
measurements, site-specific field measurements, industrial analog and/or natural 
analog research, and tests designed to replicate the range of conditions that may occur 
at the Yucca Mountain site. As appropriate, sensitivity or uncertainty analyses, used 
to support the DOE TSPA abstraction, are adequate to determine the possible need for 
additional data4. Degradation models for the processes that may be significant to the 
performance of the engineered barriers are adequate. For example, the DOE models 
consider the possible degradation of the engineered barriers, as a result of uniform and 
localized corrosion processes, stress-corrosion cracking, microbially influenced 
corrosion, hydrogen embrittlement, and incorporate the effects of fabrication 
processes, thermal aging, and phase stability. 
Quantity and Chemistry of Water Contacting Engineered Barriers and Waste Forms 
(NRC 2003 [DIRS 163274], Section 2.2.1.3.3; Canori and Leitner 2003 [DIRS 166275], 
PRD-002/T-015)  Specific requirements include quantifying the amount and chemistry of water 
contacting the waste package and the waste forms as described below: 
1. Description of the geological, hydrological, and geochemical aspects of quantity and 
chemistry of water contacting engineered barriers and waste forms, and the technical 
bases the DOE provides to support model integration across the TSPA abstractions 
2. Sufficiency of the data and parameters used to justify the model abstraction 
3. Methods the DOE uses to characterize data uncertainty, and propagate the effects of 
this uncertainty through the TSPA model abstraction 
4. Methods the DOE uses to characterize model uncertainty, and propagate the effects of 
this uncertainty through the TSPA model abstraction 

5. Approaches the DOE uses to compare TSPA output to process-level model outputs 
and empirical studies 
6. Use of expert elicitation. 
To comply with these requirements, the following acceptance criteria are identified in the TWP 
for this report (BSC 2002 [161132], Attachment C, Table C5): 
AC1: System Description and Model Integration are Adequate (This acceptance 
criterion was not specifically mentioned in the TWP (BSC 2002 [DIRS 161132]) for this 
analysis report, but is relevant to this analysis): 
1. TSPA adequately incorporates important design features, physical phenomena, and 
couplings, and uses consistent and appropriate assumptions throughout the quantity 
and chemistry of water contacting engineered barriers and waste forms abstraction 
process;. 
2. The abstraction of the quantity and chemistry of water contacting engineered barriers 
and waste forms uses assumptions, technical bases, data, and models, that are 
appropriate and consistent with other related DOE abstractions. For example, the 
assumptions used for the quantity and chemistry of water contacting engineered 
barriers and waste forms are consistent with the abstractions of Degradation of 
Engineered Barriers (Section 2.2.1.3.1); Mechanical Disruption of Engineered 
Barriers (Section 2.2.1.3.2); Radionuclide Release Rates and Solubility Limits 
(Section 2.2.1.3.4); Climate and Infiltration (Section 2.2.1.3.5); and Flow Paths in 
the Unsaturated Zone (Section 2.2.1.3.6). The descriptions and technical bases 
provide transparent and traceable support for the abstraction of quantity and chemistry 
of water contacting engineered barriers and waste forms. 
3. Important design features, such as waste package design and material selection, 
backfill, drip shield, ground support, thermal loading strategy, and degradation 
processes, are adequate to determine the initial and boundary conditions for 
calculations of the quantity and chemistry of water contacting engineered barriers and 
waste forms. 
4. Spatial and temporal abstractions appropriately address physical couplings (thermalhydrologic-
mechanical-chemical). For example, the DOE evaluates the potential for 
focusing of water flow into drifts, caused by coupled thermal-hydrologic-mechanicalchemical 
processes. 
5. Sufficient technical bases and justification are provided for TSPA assumptions and 
approximations for modeling coupled thermal-hydrologic-mechanical-chemical effects 
on seepage and flow, the waste package chemical environment, and the chemical 
environment for radionuclide release. The effects of distribution of flow on the 
amount of water contacting the engineered barriers and waste forms are consistently 
addressed, in all relevant abstractions. 

6. The expected ranges of environmental conditions within the waste package 
emplacement drifts, inside of breached waste packages, and contacting the waste 
forms and their evolution with time are identified. These ranges may be developed to 
include: (i) the effects of the drip shield and backfill on the quantity and chemistry of 
water (e.g., the potential for condensate formation and dripping from the underside of 
the shield); (ii) conditions that promote corrosion of engineered barriers and 
degradation of waste forms; (iii) irregular wet and dry cycles; (iv) gamma-radiolysis; 
and (v) size and distribution of penetrations of engineered barriers. 
7. The model abstraction for quantity and chemistry of water contacting engineered 
barriers and waste forms is consistent with the detailed information on engineered 
barrier design and other engineered features. For example, consistency is 
demonstrated for: (i) dimensionality of the abstractions; (ii) various design features 
and site characteristics; and (iii) alternative conceptual approaches. Analyses are 
adequate to demonstrate that no deleterious effects are caused by design or site 
features that the DOE does not take into account in this abstraction. 
8. Adequate technical bases are provided, including activities such as independent 
modeling, laboratory or field data, or sensitivity studies, for inclusion of any 
thermal-hydrologic-mechanical-chemical couplings and FEPs . 
9. Performance-affecting processes that have been observed in thermal-hydrologic tests 
and experiments are included into the performance assessment. For example, the DOE 
either demonstrates that liquid water will not reflux into the underground facility or 
incorporates refluxing water into the performance assessment calculation, and bounds 
the potential adverse effects of alteration of the hydraulic pathway that result from 
refluxing water. 
10. Likely modes for container corrosion (Section 2.2.1.3.1 of the Yucca Mountain 
Review Plan) are identified and considered in determining the quantity and chemistry 
of water entering the engineered barriers and contacting waste forms. For example, 
the model abstractions consistently address the role of parameters, such as pH, 
carbonate concentration, and the effect of corrosion on the quantity and chemistry of 
water contacting engineered barriers and waste forms. 
11. The abstraction of in-package criticality or external-to-package criticality, within the 
emplacement drift, provides an adequate technical basis for screening these events. If 
either event is included in the assessment, then the DOE uses acceptable technical 
bases for selecting the design criteria that mitigate the potential impact of in-package 
criticality on repository performance; identifies the FEPs that may increase the 
reactivity of the system inside the waste package; identifies the configuration classes 
and configurations that have potential for nuclear criticality; and includes changes in 
thermal conditions and degradation of engineered barriers in the abstraction of the 
quantity and chemistry of water contacting engineered barriers and waste forms. 
12. Guidance in NUREG1297 (Altman et al. 1988 [DIRS 103597]) and NUREG1298 
(Altman et al. 1988 [DIRS 103750]), or other acceptable approaches, is followed. 

AC 2: Data are Sufficient for Model Justification: 
1. Geological, hydrological, and geochemical values used in the LA are adequately 
justified. Adequate description of how the data were used, interpreted, and 
appropriately synthesized into the parameters is provided 
2. Sufficient data were collected on the characteristics of the natural system and 
engineered materials to establish initial and boundary conditions for conceptual 
models of thermal-hydrologic-mechanical-chemical coupled processes, that affect 
seepage and flow and the engineered barrier chemical environment 
3. Thermo-hydrologic tests were designed and conducted with the explicit objectives of 
observing thermal-hydrologic processes for the temperature ranges expected for 
repository conditions and making measurements for mathematical models. Data are 
sufficient to verify that thermal-hydrologic conceptual models address important 
thermal-hydrologic phenomena 
4. Sufficient information to formulate the conceptual approach(es) for analyzing water 
contact with the drip shield, engineered barriers, and waste forms is provided 
5. Sufficient data are provided to complete a nutrient- and energy-inventory calculation, 
if it has been used to justify the inclusion of the potential for microbial activity 
affecting the engineered barrier chemical environment and the chemical environment 
for radionuclide release. As necessary, data are adequate to support determination of 
the probability for microbially influenced corrosion and microbial effects, such as 
production of organic byproducts and microbially enhanced dissolution of the 
high-level radioactive waste glass form. 
4.2.3 Screening Decisions 
The NRC requires the consideration and evaluation of FEPs as part of the performance 
assessment activities. More specifically, the NRC regulations allow the exclusion of FEPs from 
the TSPA if they can be shown to be of low probability of occurrence or of low consequence. 
The specified criteria can be summarized in the form of two FEP screening statements as 
follows: 
1. The event has at least one chance in 10,000 of occurring over 10,000 years (see 
10 CFR 63.114(d) (66 FR 55732 [DIRS 156671])). 
2. The magnitude and time of the resulting radiological exposure to the RMEI, or 
radionuclide release to the accessible environment, would be significantly changed by 
its omission (see 10 CFR 63.114 (e and f) (66 FR 55732 [DIRS 156671]). 

Additionally, the Acceptance Criteria in the Yucca Mountain Review Plan 
(NRC 2003, Section 2.2.1.2.1.3 [DIRS 163274]) calls for evaluating the FEPs based 
on the regulations. This criterion can be summarized in the form of a third FEP 
screening statement. 
3. The FEP is not excluded by regulation. 
If there are affirmative conditions for all three screening criteria, the FEP is included in the 
TSPA-LA model. Any negating condition in the three screening criteria excludes the FEP from 
the TSPA-LA model. The approach used for this analysis is a combination of qualitative and 
quantitative screening of FEPs. 
Criteria presented here are addressed in each of the screening arguments and disposition 
statements presented in Section 6.2 FEPS ANALYSES of this analysis report. 
4.3 CODES AND STANDARDS 
This document was prepared to comply with NRC regulatory requirements presented in 
10 CFR 63 [DIRS 158535]. 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). 

INTENTIONALLY LEFT BLANK 

5. ASSUMPTIONS 
There are three assumptions made in screening of the waste package FEPs. These assumptions 
or combinations thereof are used throughout this report. 
Assumption 1) Assume the evolution of the geologic setting is consistent with present 
knowledge of natural processes. 
This assumption is justified based on the regulations stated in 10 CFR 63.305(c) [DIRS 158535] 
where 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 assumption affects waste package and drip shield FEPs concerned with geologic processes. 
The assumption implies that existing knowledge of natural processes is sufficient to adequately 
quantify future states of the system. 
Assumption 2) Assume that the repository will be constructed, operated, and closed according 
to the regulatory requirements applicable to the construction, operation, and 
closure period, and that deviations from design will be detected and corrected. 
This assumption is justified based on the conditions specified in 10 CFR 63.32 [DIRS 158535], 
which pertains to construction authorization and which requires 
Periodic or special reports regarding: 
(1) Progress of construction; 
(2) Any data about the site, obtained during construction, that are not within 
the predicted limits on which the facility design was based; 
(3) Any deficiencies, in design and construction that, if uncorrected, could 
adversely affect safety at any future time; 
(4) Results of research and development programs being conducted to resolve 
safety questions. 
In addition, 10 CFR 63 Subpart F [DIRS 158535] requires that a performance confirmation 
program be instituted. The focus of the program is confirmation of geotechnical and design 
parameters (Section 63.132), design testing (Section 63.133) and monitoring and testing waste 
packages (Section 63.134). In addition, under 10 CFR 63 Subpart G [DIRS 158535], quality 
assurance requirements are applied to site characterization, acquisition, control, and analyses of 
samples and data; tests and experiments; scientific studies; facility and equipment design and 
construction; facility operation; performance confirmation; permanent closure; and 
decontamination and dismantling of surface facilities. The assumption impacts waste package 
and drip shield FEPs that are affected by events occurring during the construction, operation, or 
closure period. 

Assumption 3) Assume that the design parameters for the waste package and drip shield can 
be used to justify an excluded decision. 
This assumption is justified based on the conditions specified in 10 CFR 63.142 Subpart G 
[DIRS 158535] that pertain to quality assurance. 
DOE shall establish measures to assure that applicable regulatory requirements 
and the design basis, as defined in Section 63.2 and as specified in the license 
application, for those structures, systems, and components to which this subpart 
applies, are correctly translated into specifications, drawings, procedures, and 
instructions. These measures must assure that appropriate quality standards are 
specified and included in the design documents and that deviations from such 
standards are controlled. Measures must also be established for the selection and 
review for suitability of application of materials, parts, equipment, and processes 
that are important to waste isolation and important to safety functions of the 
structures, systems and components. 
The assumption allows exclusion of FEPs when the design process specifically addresses the 
issue described by that particular FEP (refer to Section 4.2.3 for possible screening decisions). 
Note that deviation from a design process despite a set of quality controls is allowed for in the 
TSPA. None of the assumptions presented herein require confirmation. 

6. SCIENTIFIC ANALYSIS DISCUSSION 
The primary concern in this scientific analysis is to address and document the screening 
decisions for the 33 waste package and drip shield degradation FEPs listed in Table 6-1. In some 
cases, where a FEP covers multiple technical areas and is shared with other FEP analysis reports, 
this analysis report may provide only a partial technical basis for the screening of the FEP. The 
shared FEPs are identified in column 3 of Table 6-1. The full technical basis for these shared 
FEPs is addressed collectively by all of the sharing FEP analysis reports. 
Table 6-1. TSPA-LA FEP List 
LA FEP Number 
FEP Name 
Shared With 
1.1.03.01.0A 
Error in Waste Emplacement 
EBS 
2.1.03.01.0A 
General Corrosion of WPs 
2.1.03.01.0B 
General Corrosion of DSs 
2.1.03.02.0A 
SCC of WPs 
2.1.03.02.0B 
SCC of DSs 
2.1.03.03.0A 
Localized Corrosion of WPs 
2.1.03.03.0B 
Localized Corrosion of DSs 
2.1.03.04.0A 
Hydride Cracking of WPs 
2.1.03.04.0B 
Hydride Cracking of DSs 
2.1.03.05.0A 
MIC of WPs 
2.1.03.05.0B 
MIC of DSs 
2.1.03.06.0A 
Internal Corrosion of WPs Prior To Breach 
WF 
2.1.03.07.0A 
Mechanical Impact on WP 
2.1.03.07.0B 
Mechanical Impact on DS 
2.1.03.08.0A 
Early Failure of WPs 
2.1.03.08.0B 
Early Failure of DSs 
2.1.03.09.0A 
Copper Corrosion in EBS 
2.1.03.10.0A 
Healing of WPs 
2.1.03.10.0B 
Healing of DSs 
2.1.03.11.0A 
Physical Form of WP and DS 
2.1.06.06.0B 
Oxygen Embrittlement of DSs 
2.1.06.07.0B 
Mechanical Effects at EBS Component Interfaces 
EBS 
2.1.07.01.0A 
Rockfall 
EBS, Cladding 
2.1.07.05.0A 
Creep of Metallic Materials in the WP 
2.1.07.05.0B 
Creep of Metallic Materials in the DS 
2.1.09.03.0B 
Volume Increase of Corrosion Products Impacts WP 
2.1.09.09.0A 
Electrochemical Effects in EBS 
Cladding 
2.1.11.06.0A 
Thermal Sensitization of WPs 
2.1.11.06.0B 
Thermal Sensitization of DSs 
2.1.11.07.0A 
Thermal Expansion/Stress of In-Drift EBS Components 
EBS 
2.1.12.03.0A 
Gas Generation (H2) From WP Corrosion 
EBS Cladding 
2.1.13.01.0A 
Radiolysis 
EBS, WF 
2.1.13.02.0A 
Radiation Damage in EBS 
EBS, WF 
EBS = engineered barrier system, WPs = waste packages, DSs = drip shields, SSC = stress corrosion cracking, 
MIC = microbially influenced corrosion, WF  waste form 
DTN: MO0307SEPFEPS4.000 [164527] 

6.1 APPROACH 
The approach used for this analysis is a combination of qualitative and quantitative screening of 
FEPs. The analyses are based on the regulatory requirements provided by the NRC in 
10 CFR Part 63 [DIRS 158535] to determine whether or not each of the 33 waste package and 
drip shield FEPs should be included in the TSPA or excluded from further analysis. Each FEP is 
screened against the low-probability of occurrence and low-consequence criteria in Section 4.2 
of this analysis report. 
For FEPs that are excluded from the TSPA based on the NRC regulatory requirements, the 
screening argument includes a summary of the basis and results that indicate either low 
probability of occurrence or low consequence to radiological exposures to the RMEI and 
radionuclide releases to the accessible environment. As appropriate, screening arguments cite 
work done outside the YMP, such as in other scientific analyses. The Screening Argument 
section also addresses each of the descriptor phrases, which provide additional detail about the 
subject and content of the FEP beyond the FEP name and description. The screening arguments 
are documented in sufficient enough detail to meet the criteria listed in Section 4.2 of this 
analysis report. 
For FEPs that are included in the TSPA, the TSPA Disposition section of each FEP discussion 
includes a short summary of how the FEP has been incorporated in the process models or the 
TSPA models and a reference to the scientific analysis that describes the disposition in greater 
detail. The TSPA Disposition section also addresses each of the descriptor phrases, which 
provide additional detail about the subject and content of the FEP beyond the FEP name and 
description. The disposition statements are documented in sufficient enough detail to meet the 
criteria listed in Section 4.2 of this AR. 
6.2 FEATURES, EVENTS, AND PROCESSES ANALYSES 
This scientific analysis addresses the 33 FEPs that pertain to waste package and drip shield 
degradation. 
6.2.1 Error in Waste Emplacement 
6.2.1.1 Features, Events, and Process Number 
FEP 1.1.03.01.0A 
6.2.1.2 Features, Events, and Process 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 involves erroneously 
emplacing the waste packages in a saturated or wet zone of the repository. This would clearly 
impact the repository performance both by impacting waste package corrosion and radionuclide 
transport. 

6.2.1.3 Descriptor Phrases 
Waste emplacement error (mechanical impact) 
Deviations from design 
Quality control (inadequate) 
Waste emplacement error (placement in wet zone) 
Drip shield emplacement error 
6.2.1.4 Screening Decision 
Excluded  Low Consequence 
6.2.1.5 Screening Argument 
Analyses presented in Analysis of Mechanisms for Early Waste Package Failure (BSC 2003 
[DIRS 164475], Section 6) indicate that the consequences of errors in waste package 
emplacement are very low because of the administrative and procedural control measures that 
will be instituted. 
Accidental misloading of waste packages could result in thermal outputs not within the expected 
range. These misloadings could cause the waste package surface temperatures and relative 
humidities to be outside the expected ranges and impact the degradation characteristics of the 
waste package and drip shield. Analyses presented in the Analysis of Mechanisms for Early 
Waste Package Failure (BSC 2003 [DIRS 164475], Section 6.2.8), indicate that thermal 
misloads would not be expected to have any significant consequences for postclosure 
performance of the waste package/drip shield barriers. This is because the waste package has a 
large heat transfer area and therefore, the increased heat output from an overloaded waste 
package would be quickly dissipated into the emplacement drift and not alter the waste 
package/drip shield surface temperatures to an extent significant enough to affect postclosure 
performance. On this basis, the effects of thermal misloadings are excluded based on negligible 
consequence to radiological exposures to the RMEI and radionuclide releases to the accessible 
environment. 
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. These effects are excluded based on low consequence to 
radiological exposures to the RMEI and radionuclide releases to the accessible environment as 
discussed above. 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 Disposal Criticality Analysis Methodology Topical Report (YMP 2003 
[DIRS 165505]). 

The current engineered barrier design includes a titanium drip shield that is placed over the waste 
packages (BSC 2004 [DIRS 167040]). The drip shield will be continuous down the entire length 
of the drift, and will be fabricated and emplaced in segments. Each segment will slightly overlap 
the previously emplaced segment. The benefits of the drip shield could be diminished for a 
particular waste package if a drip shield segment fails to overlap with the previously emplaced 
segment, such that a large separation exists that would allow any dripping water above it to 
directly fall onto the waste package below. The Analysis of Mechanisms for Early Waste 
Package Failure (BSC 2003 [DIRS 164475], Section 6.4.7) indicates that drip shield 
emplacement errors large enough to allow dripping water to contact the underlying waste 
package would be detected. Thus, errors in drip shield emplacement can be excluded based on 
low consequence to radiological exposures to the RMEI and radionuclide releases to the 
accessible environment since drip shield emplacement errors large enough to allow dripping 
water to contact the underlying waste package would be detected. 
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 due to the fact that the drip 
shields must fail in order 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 failure 
mode of any consequence to repository performance (note that seismic effects are not considered 
in this section). As shown in the General Corrosion and Localized Corrosion of Drip Shield 
report (BSC 2003 [DIRS 161236], Tables 13 and 14) the maximum general corrosion rate which 
can be applied to the underside of drip shield is about 1.13x10-4 mm/yr and the maximum 
general corrosion rate which can be applied to the top side of drip shield is about 3.20x10-4 
mm/yr. Therefore, the earliest time at which the 15-mm thick drip shield plates can fail by 
general corrosion is about 35 thousand years. Therefore, emplacement of waste packages in a 
wet zone can be excluded based on low consequence to radiological exposures to the RMEI and 
radionuclide releases to the accessible environment. 
6.2.1.6 Total System Performance Assessment Disposition 
N/A 
6.2.1.7 Supporting Analysis Reports 
Analysis of Mechanisms for Early Waste Package Failure (BSC 2003 [DIRS 164475]) 
General Corrosion and Localized Corrosion of Drip Shield (BSC 2003 [DIRS 161236]) 
6.2.2 General Corrosion of Waste Packages 
6.2.2.1 Features, Events, and Process Number 
FEP 2.1.03.01.0A 

6.2.2.2 Features, Events, and Process Description 
General corrosion may contribute to waste package failure. 
6.2.2.3 Descriptor Phrases 
General (uniform) corrosion of waste packages 
Aqueous corrosion 
Dry-air oxidation 
Humid-air corrosion 
Effects of trace metals on general corrosion of waste packages 
Effects of corrosive gases on corrosion of waste packages 
6.2.2.4 Screening Decision 
Included 
6.2.2.5 Screening Argument 
N/A 
6.2.2.6 TSPA Disposition 
General corrosion is included in waste package degradation analysis. This includes the effects of 
corrosive gases. Because general corrosion is likely to be operative for most of the repository 
operation period, it is one of the key corrosion processes that could lead to degradation and 
failure of waste packages in the repository. General corrosion due to dry-air oxidation, aqueous 
corrosion, microbially influenced corrosion and aging and phase instability of the waste package 
outer barrier are discussed in the General Corrosion and Localized Corrosion of Waste Package 
Outer Barrier (BSC 2003 [DIRS 161235]) model report. 
It was concluded in the General Corrosion and Localized Corrosion of the Waste Package Outer 
Barrier model report (BSC 2003 [DIRS 161235], Section 6.4.2) 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 package outer barrier were exposed for 10,000 years at 350C). 
Therefore, dry oxidation does not need to be considered in TSPA analyses. 
Penetration rates for general corrosion are provided in Section 6.4.3 of the General Corrosion 
and Localized Corrosion of the Waste Package Outer Barrier model report (BSC 2003 
[DIRS 161235]). General corrosion rates of the waste package outer barrier were estimated 
using the weight-loss of Alloy 22 crevice geometry specimens after 5-year exposure in the Long 
Term Corrosion Test Facility (BSC 2003 [DIRS 161235], Section 6.4.3). General corrosion 
progresses uniformly at a time independent constant rate and the depth of penetration or thinning 
of the waste package outer barrier by general corrosion is equal to the general corrosion rate 



number of penetration openings in the waste package and drip shield as a function of time. The 
Integrated Waste Package Degradation Model is used directly in the TSPA-LA analysis. 
The TSPA-LA waste package degradation analysis simulates the behavior of a few hundred 
waste packages (BSC 2003 [DIRS 161317], 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 that 
were considered to vary over the repository are relative humidity and temperature at the waste 
package surface. In addition, potentially variable corrosion processes within a single waste 
package are represented by dividing the waste package surface into subareas called patches 
and stochastically sampling the degradation model parameter values for each patch. The use of 
patches explicitly represents the variability in degradation processes within 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 Model. For each realization, values 
are sampled for the uncertain degradation parameters and passed to the Integrated Waste 
Package Degradation Model. Each realization is a complete Integrated Waste Package 
Degradation Model simulation of a given number of waste packages, explicitly considering 
variability in the degradation processes. Accordingly, each of the Integrated Waste Package 
Degradation Model 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) which represent the potential range of the output parameters. For 
example, the waste-package failure time profiles are reported with a group of curves 
representing the cumulative probability of waste package failures as a function of time. The 
outputs of the Integrated Waste Package Degradation Model are used as input for waste form 
degradation analysis and radionuclide release analysis from failed waste packages conducted 
within the TSPA-LA model. 
6.2.2.7 Supporting Analysis Reports 
General Corrosion and Localized Corrosion of Waste Package Outer Barrier (BSC 2003 
[DIRS 161235]) 
WAPDEG Analysis of Waste Package and Drip Shield Degradation (BSC 2003 [DIRS 161317]) 
6.2.3 General Corrosion of Drip Shields 
6.2.3.1 Features, Events, and Process Number 
FEP 2.1.03.01.0B 
6.2.3.2 FEP Description 
General corrosion may contribute to drip shield failure. 

6.2.3.3 Descriptor Phrases 
General (uniform) corrosion of drip shields 
Aqueous corrosion 
Dry-air oxidation 
Humid-air corrosion 
Effects of trace metals on general corrosion of drip shields 
Effects of corrosive gases on corrosion of drip shields 
6.2.3.4 Screening Decision 
Included 
6.2.3.5 Screening Argument 
N/A 
6.2.3.6 Total System Performance Analysis 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 the General Corrosion and Localized Corrosion of the Drip 
Shield (BSC 2003 [DIRS 161236]) model report. 
It was concluded in the General Corrosion and Localized Corrosion of the Drip Shield model 
report (BSC 2003 [DIRS 161236], Section 8.2) that although dry air oxidation occurs, it results 
in a negligible amount of barrier thinning over repository time scales (only ~2,129 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 analyses. 
Penetration rates for general corrosion are provided in the General Corrosion and Localized 
Corrosion of the Drip Shield model report (BSC 2003 [DIRS 161236]) and are used in TSPA 
analyses. Both humid-air and aqueous corrosion processes are considered part of general 
corrosion (BSC 2003 [DIRS 161236], Sections 6.1 and 6.3). General corrosion rates of the drip 
shield were estimated with weight-loss data of Titanium Grade 16 samples after 1 and 5-year 
exposure in the Long Term Corrosion Test Facility (BSC 2003 [161236], Sections 6.1 and 6.3). 
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 LLNL Long Term 
Corrosion Test Facility. Samples of Titanium Grade 16, an excellent analogue for the more 
corrosion resistant Titanium Grade 7, were immersed in simulated acidified water for five years 
(DTN: LL021012712251.021 [DIRS 163112]) in the LLNL Long Term Corrosion Test Facility. 
After five years, a sample of the acidified water, including any corrosion products, was analyzed 

for trace metals using inductively coupled plasma/mass spectrometry methods. Low 
concentrations of potentially deleterious metals, including arsenic, calcium, and magnesium, 
were detected. Given that 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 2003 [DIRS 161236], Section 6.5). On this basis, the effects 
of trace metals on general corrosion of the drip shield are included in TSPA although the effects 
are not explicitly and separately modeled. 
The drip shield outer surface may be exposed to a more complicated chemistry and geometry 
than the drip shield inner surface since dust and/or mineral films (from evaporation of dripping 
water) 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 nor significant dust film formation (BSC 2003 
[DIRS 161236], Sections 6.1 and 6.3). Therefore, the general corrosion of the inner surface and 
the outer surface of the drip shield are modeled by using different sets of corrosion data 
(BSC 2003 [161236], Sections 6.1 and 6.3). Based on these test results, the corrosion (or 
oxidation) rates range from 0 to 113 nm/year with the 50th percentile being 16 nm/year for the 
underside of the drip shield (BSC 2003 [DIRS 161236], Section 6.5.3). For the outer surface of 
the drip shield, the corrosion rates range from 0 to 320 nm/year with the 50th percentile being 
about 25 nm/year (BSC 2003 [DIRS 161236], Section 6.5.3). 
The variations in the drip shield general corrosion rate distributions are considered to be entirely 
due to uncertainty (BSC 2003 [DIRS 161236], Section 6.3.4). For each realization of the 
Integrated Waste Package Degradation Model (BSC 2003 [DIRS 161317]), 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 which is applicable to 
the drip shield underside and one general corrosion rate value is sampled from the general 
corrosion rate distribution which is applicable to the drip shield outer surface). The two sampled 
values are then applied one to the drip shield outer surface and one to the Drip shield inner 
surface 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 1.13x10-4 mm/year and the maximum general 
corrosion rate for the CDF applied to the top side of the drip shield is approximately 3.2x10-4 
mm/year (BSC 2003 [DIRS 161236], Section 6.3.5); therefore, the earliest possible drip shield 
failure by general corrosion is about 35,000 years. 
The general corrosion rate distributions for the drip shields are incorporated into the Integrated 
Waste Package Degradation Model (BSC 2003 [DIRS 161317]). The output from the Integrated 
Waste Package Degradation Model 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 Model is used directly in the 
TSPA-LA analysis. 
Additional aqueous corrosion processes are addressed in other FEPs: Localized Corrosion of 
Drip Shields (FEP 2.1.03.03.0B), Stress Corrosion Cracking (SCC) of Drip Shields 
(FEP 2.1.03.02.0B), Hydride Cracking of Drip Shields (FEP 2.1.03.04.0B), and Microbially 

Influenced Corrosion (MIC) of Drip Shields (FEP 2.1.03.05.0B). These processes are described 
in BSC 2003 [DIRS 161236], BSC 2003 [DIRS 161234], and BSC 2003 [DIRS 161759]. As 
discussed in FEP 2.1.03.05.0B, Microbially Influenced Corrosion of Drip Shields, the drip shield 
is considered to be immune to Microbially Influenced Corrosion. Also, as discussed in 
FEP 2.1.11.06.0B, Thermal Sensitization of Drip Shields, aging and phase instability has no 
effect on drip shield degradation processes. 
6.2.3.7 Supporting Analysis Reports 
General Corrosion and Localized Corrosion of the Drip Shield (BSC 2003 [DIRS 161236]) 
WAPDEG Analysis of Waste Package and Drip Shield Degradation (BSC 2003 [DIRS 161317]) 
6.2.4 Stress Corrosion Cracking of Waste Packages 
6.2.4.1 Feature, Event, and Process Number 
FEP 2.1.03.02.0A 
6.2.4.2 Feature, Event, and Process Description 
At specific locations where waste packages become wet and are stressed, stress corrosion 
cracking ensues. The possibility of SCC under dry conditions or due to thermal stresses are also 
addressed as part of this FEP. 
6.2.4.3 Descriptor Phrases 
Waste package closure welds (stress corrosion cracking) 
Waste package fabrication welds (stress corrosion cracking) 
Threshold relative humidity (stress corrosion cracking) 
6.2.4.4 Screening Decision 
Included 
6.2.4.5 Screening Argument 
N/A 
6.2.4.6 Total System Performance Assessment Disposition 
SCC of the waste package outer barrier closure weld regions is included in TSPA as part of 
waste package degradation analyses. The Stress Corrosion Cracking of the Drip Shield, the 
Waste Package Outer Barrier, and the Stainless Steel Structural Material model report 
(BSC 2003 [DIRS 161234], Section 8) provides input to the TSPA for waste package 
degradation. 

As discussed in the Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer 
Barrier, and the Stainless Steel Structural Material model report (BSC 2003 [DIRS 161234], 
Section 6), the slip dissolution/film rupture model was used to assess the failure (or lack of it) of 
the waste package due to the SCC crack propagation for given manufacturing cracks and/or 
cracks initiated by the combined effects of stress and environment. The threshold stress intensity 
factor is based on the theory that there exists a threshold value for the stress intensity factor at the 
crack tip below which a pre-existing crack or flaw does not grow. The stress intensity factor 
provides a criterion for determining if a SCC crack will reach an arrest state or enter the 
propagation phase. 
The application of the SCC models to the waste package and drip shield also requires input of 
weld residual stress profiles and stress intensity factor profiles along with uncertainty and 
variability (BSC 2003 [DIRS 161234]). These input data were developed for the 25-mm outer 
lid (subjected to laser peening) and the as-welded 10-mm middle lid. This process model report 
(BSC 2003 [DIRS 161234]) also provides other needed 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, size, density and orientation distributions for manufacturing flaws or 
defects, and an estimate of crack opening size. This SCC treatment applies to both CSNF and 
co-disposed waste packages. 
Because, among other exposure condition parameters, tensile stress is required to initiate SCC 
and the waste package closure welds are the only places under such tensile stresses, only the 
waste package closure welds are considered subject to SCC (BSC 2003 [DIRS 161234], 
Section 6.4.2). Welds are the most susceptible to SCC because (1) welding can produce high 
tensile residual stress in the weld; (2) pre-existing 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 non-equilibrium brittle phases, which could enhance material susceptibility to 
SCC (BSC 2003 [DIRS 161234], Section 6.4.2). SCC of the fabrication welds of the waste 
package outer barrier will not occur due to the resistance of Alloy 22 (UNS N06022) to SCC 
when under tensile stress and because the fabrication welds will be fully annealed before waste is 
loaded into the waste containers (Plinski 2001 [DIRS 156800], Section 8.1.7). It is recognized 
that 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 the Engineered 
Barrier System Features, Events, and Processes report (BSC 2004 [DIRS 167253]). 
The Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the 
Stainless Steel Structural Material model report (BSC 2003 [DIRS 161234], Section 6.2.1) 
describes the results of SCC crack initiation measurements under constant load conditions while 
immersed in basic saturated water. Alloy 22 (UNS N06022) exhibits excellent SCC resistance 
since failure 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 60 and 90oC in the Long Term Corrosion Test Facility 
(BSC 2003 [DIRS 161234]). No evidence of SCC initiation has been observed in these U-bend 
specimens after five years of exposure. 
The presence of stable liquid water is required to initiate corrosion processes (including SCC) 
that are supported by electrochemical corrosion reactions. Threshold relative humidity is used in 
the waste package degradation analysis to simulate such a corrosion initiation condition. Under 
conditions with the relative humidity below the threshold value, SCC will not occur. 
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 Model 
(BSC 2003 [DIRS 161317]). The output from the Integrated Waste Package Degradation Model 
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 Model is used directly in the TSPA-LA analysis. 
Lead has been identified as a potential contributor to stress corrosion cracking in nickel-based 
alloys (Sakai, et al. 1992 [DIRS 154465]; Pan, et al. 2002 [DIRS 165536]; Pulvirenti et al. 2002 
[DIRS 165537]). One should note that the results of 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 investigating 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 the Stress Corrosion Cracking of the Drip Shield, the Waste Package 
Outer Barrier, and the Stainless Steel Structural Material model report (BSC 2003 
[DIRS 161234], Section 6.3.4). SCC initiation test results were obtained using Slow Strain Rate 
Tests, at 76-95C 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]). In this latter experiment a Basic Saturated 
Water solution with a pH ~ 12 at 110C was used. After 8670 hours of exposure in the lead-free 
solution,1000 ppm Pb (as PbNO3) was added to the test solution and SCC crack growth rates 
were monitored for ~1800 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 ground waters is expected to have little consequence on 
Alloy 22 (UNS N06022) SCC processes. 

6.2.4.7 Supporting Analysis Reports 
Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the 
Stainless Steel Structural Material (BSC 2003 [161234]) 
WAPDEG Analysis of Waste Package and Drip Shield Degradation (BSC 2003 [161317]) 
6.2.5 Stress Corrosion Cracking of Drip Shields 
6.2.5.1 Features, Events, and Process Number 
FEP 2.1.03.02.0B 
6.2.5.2 Features, Events, and Process Description 
At specific locations where drip shields become wet and are stressed, stress corrosion cracking 
ensues. The possibility of SCC under dry conditions or due to thermal stresses is also addressed 
as part of this FEP. 
6.2.5.3 Descriptor Phrases 
Drip shield fabrication welds (stress corrosion cracking) 
Drip shield cracks (stress corrosion cracking) 
Drip shield crack plugging 
6.2.5.4 Screening Decision 
Excluded  Low Consequence 
6.2.5.5 Screening Argument 
For the drip shields, all the 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 2003 [DIRS 161234], Section 6.3.7). However, the 
drip shields are subject to SCC under the action of seismic-induced loading and rockfalls. 
Seismic effects on drip shield degradation are discussed in FEP 2.1.07.01.0A, Rockfall and 
FEP 1.2.03.02.0A, Seismic Ground Motion Damages EBS Components (BSC 2004 
[DIRS 167253]). In the nominal case (in the absence of seismic-induced loading and rockfalls) 
even if SCC of the drip shield were to occur, cracks in passive alloys, such as Titanium Grade 7, 
tend to be tight (i.e., small crack opening displacement) (BSC 2003 [DIRS 161234], Section 
6.3.7). The opposing sides of through-wall cracks will continue to corrode at very low passive 
corrosion rates until the gap region of the tight crack opening is plugged by corrosion products 
and precipitates such as carbonate minerals. As discussed in Stress Corrosion Cracking of the 
Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material model 
document (BSC 2003 [DIRS 161234], Section 6.3.7), SCC cracks are sealed in a few hundred 
years at most when water is allowed to flow through the cracks at the expected low flow rate. 
When the cracks are bridged by water, the sealing process may take thousands of years, but no 
flow occurs since the water is held by capillary forces. Following plugging of the crack, any 

solution flow through the crack would be dominated by an efficiency factor determined by the 
ratio of solution run-off on the drip shield surface compared to through crack flow which in turn 
is determined by scale porosity/permeability. Because of the expected high density of the calcite 
deposits and lack of pressure gradient to drive water through the crack, the probability of 
solution flow through the crack would approach zero (BSC 2003 [DIRS 161234], Section 6.3.7). 
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 its intended design purpose. 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. 
6.2.5.6 Total System Performance Assessment Disposition 
N/A 
6.2.5.7 Supporting Analysis Reports 
Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the 
Stainless Steel Structural Material (BSC 2003 [DIRS 161234]) 
6.2.6 Localized Corrosion of Waste Packages 
6.2.6.1 Features, Events, and Process Number 
FEP 2.1.03.03.0A 
6.2.6.2 Features, Events, and Process Description 
Localized corrosion (pitting and crevice corrosion) could enhance degradation of the waste 
packages. 
6.2.6.3 Descriptor Phrases 
Pitting corrosion of waste packages 
Crevice corrosion of waste packages 
Effects of trace metals on localized corrosion of waste packages 
6.2.6.4 Screening Decision 
Included 
6.2.6.5 Screening Argument 
N/A 

6.2.6.6 Total System Performance Assessment Disposition 
Localized corrosion (pitting and crevice) is a corrosion mode that could lead to eventual 
compromise of waste packages in the repository. As discussed in the General Corrosion and 
Localized Corrosion of Waste Package Outer Barrier model report (BSC 2003 [DIRS 161235]), 
localized corrosion of the waste package outer barrier (Alloy 22) (UNS N06022) is not likely to 
occur under repository-relevant exposure conditions particularly during the time period when the 
drip shield performs its design function and prevents the seepage water from directly contacting 
the underlying waste package. Since the probability of occurrence of localized corrosion has not 
been quantified, localized corrosion initiation and propagation models are included in TSPA to 
be conservative. 
A possible scenario for the waste package outer barrier to be subjected to localized corrosion in 
the repository is the concurrent occurrence of the drip shield failure and direct contact of the 
waste package with seepage water during the first few hundred years of active thermal 
distribution after closure. The seepage water with a characteristic chemistry could evolve to 
highly chloride-containing brines by evaporative concentration. After the active thermal 
perturbation period, the waste package temperature slowly decreases with time. Once it cools to 
the temperature that is lower than the minimum temperature required for localized corrosion 
initiation in highly concentrated chloride-containing brines, the waste packages are completely 
immune to localized corrosion. 
Localized corrosion of the waste package outer barrier is modeled with two model components: 
initiation model and propagation model (BSC 2003 [161235], Section 6.4.4; BSC 2003 
[DIRS 161317], Section 6.3.6). The initiation model assumes that localized corrosion of the 
waste package outer barrier occurs when the open circuit potential, or corrosion potential (Ecorr), 
is equal to or greater than a certain critical 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 (e.g., Ecorr and Ecritical) could be affected by the sample 
configuration (crevice, disk, or rod), metallurgical conditions (mill or annealed or welded), and 
exposure condition (temperature, pH, chloride ion concentration, 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. The crevice repassivation model is expressed as 
follows. 
(Eq. 5) - . + = 3 NO 
rcrev 
o
rcrev rcrev E E E 
where is the crevice repassivation potential in the absence of nitrate ions (which tend to 
inhibit localized corrosion (LC) initiation), and is the crevice repassivation potential 
changes resulted from the inhibiting effect of nitrate ion in solution (BSC 2003 [DIRS 161235], 
Section 6.4.4). 
o
rcrev E 
- . 3 NO 
rcrev E 

Further, the functional form for was evaluated by fitting the crevice repassivation 
potential differences between the E 
- . 3 NO 
rp E 
rcrev data with and without nitrate ions for the same chloride 
and temperature conditions. The effect of nitrate ions on the crevice repassivation potential is 
represented as follows. 
] [ 
] [ ] [ 3 
2 3 1 
3 
-
- 
- + + = . - 
Cl 
NO b NO b b E o 
NO 
rp (Eq. 6) 
The effect of the interaction of the competing aggressive ion (chloride ion) and inhibiting ion 
(nitrate ion) on the crevice repassivation potential is represented with the ratio of the 
concentrations of the competing ions (BSC 2003 [DIRS 161235], Section 6.4.4). 
Nitrate ions have a strong inhibitive effect on localized corrosion of Alloy 22 in chloride 
containing solutions (BSC 2003 [DIRS 161235], Section 6.4.4; Dunn and Brossia 2002 
[DIRS 162213]). It is known that anions containing nitrogen, phosphorus and sulfur that are 
abundant in the repository groundwater exhibit varying degrees of inhibitive effects 
(Thomas 1994 [DIRS 120498]). An important function of inhibitive anions is to 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], page 17:54). 
A model for the effect of inhibitive anions on the crevice repassivation potentials of Alloy 22 
(UNS N06022) (i.e. increase of the crevice repassivation potentials) was developed using the 
crevice repassivation potential data measured for the test solutions containing varying 
concentrations of chloride and nitrate ions. The environments tested cover temperatures from 
60C to 130C, chloride concentrations from 2.6 m to 20.8 m, and nitrate concentrations from 
0.03 m to 2.1 m (BSC 2003 [DIRS 161235], Section 6.4.4). In the model the interaction between 
the competing aggressive anion (chloride ion) and inhibitive anion (nitrate ion) on the crevice 
repassivation potential is represented with the ratio of the concentrations of the two competing 
ions and the concentration of nitrate ion. 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 2003 [DIRS 161235], Section 8.3). 
The analysis of the crevice corrosion initiation model for the Alloy 22 (UNS N06022) waste 
package outer barrier has shown that for a condition of neutral chloride-containing brines at 95C 
(10 m chloride and pH 7), the conservative lower bound of the nitrate to chloride concentration 
ratio for the crevice corrosion immunity is 0.03, and the conservative lower bound ratio for a 
condition of highly corrosive chloride-containing brines (10 m chloride and pH 3) at 95C is 0.18 
(BSC 2003 [DIRS 161235], Section 6.4.4, Figures 6-53 and 6-54). 

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 60 and 90C in the Long Term Corrosion Test Facility. The 
nitrate to chloride concentration ratio of these solutions is about 0.5. None of the crevice 
samples have 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 2003 [DIRS 161235], Section 7.3, Table 7-2). 
A recent 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 has 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]). 
All the experimental data and model analyses that are available to date indicate that the lower 
bound of the nitrate to chloride concentration ratio for the immunity of Alloy 22 (UNS N06022) 
to localized corrosion in chloride-containing brines is about 0.2. Therefore, a nitrate to chloride 
concentration ratio of 1.0 is selected as the conservative bounding measure for total immunity of 
the Alloy 22 waste package outer barrier to crevice corrosion for the environmental conditions 
expected in the repository. 
Implementation of the localized corrosion model will be carried outside of the main TSPA-LA 
model. The output of this analysis will be one or more environmentally independent uncertainty 
distributions for the number (or fraction) of packages that experience localized corrosion. These 
uncertainty distributions for the number of localized corrosion packages will then be 
incorporated and sampled in the main TSPA-LA model. The main premise is that localized 
corrosion is a rare occurrence and need only be applied to a small number of waste packages in 
each TSPA realization. This localized corrosion treatment applies to both CSNF and 
Co-disposed waste packages. 
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 analyses. These same 
solutions were considered in developing the localized corrosion initiation criterion (BSC 2003 
[DIRS 161235], Section 6.4.4; DTN: LL021012712251.021 [DIRS 163112]). The small 
amounts of trace metals identified in the acidified waters were appropriately considered in 
analyzing localized corrosion of waste packages. 
6.2.6.7 Supporting Analysis Reports 
General Corrosion and Localized Corrosion of Waste Package Outer Barrier (BSC 2003 
[DIRS 161235]) 
WAPDEG Analysis of Waste Package and Drip Shield Degradation (BSC 2003 [DIRS 161317]) 

6.2.7 Localized Corrosion of Drip Shields 
6.2.7.1 Feature, Event, and Process Number 
FEP 2.1.03.03.0B 
6.2.7.2 Feature, Event, and Process Description 
Localized corrosion (pitting and crevice corrosion) could enhance degradation of the drip 
shields. 
6.2.7.3 Descriptor Phrases 
Pitting corrosion of drip shields 
Crevice corrosion of drip shields 
Effects of trace metals on localized corrosion of drip shields 
6.2.7.4 Screening Decision 
Excluded  Low Consequence 
6.2.7.5 Screening Argument 
As discussed in the General Corrosion and Localized Corrosion of the Drip Shield model report 
(BSC 2003 [DIRS 161236], Section 6.4), localized corrosion of the drip shield will not occur 
under the exposure conditions in the repository. A relationship between exposure parameters 
(temperature, chloride ion concentration, and pH) and the difference between the critical 
potential and the corrosion potential was developed. The localized corrosion model for the 
titanium drip shield assumes that localized attack occurs if the open circuit corrosion potential 
exceeds the threshold potential for breakdown of the passive film (BSC 2003 [DIRS 161236]). 
The critical potential versus temperature and composition model of various test media indicates 
that localized corrosion of Titanium Grade 7 would not initiate in repository-relevant 
environments even at pH values as high as 14 (BSC 2003 [DIRS 161236]). 
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 model report (BSC 2003 [DIRS 161236], Section 6.3.8, 6.4.4, and 6.4.5). The 
general corrosion rates for Titanium Grade 7 are low when in the presence of these species or 
when crevices are present on the drip shield. 
As discussed in FEP 2.1.03.01.0B, General Corrosion of Drip Shields, acidified waters that 
contain trace metals such as arsenic, calcium, and magnesium were used to develop the Titanium 
Grade 7 general corrosion rates used in the TSPA analyses. These same solutions were used in 
developing the localized corrosion initiation criterion (BSC 2003 [DIRS 161236], Section 6.4; 
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. 
Since localized corrosion of Titanium Grade 7 will not initiate under the expected repository 
conditions and general corrosion rates are low when crevices are present, this FEP can be 
excluded based on low consequence because its omission will not have a significant effect on 
radiological exposures to the RMEI and radionuclide releases to the accessible environment. 
6.2.7.6 Total System Performance Assessment Disposition 
N/A 
6.2.7.7 Supporting Analysis Reports 
General Corrosion and Localized Corrosion of the Drip Shield (BSC 2003 [DIRS 161236]) 
6.2.8 Hydride Cracking of Waste Packages 
6.2.8.1 Feature, Event, and Process Number 
FEP 2.1.03.04.0A 
6.2.8.2 Feature, Event, and Process Description 
The uptake of hydrogen and the formation of metal hydrides may mechanically weaken the 
waste packages and promote corrosion. 
6.2.8.3 Descriptor Phrases 
Hydrogen induced cracking (waste packages) 
Hydride embrittlement (waste packages) 
Galvanic coupling (drip shield-ground support) 
Hydride cracking (waste package) 
6.2.8.4 Screening Decision 
Excluded  Low Consequence 
6.2.8.5 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 (BSC 2003 [DIRS 161759], Section 6.1.1). It may be more 
appropriate to use a more general term, HIC, 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. 
HIC of the waste package outer barrier (Alloy 22) [UNS N06022])is not considered to be a 
credible degradation mechanism under repository-relevant exposure conditions. Handbook data 
(ASM International 1987 [DIRS 103753], p. 650-652) indicate that fully annealed nickel-base 
alloys, such as Alloy 22, may be immune to HIC. Similar to the drip shield material, the 
extremely low corrosion rates exhibited by nickel alloys are not sufficient to generate enough 
hydrogen to cause HIC. HIC of these alloys is generally not observed unless their yield strengths 
are increased through heavy cold working or certain heat treatments. Several Ni-Cr-Mo alloys 
with compositions and properties similar to Alloy 22 (Alloy C-276, Alloy C-4, and Alloy 625) 
maintain their resistance to HIC even when heavily cold worked to yield strengths in excess of 
1240 MPa (180 ksi) (ASM International 1987 [DIRS 103753], p. 169). Cold worked Ni-Cr-Mo 
alloys are not susceptible to HIC 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 and/or 
grain-boundary segregation of deleterious elements such as phosphorous and sulfur which can 
increase susceptibility to HIC (ASM International 1987 [DIRS 103753], p. 169). However, since 
the waste package temperature never exceeds 190C (CRWMS M&O 2001 [DIRS 154594], 
Section 6.3.1) 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 about HIC 
of the waste package are applicable to both CSNF and Co-disposed waste packages. 
As discussed in FEP 2.1.09.09.0A, Electrochemical Effects in EBS, electrical contact between 
the Alloy 22 waste package outer barrier and Titanium Grade 7 drip shield is not expected and 
would not be more detrimental to corrosion due to the similarity of these materials corrosion 
potential. 
In summary, HIC of Alloy 22 is not expected to occur under the anticipated repository 
conditions. Even in the unlikely case of HIC resulting from galvanic coupling, the corrosion 
behavior of Alloy 22 would not be affected. Therefore, this FEP can be excluded based on low 
consequence to radiological exposures to the RMEI and radionuclide releases to the accessible 
environment. 
6.2.8.6 Total System Performance Assessment Disposition 
N/A 
6.2.8.7 Supporting Analysis Reports 
Hydrogen Induced Cracking of Drip Shield (BSC 2003 [DIRS 161759]) 

6.2.9 Hydride Cracking of Drip Shields 
6.2.9.1 Feature, Event, and Process Number 
FEP 2.1.03.04.0B 
6.2.9.2 Feature, Event, and Process Description 
The uptake of hydrogen and the formation of metal hydrides may mechanically weaken the drip 
shields and promote corrosion. 
6.2.9.3 Descriptor Phrases 
Hydrogen induced cracking (drip shield) 
Hydride embrittlement (drip shield) 
Galvanic coupling (drip shield-ground support) 
Drip shield crack plugging 
Hydride cracking (drip shield) 
6.2.9.4 Screening Decision 
Excluded  Low Consequence 
6.2.9.5 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. It may be more appropriate to use a more general term, HIC, to 
refer to the impact of hydrogen on waste package and drip shield materials. HIC results from the 
combined action of hydrogen and residual or sustained applied tensile stresses. 
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 2003 
[DIRS 161759], Section 6.1.2): 
1. A mechanism for generating nascent (atomic) hydrogen on the surface. 
2. Metal temperature above approximately 80oC (175oF) where the diffusion rate of 
hydrogen into a-titanium is significant. 
3. Solution pH less than 3 or greater than 12, or impressed potentials more negative than 
-0.7 V (Saturated Calomel Electrode). 

As discussed in General Corrosion and Localized Corrosion of Drip Shield (BSC 2003 
[DIRS 161236], Section 6.4) and 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 current repository design (BSC 2004 [DIRS 167040]), 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 high dose 
rate (> 102 Gy/h) and high temperature (> 150oC) (BSC 2003 [DIRS 161759], Section 6.1.2). 
These conditions are unattainable in the repository (BSC 2003 [DIRS 161759], Section 6.1.2). 
At certain repository locations, where temperatures are high (= 80C) and concentrated 
groundwater is present, conditions two and three may also be satisfied. When all three 
conditions are present simultaneously, hydrogen absorption can be anticipated. 
As discussed in the Hydrogen Induced Cracking of Drip Shield model report (BSC 2003 
[DIRS 161759], Section 8), a simple and conservative model was developed to evaluate the 
effects of HIC 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, noticeable hydrogen induced cracking 
of Titanium Grade 7 is not expected. This is mainly due to the existence of the oxide film and 
the high critical value, Hc, due to palladium (Pd) addition. This passive film acts as a good 
transport barrier to hydrogen absorption; the impermeability of this film to hydrogen absorption 
will be improved during a period of dry thermal oxidation expected under repository conditions 
(BSC 2003 [DIRS 161759], Section 6.1.5). 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 1000 g.g-1 (BSC 2003 [DIRS 161759], 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 1000 g.g-1 and could be much higher (BSC 2003 
[DIRS 161759], Section 6.1.3). The hydrogen concentration in the drip shield from passive 
corrosion 10,000 years after emplacement is 755 g.g-1 resulting from a conservative estimate 
(BSC 2003 [DIRS 161759], Section 8). This is below the threshold concentration, and would not 
result in hydrogen induced cracking or any degradation of fracture toughness. 
In the current repository design, hydrogen generation may be caused by the galvanic couple 
formed between the titanium 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 high (= 80C) and concentrated ground waters are present, then the formation 
of locally hydrided hot spots are possible. However, the effect of these locally hydrided 
regions is negligible because (BSC 2003 [DIRS 161759], Section 6.3.2): 
1. The contact area is likely to be small, and the anode to cathode area (area of steel and 
titanium, respectively) low leading to only limited amounts of hydrogen absorption. 
2. The intermittent nature of seepage at high 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. 

3. Conditions in the repository will be oxidizing, making it less likely that the couple will 
sustain water reduction, and hence hydrogen absorption. 
4. Both the titanium drip shield and the steel component surfaces will experience a 
considerable period of dry high temperature (= 85C). This will leave both 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 Pd addition reducing the tendency for hydrogen induced 
cracking (BSC 2003 [DIRS 161759], Sections 6.1.3 and 6.1.5). 
In summary, HIC of 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 would be negligible. Therefore, this FEP can be excluded based on low consequence to 
radiological exposures to the RMEI and radionuclide releases to the accessible environment. 
6.2.9.6 Total System Performance Assessment Disposition 
N/A 
6.2.9.7 Supporting Analysis Reports 
General Corrosion and Localized Corrosion of Drip Shield (BSC 2003 [DIRS 161236]) 
Hydrogen Induced Cracking of Drip Shield (BSC 2003 [DIRS 161759]) 
6.2.10 Microbially Influenced Corrosion of Waste Packages 
6.2.10.1 Feature, Event, and Process Number 
FEP 2.1.03.05.0A 
6.2.10.2 Feature, Event, and Process 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. 
6.2.10.3 Descriptor Phrases 
Microbially influenced corrosion of waste packages 
6.2.10.4 Screening Decision 
Included 

6.2.10.5 Screening Argument 
N/A 
6.2.10.6 Total System Performance Assessment Disposition 
MIC is included in TSPA as part of the waste package degradation analysis (BSC 2003 
[DIRS 161235], Table 6-1). Waste package MIC is discussed in the General Corrosion and 
Localized Corrosion of Waste Package Outer Barrier model report (BSC 2003 [DIRS 161235]). 
The potential effect of MIC on waste package corrosion is analyzed with an enhancement factor 
approach, i.e. MIC increases the general corrosion penetration rate. In this approach, the abiotic 
corrosion rate is multiplied by the enhancement factor when the exposure conditions in the 
emplacement drift warrant significant microbial activity (BSC 2003 [DIRS 161235], 
Section 6.4.3.5 and Table 6-1). The MIC enhancement factor is taken to be uniformly 
distributed between 1 and 2 (BSC 2003 [DIRS 161235], Section 6.4.5). The MIC factor 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 2003 [DIRS 161235], 
Section 6.4.5). The general corrosion rate enhancement factor is applied to the entire waste 
package surface (BSC 2003 [DIRS 161235], Section 6.4.5) when the relative humidity threshold 
is satisfied. This treatment of MIC applies to both CSNF and co-disposed waste packages. 
In the Integrated Waste Package Degradation Model (BSC 2003 [DIRS 161317]), the general 
corrosion rate enhancement factor is sampled once per realization, i.e. the variation in the general 
corrosion rate MIC enhancement factor is entirely due to uncertainty (BSC 2003 [DIRS 161235], 
Section 6.4.5), and applied to the entire waste package surface. The output from the Integrated 
Waste Package Degradation Model 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 Model is used directly in the 
TSPA-LA analysis. 
6.2.10.7 Supporting Analysis Reports 
General Corrosion and Localized Corrosion of Waste Package Outer Barrier (BSC 2003 
[DIRS 161235]) 
WAPDEG Analysis of Waste Package and Drip Shield Degradation (BSC 2003 [DIRS 161317]) 
6.2.11 Microbially Influenced Corrosion of Drip Shields 
6.2.11.1 Feature, Event, and Process Number 
FEP 2.1.03.05.0B 
6.2.11.2 Feature, Event, and Process 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. 

6.2.11.3 Descriptor Phrases 
Microbially influenced corrosion of drip shields 
6.2.11.4 Screening Decision 
Excluded  Low Consequence 
6.2.11.5 Screening Argument 
MIC of titanium is discussed in the General Corrosion and Localized Corrosion of the Drip 
Shield report (BSC 2003 [DIRS 161236], Section 6.5.2). Corrosion handbooks and literature 
reviews generally state that titanium alloys are immune to MIC (Revie 2000 [DIRS 159370], 
Chapter 47; Little and Wagner 1996 [DIRS 131533]; Brossia et al. 2001 [DIRS 159836], 
Section 4.1.3). It is the remarkable stability of the TiO2 passive film formed on titanium alloys 
which 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). 
It is clear from Figure 20 and Figure 21 of the General Corrosion and Localized Corrosion of 
the Drip Shield (BSC 2003 [DIRS 161236], Section 6.4.3) that .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 density of defects (Shoesmith and Ikeda 1997 
[DIRS 151179], Section 3 and 6). According to Shoesmith et al. (1995 [DIRS 117892]), the 
initiation of crevice corrosion under biofilms has never been observed for titanium. Lastly, 
microbial growth in the repository will likely be limited by the availability of nutrients 
(BSC 2003 [DIRS 161235], Section 6.4.5). 
MIC is expected to have 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. 
6.2.11.5 Total System Performance Assessment Disposition 
N/A 

6.2.11.6 Supporting Analysis Reports 
General Corrosion and Localized Corrosion of the Drip Shield (BSC 2003 [DIRS 161236]) 
6.2.12 Internal Corrosion of Waste Packages Prior to Breach 
6.2.12.1 Feature, Event, and Process Number 
FEP 2.1.03.06.0A 
6.2.12.2 Feature, Event, and Process 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. 
6.2.12.3 Descriptor Phrases 
Internal corrosion of commercial spent nuclear fuel 
Internal corrosion of defense spent nuclear fuel 
Chemical effects from radiolysis in-package 
6.2.12.4 Screening Decision 
Excluded  Low Consequence 
6.2.12.5 Screening Argument 
The 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 failure in the repository. This FEP is related 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 
filled with an inert gas (Helium) prior to closure, displacing water and oxygen (necessary 
components for corrosion) from inside the package (CRWMS M&O 2000 [DIRS 123881], 
p. II-30). The inert gas environment within the package will result in a negligible amount of 
corrosion degradation prior to the breach of the waste packages. 
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 source of this residual moisture is primarily from waterlogged failed fuel 

rods. Analyses presented in the above reference indicate that the amount of moisture available to 
cause 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 2003 [DIRS 163855]) which will corrode in preference to the Alloy 22 
(UNS N06022) waste package outer barrier and the 316 stainless steel inner vessel (e.g., ASM 
International [DIRS 103753], p. 557). Thus, no significant corrosion damage to the waste 
package outer barrier and 316 stainless steel 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 interim storage. 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 interim storage. Damaged DSNF 
will be placed in high integrity 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 316 stainless steel 
inner vessel. This FEP is excluded based on low consequence to radiological exposures to the 
RMEI and radionuclide releases to the accessible environment. 
6.2.12.6 Total System Performance Assessment Disposition 
N/A 
6.2.12.7 Supporting Analysis Reports 
N/A 
6.2.13 Mechanical Impact on Waste Package 
6.2.13.1 Feature, Event, and Process Number 
FEP 2.1.03.07.0A 
6.2.13.2 Feature, Event, and Process Description 
Mechanical impact (dynamic loading) on the waste package is 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. Seismic impacts are addressed in a separate 
FEP. 
6.2.13.3 Descriptor Phrases 
Rockfall (waste package failure) 
Rockfall (waste package damage) 
Internal gas pressure (waste package damage) 
Volume increase of corrosion products (waste package damage) 
Rockfall (drip shield contacts waste package) 
6.2.13.4 Screening Decision 
Excluded  Low Consequence (Internal gas pressure and swelling of corrosion products) 
Excluded  Low Probability (Rockfall) 
6.2.13.5 Screening Argument 
The conclusions reached in this Section are equally applicable to CSNF and co-disposed 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 the ASME code 
requirements for the allowable tensile strength of the waste package material (CRWMS 
M&O 2000 [DIRS 144128], Section 6.2.2.8). 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 2003 [DIRS 161235], Section 6.4.2), indicate 
that, even under very conservative assumptions, the growth of the corrosion product (Cr2O3) 
oxide layer is not thick enough to produce enough pressure to cause mechanical damage to the 
Type 316 stainless steel inner vessel or the Alloy 22 (UNS N06022) outer barrier. Therefore, 
waste package damage from swelling corrosion products is excluded based on low consequence 
to radiological exposures to the RMEI and radionuclide releases to the accessible environment. 
Rockfall: Mechanical damage of the waste package by rockfall is discussed in greater detail 
under FEP 2.1.07.01.0A, Rockfall. The Emplacement Drift System design criteria require that 
the drip shield protect the waste package from rockfalls during postclosure (BSC 2002 
[DIRS 159292], Section 2.1.3.2). Because the drip shield provides adequate protection to the 

waste packages from rockfall (refer to 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 
the Engineered Barrier System Features, Events, and Processes report (BSC 2004 
[DIRS 167253]). 
6.2.13.6 Total System Performance Assessment Disposition 
N/A 
6.2.13.7 Supporting Analysis Reports 
General Corrosion and Localized Corrosion of Waste Package Outer Barrier (BSC 2003 
[DIRS 161235]) 
6.2.14 Mechanical Impact on Drip Shield 
6.2.14.1 Feature, Event, and Process Number 
FEP 2.1.03.07.0B 
6.2.14.2 Feature, Event, and Process Description 
Mechanical impact (dynamic loading) on the drip shield is caused by forces such as rockfall, and 
possible waste package or drip shield movement. Seismic induced impacts are addressed in a 
separate FEP. 
6.2.14.3 Descriptor Phrases 
Rockfall (drip shield separation) 
Rockfall (drip shield damage) 
Rockfall (drip shield contacts waste package) 
6.2.14.4 Screening Decision 
Excluded  Low Consequence 
6.2.14.5 Screening Argument 
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 requires that 

the drip shield protect the waste package from rockfalls during postclosure (BSC 2002 
[DIRS 159292], Section 2.1.3.2). According to the Drip Shield Structural Response to Rock Fall 
calculation (BSC 2003 [DIRS 162598], Section 6), LS-DYNA analysis shows that the deflection 
of the drip shield due to a 11.5 MT 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 MT rockfall event is 254 mm (BSC 2003 
[DIRS 168489]). The minimum gap between the drip shield and waste package outer barrier was 
calculated to be about 367 mm based on the following equation: 
2 
d 
- dist - h wp 
inv ds int = Gap (Eq. 7) 
where 
hint ds = interior height of drip shield (BSC 2003 [168489], Table 1) (2716. mm) 
distinv = distance from top of invert to centerline of waste package for 5 DHLW (BSC 2004 
[DIRS 167040]) (1286.1 mm) 
dwp = diameter of waste package (5 DHLW/DOE SNF  Short) (BSC 2004 
[DIRS 167207], Table 1) (2126 mm) 
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 the Engineered 
Barrier System Features, Events, and Processes report (BSC 2004 [DIRS 167253]). 
6.2.14.6 Total System Performance Assessment Disposition 
N/A 
6.2.14.7 Supporting Analysis Reports 
N/A 
6.2.15 Early Failure of Waste Packages 
6.2.15.1 Feature, Event, and Process Number 
FEP 2.1.03.08.0A 
6.2.15.2 Feature, Event, and Process Description 
Waste packages may fail prematurely because of manufacturing defects, improper sealing, or 
other factors related to quality control during manufacture and emplacement. 

6.2.15.3 Descriptor Phrases 
Waste package closure welds (stress corrosion cracking) 
Waste package fabrication welds (stress corrosion cracking) 
Waste package emplacement error 
6.2.15.4 Screening Decision 
Included 
6.2.15.5 Screening Argument 
N/A 
6.2.15.6 Total System Performance Assessment Disposition 
The Analysis of Mechanisms for Early Waste Package/Drip Shield Failure (BSC 2003 
[DIRS 164475]) evaluates several mechanisms for early failure of the waste package. Of these 
mechanisms, weld flaws, improper heat treatment, improper laser peening, and improper 
handling of waste packages were determined to be necessary for inclusion in TSPA models 
(BSC 2003 [DIRS 164475], Section 8). 
As discussed in FEP2.1.03.02.0A, Stress Corrosion Cracking of Waste Packages, manufacturing 
defects (weld flaws) on waste packages act as sites for initiation of SCC (BSC 2003 
[DIRS 161234]; BSC 2003 [DIRS 161317]). Manufacturing defects are included in TSPA 
analysis through the SCC analysis of waste packages. 
Early failure (due to improper heat treatment, improper laser peening, and improper handling of 
waste packages) is included in the waste package performance analysis. Improper heat treatment 
results primarily from improper stress relief annealing and the consequence of improper heat 
treatment is assumed to be immediate failure upon initiation of degradation processes (BSC 2003 
[DIRS 164475], Section 6.4.3). 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 (BSC 2003 [DIRS 164475], Section 6.4.4). Improper handling of the waste 
packages may lead to gouges in the waste package outer surface and provide sites for stress 
corrosion cracks (BSC 2003 [DIRS 164475], Section 6.4.6). Early failure (due to improper heat 
treatment, improper laser peening, and improper handling of waste packages) is included in 
TSPA analysis. 
The number of early failed waste packages per realization is given by a Poisson distribution with 
an uncertain intensity. The Poisson intensity is sampled from a log normal distribution with a 
median of 7.2x10-6 and an error factor of 15 (BSC 2003 [DIRS 164475], Section 7, Table 20). 
Since an improperly heat treated waste package might be susceptible to aging and phase stability, 
it is not possible to identify a single and specific mechanism of degradation. For these reasons, 
the following recommendations are made in the Analysis of Mechanisms for Early Waste 

Package/Drip Shield Failure (BSC 2003 [DIRS 164475], Section 6.4.8) for evaluating waste 
package early failure 
 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. 
 The entire waste package surface area should be considered affected by waste package 
early failure. 
This early failure treatment applies to both CSNF and Co-disposed 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 the Engineered Barrier System Features, Events, and Processes report (BSC 2004 
[DIRS 167253]). 
6.2.15.7 Supporting Analysis Reports 
Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the 
Stainless Steel Structural Material (BSC 2003 [DIRS 161234]) (This AR did not explicitly list 
this FEP as an Included FEP (BSC 2003 [DIRS 161234], Section 1.1). However, this Analysis 
Report does provide relevant supporting information describing the implementation of this FEP 
in TSPA-LA). 
WAPDEG Analysis of Waste Package and Drip Shield Degradation (BSC 2003 [DIRS 161317]) 
6.2.16 Early Failure of Drip Shields 
6.2.16.1 Feature, Event, and Process Number 
FEP 2.1.03.08.0B 
6.2.16.2 Feature, Event, and Process Description: 
Drip shields may fail prematurely because of manufacturing defects, improper sealing, or other 
factors related to quality control during manufacture and emplacement. 
6.2.16.3 Descriptor Phrases 
Drip shield fabrication welds (stress corrosion cracking) 
Drip shield emplacement error 

6.2.16.4 Screening Decision 
Excluded  Low Consequence 
6.2.16.5 Screening Argument 
The Analysis of Mechanisms for Early Waste Package/Drip Shield Failure (BSC 2003 
[DIRS 164475]) evaluates 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. All of these mechanisms 
have a low consequence to radiological exposures to the RMEI and radionuclide releases to the 
accessible environment (BSC 2003 [DIRS 164475], Section 6.3). 
The consequence of weld flaws, base metal flaws, improper heat treatment, and damage by 
mishandling is SCC (BSC 2003 [DIRS 164475], Section 6.4). Among other exposure condition 
parameters, the surface stress at the tip of a SCC crack must exceed the critical threshold stress in 
order to initiate SCC (BSC 2003 [DIRS 161234], Section 6.2.1). 
For the drip shields, all the 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 2003 [DIRS 161234], Section 6.3.7). However, the 
drip shields are subject to SCC under the action 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 167253]). In the nominal case (in the absence of seismic-induced loading and 
rockfalls) even if SCC of the drip shield were to occur, cracks in passive alloys, such as Titanium 
Grade 7, tend to be tight (i.e., small crack opening displacement) (BSC 2003 [DIRS 161234], 
Section 6.3.7). As the crack grows through-wall, the tensile stresses normal to the crack walls 
are relieved, and the resulting crack faces continue to corrode at very low passive corrosion rates 
until the gap region of the tight crack opening is plugged by corrosion products and 
precipitates such as carbonate minerals. As discussed in Stress Corrosion Cracking of the Drip 
Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material model 
report (BSC 2003, [DIRS 161234], Section 6.3.7), SCC cracks are sealed in a few hundred years 
at most when water is allowed to flow through the cracks at the expected low flow rate. When 
the cracks are bridged by water, the sealing process may take thousands of years, but no flow 
occurs since the water is held by capillary forces. Following plugging of the crack, any solution 
flow through the crack would be dominated by an efficiency factor determined by the ratio of 
solution run-off on the drip shield surface compared to through crack flow which in turn is 
determined by scale porosity/permeability. Because of the expected high density of the calcite 
deposits and lack of pressure gradient to drive water through the crack, the probability of 
solution flow through the crack would approach zero (BSC 2003 [DIRS 161234], Section 6.3.7). 
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 its intended design purpose. 

The use of improper weld or base metal material is possible in the repository (BSC 2003 
[DIRS 164475], Section 6.3.3), however, due to the strict controls that will govern the 
fabrication of the DS, it is expected that the material composition of the improper weld or base 
metal material will differ only slightly from the intended composition (BSC 2003 
[DIRS 164475], 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 2003 
[DIRS 164475], Section 6.4.2). 
The probability of drip shield surface contamination is also evaluated in the Analysis of 
Mechanisms for Early Waste Package/Drip Shield Failure (BSC 2003 [DIRS 164475], 
Section 6.3.5). It is found that the consequence of drip shield surface contamination is not 
significant from a corrosion standpoint (BSC 2003 [DIRS 164475], Section 6.4.5). On this basis, 
drip shield surface contamination is not significant. 
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. 
Thus, manufacturing defects in the drip shield and early failure mechanisms (not including drip 
shield emplacement error, see 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. 
6.2.16.6 Total System Performance Assessment Disposition 
N/A 
6.2.15.7 Supporting Analysis Reports 
Analysis of Mechanisms for Early Waste Package/Drip Shield Failure (BSC 2003 
[DIRS 164475]) 
Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the 
Stainless Steel Structural Material (BSC 2003 [DIRS 161234]) 
6.2.17 Copper Corrosion in Engineered Barrier System 
6.2.17.1 Feature, Event, and Process Number 
FEP 2.1.03.09.0A 
6.2.17.2 Feature, Event, and Process Description 
Chemical reactions involving copper corrosion have been identified as being of potential interest 
for repository programs considering the use of copper containers. 

6.2.17.3 Descriptor Phrases 
Corrosion of gantry rail system 
6.2.17.4 Screening Decision 
Excluded  Low Consequence 
6.2.17.5 Screening Argument 
The repository does not use copper containers. However, a small amount of copper may be 
present as part of the gantry rail system (BSC 2003 [DIRS 164101], Table 3). This will 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 2003 [DIRS 168489], Table 6) and is designed to keep the waste 
package from contacting other dissimilar metals. 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 2003 [DIRS 168489], Table 5). 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 with titanium and hydrogen absorption. The 
potential for hydrogen absorption, or hydrogen induced cracking, is considered low since 
Titanium Grade 7 contains Pd and since a protective oxide film forms on the surface of the drip 
shield, which hinders hydrogen absorption (BSC 2003 [DIRS 161759], Section 6.1.3). When 
Titanium Grade 7 is cathodically polarized at -1.2 V (vs. 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 2003 
[DIRS 161759], 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. 
6.2.17.6 Total System Performance Assessment Disposition 
N/A 
6.2.17.7 Supporting Analysis Reports 
Hydrogen Induced Cracking of Drip Shield (BSC 2003 [DIRS 161759]) 
6.2.18 Healing of Waste Packages 
6.2.18.1 Feature, Event, and Process Number 
FEP 2.1.03.10.0A 

6.2.18.2 Feature, Event, and Process Description 
Pits, cracks, and holes in waste packages could be partially or fully plugged by chemical or 
physical reactions during or after their formation, affecting corrosion processes and water flow 
and radionuclide transport through the breached waste package. Passivation by corrosion 
products is a potential mechanism for waste package healing. 
6.2.18.3 Descriptor Phrases 
Waste package crack plugging (precipitates, corrosion products) 
Passivation by corrosion products 
6.2.18.4 Screening Decision 
Excluded  Low Consequence 
6.2.18.5 Screening Argument 
Plugging (or healing) of cracks and corrosion holes or pits in waste package by corrosion 
products and mineral precipitates is a possible process in the repository. However, the effect of 
plugging could only be to retard the release rate of radionuclides from breached waste packages 
through retardation of water flow through openings. On this basis, it is conservative to assume 
that cracks and corrosion holes or pits in the waste package do not heal. Because of this, 
potential performance credit from the plugging (or healing) of the corrosion penetration openings 
is not taken into account in TSPA analysis. This treatment of healing of waste packages applies 
to both CSNF and Co-disposed waste packages. This FEP is excluded on the basis of low 
consequence because it has no adverse effects on performance. 
6.2.18.6 Total System Performance Assessment Disposition 
N/A 
6.2.18.7 Supporting Analysis Reports 
N/A 
6.2.19 Healing of Drip Shields 
6.2.19.1 Feature, Event, and Process Number 
FEP 2.1.03.10.0B 
6.2.19.2 Feature, Event, and Process Description 
Pits, cracks, and holes in drip shields could be partially or fully plugged by chemical or physical 
reactions during or after their formation, affecting corrosion processes and water flow through 
the drip shield. 

6.2.19.3 Descriptor Phrases 
Drip shield crack plugging (precipitates, corrosion products) 
6.2.19.4 Screening Decision 
Included 
6.2.19.5 Screening Argument 
N/A 
6.2.19.6 Total System Performance Assessment Disposition 
Plugging (or healing) of cracks, holes or pits in drip shields by corrosion products and mineral 
precipitates could occur in the repository. 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, pits (assumed to be equivalent to holes) will not form on 
the drip shield material. As discussed in FEP 2.1.03.02.0A, Stress Corrosion Cracking (SCC) of 
Drip Shields, plugging of any cracks developed by SCC processes is expected. Drip shields are 
subject to SCC under the action of seismic-induced loading and rockfalls. In the nominal case 
(in the absence of seismic-induced loading and rockfalls) even if SCC of the drip shield were to 
occur, cracks in passive alloys, such as Titanium Grade 7, tend to be tight (i.e., small crack 
opening displacement) (BSC 2003 [DIRS 161234], Section 6.3.7; BSC 2003 [DIRS 161317], 
Section 6.3.7). As the crack grows through-wall, the tensile stresses normal to the crack walls 
are relieved, and the resulting crack faces continue to corrode at very low passive corrosion rates 
until the gap region of the tight crack opening is plugged by corrosion products and 
precipitates such as carbonate minerals. As discussed in Stress Corrosion Cracking of the Drip 
Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material model 
report (BSC 2003 [DIRS 161234], Section 6.3.7), SCC cracks are sealed in a few hundred years 
at most when water is allowed to flow through the cracks at the expected low flow rate. When 
the cracks are bridged by water, the sealing process may take thousands of years, but no flow 
occurs since the water is held by capillary forces. Following plugging of the crack, any solution 
flow through the crack would be dominated by an efficiency factor determined by the ratio of 
solution run-off on the drip shield surface compared to through crack flow which in turn is 
determined by scale porosity/permeability. Because of the expected high density of the calcite 
deposits and lack of pressure gradient to drive water through the crack, the probability of 
solution flow through the crack would approach zero (BSC 2003 [DIRS 161234], Section 6.3.7). 
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. The 
development of pits or holes in the drip shield from localized corrosion is not expected. 
Since the formation of corrosion products and precipitates precludes water flow through the drip 
shield, performance credit is taken for the ability of the drip shield to prevent water flow and 
protect the waste package. Healing of drip shields is included in TSPA as part of waste package 
degradation analyses in that stress corrosion crack openings are not considered to compromise 
the water diversion design function of the drip shield. 

6.2.19.7 Supporting Analysis Reports 
Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the 
Stainless Steel Structural Material (BSC 2003 [DIRS 161234]) (This AR did not explicitly list 
this FEP as an Included FEP (BSC 2003 [DIRS 161234], Section 1.1). However, this analysis 
report does provide relevant supporting information describing the implementation of this FEP in 
TSPA LA). 
WAPDEG Analysis of Waste Package and Drip Shield Degradation (BSC 2003 [DIRS 161317]) 
6.2.20 Physical Form of Waste Package and Drip Shield 
6.2.20.1 Feature, Event, and Process Number 
FEP 2.1.03.11.0A 
6.2.20.2 Feature, Event, and Process 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 container 
and affect heat dissipation through container 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. 
6.2.20.3 Descriptor Phrases 
Effects of commercial spent nuclear fuel waste package form on heat distribution 
Effects of co-disposed waste package form on heat distribution 
Effects of degraded waste package on flow 
Effects of degraded waste package on transport 
Effects of degraded drip shield on flow 
6.2.20.4 Screening Decision 
Included 
6.2.20.5 Screening Argument 
N/A 

6.2.20.6 Total System Performance Analysis Disposition 
The waste package, drip shield, and repository design are standardized for the YMP (BSC 2004 
[DIRS 167040]). While there is more than one waste package configuration expected to 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. Effects of different 
waste forms (CSNF, DSNF, and DHLW) on heat dissipation and physical and chemical 
conditions in the vicinity of the waste packages are indirectly included in the TSPA analysis 
through different thermal-hydrologic-geochemical responses and their impacts on corrosion 
processes. Waste package and drip shield degradation modes are modeled in the General 
Corrosion and Localized Corrosion of Waste Package Outer Barrier (BSC 2003 
[DIRS 161235]), the General Corrosion and Localized Corrosion of the Drip Shield (BSC 2003 
[DIRS 161236]), and the WAPDEG Analysis of Waste Package and Drip Shield Degradation 
(BSC 2003 [DIRS 161317]) model reports. 
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 
have an explicit effect because the flow pathways are modeled without regard to the detailed 
mechanisms of flow (refer to FEP 2.1.08.07.0A, Unsaturated Flow in the EBS) (BSC 2004 
[DIRS 167253]). 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 the Engineered Barrier System Features, Events and 
Processes report (BSC 2004 [DIRS 167253]). 
6.2.20.7 Supporting Analysis Reports 
General Corrosion and Localized Corrosion of Waste Package Outer Barrier (BSC 2003 
[DIRS 161235]) (This analysis report did not explicitly list this FEP as an Included FEP 
(BSC 2003 [DIRS 161235], Table 6-1). However, this AR does provide relevant supporting 
information describing the implementation of this FEP in TSPA-LA). 
General Corrosion and Localized Corrosion of the Drip Shield (BSC 2003 [DIRS 161236]) 
(This AR did not explicitly list this FEP as an Included FEP (BSC 2003 [DIRS 161236], 
Table I-1). However, this analysis report does provide relevant supporting information 
describing the implementation of this FEP in TSPA-LA). 
WAPDEG Analysis of Waste Package and Drip Shield Degradation (BSC 2003 [DIRS 161317]) 
6.2.21 Oxygen Embrittlement of Drip Shields 
6.2.21.1 Feature, Event, and Process Number 
FEP 2.1.06.06.0B 

6.2.21.2 Feature, Event, and Process Description 
A potential failure mechanism for drip shields is oxygen embrittlement, resulting from the 
diffusion of interstitial oxygen in the titanium at high temperatures. 
6.2.21.2 Descriptor Phrases 
Oxygen embrittlement (drip shield) 
6.2.21.3 Screening Decision 
Excluded  Low Probability 
6.2.21.4 Screening Argument 
Oxygen embrittlement of titanium results from diffusion of interstitial oxygen into the metal at 
higher temperatures (> 340C) (ASM International 1987 [DIRS 103753], p. 681). The time to 
failure depends on the alloy composition, material thickness, and stress state. For the thermal 
hydrologic time history files used in the TSPA analyses, the drip shield surface temperatures 
never exceed 157C , which is less than the threshold temperature for oxygen embrittlement of 
340C (CRWMS M&O 2001 [DIRS 154594], Section 6.3.5). Therefore, oxygen embrittlement 
of the titanium drip shields is excluded on the basis of low probability of occurrence under the 
exposure conditions in the repository. 
6.2.21.5 Total System Performance Assessment Disposition 
N/A 
6.2.21.6 Supporting Analysis Reports 
N/A 
6.2.22 Mechanical Effects at Engineered Barrier System Component Interfaces 
6.2.22.1 Feature, Event, and Process Number 
FEP 2.1.06.07.0B 
6.2.22.2 Feature, Event, and Process 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. 
6.2.22.3 Descriptor Phrases 
Rockfall - (drip shield contacts waste package) 
Drip shield  pallet contact (mechanical effects) 

Waste package  pallet contact (mechanical effects) 
6.2.21.4 Screening Decision 
Excluded - Low Consequence 
6.2.22.5 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 167040]). The pallet is also designed to 
keep the waste package supported in a horizontal position, and away from the invert and ground 
support under non-seismic scenarios. 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. 
There is some potential for the drip shield to contact the waste package due to mechanical 
damage caused by rockfall. This is, however, excluded as discussed in FEP 2.1.03.07.0B, 
Mechanical Impact on Drip Shield. 
Mechanical loading at the waste package/(degraded) pallet interfaces has been analyzed 
(BSC 2003 [DIRS 168489]). The contact stresses are shown to be much less (maximum stress 
intensity ~150 MPa) than the stress threshold for initiation of stress corrosion cracking 
(~286 MPa; (BSC 2003 [DIRS 161234], Section 8.3). On this basis, no enhanced degradation 
due to mechanical loading at the waste package/pallet interfaces is expected. One should note 
that waste package and drip shield corrosion degradation analyses include the effects of material 
interfaces in the repository on thermal-hydrologic-geochemical 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 waste package, drip shield and backfill (if used), which are 
described in the General Corrosion and Localized Corrosion of Waste Package Outer Barrier 
(BSC 2003 [DIRS 161235]) and General Corrosion and Localized Corrosion of the Drip Shield 
(BSC 2003 [DIRS 161236]) model reports. 
This treatment of mechanical effects at EBS component interfaces applies to both CSNF and 
Co-disposed waste packages. 
This FEP is excluded based on low consequence to radiological exposures to the RMEI and 
radionuclide releases to the accessible environment. 
6.2.22.6 Total System Performance Assessment Disposition 
N/A 
6.2.22.7 Supporting Analysis Reports 
General Corrosion and Localized Corrosion of Waste Package Outer Barrier (BSC 2003 
[DIRS 161235]) 

General Corrosion and Localized Corrosion of the Drip Shield (BSC 2003 [DIRS 161236]) 
6.2.23 Rockfall 
6.2.23.1 Feature, Event, and Process Number 
FEP 2.1.07.01.0A 
6.2.23.2 Feature, Event, and Process 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. 
6.2.23.3 Descriptor Phrases 
Rockfall (geometry, key blocks, time dependence) 
Rockfall (drip shield damage) 
Rockfall (drip shield stress corrosion cracking) 
Rockfall (waste package damage) 
Rockfall (drip shield contacts waste package) 
Drip shield crack plugging 
6.2.23.4 Screening Decision 
Excluded  Low Consequence (drip shield) 
Excluded  Low Probability (waste package) 
6.2.23.5 Screening Argument 
It should be noted that seismic induced rockfalls/drift degradation are not treated within this 
FEP. A full discussion of seismic effects is contained in FEPs 1.2.03.02.0A, Seismic Ground 
Motion Damages EBS Components; 1.2.03.02.0B, Seismic Induced Rockfall Damages EBS 
Components; and 1.2.03.02.0C, Seismic Induced Drift Collapse Damages EBS Components, 
treated in the Engineered Barrier System Features, Events, and Processes report (BSC 2004 
[DIRS 167253]). 
According to the Drip Shield Structural Response to Rock Fall calculation (BSC 2003 
[DIRS 162598], Section 6), LS-DYNA analysis 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 a 
11.5 MT rockfall (see Section 6.2.14) without contacting the waste package. The maximum 
displacement from the 11.5 MT rockfall event is 254 mm (BSC 2003 [DIRS 168489]) and the 
minimum gap between the drip shield and waste package outer barrier is 367 mm (refer to 

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 the Stress Corrosion 
Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural 
Material report (BSC 2003 [DIRS 161234], Section 6.3.7). It is concluded that a crack will take 
a minimum of 40 years to grow through the 15-mm drip shield wall (BSC 2003 [DIRS 161234], 
Section 6.3.7). These cracks are extremely tight and with time, become plugged with corrosion 
products and other mineral precipitates (BSC 2003 [DIRS 161234], 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. 
Analyses related to multiple rockfalls are beyond the scope of this document. Bounding effects 
of multiple rockfalls and drift degradation are addressed as part of the seismic consequences in 
the Engineered Barrier System Features, Events, and Processes report (BSC 2004 
[DIRS 167253]). 
It is feasible that 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, it 
is reasonable to predict that the drip shield continues to divert water from falling on the waste 
package until the drip shield fails after 10,000 years. Therefore, increased inflow of water 
related to rockfall is excluded from the TSPA analysis based on low consequence to radiological 
exposures to the RMEI and radionuclide releases to the accessible environment. 
The Emplacement Drift System design criteria requires that the drip shield protect the waste 
package from rockfalls during postclosure (BSC 2002 [DIRS 159292], Section 2.1.3.2). 
Furthermore, as discussed in FEP 2.1.03.07.0B, Mechanical Impact on Drip Shield, the 
maximum displacement from a 11.5 metric ton rockfall (which produces the maximum vertical 
displacement in the drip shield components) is 254 mm. This is less than the minimum gap 
between the drip shield and waste package outer barrier, which was calculated to be about 
367 mm. Therefore, 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 of emplacement). 
6.2.23.6 Total System Performance Assessment Disposition 
N/A 
6.2.23.7 Supporting Analysis Reports 
Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the 
Stainless Steel Structural Material (BSC 2003 [DIRS 161234]) 

6.2.24 Creep of Metallic Materials in the Waste Package 
6.2.24.1 Feature, Event, and Process Number 
FEP 2.1.07.05.0A 
6.2.24.2 Feature, Event, and Process Description: 
Metals used in the waste package may deform by creep processes in response to deviatoric stress 
or internal void space. 
6.2.24.3 Descriptor Phrases 
Thermally-induced waste package creep 
Waste package creep from deviatoric stress (rockfall, internal gas pressure) 
Waste package creep from internal void space 
6.2.24.4 Screening Decision 
Excluded  Low Probability 
6.2.24.5 Screening Argument 
The maximum temperature at the waste package surface will be about 190oC (CRWMS M&O 
2001 [DIRS 154594], Section 6.3.1). Elevated-temperature behavior (i.e., creep deformation or 
creep-fracture) of nickel-based alloys is not expected at temperatures under 650C (Boyer and 
Gall [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 1370oC (1643 K) (Haynes International 1988 [DIRS 101995]) and the maximum 
surface temperature (about 190C or 463 K) is only about 28% of the melting temperature. This 
treatment of creep of metallic materials in the waste package applies to both CSNF and 
Co-disposed waste packages. Creep of Alloy 22 at such low temperatures is not expected. 
Therefore, high temperature creep has a low probability of occurrence. 
External stress, by rock displacements or ground motion for example, may lead to plastic 
deformations and mechanical damage of the container and subsequent leakage of radionuclides. 
The drip shield is designed to be protective of the waste package during rockfall and ground 
motion events (refer to 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 above 650oC (Boyer and Gall 
[DIRS 155318], Section 32). In view of the above rationale, this FEP is excluded based on low 
probability of occurrence. 
6.2.24.5 Total System Performance Assessment Disposition 
N/A 

6.2.24.6 Supporting Analysis Reports 
N/A 
6.2.25 Creep of Metallic Materials in the Drip Shield 
6.2.25.1 Feature, Event, and Process Number 
FEP 2.1.07.05.0B 
6.2.25.2 Feature, Event, and Process Description 
Metals used in the drip shield may deform by creep processes in response to deviatoric stress. 
6.2.25.3 Descriptor Phrases 
Thermally-induced drip shield creep 
Drip shield creep from deviatoric stress (rockfall) 
6.2.25.4 Screening Decision 
Excluded  Low Probability 
6.2.25.5 Screening Argument 
Based on the current analyses, the maximum surface temperatures at the drip shield will be about 
157oC (CRWMS M&O 2001 [DIRS 154594], Section 6.3.5). Literature indicates that between 
200 and 315oC (400 and 600oF), 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 postclosure (BSC 2002 [DIRS 159292], Section 2.1.3.2). 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 the Engineered Barrier System Features, Events and Processes report (BSC 2004 
[DIRS 167253]). 
In view of the above rationale, this FEP is excluded based on low probability of occurrence. 

6.2.25.6 Total System Performance Assessment Disposition 
N/A 
6.2.25.7 Supporting Analysis Reports 
N/A 
6.2.26 Volume Increase of Corrosion Products Impacts Waste Package 
6.2.26.1 Feature, Event, and Process Number 
FEP 2.1.09.03.0B 
6.2.26.2 Feature, Event, and Process 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, leading to waste package damage. 
6.2.26.2 Descriptor Phrases 
Volume increase of corrosion products (waste package damage) 
6.2.26.3 Screening Decision 
Excluded  Low Consequence 
6.2.26.4 Screening Argument 
In general, corrosion products have greater volume than the metal. When the corrosion products 
form in a tightly confined space, the volume increase by the corrosion products 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 due to the pressure exerted by 
the corrosion product (Cr2O3) of the inner vessel (Type 316 stainless steel) will not occur. 
Analyses cited in General Corrosion and Localized Corrosion of Waste Package Outer Barrier 
(BSC 2003 [DIRS 161235], Section 6.4.2), indicate that for chromia (Cr2O3) scale-forming 
alloys (e.g., Alloy 22 and 316 stainless steel), 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 Alloy 22 waste 
package. In the current design of waste package and engineered barrier system in the 
emplacement drift (BSC 2004 [DIRS 167040]), 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 consequence to radiological exposures to the RMEI 
and radionuclide releases to the accessible environment. 

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 166926]). 
6.2.26.5 Total System Performance Assessment Disposition 
N/A 
6.2.26.6 Supporting Analysis Reports 
General Corrosion and Localized Corrosion of Waste Package Outer Barrier (BSC 2003 
[DIRS 161235]) 
6.2.27 Electrochemical Effects in Engineered Barrier System 
6.2.27.1 Feature, Event, and Process Number 
FEP 2.1.09.09.0A 
6.2.27.2 Feature, Event, and Process Description 
Electrochemical effects may establish an electric potential within the drift or between materials 
in the drift and more distant metallic materials in the natural system. 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. 
6.2.27.3 Descriptor Phrases 
Electrophoresis/electro-osmosis 
Galvanic coupling (waste package) 
Galvanic coupling (drip shield) 
6.2.27.4 Screening Decision 
Excluded  Low Consequence 
6.2.27.5 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 electrochemical effects 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, which is 
constructed of Alloy 22 (UNS N06022) and is designed to keep the waste package from 

contacting other dissimilar metals (BSC 2004 [DIRS 167040]). The pallet is also designed to 
keep the waste package supported in a horizontal position, and away from the invert and ground 
support under non-seismic scenarios. 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 current waste package design (BSC 2004 [DIRS 167040]) includes an outer barrier of 
Alloy 22 (UNS N06022) over a 316 stainless steel inner vessel. In addition, a titanium drip 
shield is added to this design to provide defense in depth. Although the stainless steel 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 316 stainless steel are very close to each other under typical 
exposure conditions (e.g., ASM International [DIRS 103753], p. 557) with Alloy 22 
(UNS N06022) slightly more noble than 316 stainless steel. After failure of the Alloy 22 
(UNS N06022) waste package outer barrier, electrochemical coupling of the Alloy 22 
(UNS N06022) waste package outer barrier with the 316 stainless steel waste package inner 
vessel could occur. Due to the similarity in corrosion potential of Alloy 22 (UNS N06022) and 
316 stainless steel, any enhanced degradation of either material would be negligible. Therefore, 
electrochemical coupling of the Alloy 22 (UNS N06022) waste package outer barrier and the 316 
stainless steel waste package inner vessel is of low consequence to radiological exposures to the 
RMEI and radionuclide releases to the accessible environment. 
There is some potential for the drip shield to contact the waste package due to mechanical 
damage caused by rockfall. This is, however, unlikely 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 (e.g., ASM International [DIRS 103753], p. 557), which indicates that there is very low 
probability for galvanic interaction between the two materials. 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. 
6.2.27.6 Total System Performance Assessment Disposition 
N/A 
6.2.27.7 Supporting Analysis Reports 
N/A 
6.2.28 Thermal Sensitization of Waste Packages 
6.2.28.1 Feature, Event, and Process Number 
FEP 2.1.11.06.0A 

6.2.28.2 Feature, Event, and Process Description: 
Phase changes in waste package materials can result from long-term storage at moderately hot 
temperatures in the repository. SCC, intergranular corrosion, or mechanical degradation may 
ensue. 
6.2.28.3 Descriptor Phrases 
Aging-phase instability (waste package) 
Decreased resistance to waste package corrosion 
6.2.28.4 Screening Decision 
Excluded  Low Probability 
6.2.28.5 Screening Argument 
Alloy 22 (UNS N06022) could to be subject to aging and phase instability when exposed to 
elevated temperatures. 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 2003 
[DIRS 162199]). 
Before waste loading, the waste containers (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 container (Plinski 2001 [DIRS 156800], Section 8.1.8). Analyses 
presented in the General Corrosion and Localized Corrosion of Waste Package Outer Barrier 
report (BSC 2003 [DIRS 161235], Section 6.4.6) show that the corrosion performance of 
Alloy 22 (UNS N06022) base metal is not affected by the aging and phase instability as long as 
the waste package surface temperature is kept below 200C under the exposure conditions 
expected in the repository. The maximum temperature at the waste package surface will never 
exceed 190oC (CRWMS M&O 2001 [DIRS 154594], Section 6.3.1). 
The closure lid welds cannot be solution annealed without risking damage to the waste form. 
Therefore, the closure welds of the waste package out barrier could be more prone to thermal 
aging and phase instability than the base metal under long-term thermal exposure in the 
repository (BSC 2003 [DIRS 161235], Section 6.4.6). Analyses conducted in the report entitled 
General Corrosion and Localized Corrosion of Waste Package Outer Barrier (BSC 2003 
[DIRS 161235], 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 the test 
temperatures varying from 45 to 120C. Comparison of the calculated 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 2003 [DIRS 161235], 
Section 6.4.6) and the corrosion performance of the waste package outer barrier is not expected 

to be affected by the aging and phase stability in the repository. This treatment of thermal 
sensitization of waste packages applies to both CSNF and co-disposed waste packages. Thermal 
sensitization of waste packages is excluded on the basis of low probability of occurrence. 
6.2.28.5 Total System Performance Assessment Disposition 
N/A 
6.2.28.6 Supporting Analysis Reports 
Aging and Phase Stability of Waste Package Outer Barrier (BSC 2003 [DIRS 162199]) 
General Corrosion and Localized Corrosion of Waste Package Outer Barrier (BSC 2003 
[DIRS 161235]) 
Note that the BSC 2003 [DIRS 161235] lists this FEP in a Table entitled List of Included FEPs 
Relevant to This Model Report (BSC 2003 [DIRS 161235], Table 1.1) and also in a Table 
entitled Included FEPs for This Model Report and Their Disposition in TSPA-LA (BSC 2003 
[DIRS 161235], Table 6-1). Regardless of the titles of the Tables, the associated text in fact 
supports the exclusion of the FEP on the basis of low probability of occurrence (BSC 2003 
[DIRS 161235], Table 6-1). 
6.2.29 Thermal Sensitization of Drip Shields 
6.2.29.1 Feature, Event, and Process Number 
FEP 2.1.11.06.0B 
6.2.29.2 Feature, Event, and Process Description 
Phase changes in drip shield materials can result from long-term storage at moderately hot 
temperatures in the repository. Stress-corrosion cracking, intergranular corrosion, or mechanical 
degradation may ensue. 
6.2.29.3 Descriptor Phrases 
Aging-phase instability (drip shield) 
Decreased resistance to drip shield corrosion 
6.2.29.4 Screening Decision 
Excluded  Low Probability 
6.2.29.5 Screening Argument 
Aging and Phase stability of the drip shield is considered in Section 6.5.3 of the report entitled 
General Corrosion and Localized Corrosion of the Drip Shield (BSC 2003 [DIRS 161236]). In 
the report, it is observed that Titanium Grade 7 is a stabilized alpha (a) phase alloy and 

possesses outstanding phase stability. While Titanium Grade 7 does contain small amounts of 
alloying elements (DTN: MO0003RIB00073.000 [DIRS 152926]), most notably Pd, it is 
essentially a pure titanium alloy which has no capability of forming intermetallic compounds 
under the thermal exposure conditions in the repository. 
The solubility of Pd in Titanium Grade 7 is about 1 weight percent at 400C. The nominal 
concentration of Pd in Titanium Grade 7 is well below the solubility limit at this temperature 
(Gdowski 1997 [DIRS 102789], pp. 1-8). Titanium-palladium intermetallic compounds capable 
of being formed in this 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 60, 70, 80, 90, 100 and 105oC 
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 (HAZ) and between 
the HAZ and base metal were not visibly attacked. Therefore, based on the above experimental 
evidence, thermal sensitization of the drip shield can be excluded based on low probability of 
occurrence. 
6.2.29.6 Total System Performance Assessment Disposition 
N/A 
6.2.29.6 Supporting Analysis Reports 
General Corrosion and Localized Corrosion of the Drip Shield (BSC 2003 [DIRS 161236]) 
6.2.30 Thermal Expansion/Stress of In-Drift Engineered Barrier System Components 
6.2.30.1 Feature, Event, and Process Description 
FEP 2.1.11.07.0A 
6.2.30.2 Feature, Event, and Process 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. 
6.2.30.3 Descriptor Phrases 
Thermal-mechanical effects on waste packages 
Thermal expansion of drip shield 

6.2.30.4 Screening Decision 
Excluded - Low Consequence 
6.2.30.5 Screening Argument 
The coefficient of thermal expansion for Type 316L stainless steel (an analogue for the 316 
stainless steel used for the waste package inner vessel) 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 the calculation entitled 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 barrier gap size to be at least 1 mm (BSC 2004 [DIRS 166694]). 
The 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 (316 stainless steel) to accommodate the differing thermal expansion coefficients. 
Typical waste package designs also require large longitudinal barrier gaps (~30 mm) (BSC 2001 
[DIRS 157816], page 3). Therefore, although thermal expansion of waste package components 
does occur, 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 current drip shield design (BSC 2003 [DIRS 161519]), the drip shield connectors are 
designed in such a way that allows 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 (367 mm) (refer to 
Equation 7 in Section 6.2.14.5). The space needed 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. 
6.2.30.6 Total System Performance Assessment Disposition 
N/A 
6.2.30.7 Supporting Analysis Reports 
N/A 

6.2.31 Gas Generation (H2) from Waste Package Corrosion 
6.2.31.1 Feature, Event, and Process Number 
FEP 2.1.12.03.0A 
6.2.31.2 Feature, Event, and Process 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 containers, 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. 
6.2.31.3 Descriptor Phrases 
Internal gas pressure from H2 (waste package damage) 
Hydride cracking (drip shield) 
Hydride cracking (waste package) 
Repository pressurization 
Chemical effects from H2 generation 
6.2.31.4 Screening Decision 
Excluded  Low Consequence 
6.2.31.5 Screening Argument 
The materials selected for waste package outer barrier and the drip shield are highly corrosion 
resistant materials. 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 Yucca Mountain 
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 (HIC). HIC results 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. Hydrogen generation and absorption 

in the drip shield due to corrosion and galvanic interaction with ground support materials has 
been addressed in FEP 2.1.03.04.0B, Hydride Cracking of Drip Shields. This treatment of gas 
generation from waste package corrosion applies to both CSNF and co-disposed 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. 
6.2.31.6 Total System Performance Assessment Disposition 
N/A 
6.2.31.7 Supporting Analysis Reports 
N/A 
6.2.32 Radiolysis 
6.2.32.1 Feature, Event, and Process Number 
FEP 2.1.13.01.0A 
6.2.32.2 Feature, Event, and Process 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). 
6.2.32.3 Descriptor Phrases 
Gas generation from radiolysis 
Chemical effects from radiolysis in-drift 
6.2.32.4 Screening Decision 
Excluded  Low Consequence 
6.2.32.5 Screening Argument 
This section addresses only the effects of radiolysis on the waste package outer barrier. 
Radiolysis effects on waste form are addressed in Miscellaneous Waste-Form FEPs (BSC 2004 
[DIRS 167252]). Chemical effects from radiolysis (cement degradation) are addressed in the 
Engineered Barrier System Features, Events, and Processes report (BSC 2004 [DIRS 167253]). 
Gamma radiation is the dominant contributor to dose rate at the waste package surface 
(BSC 2003 [DIRS 165269], Tables 59 and 60) because alpha and beta radiation will not 
penetrate the container and the neutron dose rate is relatively small. The effects of radiation on 
waste package materials corrosion differ depending on the amount of liquid present on their 
surfaces (i.e., humid air or aqueous conditions). Under humid air conditions, a thin film of liquid 
forms that may contain 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. These shifts in potential have been shown to be 
due to the formation of hydrogen peroxide (BSC 2003 [DIRS 161235], Section 6.5.1). 
Calculations of the expected radiation levels at the surface of the waste package have been 
performed (BSC 2003 [DIRS 165269], Table 60). For a bounding case waste package 
containing 21 PWR spent fuel assemblies (75 GWd/MTU burnup, and 5 year decay), the 
maximum surface radiation level was calculated to be about 1100 rad/hr at the outer surface of 
the waste package barrier and 1550 rad/hr at the bottom lid of the barrier (BSC 2003 
[DIRS 165269], Table 60). During the ventilation period of 50 years, no aqueous or humid air 
environment, and therefore no radiolysis, is expected. After 50 years, the maximum surface 
radiation level decreases to less than 100 rad/hr for the outer surface and 70 rad/hr for the bottom 
lid (BSC 2003 [DIRS 165269], Table 60). These values correspond to reduction factors of 0.09 
for the outer surface and 0.045 for the bottom lid (BSC 2003 [DIRS 165269], Table 60). One 
hundred years after emplacement, the maximum calculated levels reduce to about 30 rad/hr for 
the outer surface and 20 rad/hr for the bottom lid region (BSC 2003 [DIRS 165269], Table 60). 
These values correspond to reduction factors of 0.03 for the outer surface and 0.015 for the 
bottom lid (BSC 2003 [DIRS 165269], Table 60). It is to be noted that there are bounding 
radiation levels for the highest burnup spent fuel and these are well below the levels at which 
some 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 that 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 
[DIRS 112178], p. 30). Based on this data it is concluded that, even in aggressive MgCl2 brines, 
the 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, the effects of 
radiolysis are excluded based on low consequence to radiological exposures to the RMEI and 
radionuclide releases to the accessible environment. 
6.2.32.6 Total System Performance Assessment Disposition 
N/A 
6.2.32.7 Supporting Analysis Reports 
General Corrosion and Localized Corrosion of the Drip Shield (BSC 2003 [DIRS 161236]) 
6.2.33 Radiation Damage in Engineered Barrier System 
6.2.33.1 Feature, Event, and Process Number 
FEP 2.1.13.02.0A 

6.2.33.2 Feature, Event, and Process Description 
Radiolysis due to the alpha, beta, gamma-ray, and neutron irradiation of water could result in the 
enhancement for the movement of the radionuclides from the surface of a degraded waste form 
into groundwater flow. When radionuclides decay, the emitted high-energy particle could result 
in the production of radicals in the water or air surrounding the SNF. If these radicals migrate 
(diffuse) to the surface of the fuel, they may then enhance the degradation/corrosion rate of the 
fuel (UO2). This effect would increase the dissolution rate for radionuclides from the fuel 
material (fuel meat) into the groundwater flow. Strong radiation fields could lead to radiation 
damage to the waste forms and containers (CSNF, DSNF, DHLW), drip shield, seals and 
surrounding rock. 
6.2.33.3 Descriptor Phrases 
Radiation damage in-drift (waste package) 
Radiation damage in-drift (drip shield, pallet, invert, seals, EDZ) 
6.2.33.4 Screening Decision 
Excluded  Low Consequence 
6.2.33.5 Screening Argument 
The enhanced degradation/corrosion of the fuel is covered in the Miscellaneous Waste-Form 
FEPs (BSC 2004 [DIRS 167252]). The Waste Package Materials Performance Peer Review 
Panel addressed the possibility of radiation damage in the repository in their Final Report 
(Beavers, et al. [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. These fluxes 
could cause the following types of damage: 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. [DIRS 158781], Section 3.10). The peak neutron flux 
has been calculated to be about 5x104 
n/cm2/sec) in the repository environment (Beavers et al. 
[DIRS 158781], Section 3.10). The total neutron fluence, taking the most conservative estimate 
with no nuclear decay of the waste, will be about 1.5x1016 n/cm2 in 10,000 years in the 
repository environment (Beavers et al. [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. [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], p. 971-974), 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. 
6.2.33.6 Total System Performance Assessment Disposition 
N/A 
6.2.33.7 Supporting Analysis Reports 
N/A 

7. CONCLUSIONS 
The analyses documented in this scientific analysis are for the current LA design where a drip 
shield is placed over the waste package (BSC 2004 [DIRS 167040]). Repository designs that are 
lower in temperature than the current LA design will not affect the screening decision of any of 
the waste package FEPs. Most of the waste package degradation mechanisms have no 
temperature dependence and those that do, have a positive correlation with rising temperatures 
and increased degradation rates (see screening arguments and TSPA Disposition discussions 
listed in this document). This is true for designs that only affect the repository temperature and 
not any other aspects of the design, such as waste package spacing in the repository, drift 
orientation, etc. 
Thirty-three FEPs relevant to waste package and drip shield degradation processes were screened 
against criteria presented in Section 4.2 of this report. The results of this screening process are 
documented in the screening arguments and the TSPA disposition statements in Section 6.2 of 
this report. The screening basis for each FEP is summarized below in Table 7-1. This table 
shows the FEP number, FEP name, screening decision (included/excluded in TSPA) 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 
Screening Basis 
1.1.03.01.0A 
Error in Waste Emplacement 
Excluded 
Low consequence 
2.1.03.01.0A 
General Corrosion of WPs 
Included 
N/A 
2.1.03.01.0B 
General Corrosion of DSs 
Included 
N/A 
2.1.03.02.0A 
SCC of W Ps 
Included 
N/A 
2.1.03.02.0B 
SCC of DSs 
Excluded 
Low consequence 
2.1.03.03.0A 
Localized Corrosion of WPs 
Included 
N/A 
2.1.03.03.0B 
Localized Corrosion of DSs 
Excluded 
Low consequence 
2.1.03.04.0A 
Hydride Cracking of WPs 
Excluded 
Low consequence 
2.1.03.04.0B 
Hydride Cracking of Drip Shields 
Excluded 
Low consequence 
2.1.03.05.0A 
MIC of WPs 
Included 
N/A 
2.1.03.05.0B 
MIC of DSs 
Excluded 
Low consequence 
2.1.03.06.0A 
Internal Corrosion of WPs Prior To Breach 
Excluded 
Low consequence 
2.1.03.07.0A 
Mechanical Impact on WP 
Excluded 
Low consequence / 
Low probability 
2.1.03.07.0B 
Mechanical Impact on DS 
Excluded 
Low consequence 
2.1.03.08.0A 
Early Failure of WPs 
Included 
N/A 
2.1.03.08.0B 
Early Failure of DSs 
Excluded 
Low consequence 
2.1.03.09.0A 
Copper Corrosion in EBS 
Excluded 
Low consequence 
2.1.03.10.0A 
Healing of WPs 
Excluded 
Low consequence 
2.1.03.10.0B 
Healing of DSs 
Included 
N/A 
2.1.03.11.0A 
Physical Form of WP and DS 
Included 
N/A 
2.1.06.06.0B 
Oxygen Embrittlement of DSs 
Excluded 
Low probability 
2.1.06.07.0B 
Mechanical Effects at EBS Component Interfaces 
Excluded 
Low consequence 

Table 7-1. Summary of Waste Package FEPs (Continued) 
FEP Number 
FEP Name 
Screening 
Decision 
Screening Basis 
2.1.07.01.0A 
Rockfall 
Excluded 
Low consequence / 
Low probability 
2.1.07.05.0A 
Creep of Metallic Materials in the WP 
Excluded 
Low probability 
2.1.07.05.0B 
Creep of Metallic Materials in the DS 
Excluded 
Low probability 
2.1.09.03.0B 
Volume Increase of Corrosion Products Impacts 
WP 
Excluded 
Low consequence 
2.1.09.09.0A 
Electrochemical Effects in EBS 
Excluded 
Low consequence 
2.1.11.06.0A 
Thermal Sensitization of WPs 
Excluded 
Low probability 
2.1.11.06.0B 
Thermal Sensitization of DSs 
Excluded 
Low probability 
2.1.11.07.0A 
Thermal Expansion/Stress of In-Drift EBS 
Components 
Excluded 
Low consequence 
2.1.12.03.0A 
Gas Generation (H2) From WP Corrosion 
Excluded 
Low consequence 
2.1.13.01.0A 
Radiolysis 
Excluded 
Low consequence 
2.1.13.02.0A 
Radiation Damage in EBS 
Excluded 
Low consequence 
DS = drip shield, WP = waste package, SCC = stress corrosion cracking, MIC = microbially influences corrosion 
It is to be noted that the screening decisions documented herein do not consider the 
consequences of combinations of FEPs. This is beyond the scope of this document at this time. 
If a decision is made to address combined FEPs, they will be considered on a case-by-case basis 
and any impact on screening decisions will be appropriately documented. 

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8.2 CODES, STANDARDS, REGULATIONS, AND PROCEDURES 
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AP-3.15Q, Rev. 4, ICN 2. Managing Technical Product Inputs. Washington, 
D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste 
Management. ACC: DOC.20030627.0002. 
164786 
AP-2.22Q, Rev. 1, ICN 0. Classification Analyses and Maintenance of the QList. 
Washington, D.C.: U.S. Department of Energy, Office of Civilian 
Radioactive Waste Management. ACC: DOC.20030807.0002. 
165023 
AP-SI.1Q, Rev. 5, ICN 2. Software Management. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20030902.0003. 
165065 
AP-2.14Q, Rev. 3, ICN 0. Document Review. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20030827.0018. 
166252 
AP-SIII.9Q, Rev. 1, ICN 2. Scientific Analyses. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20031126.0001. 

8.3 SOURCE DATA, LISTED BY DATA TRACKING NUMBER 
152926 
MO0003RIB00073.000. Physical and Chemical Characteristics of TI Grades 7 
and 16. Submittal date: 03/13/2000. 
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. 
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. 
164527 
MO0307SEPFEPS4.000. LA FEP List. Submittal date: 07/31/2003.