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November 2004

1. PURPOSE 
1.1 PURPOSE OF THE MODEL REPORT 
The purpose of this model report is to abstract a model that is suitable for use in the TSPA-LA 
model to describe the degradation rate of spent nuclear fuel (SNF) in the possession of the U.S. 
Department of Energy (DOE). This nuclear fuel is referred to as DOE-owned spent nuclear fuel 
(DSNF). 
1.2 BACKGROUND 
Several hundred distinct types of DSNF may eventually be emplaced in the repository 
(DOE 2002 [DIRS 158405], Section 5.2 and Appendix D). It is not practical to attempt to 
determine the impact of each individual type on any resulting site dose. Eleven general groups 
have been established to represent the entire inventory for the TSPA-LA model. To enable 
analyses of a limited number of types to represent, or bound, the behavior of all DSNF disposed 
in the repository, the DOE Office of Civilian Radioactive Waste Management (OCRWM) and 
the National Spent Nuclear Fuel Program (NSNFP) collaborated to identify the DSNF groups. 
These groups are individual categories of DSNF for TSPA analysis. The description of, and 
justification for, these DSNF groupings for repository criticality, design basis events, and TSPA 
analysis purposes are in the NSNFP report, DOE Spent Nuclear Fuel Grouping in Support of 
Criticality, DBE, TSPA-LA (DOE 2000 [DIRS 118968]). Also, during the Site Recommendation 
period, the NSNFP compiled a report, DOE Spent Nuclear Fuel Information in Support of 
TSPA-SR (DOE 2002 [DIRS 158405]), which contains postclosure performance characteristics 
of the groups and suggests models for describing the degradation rates of waste-forms in each 
group. 
The TSPA-LA groups of DSNF contained in the NSNFP documents and described in 
Sections 6.1.1 through 6.3.12 of this document follow, along with a typical type of DSNF in each 
group are: 
 Group 1 - Naval SNF 
 Group 2 - Plutonium/uranium alloy (Fermi 1 SNF) 
 Group 3 - Plutonium/uranium carbide (Fast Flux Test Facility-Test Fuel Assembly SNF) 
 Group 4 - Mixed oxide and plutonium oxide (Fast Flux Test Facility-Demonstration 
Fuel Assembly/Fast Flux Test Facility-Test Demonstration Fuel Assembly SNF) 
 Group 5 - Thorium/uranium carbide (Fort St. Vrain SNF) 
 Group 6 - Thorium/uranium oxide (Shippingport light water breeder reactor SNF) 
 Group 7 - Uranium metal (N Reactor SNF) 
 Group 8 - Uranium oxide (Three Mile Island-2 core debris) 
 Group 9 - Aluminum-based SNF (Foreign Research Reactor SNF) 
 Group 10 - Miscellaneous Fuel 
 Group 11 - Uranium-zirconium hydride (Training Research IsotopesGeneral Atomics 
SNF). 

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1.3 OBJECTIVE 
The objective of this model report is to use the group-related studies, published analyses, and 
results of experimental degradation tests to abstract a model for the degradation rate of the DSNF 
inventory based on consideration of available information for the fuels in the DSNF groups. The 
model is applicable to the conditions after waste package breach when the waste forms are 
exposed to water or water vapor and air. The model is also to be suitable for use in a TSPA-LA 
approach that does not differentiate between codisposal waste packages containing different 
DSNF types. 
This activity is covered under Technical Work Plan for: Regulatory Integration Modeling and 
Analysis of the Waste Form and Waste Package (BSC 2004 [DIRS 171583]). 
This model report does not use input from any other analysis reports or model reports. This 
model report provides direct input to the TSPA-LA model by specifying that all DSNF (with the 
exception of naval DSNF) should be modeled as degrading instantaneously. Naval DSNF should 
be modeled using the degradation rate of commercial spent nuclear fuel (CSNF). 

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2. QUALITY ASSURANCE 
Abstraction of the DSNF model and the supporting activities have been determined to be subject 
to Quality Assurance Requirements Description (DOE 2004 [DIRS 171539]), in accordance with 
Section 8 of Technical Work Plan for: Regulatory Integration Modeling and Analysis of the 
Waste Form and Waste Package (BSC 2004 [DIRS 171583]). This report does not address items 
or barriers on the Q-List. 
Control of the electronic management of data was accomplished in accordance with the controls 
specified by Appendix A of Technical Work Plan for: Regulatory Integration Modeling and 
Analysis of the Waste Form and Waste Package (BSC 2004 [DIRS 171583]). 

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3. USE OF SOFTWARE 
No computer software was used in the development of the abstracted upper-limit DSNF model 
described in this report. 

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4. INPUTS 
4.1 DIRECT INPUT 
There are no direct inputs to the abstraction of the DSNF upper-limit model developed in this 
model report. Information and experimental data from sources in the open literature are used as 
indirect input in Section 6 to describe the basis for the abstracted model. 
No parameter values are used as direct input to this model report. 
4.2 CRITERIA 
Section 3.4 of Project Requirements Document (Canori and Leitner 2003 [DIRS 166275]) 
contains requirements relevant to repository postclosure modeling. The key requirements 
(referred to by their PRD requirement identifier) relevant to this model report are: 
1. PRD-002/T-014 (Canori and Leitner 2003 [DIRS 166275]) Performance Objectives 
for the Geologic Repository After Permanent Closure. 
This specifies the repository performance objectives that must be met following 
permanent closure. It includes a requirement for multiple barriers and limits on 
radiological exposure. 
2. PRD-002/T-015 (Canori and Leitner 2003 [DIRS 166275]) Requirements for 
Performance Assessment. 
This specifies the technical requirements to be used in performing a performance 
assessment. It includes requirements for calculations, including data related to site 
geology, hydrology, variability in the models, and deterioration or degradation 
processes, including waste form degradation. 
3. PRD-002/T-016 (Canori and Leitner 2003 [DIRS 166275]) Requirements for 
Multiple Barriers. 
This specifies the requirements for the identification of the repository multiple 
barriers, describing the capabilities of the barriers to isolate waste, and provide the 
technical bases for the capability descriptions. 
The technical work plan for this model report (BSC 2004 [DIRS 171583], Table 3-1) contains a 
table of acceptance criteria based on the requirements in Yucca Mountain Review Plan, Final 
Report (NRC 2003 [DIRS 163274]). Applicable Acceptance Criteria are presented below. 
Section 8.3 quotes the full text that pertains to the specific criteria with pointers to the 
information within this report that pertains to the criteria. 

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System Description and Demonstration of Multiple Barriers Acceptance Criteria 
The following Acceptance Criteria are from Section 2.2.1.1.3 of Yucca Mountain Review Plan, 
Final Report (NRC 2003 [DIRS 163274]). 
 Acceptance Criterion 1Identification of Barriers Is Adequate 
 Acceptance Criterion 2Description of Barrier Capability to Isolate Waste Is 
Acceptable 
 Acceptance Criterion 3Technical Basis for Barrier Capability Is Adequately 
Presented. 
Degradation of Engineered Barriers 
The following Acceptance Criteria are from Section 2.2.1.3.1.3 of Yucca Mountain Review Plan, 
Final Report (NRC 2003 [DIRS 163274]). 
 Acceptance Criterion 1System Description and Model Integration are Adequate 
 Acceptance Criterion 2Data Are Sufficient for Model Justification 
 Acceptance Criterion 3Data Uncertainty is Characterized and Propagated through the 
Model Abstraction 
 Acceptance Criterion 4Model Uncertainty is Characterized and Propagated Through 
the Model Abstraction 
 Acceptance Criterion 5Model Abstraction Output Is Supported By Objective 
Comparisons. 
Radionuclide Release Rates and Solubility Limits Acceptance Criteria 
The following Acceptance Criteria are from Section 2.2.1.3.4.3 of Yucca Mountain Review Plan, 
Final Report (NRC 2003 [DIRS 163274]). 
 Acceptance Criterion 1System Description and Model Integration Are Adequate 
 Acceptance Criterion 2Data Are Sufficient for Model Justification 
 Acceptance Criterion 3Data Uncertainty Is Characterized and Propagated Through the 
Model Abstraction 
 Acceptance Criterion 4Model Uncertainty Is Characterized and Propagated Through 
the Model Abstraction 
 Acceptance Criterion 5Model Abstraction Output Is Supported by Objective 
Comparisons. 
4.3 CODES AND STANDARDS 
American Society for Testing and Materials (ASTM) Standard C 1174-97 [DIRS 105725], 
Standard Practice for Prediction of the Long-Term Behavior of Materials, Including Waste 
Forms, Used in Engineered Barrier Systems (EBS) for Geologic Disposal of High-Level 
Radioactive Waste, is used to support the degradation model development methodology. 

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5. ASSUMPTIONS 
One assumption is used in this document to support the output. 
5.1 COMMERCIAL SNF AS SURROGATE FOR NAVAL SNF 
Assumption: It is assumed that commercial light water reactor SNF can be used as a surrogate 
for naval SNF under the range of expected repository environmental conditions. 
Rationale: To provide a conservative simplification for the TSPA-LA model, the commercial 
SNF radionuclide release model is used to model release rates from naval SNF. Naval SNF 
represents a very small fraction of the SNF MTHM inventory (DOE 2002 [DIRS 158405], 
Table D-1). Releases from naval SNF waste packages are expected to be lower than for 
commercial SNF (BSC 2001 [DIRS 152059], Section 6.5). Because of its design, the 
radionuclide releases from naval SNF waste packages are considerably less than releases from 
commercial light water reactor SNF waste packages; accordingly, this assumption is 
conservative. This assertion is supported by the demonstration by the Naval Nuclear Propulsion 
Program (NNPP) that doses from the radionuclides released from naval SNF are less than the 
doses from radionuclides released from commercial SNF. There are 300 waste packages 
containing naval SNF compared to the commercial spent nuclear fuel (CSNF) inventory of 
approximately 7,500 waste packages. A comparison with an equivalent amount of Zircaloy-clad 
CSNF indicates that the total dose for the first one million years from the TSPA simulation, 
using the commercial-fuel equivalent, is more than four orders of magnitude higher than the total 
dose from the source-term simulation for naval SNF (BSC 2001 [DIRS 152059], p. 36 and 
Figure 6.1-2). Therefore, it is conservative to model naval SNF as CSNF. 
Confirmation Status: This assumption will be appropriately confirmed by the naval SNF 
disposition program in a planned classified Naval Nuclear Propulsion Program Addendum to the 
License Application. 
Use in the Model: This assumption is used in Section 6 to support the use of light water reactor 
SNF as a conservative surrogate for the naval SNF. 

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6. MODEL DISCUSSION 
10 CFR Part 63 [DIRS 156605] requires the DOE to perform a systematic analysis of features, 
events, and processes that could potentially affect the performance of the geologic repository. 
These analyses provide the technical basis for the inclusion or exclusion of specific features, 
events, and processes in the TSPA. The evaluation methodology provides an approach for 
considering the possible future states of the repository system, and begins with identification of 
features, events, and processes, followed by a rigorous screening that identifies those features, 
events, and processes that would be directly addressed in the TSPA-LA model. 
The development of a comprehensive list of features, events, and processes potentially relevant 
to postclosure performance of the potential Yucca Mountain repository is an ongoing, iterative 
process based on site-specific information, design, and regulations. The approach for developing 
an initial list in support of Total System Performance Assessment for the Site Recommendation 
(CRWMS M&O 2000 [DIRS 153246]) was documented in The Development of Information 
Catalogued in REV00 of the YMP FEP Database (Freeze et al. 2001 [DIRS 154365]). The 
initial list contained 328 features, events, and processes, of which 176 were included in 
TSPA-SR models (CRWMS M&2000 [DIRS 153246], Tables B-9 through B-17). To support 
the TSPA-LA model, the list was reevaluated in accordance with The Enhanced Plan for 
Features, Events, and Processes (FEPs) at Yucca Mountain (BSC 2002 [DIRS 158966], 
Section 3.2). Table 6-1 provides a list of features, events, and processes included in TSPA-LA 
models that are described in this model report. The table also summarizes their implementation 
in the TSPA-LA model and provides specific references to sections within this document. 
Two features, events, and processes are primarily addressed in this report: FEP 2.1.02.01.0A, 
DSNF Degradation (Alteration, Dissolution, and Radionuclide Release), and FEP 2.1.02.28.0A, 
Grouping of DSNF Types into Categories. 
Table 6-1. Included Features, Events, and Processes addressed in this Model Report 
FEP No. FEP Name 
Section Where the FEP 
is Addressed 
2.1.02.01.0A DSNF Degradation (Alteration, Dissolution, and Radionuclide Release) 6.1 
2.1.02.28.0A Grouping of DSNF Waste Types into Categories 1.2 
Because an upper-limit bounding model is to be used to describe the rate of degradation of 
DSNF in the TSPA-LA model (Section 8), the effects of potential interactions between 
codisposed waste on the DSNF degradation rate (i.e., the part of FEP 2.1.01.02.0B that pertains 
to this report) are not explicitly considered. However, the DSNF instantaneous degradation 
model, which is to be used in the TSPA-LA model, conservatively bounds the effects of such 
interactions on the DSNF degradation rate. 
For each DSNF category or group this section examines two types of DSNF degradation models 
for possible TSPA-LA application purposes: best-estimate and upper limit. 
Preliminary analyses of the DSNF in the repository (Thornton 1998 [DIRS 125082]; 
Thornton 1998 [DIRS 107796]) had indicated that for large incremental time steps (>1,000 
years), the dose at the compliance point on the site boundary resulting from the failure of waste 

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packages containing DSNF is very insensitive to the release rates from the DSNF waste form. 
Subsequent, more detailed, analyses (CRWMS M&O 2000 [DIRS 125038], Section 7, bullets 11 
to 16; CRWMS M&O 2000 [DIRS 153595], Section 6; BSC 2003 [DIRS 168796], Section 
3.3.6) have supported the conclusion that even complete and instantaneous degradation of the 
DSNF waste results in calculated doses at the compliance point on the site boundary that are 
below established limits. Also, as discussed below, examination of the available experimental or 
analytical degradation rate behavior data for the DSNF groups shows that, in general, insufficient 
qualified data exists to formulate or abstract separate models for each group. 
6.1 MODEL SELECTION AND ABSTRACTION 
In the following sections, the best-estimate model for each of the DSNF groups is examined to 
assess both the models defensibility and the barrier capability of DSNF, as indicated by the 
best-estimate model. The degradation modes of interest are oxidation, corrosion, and dissolution 
of the waste forms in air, water vapor, and liquid water. These modes, as well as their associated 
radionuclide release rates, are encompassed under the general heading of degradation model. 
The radionuclide release rate is related to the degradation rate models for each waste form 
through the simplification that all components of the DSNF waste form are released congruently 
with the degradation of the matrix. The release rate of radionuclide i (Ri) is related to the rate of 
waste form degradation as follows: 
Ri = D  A  F (Eq. 1) 
where 
D = degradation rate (mg/m2day) 
A = total surface area of waste form exposed to water (m2) 
F = mass fraction of species i 
In using the best-estimate models, the release rate of radionuclides from the DSNF is calculated 
by multiplying the dissolution rate provided by the model(s) by the total surface area of the 
DSNF exposed to water and by the inventory of the radionuclides. 
A list of information sources used to provide the rationale for using an upper-bound 
instantaneous release model for DSNF degradation in the TSPA-LA model follows: 
 Several reports (Thornton 1998 [DIRS 125082]; Thornton 1998 [DIRS 107796]; 
CRWMS M&O 2000 [DIRS 125038], Sections 6.7.1 and 7, Bullets 11-16; BSC 2003 
[DIRS 168796], Section 3.3.6) provide analyses demonstrating that the contribution of 
the DSNF to the postclosure site dose is negligible, even when extremely conservative 
degradation and pyrophoric behavior of the DSNF is used. Performance Assessment 
Sensitivity Analyses of Selected U.S. Department of Energy Spent Fuels 
(CRWMS M&O 2000 [DIRS 125038], Section 6.7.1 and 7, Bullets 11 to 16) provides 
analyses that show that the dose rates at the site boundary resulting from the failure of 
DSNF-containing waste packages are insensitive to the degradation rates of the DSNF 
waste forms. Input values for the degradation rate used in the sensitivity analyses did 
not require qualification because it is a parametric study that used the maximum 
conceivable degradation rates of the waste form to perform the analysis. The analyses in 

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these documents do not require qualification because they do not provide direct input to 
the abstraction of any of the degradation models. 
 DOE Spent Nuclear Fuel Grouping in Support of Criticality, DBE, TSPA-LA 
(DOE 2000 [DIRS 118968]) establishes and justifies the DSNF groups to be used for 
TSPA-LA analysis purposes and proposes that the degradation behavior of each DSNF 
group can be adequately represented by the behavior of a surrogate SNF type within the 
group. The conclusions are supported by TSPA analyses (Thornton 1998 
[DIRS 125082]; Thornton 1998 [DIRS 107796]; CRWMS M&O 2000 [DIRS 125038], 
Sections 6.7.1 and 7, Bullets 11 to 16; CRWMS M&O 2000 [DIRS 153595], Section 6; 
BSC 2003 [DIRS 168796], Section 3.3.6). These analyses show that the 
postcontainment site-boundary dose is insensitive to the degradation rate of the DSNF or 
individual components of the DSNF. This information does not require qualification 
because no parameter is directly used in selecting or abstracting degradation models. 
Appendix D of DOE Spent Nuclear Fuel Grouping in Support of Criticality, DBE, 
TSPA-LA (DOE 2000 [DIRS 118968]) contains the equivalent metric tons of heavy 
metal (MTHM) inventory for each of the DSNF groups in the Total System Performance 
Assessment for the Site Recommendation (TSPA-SR). The data used in this model 
report compares the relative importance of each group in formulating a composite 
degradation model for all DSNF. While the data are required to be qualified for other 
documents supporting the TSPA-LA model, they are used here to indicate the relative 
quantities of the DSNF groups only. Therefore, they do not require qualification for use 
in abstracting group dissolution models in this report. 
 DOE reports on SNF (DOE 1999 [DIRS 107790], DOE 2000 [DIRS 152658], 
DOE 2002 [DIRS 158405]) contain proposed oxidation rate and dissolution models and 
MTHM inventory for each of the DSNF TSPA groups identified in DOE Spent Nuclear 
Fuel Grouping in Support of Criticality, DBE, TSPA-LA (DOE 2000 [DIRS 118968]). 
The suggested models were evaluated in this model report. DOE Spent Nuclear Fuel 
Information in Support of TSPA-SR (DOE 1999 [DIRS 107790], Appendix B) and 
DOE Spent Nuclear Fuel Information in Support of TSPA-SR (DOE 2002 [DIRS 
158405], Appendix B) contain information on the surface area for each of the TSPA-LA 
DSNF groups, and also contain the equivalent MTHM inventory. The information in 
DOE Spent Nuclear Fuel Information in Support of TSPA-SR (DOE 1999 [DIRS 
107790]) is used for comparative purposes only, to indicate the conservatism inherent in 
the upper limit dissolution model analyzed in this report. 
The following sections describe the upper-limit and best-estimate dissolution models developed 
for each DSNF group using the available information sources. Table 6-2 contains a summary of 
these models. This table shows the two types of models for each DSNF group, the surrogate 
material for which the model was developed, and the literature or document source reference for 
the model. A discussion of the individual models for each fuel group follows. 

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Table 6-2. DSNF, Naval SNF, Plutonium Disposition Release/Degradation Models 
Upper-Limit Model Best-Estimate Model 
DSNF Group Surrogate Model Ref Surrogate Model Ref. 
1. Naval N/A Section 6.3.1 N/A Section 6.3.1 Section 6.3.1 N/A 
2. Plutonium / 
Uranium Alloy 
N/A Instantaneous release 
upon exposure to 
groundwater 
N/A uranium - molybdenum (semi-empirical) 
rate (mg metal/cm2/h) = 1.15  108 exp{(- 
66,500  12,200 J/mol)/RT} 
[100178C] 
rate (mg metal/cm2/h) = 1.58  106 exp{(- 
80,500  10,600 J/mol)/RT} [304-440C] 
(Linear interpolation between 178C and 
304C) 
DOE 2002 
[DIRS 158405], 
Section 6.2, 
Equations 6-1 
and 6-2 
3. Plutonium / 
Uranium Carbide 
N/A Instantaneous release 
upon exposure to 
groundwater 
N/A uranium metal 100  Unirradiated uranium metal bestestimate: 
k (mg/m2-day) = 100  {1.21  1015 exp(- 
66.4  2.0 kJ/mol /RT)} 
DOE 2002 
[DIRS 158405], 
Section 6.7, 
Equation 6-11 
4. Mixed Oxide and 
Plutonium Oxide 
N/A Instantaneous release 
upon exposure to 
groundwater 
N/A light water reactor SNF (semi-empirical) 
uranium oxide best-estimate model 
DOE 1999 
[DIRS 107790], 
Section 6.4 
5. Thorium / 
Uranium Carbide 
N/A Instantaneous release 
upon exposure to 
groundwater 
N/A SiC (semi-empirical) 
R (kg/m2-s) = 0.6  10-12 
DOE 1999 
[DIRS 107790], 
Section 6.5 and 
Figure 6-3 
6. Thorium / 
Uranium Oxide 
N/A Instantaneous release 
upon exposure to 
groundwater 
N/A Synroc (semi-empirical) 
k (mg/m2day) = 
82.0  10(-1,000/TK) 
(1-3) 
7. Uranium Metal- 
Based 
N/A Instantaneous release 
upon exposure to 
groundwater 
N/A N Reactor (semi-empirical) 
2.52  1010exp 
(-66,400/RT) 
mg/cm2-hr 
R = 8.314 J/mol-K 
Section 6.7; 
DOE 2002 
[DIRS 158405], 
Section 6.7, 
Equation 6-11 
8a. Intact Uranium 
Oxide 
N/A Instantaneous release 
upon exposure to 
groundwater 
N/A light water reactor SNF (semi-empirical) 
uranium oxide best-estimate model 
DOE 1999 
[DIRS 107790], 
Sections 6.4 and 
6.8 

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Table 6-2. DSNF, Naval SNF, Plutonium Disposition Release/Degradation Models (Continued) 
Upper-Limit Model Best-Estimate Model 
DSNF Group Surrogate Model Ref Surrogate Model Ref. 
8b. Damaged 
Uranium Oxide 
N/A Instantaneous release 
upon exposure to 
groundwater 
Three Mile Island-2 debris (surface area enhancement factor of 100 is 
based on professional judgment) 
100  uranium oxide best-estimate 
Section 6.3.7 
9. Aluminum-based N/A Instantaneous release 
upon exposure to 
groundwater 
N/A Savannah River Site 
uranium/ aluminum SNF in 
J-13 well water 
(empirical) 
1.38 mg metal/m2day at 25C 
13.80 mg metal/m2day at 90C 
Wiersma and 
Mickalonis 1998 
[DIRS 107984], 
Tables 3 and 4 
10. Miscellaneous N/A Instantaneous release 
upon exposure to 
groundwater 
N/A N/A (empirical) 
rate (mg metal/cm2/h) = 1.15  108 exp{(- 
66,500  12,200 J/mol)/RT} [100178C] 
rate (mg metal/cm2/h) = 1.58  106 
exp{(-80,500  10,600 J/mol)/RT} 
[304C to 440C] 
DOE 2002 
[DIRS 158405], 
Section 6.2, 
Equations 6-1 
and 6-2 
11. Uranium- 
Zirconium Hydride 
N/A Instantaneous release 
upon exposure to 
groundwater 
N/A Training Research 
IsotopesGeneral Atomic 
(empirical) 
0.1  uranium oxide best estimate 
DOE 1999 
[DIRS 107790], 
Section 6.11; 
LMITCO 1997 
[DIRS 125091], 
Section 2.1.8 

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6.1.1 DSNF Group 1 (Naval SNF) 
The Naval Nuclear Propulsion Program has demonstrated that the dissolution rate of naval SNF 
is lower than the dissolution rate of commercial SNF (BSC 2001 [DIRS 152059], Section 6.5). 
A planned classified Naval Nuclear Propulsion Program Addendum to the License Application 
will provide the details of this evaluation. To provide a conservative simplification for the 
TSPA-LA model, the commercial SNF dissolution model will be used for the dissolution of the 
naval SNF (Assumption 5.1). 
6.1.2 DSNF Group 2 (Plutonium/Uranium Alloy) 
There are several individual types of plutonium/uranium alloy-based DSNF primarily comprised 
of uranium-molybdenum, although smaller quantities of uranium-thorium alloy are part of this 
group (DOE 2002 [DIRS 158405], Sections 5.3.1.2 and 6.2). These alloy fuels are generally 
clad in zirconium alloy, but some small quantities have aluminum, stainless steel, or tantalum 
alloy cladding. The largest single fuel type in this group, comprising slightly over 90% by 
weight of uranium of the total Group 2 inventory, are the zirconium-clad uranium-molybdenum 
Fermi Unit 1 SNF. Studies of the dissolution behavior of uranium-molybdenum and uraniumzirconium 
alloys, reported in Section 6.2 of DOE Spent Nuclear Fuel Information in Support of 
TSPA-SR (DOE 2002 [DIRS 158405]), give dissolution rates for the alloy that depend on the 
amount of alloying molybdenum and zirconium. The corrosion behavior of uraniummolybdenum 
and uranium-zirconium alloys reported in DOE Spent Nuclear Fuel Information in 
Support of TSPA-SR (DOE 2002 [DIRS 158405]) supported the following correlations for the 
corrosion rate (in mg metal/cm2h) of Group 2 SNF based on uranium-molybdenum alloy: 
For the range of temperatures between 100C and 178C: 
rate = 1.15  108 exp[(-66,500  12,200 J/mol)/RTk] (Eq. 2) 
For the range of temperatures between 304C and 440C: 
rate = 1.58  106 exp[(-80,500  10,600 J/mol)/RTk] (Eq. 3) 
Equation 3 covers a temperature range higher than that expected for waste form temperature in 
codisposal packages. However, it can be used to interpolate rates in the interval between 178C 
and 304C for waste package temperatures in that range. These equations are based on uraniummolybdenum 
alloys with, respectively, molybdenum less than 8% by weight and molybdenum 
greater than 8% by weight. They predict corrosion rates approximately a factor of 40 lower than 
rates predicted for unirradiated metallic uranium and a factor of 200 lower than the rates 
predicted for N Reactor SNF. Section 6.2 of DOE Spent Nuclear Fuel Information in Support of 
TSPA-SR (DOE 2002 [DIRS 158405]) also noted that the observed corrosion rate for uraniummolybdenum 
alloys was 1.5 to 5 orders of magnitude lower than that of bare uranium metal 
(Group 7). The information in Figures 6-1 and 6-2 of DOE Spent Nuclear Fuel Information in 
Support of TSPA-SR (DOE 2002 [DIRS 158405]) showed that the corrosion rate of uraniumzirconium 
alloys is over two orders of magnitude lower than that of the uranium-molybdenum 
alloy from 75C to 200C. 

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Equations 2 and 3 should be used for the best-estimate models for the Group 2 SNF. It should 
also be noted that the total inventory of this SNF waste form is small compared to most other 
DSNF types (Table 6-3). These equations were derived from data in the open literature for the 
corrosion of unirradiated metallic uranium and uranium alloy, and the suggested best-estimate 
degradation model for Group 2 SNF for unirradiated uranium alloys was obtained from 
Section 6.2 of DOE Spent Nuclear Fuel Information in Support of TSPA-SR (DOE 2002 
[DIRS 158405]). However, the uranium-molybdenum and uranium-zirconium corrosion data 
used for the best-estimate model did not include irradiation effects, for which there are no known 
data. Because the data upon which the models in this reference are based is from open literature 
sources, it is from uncontrolled sources and, thus, would need to be qualified to support direct 
use of these models in the TSPA-LA model. 
Table 6-3. Approximate Dissolution Rates at pH 8.5, 0.002 Molar CO3
2-, and 0.20 Atmospheres Oxygen 
Calculated from the Best-Estimate Models 
DSNF GROUP 
Approximate Fuel 
Matrix Density,. a in 
g/cm3 (./.Group 7) 
Inventory 
(MTHM)b 
Best-Estimate 
Corrosion Rate c at 
50C (mg/m2day) 
(adjusted for density) 
Group 2 - Plutonium/Uranium Alloy 19.05d (1) 8.5 492 
Group 3 - Plutonium/Uranium Carbide 11.28 (0.59) 0.1 1.09  106 
Group 4 - Mixed Oxide and Plutonium Oxide 11.03 (0.58) 11.59 5.41 
Group 5 - Thorium/Uranium Carbide 9.35 (0.49) 24.52 0.025 
Group 6 - Thorium/Uranium Oxide 9.87 (0.52) 46.98 0.034 
Group 7 - Uranium Metal-Based 19.05 (1) 1,984.81 1.1  105 
Group 8a - Intact Uranium Oxide e 9.86 f (0.52) 166.2 (8a + 8b) 4.83 
Group 8b - Damaged Uranium Oxide 9.86 (0.52)  483 
Group 9 - Aluminum-Based 2.70 g (0.14) 19.54 0.19 at 25C 
Group 10 - Miscellaneous SNF 19.05 d (1) 4.24 492 
Group 11 - Uranium-Zirconium Hydride 6.88 (0.36) 1.51 0.33 
NOTES: a DOE 2002 [DIRS 158405], Appendix C, Attachment A. 
b DOE 2002 [DIRS 158405], Appendix D, Radionuclide Inventory Summary of DOE SNF. 
c The tabulated rates are adjusted for density as follows rate = rategroup i  (.i/.group 7). 
d Conservatively taken to be that of metallic uranium. 
e Using Equation 7 with nominal burnup of 40 MWd/kgU. 
f Estimating that the actual density after irradiation of the oxide fuel is 90% of the theoretical density of 
10.96 g/cm3. 
g For the purpose of this comparison the dissolution rate of the aluminum matrix is used. 
6.1.3 DSNF Group 3 (Plutonium/Uranium Carbide SNF) 
Group 3 SNF consists primarily of fuel from the Fast Flux Test Facility-Test Fuel Assembly with 
most of the balance from the sodium reactor experiment. Both consist of mixed-carbide-fissile 
fuel particles in a nongraphite matrix. Only a very limited amount of information concerning the 
chemical reactivity of the DSNF in this group is available. Because of the lack of specific 
information concerning degradation behavior and the indication that the fissile particles could be 
very reactive, the best-estimate rate models should be a rate that is very high, arbitrarily taken as 

DSNF and Other Waste Form Degradation Abstraction 
ANL-WIS-MD-000004 REV 04 6-8 November 2004 
100 times the dissolution rate of the uranium metal SNF (DOE 2002 [DIRS 158405], 
Section 6.3). In applying the model, the rate of dissolution should be taken as the congruent 
dissolution of the composite carbide matrix and fuel particles with the radionuclide content of its 
particles released as the composite degrades. 
The degradation model for Group 3 SNF was obtained from Section 6.3 of DOE Spent Nuclear 
Fuel Information in Support of TSPA-SR (DOE 2002 [DIRS 158405]). The model in this 
reference is based on sparse data or information from old open literature sources and would need 
to be qualified to support direct use of these models in the TSPA-LA model. 
6.1.4 DSNF Group 4 (Mixed Oxide and Plutonium Oxide SNF) 
Group 4 SNF is composed of a mixture of uranium and plutonium oxides with various cladding 
materials. Although several dozen SNF types are in this category, the predominant types are the 
Fast Flux Test Facility, Demonstration Fuel Assembly, and Test Demonstration Fuel Assembly 
fuel, contributing over 83% of the mass of heavy metal (DOE 2002 [DIRS 158405], 
Section 5.3.1.4). Since the fuel material is either uranium oxide or plutonium oxide, the 
dissolution kinetics of the fuel form is not expected to be materially different from that for 
commercial light water reactor SNF. The best-estimate models abstracted are those for the light 
water reactor SNF. 
The degradation model for Group 4 SNF is abstracted in CSNF Waste Form Degradation: 
Summary Abstraction (CRWMS M&O 2000 [DIRS 136060]) for commercial UO2 fuel. The 
data or information upon which the suggested model in this reference is based is qualified data. 
6.1.5 DSNF Group 5 (Thorium/Uranium Carbide SNF) 
This SNF group consists primarily of thorium or uranium-carbide particles coated with pyrolytic 
carbon or silicon carbide embedded in a carbonaceous matrix. Over 90% by weight of MTHM 
of this group is Fort St. Vrain SNF, with the remainder being Peach Bottom Unit 1 SNF 
(DOE 2002 [DIRS 158405], Section 5.3.1.5). The Peach Bottom fuel may be more damaged 
than the Fort St. Vrain fuel, although there is little qualified information concerning the condition 
of either. Fuel in this group, whose protective silicon carbide coatings and matrix are intact, 
would be expected to have dissolution kinetics similar to those of pure silicon carbide, and this 
would provide the basis for the best-estimate model for this group. Spent nuclear fuel with a 
damaged coating or matrix would be expected to have dissolution kinetics similar to uranium 
carbide (DOE 2002 [DIRS 158405], Section 6.5; CRWMS M&O 1998 [DIRS 100362], 
Section 6.3.2.1), and this should provide the basis for a conservative model for this group. 
The suggested degradation model for Group 5 SNF was obtained from Section 6.5 of DOE Spent 
Nuclear Fuel Information in Support of TSPA-SR (DOE 2002 [DIRS 158405]). The data or 
information upon which the suggested model in this reference is based is from open literature 
sources and would need to be qualified to support direct use of these models in the TSPA-LA 
model. The specific surface area used should be taken as 0.022 m2/g (DOE 2002 
[DIRS 158405], Appendix B). 

DSNF and Other Waste Form Degradation Abstraction 
ANL-WIS-MD-000004 REV 04 6-9 November 2004 
6.1.6 DSNF Group 6 (Thorium/Uranium Oxide SNF) 
Thorium/uranium oxide SNF in Group 6 primarily consists of the Shippingport light-water 
breeder reactor reflector SNF with the remainder from the Dresden Unit 1 and the Experimental 
Research Reactor thorium/uranium oxide SNF. The Shippingport fuel was clad in zirconium 
alloy, and the Dresden and Experimental Research Reactor fuel was clad in stainless steel. The 
thorium/uranium oxide fuel consisted of sintered pellets similar to commercial light water reactor 
fuel pellets. 
Several reports discussed in DOE Spent Nuclear Fuel Information in Support of TSPA-SR 
(DOE 1999 [DIRS 107790]) and Section 6.6 of DOE Spent Nuclear Fuel Information in Support 
of TSPA-SR (DOE 2002 [DIRS 158405]) indicate that the mixed thorium uranium was more 
corrosion resistant than pure uranium dioxide. The approach taken here, for the best estimate 
model, is to use a ceramic Synroc release model that conservatively bounded data for 
unirradiated light water breeder reactor fuel (DOE 2002 [DIRS 158405], Section 6.6). This 
model and the supporting data would need to be qualified to support direct use of the model in 
the TSPA-LA model 
6.1.7 DSNF Group 7 (Uranium Metal SNF) 
The zirconium-clad N Reactor SNF constitutes over 95% of the mass of this group, with small 
quantities of aluminum-clad-single-pass reactor and experimental breeder reactor metallic 
uranium SNF (DOE 2002 [DIRS 158405], Section 5.3.1.7). Due to the small quantity of other 
uranium metal-based DSNF compared to the N Reactor SNF, the degradation of the other SNF 
may be taken as similar to the N Reactor SNF. A significant fraction of the N Reactor fuel is 
visibly damaged, and much of the rest could have small pinholes/cracks in the cladding. The 
exposed uranium metal surfaces of the N Reactor fuel elements show extensive corrosion 
resulting from the many years of direct exposure to the K-basin water (Abrefah et al. 1995 
[DIRS 151125], Section 3.1 and Figures 3.1-3.5; Abrefah et al. 1999 [DIRS 151226], Figure 3.1; 
Welsh et al. 1997 [DIRS 151225], Section 1, Figures 3.3 to 3.7). N Reactor SNF comprises 
approximately 85% by weight in MTHM of the total quantity of DSNF (DOE 2002 
[DIRS 158405], Appendix D). 
Experimental studies have been conducted in the past on unirradiated and uncorroded uranium 
metal and uranium-metal alloy, and some of this work is summarized in Section 6.7 of 
DOE Spent Nuclear Fuel Information in Support of TSPA-SR (DOE 2002 [DIRS 158405]). 
Uranium metal degradation behavior is obtained from the following references: 
 The information in DOE Spent Nuclear Fuel Information in Support of TSPA-SR 
(DOE 2002 [DIRS 158405], Section 6.7) is used to abstract a temperature dependence for 
the Arrhenius form of the Group 7 dissolution model. The analyses of the dissolution or 
oxidation reaction kinetics of uranium metal found in DOE Spent Nuclear Fuel 
Information in Support of TSPA-SR (DOE 2002 [DIRS 158405]) include a 
comprehensive evaluation of the available literature concerning the temperature 
dependence of the oxidation and dissolution rates of bare uranium metal. These analyses 
were conducted under an NSNFP quality assurance program similar to Quality Assurance 
Requirements and Description (DOE 2004 [DIRS 171539]). The temperature 

DSNF and Other Waste Form Degradation Abstraction 
ANL-WIS-MD-000004 REV 04 6-10 November 2004 
dependence developed in the report is used to abstract the temperature dependence for the 
dissolution of the Group 7 DSNF. The temperature dependence of these reactions is 
developed in DOE Spent Nuclear Fuel Information in Support of TSPA-SR (DOE 2002 
[DIRS 158405]). This NSNFP report meets the definition of product output for use in 
this model report in Attachment 3 of AP-3.15Q, Managing Technical Product Inputs. 
 Gray and Einziger (1998 [DIRS 109691]) report the results of dissolution testing of 
irradiated N Reactor SNF. 
 Abrefah et al. (1996 [DIRS 162037], Section 3.4), Abrefah, Buchanan, and Marschman 
(1998 [DIRS 159159], Section 3.2), Abrefah et al. (1999 [DIRS 151226]), Abrefah and 
Sell (1999 [DIRS 118528], Section 4), Abrefah et al. (2000 [DIRS 153644]), Goldberg 
(2000 [DIRS 159482]), Kaminski (2001 [DIRS 161679]), Totemeier et al. (1999 
[DIRS 159408]), and Tyfield (1988 [DIRS 159597]), along with DOE Spent Nuclear 
Fuel Information in Support of TSPA-SR (DOE 2002 [DIRS 158405]), provide 
information on the oxidation rate in moist air and the aqueous dissolution rate and 
chemical reactivity of uranium metal-based fuel, both unirradiated and irradiated. 
 Corrosion of Reactor Grade Uranium in Aqueous Solutions Relevant to Storage and 
Transport. (Tyfield 1988 [DIRS 159597]) contains the results of a study of the 
dissolution kinetics of Magnox SNF in aqueous environments under various pH 
conditions. 
Tests by Tyfield (1988 [DIRS 159597]) were performed on samples of Magnox SNF, a nuclear 
fuel design similar to the N Reactor fuel, except that the Magnox cladding was aluminum, rather 
than zirconium, alloy. The Magnox SNF did not exhibit the kind of physical damage that much 
of the N Reactor SNF showed, and therefore the uranium metal did not have the prolonged direct 
exposure to water that the N Reactor SNF has had. This fuel had burnups ranging from 1,100 to 
9,850 MWd/t (reasonably typical of the N Reactor SNF), but had displayed more extensive 
localized irradiation swelling (ranging from 5% to 50%) compared to swelling of approximately 
a few percent observed on the N Reactor SNF (Abrefah et al. 1995 [DIRS 151125], 
Section 3.2.1). The tests in the study by Tyfield (1988 [DIRS 159597]) also used samples with 
initially uncorroded surfaces, and were conducted in both fluoride-free and 1 kg/m3 fluoridedoped 
water solutions. Similarly, the studies by Abrefah, Buchanan, and Marschman (1998 
[DIRS 159159]) and Abrefah and Sell (1999 [DIRS 118528]) were performed on irradiated N 
Reactor fuel samples taken from a section of the assembly that had not been in direct contact 
with the K-basin water, and therefore did not have corroded uranium metal surfaces. The study 
by Kaminski (2001 [DIRS 161679]) was performed in EJ-13 water on samples of unirradiated N 
Reactor fuel obtained from archive materials. And unlike the studies by Tyfield (1988 
[DIRS 159597]), Abrefah, Buchanan, and Marschman (1998 [DIRS 159159]) and Abrefah and 
Sell (1999 [DIRS 118528]), the Kaminski (2001 [DIRS 161679]) tests were performed under 
anoxic water conditions, whereas the repository environment is expected to be oxic. Other 
studies had indicated that the rate of degradation of uranium metal in an anoxic water 
environment is greater than in an oxic environment (DOE 2000 [DIRS 152658], Figures 2-9 and 
2-10). 

DSNF and Other Waste Form Degradation Abstraction 
ANL-WIS-MD-000004 REV 04 6-11 November 2004 
Additionally, the Tyfield (1988 [DIRS 159597]) study reported an activation energy for 
oxidation of 60 kJ/mol over the temperature range of 75C to 145C. The study by Abrefah, 
Buchanan, and Marschman (1998 [DIRS 159159], Section 3.2) had indicated an activation 
energy for oxidation in moist air of 67 kJ/mol, and Abrefah and Sell (1999 [DIRS 118528], 
Equation 4.3) had indicated an activation energy for oxidation in moist helium of 13.3 kcal/mol 
(55.7 kJ/mol). These values are very comparable with the value of 66.4  2 kJ/mol (DOE 2002 
[DIRS 158405]) for the Group 7 models discussed below. 
The degradation of all uranium metal-based DSNF is taken to be similar to that observed by 
Gray and Einziger (1998 [DIRS 109691]) for N Reactor SNF specimens. This dissolution rate is 
further supported by a study by Abrefah et al. (2000 [DIRS 153644]) concerning the dissolution 
behavior of unirradiated and uncorroded N Reactor fuel in which they concluded that the 
dissolution behavior of the uranium metal in the fuel was qualitatively similar to, but 
significantly less than, that for the N Reactor SNF observed by Gray and Einziger (1998 
[DIRS 109691]). The data contained in the report were generated under Pacific Northwest 
National Laboratorys (PNNLs) quality assurance requirements and have been widely accepted 
and used in the analysis of N Reactor SNF behavior, in particular by the NSNFP (DOE 2002 
[DIRS 158405], Section 6.7; DOE 2000 [DIRS 152658], Section 2.3). Although there is an 
extensive database for the dissolution of metallic uranium in aqueous environments, the data 
contained in this report are the only known experimental data specifically for the dissolution 
behavior of N Reactor SNF. The database for the oxidative dissolution of uranium metal was 
analyzed by the NSNFP and summarized in (DOE 2000 [DIRS 152658]), and it was concluded 
that the basic form of the dissolution model, taking account of specific surface area differences, 
would apply generally to metallic uranium-based SNF inventory. Additionally, the fact that the 
amount of metallic uranium-based SNF that is not N Reactor fuel is very small (DOE 2002 
[DIRS 158405], Section 5.3.1.7) also supports the use of the N Reactor SNF model for all 
uranium metal-based SNF. Additionally, a study (Tyfield 1988 [DIRS 159597]) concerning the 
dissolution behavior of Magnox SNF, a metallic uranium-based nuclear fuel similar to the N 
Reactor SNF, showed similar dissolution kinetics. 
The degradation rate for uranium metal-based (i.e., N Reactor) DSNF in water can be regarded 
as a simple enhancement of the oxidation rate of unirradiated and uncorroded uranium metal 
given in NSNFP reports (DOE 2000 [DIRS 152658], Section 2.3; DOE 2002 [DIRS 158405], 
Section 6.7). This rate is supported by test results reported by Abrefah et al. (2000 
[DIRS 153644]) concerning the dissolution behavior of unirradiated and uncorroded N Reactor 
fuel. From the testing results they concluded that the dissolution behavior of the uranium metal 
in the fuel was qualitatively similar to, but significantly less than, that for the N Reactor SNF 
observed by Gray and Einziger (1998 [DIRS 109691]). Review of Oxidation Rates of DOE Spent 
Nuclear Fuel, Part 1: Metallic Fuel (DOE 2000 [DIRS 152658]) and DOE Spent Nuclear Fuel 
Information in Support of TSPA-SR (DOE 2002 [DIRS 158405]) are reports generated by the 
NSNFP that contain analyses of the oxidation rates of metallic uranium in dry air, humid air, 
saturated water vapor, and liquid water. The data analyzed in the NSNFP reports represent a 
comprehensive compilation of the available literature on uranium metal oxidation. The basis for 
using this rate is the similarity of the temperature dependence for the oxidation given in the DOE 
report and that for N Reactor SNF shown in Gray and Einziger (1998 [DIRS 109691]). The 
NSNFP reports compile reference data from numerous publications that have been used 
extensively in the past for the analysis of the behavior of metallic uranium. The reports also 

DSNF and Other Waste Form Degradation Abstraction 
ANL-WIS-MD-000004 REV 04 6-12 November 2004 
provide a correlation for the oxidation rates of metallic uranium based on literature data for 
unirradiated and uncorroded uranium metal. The correlation provided in Section 6.7 of DOE 
Spent Nuclear Fuel Information in Support of TSPA-SR (DOE 2002 [DIRS 158405]) forms a 
partial basis for the best-estimate dissolution rate expression for N Reactor SNF used in this 
report. 
The NSNFP has conducted an analysis of the kinetics of uranium metal oxidation from several 
literature sources (DOE 2000 [DIRS 152658], Sections 2.2.3 and 2.2.4; DOE 2002 
[DIRS 158405], Equation 6-11; DTN: MO0210SPAOXIDA.001 [DIRS 160443]) from which 
the Equation 4 expression for the dissolution kinetics of unirradiated and uncorroded uranium 
metal in water was derived from Equation 6-11 of DOE Spent Nuclear Fuel Information in 
Support of TSPA-SR (DOE 2002 [DIRS 158405]): 
k (mg U/cm2hr) = 5.03  109 exp(-A/RT), or 
k (mg U/m2day) = 1.21  1015 exp(-A/RT) (20C to 200C) (Eq. 4) 
where 
k = the corrosion rate 
A = the activation energy for dissolution (66.4  2.0 kJ/mol) 
R = the gas constant (8.314 J/mol-K) 
T = the temperature in kelvins 
The Arrhenius temperature dependence of these reactions is developed in DOE Spent Nuclear 
Fuel Information in Support of TSPA-SR (DOE 2002 [DIRS 158405]). Additionally, PNNL has 
conducted flow-through dissolution tests on samples of the N Reactor SNF. These tests 
generally indicate that the rates of dissolution of this SNF exceed that of the unirradiated or 
uncorroded uranium metal or alloy. The PNNL experimental work was performed on N Reactor 
SNF samples cut from an undamaged/uncorroded area of an N Reactor fuel element (Gray and 
Einziger 1998 [DIRS 109691], Section 2.2 and Figure 1). The test results indicated that there 
were two stages in the dissolution behavior of the N Reactor SNF samples tested: an initial 
Stage 1 rate and, after an incubation period, a significantly higher Stage 2 dissolution rate. 
This Stage 1 dissolution of the matrix was correlated with the Equation 5 expression (Gray and 
Einziger 1998 [DIRS 109691], Equation 3): 
log10(R) [mg/m2day] = 8.52 + 0.347 log10([CO3
2-]) + 0.088 pH  1929/T (Eq. 5) 
where 
R = the dissolution rate expressed in mg/m2day, 
[CO3
2-] = the molar concentration of carbonate in the contacting solution, 
T = the absolute temperature. 
Unsaturated testing of metallic uranium showed that alteration phases form that can retard the 
release of solutes from the N Reactor SNF (Goldberg 2000 [DIRS 159482], unnumbered 
paragraphs labeled Unsaturated Corrosion Tests and Discussion and Summary). This may 
explain the lower Stage 1 rates. 

DSNF and Other Waste Form Degradation Abstraction 
ANL-WIS-MD-000004 REV 04 6-13 November 2004 
During Stage 2, a sludge-like corrosion product material was formed consisting primarily of 
U4O9 and the mineral form schoepite (UO3 2H2O) (Gray and Einziger 1998 [DIRS 109691], 
Section 4.2). The rates of dissolution that were observed for Stage 2 reached 13,000 and 180,000 
mg/m2day at 25C and 75C, respectively (Gray and Einziger 1998 [DIRS 109691], Section 
4.3). The temperature dependence of the dissolution was similar to that for unirradiated and 
uncorroded uranium metal in Section 2.2 of Review of Oxidation Rates of DOE Spent Nuclear 
Fuel, Part 1: Metallic Fuel (DOE 2000 [DIRS 152658]), although the absolute rates were higher 
for the N Reactor SNF (see Table 6-4). Table 6-5 gives values for the Stage 1 dissolution rate as 
calculated from Equation 5 and for the experimental Stage 2 dissolution rates of the PNNL study. 
In most of the cases where Stage 2 dissolution was initiated, it began around 60 days into the 
test, a very short period in terms of TSPA analyses. 
Table 6-4. Comparison of Unirradiated/Uncorroded Uranium Metal and N Reactor SNF Corrosion Rates 
Temp 
(C) 
N Reactor SNFa 
(mg/m2day) 
Uranium Metalb 
(mg/m2day) 
Uranium Metalc 
(mg/m2day) 
5 Uranium Metalb 
(mg/m2day) 
25 Uranium Metalb 
(mg/m2day) 
25 1.3  104 2.8  103 6.2  103 1.4  104 6.9  104 
75 1.8  105 1.3  105 2.6  105 6.5  105 3.3  106 
NOTES: a From Gray and Einziger 1998 [DIRS 109691], Section 4.3. 
b From Equation 4 using the nominal activation energy of 66.4 kJ/mol. 
c From Equation 4 using the 1s minimum activation energy of 64.4 kJ/mol. 
Table 6-5. Summary and Comparison of N Reactor SNF Stage 1 and Stage 2 Dissolution Rates 
T(C) 
CO3
2- 
(molar) pH 
HNO3 
(molar) 
J-13 well 
water 
Measured a 
Stage 1 Rate 
(mg/m2day) 
Calculated b 
Stage 1 Rate 
(mg/m2day) 
Measured c 
Maximum Stage 2 
Rate (mg/m2day) 
25 2  10-2 5   63 79  e 
25 2  10-2 8   160 145 220 
25 2  10-4 10   50 44 100 
75 2  10-2 10   2,100 1,851 4,000 
75 2  10-4 5   200 142 23,000 
75 2  10-4 8   150 250 180,000 
25 N/A 5 1  10-5  38   e 
25 N/A 3 1  10-3  130   e 
25 2  10-3 8.5  X  d 72 13,000 
75 2  10-3 8.5  X  d 614 52,000 
NOTES: a As given in Table 1 of Gray and Einziger 1998 [DIRS 109691]. 
b Calculated using Equation 5: log rate (mg/m2day) = 8.52 + 0.347 log[CO3
2-] + 0.088 pH  
1929/TK. 
c As estimated from the cesium release plots from Figures 3 through 15 of Gray and Einziger 1998 
[DIRS 109691]. 
d These samples showed only Stage 2 dissolution behavior. 
e These samples showed only Stage 1 dissolution behavior. 
A significant fraction of the N Reactor fuel has been stored in a damaged condition under water 
at the K-basins at Hanford, and the metallic uranium has been exposed to the water environment. 
Therefore, the degradation models for unirradiated and uncorroded uranium metal derived in 
Section 6.7 of DOE Spent Nuclear Fuel Information in Support of TSPA-SR (DOE 2002 [DIRS 

DSNF and Other Waste Form Degradation Abstraction 
ANL-WIS-MD-000004 REV 04 6-14 November 2004 
158405]) are not recommended for the N Reactor SNF that forms the basis for Group 7. 
Although the PNNL experimental data is limited, it indicates that Stage 2 dissolution kinetics 
should be used to predict the repository performance of the N Reactor SNF. Based on these 
analyses, the abstracted best-estimate degradation rate model for the uranium metal SNF is 
selected so as to encompass the highest directly observed dissolution rates based on cesium 
release in the PNNL studies of 1.8  105 mg/m2day (0.75 mg/cm2hr) at 75C and 
13,000 mg/m2day (0.05 mg/cm2hr) at 25C (Gray and Einziger 1998 [DIRS 109691], Section 
4.3; DTN: MO0210SPAOXIDA.001 [DIRS 160443]). The samples of N Reactor SNF tested in 
the flow-through testing at PNNL were from an undamaged and uncorroded portion of an N 
Reactor fuel element. No dissolution testing has been conducted on samples of damaged or 
corroded N Reactor SNF, although this condition would presumably be the condition of exposure 
after breach of the waste package in the repository. In other studies at PNNL concerning the 
ignition properties (Abrefah et al. 1999 [DIRS 151226], Table S.1) and oxidation kinetics 
(Abrefah et al. 1998 [DIRS 151126], Summary) of N Reactor SNF, and the dissolution behavior 
of unirradiated N Reactor fuel (Abrefah et al. 2000 [DIRS 153644]), samples of N Reactor SNF 
taken from the damaged and corroded areas of the N Reactor fuel elements showed higher 
oxidation rates than undamaged and uncorroded samples. 
Table 6-4 compares values for Stage 2 dissolution rates of N Reactor SNF obtained 
experimentally by Gray and Einziger (1998 [DIRS 109691]) with rates calculated from the 
correlation for water dissolution of uranium metal given by Equation 4, and 5 times and 25 times 
this value. Note that the measured reaction rates given for the N Reactor SNF are approximately 
a factor of two to five higher than those given using Equation 4. Note also that use of the lower 
(64.4 kJ/mol) one standard deviation (1s) uncertainty given in the Equation 4 expression for the 
activation energy (66.4  2.0 kJ/mol) would result in another factor of 2 to 3 in the reaction rate 
in this temperature range. Thus, the combination of these two factors would result in 
approximately a factor of 5 increase on the nominal Equation 4 reaction rate, which would 
represent the most defensible best-estimate model kinetics. 
In the PNNL N Reactor SNF experimental work (Gray and Einziger 1998 [DIRS 109691], 
Section 4.3; DTN: MO0210SPAOXIDA.001 [DIRS 160443]), the highest dissolution rate 
inferred from cesium releases and corrected for the cesium that was found in the corrosion 
product was 290,000 mg/m2day (1.21 mg/cm2hr). This is a factor of 1.6 higher than the 
maximum value of 180,000 mg/m2day observed for dissolution rates based solely on cesium 
releases to solution. Combining this inferred rate for the N Reactor SNF and a 2s uncertainty 
(95%) lower activation energy of 62.4 kJ/mol would result in a higher dissolution rate by factor 
of 5 over the best-estimate dissolution rate for use in the conservative model. 
From the above discussion, the dissolution rate for the best-estimate model for N Reactor SNF 
and the much smaller quantity of other uranium metal-based DSNF should be taken as 5 times, 
and a conservative model be taken as 25 times, the model represented by Equation 4 (DOE 2000 
[DIRS 152658], Section 2.2; DOE 2002 [DIRS 158405], Section 6.7). Input parameter values 
used in these models are shown in Table 6-6. 

DSNF and Other Waste Form Degradation Abstraction 
ANL-WIS-MD-000004 REV 04 6-15 November 2004 
Table 6-6. Input Parameters 
Parameter Name Parameter Source DTN 
Parameter 
Value Units 
Maximum N Reactor SNF 
dissolution rate observed at 75C 
Gray and Einziger 1998 
[DIRS 109691] 
MO0210SPAOXIDA.001 
[DIRS 160443] 
1.8  105 mg/m2day 
Maximum N Reactor SNF 
dissolution rate observed at 25C 
Gray and Einziger 1998 
[DIRS 109691] 
MO0210SPAOXIDA.001 
[DIRS 160443] 
1.3  104 mg/m2day 
Arrhenius activation energy for the 
oxic aqueous dissolution of uranium 
metal 
DOE 2002 [DIRS 158405], 
Section 6.7 
MO0210SPAOXIDA.001 
[DIRS 160443] 
66.4 kJ/mol 
Effective geometric specific surface 
area and volume (in a waste 
package) for N Reactor SNF 
DOE 2002 [DIRS 158405], 
Appendices B and C 
N/A 7.0  10-5 
1.0 
m2/g 
m3/pkg 
The Arrhenius temperature dependence of the corrosion of uranium metal was the primary 
information used in developing and corroborating this model. This data set 
(DTN: MO0210SPAOXIDA.001 [DIRS 160443]) was developed by the NSNFP, and bounding 
values for the dissolution rate of N Reactor SNF were developed at PNNL. 
The corroborative information consists of: 
1. Data from Tyfield (1988 [DIRS 159597]) on the corrosion of irradiated uranium 
metal-based Magnox fuel 
2. Data from Kaminski (2001 [DIRS 161679]) on the 90C batch test corrosion of 
unirradiated uranium metal in EJ-13 water under anoxic conditions 
3. Activation energies for the moist air oxidation and moist helium oxidation of 
N Reactor SNF from Abrefah, Buchanan, and Marschman (1998 [DIRS 159159], 
Section 3.2) and Abrefah and Sell (1999 [DIRS 118528], Section 4) 
4. The results of dissolution testing on unirradiated N Reactor fuel samples by 
Abrefah et al. (2000 [DIRS 153644]) 
5. Qualitative observations and examinations of uranium hydride inclusions under the 
corroded exposed uranium surfaces of the N Reactor SNF (Abrefah et al. 1996 
[DIRS 162037], Section 3.4). 
Table 6-7 provides a comparison between the best-estimate model and the models by Tyfield 
(1988 [DIRS 159597]) and Kaminski (2001 [DIRS 161679]). The Tyfield and best-estimate 
values are similar at both 30C and 90C. Both these correlations relate to irradiated uranium 
metal dissolution. The Kaminski (2001 [DIRS 161679]) value, which applies to anoxic 
conditions, is lower by about an order of magnitude. 

DSNF and Other Waste Form Degradation Abstraction 
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Table 6-7. Comparison of Irradiated and Unirradiated Uranium Metal SNF Dissolution Rates 
Degradation Rate 
mgU/cm2hr 
Tyfield 1988 
[DIRS 159597]a 
Kaminski 2001 
[DIRS 161679]b 
Conservative 
Model c 
Best-Estimate 
Model c 
30C, pH 7, fluoride-free 0.045  0.18 0.037 
30C, pH 11.5, fluoride-free 0.016  0.18 0.037 
30C, pH 7, fluoride-dosed 0.081    
30C, pH 11.5, fluoride-dosed 0.055    
90C, pH~6, EJ-13 water 2.31 0.34 35.1 7.02 
NOTES: a These data points are approximated from Figure 6 in Tyfield 1988 [DIRS 159597]. The degradation 
rate values in that figure are given in units of mgU/m2hr and are converted to mgU/cm2hr by 
multiplying by 10-4. The 90C (363 K) value is extrapolated from the 30C (303 K) pH 7 value using 
the reported Arrhenius activation energy of 60 kJ/mol as follows: k2/k1 = exp{(-60,000  
J/mol/R)(1/T2-1/T1)} = exp{(-60,000 J/mol/8.314 J/mol-K)(1/363 -1/303)} = 51.25 so, k2 = 0.045  
51.25 = 2.31 mgU/cm2hr. 
b The 90C pH ~6 data point is obtained from the uranium corrosion rate of 1.9 mg U/day and sample 
surface area of 23.2 mm2 (0.232 cm2) given in Kaminski (2001 [DIRS 161679], pp. 13 and 14). The 
normalized rate therefore is (1.9 mg U/day)(1/24 day/hr)/(0.232 cm2) = 0.34 mg U/cm2hr. 
c Table 6-2, Group 7 models. 
It has been observed in other analyses of N Reactor SNF that samples having corroded surfaces 
(similar to the uranium surfaces that would be in contact with an air/water environment typical of 
repository conditions) exhibit enhanced chemical reactivity-related properties exceeding those of 
irradiated, but uncorroded, samples (Abrefah et al. 1999 [DIRS 151226]; Abrefah et al. 2000 
[DIRS 153644]). Enhanced chemical effects such as these are generally attributed in the 
literature to the formation of uranium hydride precipitates in the uranium metal during corrosion. 
The presence of these hydrides has been shown to significantly enhance such aspects of chemical 
reactivity as oxidation rate and pyrophoricity (e.g., ignition temperature) of corroded uranium 
surfaces (Abrefah et al. 1999 [DIRS 151226]; Totemeier et al. 1999 [DIRS 159408]). 
Metallographic examinations of the corroded uranium metal areas of N Reactor SNF have 
revealed significant formations of uranium hydride precipitates in the metallic uranium matrix 
(Abrefah et al. 1996 [DIRS 162037], Section 3.4.1; Abrefah et al. 1999 [DIRS 151226], 
Section 2.4; Goldberg 2000 [DIRS 159482]). These precipitates were not mentioned in the 
samples of N Reactor SNF used in the dissolution tests by Gray and Einziger (1998 
[DIRS 109691]), which were taken from an area of an N Reactor SNF element far removed from 
the corroded end of the element. The presence of uranium hydride in the N Reactor SNF 
samples suggests that the reaction proceeds through uranium hydride and may explain the slow 
reaction in Stage 1, since the formation of uranium hydride is known to have an induction period 
that would be consistent with these results. 
The studies by Tyfield (1988 [DIRS 159597]), Abrefah et al. (2000 [DIRS 153644]), and 
Kaminski (2001 [DIRS 161679]) used herein did not include tests with samples of N Reactor 
SNF with highly corroded metallic uranium surfaces, and only the Tyfield (1988 
[DIRS 159597]) study included the effect of pH on the dissolution of DSNF. Abrefah et al. 
(2000 [DIRS 153644]) testing on unirradiated N Reactor fuel samples included tests at pH values 
from pH 3 to pH 8, and various water chemistry conditions (Abrefah et al. 2000 [DIRS 153644], 
Table 1), but the water chemistry effect was minor compared to the SNF results of Gray and 
Einziger (1998 [DIRS 109691]). 

DSNF and Other Waste Form Degradation Abstraction 
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The best-estimate N Reactor SNF release-rate model (as adjusted for the relative density of the 
particular SNF with respect to that of metallic uranium) is: 
Rate in units of mg/cm2hr = 2.52  1010 exp(-66,400/RT)(.matrix/.U-metal) 
Rate in units of mg/m2day = 6.05  1015 exp(-66,400/RT)(.matrix/.U-metal) (Eq. 6) 
where 
R = the gas constant (8.314 J/mol-K) 
.matrix = the density of the SNF matrix 
.U-metal = the density of uranium metal 
Best-estimate calculation of the overall release of radionuclides from a waste package containing 
N Reactor SNF (i.e., multicanister overpacks), should use the specific surface area for N Reactor 
SNF (7.0  10-5 m2/g). In TSPA-LA analyses that do not differentiate between codisposal waste 
packages containing N Reactor SNF and other DSNF types, the waste form specific area of the 
N Reactor SNF (7.0  10-5 m2/g) should be used. The volume of DSNF in a codisposal waste 
package should be taken to be the initial volume of N Reactor SNF, or approximately 1.0 m3/pkg 
(DOE 2002 [DIRS 158405], Appendix C, Section 5.2). 
6.1.8 DSNF Group 8 (Uranium Oxide SNF) 
TSPA Group 8 consists of uranium dioxide-based SNF removed from commercial light water 
reactors or similar SNF from test reactors. About half of the total inventory of approximately 
162 MTHM of Group 8 DSNF comes from the Three Mile Island-2 reactor and, therefore, is not 
in the form of intact fuel rods or assemblies. The other half is substantially undamaged SNF 
from other commercial reactors. A model (Equation 7) for the dissolution kinetics of light water 
reactor SNF under neutral-to-alkaline conditions (pH >7) has been developed in CSNF Waste 
Form Degradation: Summary Abstraction (CRWMS M&O 2000 [DIRS 136060] Equation 11, 
Table 14) based on a functional form of the model originally proposed in Section 6.3.1.3.2 of 
Total System Performance Assessment-Viability Assessment (TSPA-VA) Analyses Technical 
Basis Document (CRWMS M&O 1998 [DIRS 100362]). The model was derived from qualified 
flow-through test data for the dissolution kinetics of sintered UO2 SNF over the pH range 7 
to 12. A model (Equation 8) for the dissolution of light water reactor SNF under acidic 
conditions (pH = 7) was also developed (CRWMS M&O 2000 [DIRS 136060], Equation 18), but 
from a much more limited data set than the alkaline model. 

DSNF and Other Waste Form Degradation Abstraction 
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For pH >7: 
log10 R = 5.479057 -2457.050662(1/T) - 1.510878log10([CO3
2-]) + 1.729906 
log10(O2) + 0.234718(pH) - 0.799526log10(BU) - 
400.755947{log10(O2)}{1/T} + 780.806133{log10(BU)}{1/T} + 
0.172305{log10(BU)}{-log10([CO3
2-])} + 
0.174428{log10(BU)}{-log10(O2)} - 0.271203{log10(BU)}{pH} - 
0.339535 {-log10([CO3
2-])}2 (Eq. 7) 
For pH =7: 
log10(R) = 7.13  1,085/T  0.32 (O2)  0.41 pH (Eq. 8) 
where 
R is the dissolution rate in mg/m2day, 
T is the absolute temperature, 
[CO3
2-] is the molar concentration of carbonate ion in the liquid phase, 
BU is the SNF burnup in megawatt days (MWd)/kgU, and 
O2 is the oxygen partial pressure in atmospheres in the gas phase. 
Section 6.8 of DOE Spent Nuclear Fuel Information in Support of TSPA-SR (DOE 2002 
[DIRS 158405]) suggests a surface-area enhancement factor of 100 for a release model 
representing the damaged Group 8 SNF. Thus, the abstracted best-estimate degradation model 
for intact uranium dioxide based Group 8 SNF should be the same as the degradation model for 
commercial light water reactor SNF (Equations 7 and 8) and the abstracted best-estimate model 
for the nonintact Group 8 SNF is 100 the intact fuel best-estimate model. 
6.1.9 DSNF Group 9 (Aluminum-based SNF) 
This group consists of fuels based on a uranium aluminide, uranium silicide, or uranium oxide 
particle phase dispersed in a continuous aluminum alloy matrix. The fission product 
radionuclides generally remain in the dispersed phase; therefore, the dissolution of the dispersed 
phase material is the parameter most germane to the release of the radionuclides upon contact 
with groundwater. The dissolution rate of interest is expressed in terms of mgU/m2day or 
mg metal/m2day. 
Approximately 36% of this DSNF is from foreign research reactor sources, with the balance 
from domestic research reactors such as the Advanced Test Reactor (DOE 2002 [DIRS 158405], 
Section 5.3.1.9). The Savannah River Site has conducted dissolution studies on Savannah River 
Site reactor UAlx, UAl, U3O8, and U3Si2 dispersion SNF samples at 25C in J-13 well water and 
bicarbonate solutions (Wiersma and Mickalonis 1998 [DIRS 107984], Table 3) and on 
unirradiated samples at 25C and 90C (Wiersma and Mickalonis 1998 [DIRS 107984], 
Table 4). Flow-through dissolution tests at 25C subsequently performed at PNNL on similar 
samples gave results similar to the Savannah River Site study (Gray 2000 [DIRS 151876], 
Table 7). The Savannah River Site data showed dissolution rates for the uranium in the 
irradiated SNF of 0.14 to 0.22 mgU/m2day in J-13 well water and 22 to 36 mgU/m2day in 
bicarbonate solution. 

DSNF and Other Waste Form Degradation Abstraction 
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The NSNFP has conducted an analysis (DOE 2000 [DIRS 152658], Sections 4.2.1 and 4.3.2) of 
the available literature concerning the corrosion rates in water of aluminum alloys and 
unirradiated UAlx fuel typical of the aluminum-based SNF. As a result of this analysis, the 
NSNFP recommended the following correlations for the aqueous corrosion of the aluminum 
alloys and UAlx: 
Aluminum alloy: rate (mg metal/cm2hr) = 4.29 exp[(-32,800  1,800 J/mol)/RT] 
or rate(mg metal /m2day) = 1.03  106 exp[(-32,800  1,800 J/mol)/RT] (25-360C) 
(Eq. 9) 
UAlx: rate (mg metal/cm2hr) = 5.20 exp[(-34,900  2,300 J/mol)/RT] 
or rate (mg metal/m2day) = 1.25  106 exp[(-34,900  2,300 J/mol)/RT] (178-350C) 
(Eq. 10) 
where R = 8.314 J/mol/K and T is in kelvins. 
Table 6-8 compares the uranium dissolution results obtained experimentally for irradiated and 
unirradiated Savannah River Site fuels with the results using the NSNFP correlations. For the 
purpose of this comparison, the UAlx correlation, Equation 10, is extrapolated to lower 
temperatures than the data from which it was derived (DOE 2000 [DIRS 152658], Figure 4-5). 
This can be done because the aluminum alloy correlation, Equation 9, shows the same log-linear 
relationship with temperature over a greater temperature range (DOE 2000 [DIRS 152658], 
Figure 4-4), and the oxidation rates of both the aluminum alloy and the UAlx is considered to be 
controlled by the oxidation of the aluminum metal matrix (DOE 2000 [DIRS 152658], 
Section 4.1). The NSNFP correlations predict an increase in the uranium dissolution rate at 90C 
of approximately a factor of 10 over the rate at 25C, and this is supported by the data for the 
unirradiated uranium/aluminum alloy fuel samples in J-13 water in the Savannah River Site 
study. 
The best-estimate dissolution model for the aluminum-based SNF is based on values of 
0.19 mgU/m2day at 25C and ten times that dissolution rate at 90C for a nominal uranium 
content of 13.8% (Wiersma and Mickalonis 1998 [DIRS 107984], Tables 1 and 3). These values 
translate to a total matrix dissolution rate of 1.38 mg metal/m2day at 25C and 
13.80 mg metal/m2day at 90C. This model is primarily based on the Savannah River Site test 
results for irradiated UAlx fuel in J-13 water because (1) when normalized to uranium content the 
UAlx fuel at 25C dissolved more rapidly than the U-Al, U3Si2, or U3O8 dispersion fuel 
(Wiersma and Mickalonis 1998 [DIRS 107984], Table 3), (2) the groundwater chemistry at the 
time of waste-package failure is expected to be approximately that of the J-13 well water, rather 
than the other very low and high pH conditions tested, and (3) the factor of 10 enhancement in 
the dissolution rate at 90C is supported by both the unirradiated uranium-aluminum Savannah 
River Site test results and the NSNFP UAlx dissolution correlations. In implementing the model, 
the dissolution rate for a single component should be taken as the total matrix rate multiplied by 
the inventory fraction of the component. 

DSNF and Other Waste Form Degradation Abstraction 
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Table 6-8. Comparison of Measured versus Calculated Uranium Dissolution Rates for Aluminum-Based 
SNF (in units of mg/m2day) 
25C 90C 
Fuel Type 
Uranium 
Content 
Fraction 
Calculated J-13 Low 
pH 
High 
pH 
Calculated J-13 Low pH High 
pH 
UAlx (irradiated) 0.138a 0.13 b 0.19c 28c 22c 1.6b    
U-Al (irradiated) 0.144a 0.27d 0.17c 99c 33c 2.8d    
U-Al (unirradiated) 0.132e 0.24d 1.7b 139b  2.6d 24.2e 215e 86e 
U3O8 dispersion 0.134a  0.14c 31c 33c     
U3Si2 dispersion 0.449a  0.22c 36c 36c     
NOTES: a From Wiersma and Mickalonis 1998 [DIRS 107984], Table 1. 
b Using Equation 10 with the uranium fraction multiplier. 
c From Wiersma and Mickalonis 1998 [DIRS 107984], Table 3. 
d Using Equation 9 with the uranium fraction multiplier. 
e From Wiersma and Mickalonis 1998 [DIRS 107984], Table 4. 
The data for the dissolution of UAlx in bicarbonate solution documented by Wiersma and 
Mickalonis (1998 [DIRS 107984], Table 3), and the factor of 10 increase between 25C and 
90C should be used as the basis for a conservative dissolution model. These values are 22 
mgU/m2day at 25C and 220 mgU/m2day at 90C for a nominal uranium content of 13.8% 
(Wiersma and Mickalonis 1998 [DIRS 107984], Tables 1 and 3). These values translate to a 
total matrix dissolution rate of about 160 mg metal/m2day at 25C and 1,600 mg metal/m2day at 
90C, with linear interpolation of these values for temperatures between 25C and 90C. 
6.1.10 DSNF Group 10 (Miscellaneous SNF) 
Group 10 SNF consists of a small amount of uranium nitride SNF and fuel with unknown 
matrices (DOE 1999 [DIRS 107790], Section 6.10; DOE 2002 [DIRS 158405], Section 6.10). 
This group consists of only about 0.2% of the total inventory of DSNF in MTHM. Because of 
the unknown fuel matrix material, there are no degradation data or analyses specific to this SNF 
upon which to base a degradation model, and because of the small inventory the impact of the 
dissolution kinetics on the overall TSPA-LA model is very low. In the absence of information 
concerning this DSNF type, the dissolution kinetics of Group 10 DSNF is likely to be 
conservatively represented by the corrosion rate of unirradiated uranium metal. Therefore, the 
best-estimate and conservative models for this group should be taken as the same as for the 
Group 2 SNF. 
6.1.11 DSNF Group 11 (Uranium-Zirconium Hydride) 
Test reactor fuel from Training Research IsotopesGeneral Atomic comprises approximately 
97% of the total Group 11 SNF inventory of approximately 1.5 MTHM (DOE 2002 
[DIRS 158405], Section 5.3.1.11 and Appendix D). The Training Research IsotopesGeneral 
Atomic fuel consists of a dispersion of fine particles of metallic uranium dispersed in a 
zirconium hydride matrix (U-Zr-Hx). The fuel matrix is clad in aluminum, stainless steel, or 
Alloy 800 (LMITCO 1997 [DIRS 125091], Table 2-1). Uranium loadings varied from 
approximately 8% to 45% by weight (LMITCO 1997 [DIRS 125091], Section 1). 

DSNF and Other Waste Form Degradation Abstraction 
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Unirradiated uranium-zirconium hydride fuel has been shown to have good elevated temperature 
corrosion resistance, but there is no known qualified data for the dissolution rate of this material 
in repository-relevant water and temperature conditions. However, Uranium-Zirconium Hydride 
Fuels for TRIGA Reactors (LMITCO 1997 [DIRS 125091], p. 2-5) reports a corrosion 
penetration depth of 2 mils (0.05 mm) for the zirconium hydride matrix after 400 hours in 
pressurized water at 570F (300C). This corresponds to a penetration rate of approximately 
1 mm/yr at 300C. When multiplied by the density of uranium-zirconium hydride of 
approximately 6.0 g/cm3 (LMITCO 1997 [DIRS 125091], Section 2.1.5), a penetration rate 
of 1 mm/yr corresponds to a mass corrosion rate of approximately 16,400 mg/m2day. In 
comparison, the Metals Handbook (ASM International 1987 [DIRS 103753], section on the 
Corrosion of Zirconium and Hafnium, Figure 1) gives an oxygen weight-gain corrosion rate 
for zirconium metal in pressurized water or steam at 315C of about 2 g(oxygen)/m2yr. 
Accounting for the ZrO2 stoichiometry of the corrosion product and using 91.2 as the atomic 
weight of zirconium, the Metals Handbook corrosion rate converts to [2  (91/16)] or 
5.7 g(Zr)/m2yr or approximately 14.5 mg(Zr)/m2day. This would indicate that the corrosion 
rate of the uranium-zirconium hydride is significantly greater than that of pure zirconium. 
By way of further qualitative comparison, the corrosion rate of zirconium alloys in various 
chemical media (such as sodium hydroxide, potassium hydroxide, nitric acid, etc.) at 
temperatures above approximately 100C is reported in the Metals Handbook 
(ASM International 1987 [DIRS 103753], section on the Corrosion of Zirconium and 
Hafnium, Table 6) to be generally less than 0.025 mm/yr. A penetration rate of 0.025 mm/yr 
converts to a corrosion rate of approximately 450 mg/m2day. Only very aggressive media (not 
typical of the expected repository environment) such as highly concentrated acids show 
corrosion layer penetration rates greater than 0.1 mm/yr. 
The NSNFP DSNF information report (DOE 2002 [DIRS 158405], Section 6.11) recommends 
that the dissolution rate for this DSNF form should be taken as 0.1 times the uranium oxide SNF 
dissolution rate based on inspections of the fuel assemblies that had shown no observable 
corrosion or cladding breach, and the judgment that the zirconium fuel matrix should be more 
stable than uranium oxide. 
The NSNFP-recommended value of 0.1 times the dissolution rate of the Group 8 (uranium 
oxide) degradation rate is used as the best-estimate model, since it more realistically represents 
the condition of intact cladding, and because the low total inventory in MTHM of the uraniumzirconium 
hydride SNF makes the dose consequence to the reasonably maximally exposed 
individual insensitive to the degradation model for this waste form. 
The single value rate of 16,400 mg/m2day should be used for a conservative model for the 
Group 11 DSNF. This value is chosen because it reflects the corrosion data from Uranium- 
Zirconium Hydride Fuels for TRIGA Reactors (LMITCO 1997 [DIRS 125091]), because it may 
be expected to conservatively bound the expected temperature and chemical environment ranges 
of the repository, and because it would better represent a breached cladding condition in which 
water contacts the uranium-zirconium hydride fuel matrix. 

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6.1.12 Surface Area of DSNF Groups 
The best-estimate models for the degradation rates of the DSNF groups are generally expressed 
in terms of the weight of material released per unit surface area of SNF exposed to water per unit 
time, typically in units of mg/m2day. In order to use the best-estimate models abstracted in this 
report to calculate the actual release rate from a waste form exposed to water after failure of a 
waste package, it is, therefore, necessary to use an estimate for the exposed surface area of the 
fuel. It is not necessary to use a value for the surface area for the upper-limit model since this 
model, which is total release of the inventory, is not dependent on exposed surface area. 
The N Reactor SNF, represented in Group 7, constitutes approximately 85% by weight of the 
total DSNF inventory, and is known to have extensively damaged cladding, in terms of both 
gross damage and cracking. Appendix B of DOE Spent Nuclear Fuel Information in Support of 
TSPA-SR (DOE 2002 [DIRS 158405]) calculated the specific exposed metallic uranium surface 
area for the N Reactor SNF to be 7.0  10-5 m2/g. Appendix B of DOE Spent Nuclear Fuel 
Information in Support of TSPA-SR (DOE 2002 [DIRS 158405]) calculated the geometric 
surface area of potentially exposed fuel based on the geometric characteristics of representative 
fuels within the groups. Where the fuel group contained damaged fuel, or fuel suspected to be 
damaged, a roughness factor for the surface area of 5 was applied. The result is the estimate of 
exposed surface area per gram of fuel group given in the third column of Table 6-9, and the 
estimate of the total exposed surface area for each group in the fourth column of the same table. 
When the best-estimate degradation models for each DSNF group are used, those models should 
use the estimated surface area for the individual groups in calculating overall release rates for 
waste packages containing those groups. Due to the great preponderance of the N Reactor SNF 
in the DSNF inventorythe specific surface area of the N Reactor SNF (7.0  10-5 m2/g) should 
be used in analyses that do not differentiate between waste packages containing N Reactor SNF 
and those containing other types of DSNF. The total surface area of DSNF available for reaction 
in each TSPA-LA time step should be taken to decrease in proportion to the mass of DSNF that 
has degraded in all previous time steps. 
6.2 ABSTRACTION SUMMARY - DSNF DEGRADATION MODELS 
A summary of the selected/abstracted group-specific upper-limit, and best-estimate degradation 
models is given in Table 6-2. Table 6-3 gives a comparison of the dissolution rates calculated by 
the best-estimate models for each fuel group at 50C (328 K) and for pH 8, 0.002 molar CO3
2- in 
the contacting water and 0.20 atmospheres of oxygen in the gas phase. Also included in this 
table is the approximate inventory in MTHM of each of the DSNF groups. Table 6-9 further 
gives an example of the mass fractional degradation rate at 50C that demonstrates the high 
corrosion rates of these waste forms compared to the long time frames of the TSPA-LA 
representation. Also, as shown in Table 6-9, the degradation rates of the DSNF types (with the 
exception of groups 5, 6, 9, and 11, for which the total inventories are small compared to 
group 7) are high enough to be effectively instantaneous. 
The TSPA-LA model requires that the DSNF groups be treated as a single waste type. 
Therefore, for all 11 DSNF groups, other than the naval DSNF group, the upper-limit model to 
be used in TSPA-LA analyses is instantaneous degradation of the waste form upon exposure of 
the waste form to groundwater. This upper-limit model is chosen because most of the best

DSNF and Other Waste Form Degradation Abstraction 
ANL-WIS-MD-000004 REV 04 6-23 November 2004 
estimate models (other than the Group 7 and the naval SNF models) are currently based on 
limited and unqualified corrosion, dissolution, or oxidation data, and because, as discussed in the 
preceding paragraph, the degradation of most of the DSNF inventory is effectively instantaneous. 
The upper-limit model is also chosen in part because earlier TSPA analyses performed for the 
DSNF (Thornton 1998 [DIRS 125082]; Thornton 1998 [DIRS 107796]) have shown that the 
overall effect of the failure of DSNF-containing waste packages does not significantly contribute 
to the dose at the compliance point on the repository site boundary. 
Table 6-9. Fractional DSNF Waste Form Dissolution Rates at 50C, pH 8.5, 0.002 Molar CO3
2-, and 
0.20 Atmospheres Oxygen Calculated for Best-Estimate Models 
DSNF Group 
Best-Estimate 
Release Rate 
(mg/m2day) d 
Exposed 
Specific Surface 
Area (m2/g) a 
Exposed 
Total Surface 
Area (m2) b 
Fractional 
Corrosion 
Rate (d-1) 
Group 2 - Plutonium/Uranium Alloy 492 1.2  10-3 1.02  104 5.9  10-4 
Group 3 - Plutonium/Uranium Carbide 1.09  105 2.7  10-3 2.61  102 0.174 
Group 4 - Mixed Oxide and Plutonium Oxide 5.41 4.0  10-3 d 4.63  104 2.2  10-5 
Group 5 - Thorium/Uranium Carbide 0.025 2.2  10-2 0.54  105 5.5  10-7 
Group 6 - Thorium/Uranium Oxide 0.034 3.6  10-4 1.69  104 1.2  10-8 
Group 7 - Uranium Metal-Based 1.1  105 7.0  10-5 1.39  105 7.7  10-3 
Group 8a - Intact Uranium Oxide 4.83 4.0  10-3 6.65  105 1.9  10-5 
Group 8b - Damaged Uranium Oxide 483 4.0  10-1  0.19 
Group 9 - Aluminum-Based 0.19 at 25C 6.5  10-3 1.27  105 1.2  10-6 
Group 10Miscellaneous SNF 492 4.0  10-1 1.69  106 0.2 
Group 11Uranium-Zirconium Hydride 0.33 1.0  10-4 (c) 1.51  102 3.3  10-8 
NOTES: a DOE 2002 [DIRS 158405], Appendix B, Attachment A and pages 125-126. 
b Obtained by multiplying the specific surface area given in column 3 in m2 per gram by the inventory of the 
DSNF group given in Table 6-3; using 1 MTHM = 1,000 kg = 1,000,000 g. The total MTHM from column 3 
of Table 6-3 is approximately 2,333 MTHM or 2.33  109 g. Thus the weighted average specific surface 
area is 2.71  106 m2/ 2.33  109 g . 0.001 m2/g. 
c Conservatively taken to be the same as that of damaged uranium-oxide. 
d The tabulated rates are adjusted for density as follows rate = rategroup i  (.i/.group 7). 
Commercial light water reactor SNF will be used as the surrogate for naval SNF in repository 
performance analyses. Expected releases from naval SNF waste packages were provided in 
Performance Assessment of U.S. Department of Energy Spent Fuels in Support of Site 
Recommendation (BSC 2001 [DIRS 152059]). Because of the robust design of naval SNF, the 
radionuclide releases from naval SNF waste packages are considerably less than releases from 
commercial light water reactor SNF waste packages; accordingly, the use of this surrogate is 
conservative. Naval SNF represents approximately one-tenth of one percent of the SNF MTHM 
inventory. The information contained in Revised Source Term for Naval Spent Nuclear Fuel and 
Special Case Waste Packages (Mowbray 2000 [DIRS 152077]) is qualified data. 
Model UncertaintiesThe application of the degradation models abstracted in this document 
involves the extrapolation of data over periods of time that are orders of magnitude greater than 
the experimental test periods used to generate the data. The American Society for Testing and 
Materials recommends that uncertainties in the extrapolation of such models be minimized 
through the use of models whose mathematical forms are as mechanistic as possible 
(ASTM C1174-97 1998 [DIRS 105725], Section 24). However, it can be seen from the 
abstractions above that the lack of any directly relevant experimental dissolution/degradation 

DSNF and Other Waste Form Degradation Abstraction 
ANL-WIS-MD-000004 REV 04 6-24 November 2004 
data for many of the DSNF waste forms, and the relatively small amount of data for those that 
have been tested, makes the generation of mechanistic models problematic. Additionally, 
uncertainties in the data used to generate the modelssuch as in the surface area measurements 
used to calculate normalized dissolution rates (Gray and Einziger 1998 [DIRS 109691], 
Section 2.5)produce significant uncertainties even in the short-term application of the models. 
For this reason, and because earlier TSPA analyses (Thornton 1998 [DIRS 125082]; Thornton 
1998 [DIRS 107796]) have shown that performance of the repository is very insensitive to the 
degradation rate of the DSNF, a general upper-limit or bounding degradation model, applicable 
to all but the naval DSNF, is used in the TSPA-LA analyses. Because the instantaneous release 
model to be used in the TSPA-LA modelfor all groups except Group 1 (i.e., naval SNF) 
represents the maximum possible degradation rate of the DSNF inventory, there is no uncertainty 
associated with its use as a conservative upper-limit model. 
6.3 ALTERNATIVE CONCEPTUAL MODELS 
The individual DSNF group best-estimate submodels examined in Section 6.1 are themselves 
alternative conceptual models for the DSNF degradation. Each model could be applied 
separately in the TSPA-LA model to waste packages containing that particular type of DSNF. A 
comparison of the corrosion rates predicted from each best-estimate model at 50C is given in 
Table 6-3. These alternative models are not used in the TSPA-LA model because insufficient 
qualified data on the corrosion rates and the surface areas of the fuel in each group are available 
to justify their use. 
6.4 DESCRIPTION OF BARRIER CAPABILITY 
The DSNF waste forms provide a barrier to the release of radionuclides because the fuel matrices 
must degrade before radionuclides are made available to be transported by groundwater. The 
DNSF waste form will immobilize radionuclides until the waste form is contacted by water and 
begins to degrade. The DSNF degradation models developed in this model report bound the rate 
at which radionuclides are made available for transport as either dissolved or colloidal species. 
The primary focus of this model report is the development of an upper bound model; best 
estimate mathematical models are used to calculate the degradation rate for DSNF exposed to 
humid or aqueous exposure conditions and to show that the choice of an upper-bound model is a 
reasonable model. 
Best-estimate degradation rates indicate that the degradation rate of the N Reactor DSNF, which 
constitutes most DSNF inventory, is very high. On this basis, most of the DSNF inventory is 
expected to have very limited barrier capability after the fuel is exposed to the repository 
environment. This, together with the uncertainties in the best-estimate models, leads to the 
choice of the upper-limit model for use in the TSPA-LA model. Use of this model 
conservatively indicates that the DSNF does not provide a barrier to the release of radionuclides. 

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7. VALIDATION 
The technical work plan (BSC 2004 [DIRS 171583], Table 2-1) states that This model does not 
require validation or confirmatory testing because it represents essentially the instantaneous 
mobilization of all the available radionuclides from the DSNF within a waste package, and thus 
the model is a conceptual model not implemented in a mathematical model. However, the 
instantaneous release model is mathematically implemented within TSPA and, therefore, it is 
appropriate to validate the model and an exception is taken to the technical work plan (BSC 2004 
[DIRS 171583]). 
The model is validated per the method described in Section 5.3.2 (c)(2) of AP-SIII.10Q, Models. 
This validation method consists of corroborating model results with alternative mathematical 
models. The mathematical models developed in Section 6.1, which result in the fractional 
dissolution rates in Table 6-9, show that Group 7 uranium metal-based fuels are expected to be 
completely degraded in less than one year of exposure. Group 7 comprises approximately 85% 
of the total DSNF MTHM and is, thus, representative of the DSNF. Because one year is an 
insignificant timeframe relative to geologic timeframes, the instantaneous release model is 
bounding, yet reasonable compared to the alternative mathematical model and is, therefore, 
validated through corroboration. 
Because the technical work plan (BSC 2004 [DIRS 171583]) did not require validation of the 
model, it also does not provide for an acceptability criterion. For the other Level 1 models, the 
technical work plan (BSC 2004 [DIRS 171583]) specified a validation metric of corroborating 
data must match qualitatively and must be bounded by model predictions. Therefore, the 
validation criterion is formed by substituting comparison to alternative mathematical models 
for corroborating data in the technical work plan (BSC 2004 [DIRS 171583]) criterion for 
Level 1 models. This results in the applicable criterion: alternative mathematical models must 
match qualitatively and must be bounded by model predictions. As shown above, this criterion 
is met and the model is considered to be fully validated. 

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8. CONCLUSIONS 
8.1 RECOMMENDED DSNF MODEL FOR THE TSPA-LA MODEL 
This document assessed the data and models that are available for modeling the degradation rates 
and associated radionuclide release rates from DSNF. Based on this assessment it is concluded 
that a conservative bounding approach should be implemented in the TSPA-LA model. The 
recommended model is an upper-limit model for each DSNF group of instantaneous degradation 
and complete release of radionuclides upon exposure to groundwater. For Group 7 (metallic 
uranium SNF), which represents most of the DSNF inventory, the degradation and radionuclide 
release rates are effectively instantaneous (i.e., occur within a TSPA time step). 
Based on the discussion in Section 6.1.12, the volume of DSNF in a codisposal waste package is 
taken to be the initial volume of N Reactor SNF, or approximately 1.0 m3/pkg (DOE 2002 
[DIRS 158405], Appendix C, Section 5.2). 
The outputs from this report are summarized in Table 8-1. 
Naval SNF represents approximately one-tenth of one percent of the SNF MTHM inventory. 
Because of its robust design, the radionuclide releases from naval SNF waste packages would be 
considerably less than releases from commercial light water reactor SNF waste packages. For 
this reason commercial light water reactor SNF is used as a conservative surrogate for naval SNF 
in repository performance analyses. 
Table 8-1. Summary of Outputs 
Output Description 
Degradation and Radionuclide 
Release Model for DSNF Groups 2-11 
An upper-limit model is the instantaneous release of radionuclides upon 
exposure to groundwater is used in the TSPA-LA model for each DSNF 
group, and for the DSNF inventory as a whole. 
The volume of DSNF in a codisposal 
waste package 
The volume of DSNF in a codisposal waste package is taken to be the 
initial volume of N Reactor SNF, or approximately 1.0 m3/pkg (DOE 2002 
[DIRS 158405], Appendix C Section 5.2). 
8.2 MODEL AND PARAMETER UNCERTAINTIES 
The emphasis in this document is on the basis for using an instantaneous upper-limit degradation 
modelapplicable to all but the naval DSNFfor use in the TSPA-LA model analyses. Because 
the instantaneous release of the DSNF inventory used in the TSPA-LA model for all groups 
except Group 1 (i.e., naval SNF) represents the maximum possible rate there is no uncertainty 
associated with the use of the instantaneous release model as a conservative upper-limit model. 
8.3 KEY TECHNICAL ISSUE RESOLUTION AND ACCEPTANCE CRITERIA 
The NRC Acceptance Criteria (NRC 2003 [DIRS 163274]) applicable to this model report are 
identified in Table 3-1 of Technical Work Plan for: Regulatory Integration Modeling and 
Analysis of the Waste Form and Waste Package (BSC 2004 [DIRS 171583]). For each 
applicable criterion, the criterion is quoted in italics, followed by pointers to where within the 
model report information addressing the criterion can be found. In some cases, the criterion is 

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only partially addressed in this report. A demonstration of full compliance requires a review of 
multiple reports. 
8.3.1 System Description and Demonstration of Multiple Barriers 
The acceptance criteria in this section are found in Section 2.2.1.1.3 of Yucca Mountain Review 
Plan, Final Report (NRC 2003 [DIRS 163274]). 
Acceptance Criterion 1Identification of Barriers Is Adequate. 
Barriers relied on to achieve compliance with 10 CFR 63.113(b), as 
demonstrated in the total system performance assessment, are adequately 
identified, and are clearly linked to their capability. 
As described in Sections 6.4 and 8.1, the TSPA-LA DSNF model conservatively 
indicates that the DSNF (with the exception of naval SNF) provides no barrier function in 
limiting radionuclide release. The basis for modeling the naval SNF is provided 
elsewhere. 
Acceptance Criterion 2Description of Barrier 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; and 
(4) The described capabilities are consistent with the definition of a barrier at 10 
CFR 63.2. 
As described in Sections 6.4 and 8.1, the TSPA-LA DSNF model conservatively 
indicates that the DSNF (with the exception of naval SNF) provides no barrier function in 
limiting radionuclide release. The basis for modeling the naval SNF is provided 
elsewhere. 
Acceptance Criterion 3Technical Basis for Barrier Capability Is Adequately Presented. 
The technical bases are consistent with the technical basis for the performance 
assessment. The technical basis for assertions of barrier capability is 

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commensurate with the importance of each barriers capability and the 
associated uncertainties. 
The technical bases for evaluating the barrier capabilities of the DSNF groups are 
presented in Section 6. 
8.3.2 Degradation of Engineered Barriers 
The acceptance criteria in this section are found in Section 2.2.1.3.1.3 of Yucca Mountain Review 
Plan, Final Report (NRC 2003 [DIRS 163274]). 
Acceptance Criterion 1System Description and Model Integration are Adequate 
(1) TSPA adequately incorporates important design features, physical phenomena 
and couplings and uses consistent assumptions throughout the degradation of 
engineered barriers abstraction process; 
The appropriate design features and applicable physical phenomena, as well as environmental 
factors are discussed in Section 6. 
(2) Abstraction uses assumptions, technical bases, data and models that are 
appropriate and consistent with [those used] in other abstractions. 
This model report uses one assumption that is to be confirmed by the naval SNF disposition 
program. The technical bases for using an upper-limit model in the TSPA-LA model are 
described in Section 6. 
(3) The descriptions of the engineered barriers, design features, degradation 
processes, physical phenomena, and couplings that may affect the degradation of 
the engineered barriers are adequate. 
Descriptions of the relevant DSNF design features and degradation processes (Sections 6.1.1 
through 6.1.12), and the physical phenomena and their couplings that may affect the DSNF 
degradation are provided. 
(4) Initial and boundary conditions are propagated consistently throughout the 
abstraction process. 
The DSNF initial (Sections 6.1.1 through 6.1.12) and relevant environmental boundary 
conditions are described and used as a basis for selecting an upper-limit bounding (instantaneous 
degradation) model for use in the TSPA-LA model for all DSNF groups except Group 1. 
(5) Sufficient technical basis for the inclusion and exclusion of FEPs are 
provided; 
The features, events, and processes (FEPs) relevant to assessing DSNF degradation and 
radionuclide release are identified and discussed in Section 6. 

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(7) Guidance in NUREG 1297 and NUREG 1298 [re: Expert Elicitation] are 
followed. 
Not applicable to this model report because expert elicitation is not used. 
Acceptance Criterion 2Data Are Sufficient for Model Justification 
(1) Parameters used to evaluate the degradation of EBS are adequately justified; 
Input data are not needed to justify use of the instantaneous degradation model for use in 
the TSPA-LA model. The basis for modeling the naval SNF is provided elsewhere. 
(2) Sufficient data have been collected to establish initial and boundary 
conditions; 
Input data are not needed to justify use of the instantaneous degradation model for use in 
the TSPA-LA model. The basis for modeling the naval SNF is provided elsewhere. 
(3) Data on the degradation of the engineered barriers (e.g. general and localized 
corrosion, microbially induced corrosion, galvanic interactions, hydrogen 
embrittlement and phase stability) are based on laboratory measurements, sitespecific 
field measurements, industrial and/or natural analogs and tests designed 
to replicate anticipated conditions. As appropriate, sensitivity or uncertainty 
analyses are provided and are shown to be adequate. 
Data related to the various potential degradation modes for the DSNF waste form are discussed 
in detail in Section 6. 
Acceptance Criterion 3Data Uncertainty is Characterized and Propagated through the 
Model Abstraction 
(1) Models use parameter values, assumed ranges, probability distributions 
and/or bounding assumptions that are technically defensible, reasonably account 
for uncertainties and variabilities, and do not result in under-representation of 
the risk estimate. 
Not applicable to this report; a conservative upper-limit model is to be used in the TSPA-LA 
model. 
(2) Appropriate parameters, based on techniques that may include laboratory 
experiments, field measurements, and industrial analogs are used. 
Not applicable to this report; a conservative upper-limit model is to be used in the TSPA-LA 
model. 

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(3) Assumed range of values and probability distributions for parameters used in 
conceptual and process-level models are not likely to underestimate the actual 
degradation and failure of engineered barriers. 
Not applicable to this report; a conservative upper-limit model is to be used in the TSPA-LA 
model. 
(5) Where sufficient data do not exist, the definition of parameter values and 
conceptual models is based on appropriate use of other sources, such as expert 
elicitation. 
Not applicable to this report; a conservative upper-limit model is to be used in the TSPA-LA 
model. 
Acceptance Criterion 4Model Uncertainty is Characterized and Propagated Through the 
Model Abstraction 
(1) Alternative modeling approaches are considered and are consistent with 
available data and current scientific understanding. 
Alternative modeling approaches that are consistent with available data and current scientific 
understanding are considered and discussed in Section 6.3. Alternative models are screened out 
for use in the TSPA-LA model. 
(2) Consideration of conceptual model uncertainty is consistent with available 
site characterization data, laboratory experiments,  and the treatment of 
uncertainty does not result in under-estimation of the risk estimate. 
Consideration of conceptual and other uncertainties is addressed by using a conservative upperlimit 
model is to be used in the TSPA-LA model. 
(3) Alternative modeling approaches, consistent with available data and current 
scientific understanding, are used and the modeling results are evaluated using 
tests that are sensitive to the processes modeled. 
Alternative modeling approaches that are consistent with available data and current scientific 
understanding are considered and discussed in Section 6.3. Alternative models are screened out 
for use in the TSPA-LA model. 
Acceptance Criterion 5Model Abstraction Output Is Supported By Objective 
Comparisons 
(1) Models implemented in this total system performance assessment abstraction 
provide results consistent with output from detailed process-level models and or 
empirical observations (laboratory and field testing, and/or natural analogs). 
Not applicable to this report; a conservative upper-limit model is to be used in the TSPA-LA 
model. 

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(3) Evidence is sufficient to show that models will not underestimate the actual 
degradation and failure of engineered barriers. 
A conservative upper-limit (instantaneous degradation) model is to be used in the TSPA-LA 
model. 
(4) Mathematical degradation models are based on the same environmental 
parameters, material factors, assumptions and approximations shown to be 
appropriate for closely analogous applications. 
Not applicable to this report; a conservative upper-limit model is to be used in the TSPA-LA 
model. 
(5) Accepted and well documented procedures are used to construct and test the 
numerical models that simulate the EB chemical environment and degradation of 
EB; 
Not applicable to this report. 
8.3.3 Radionuclide Release Rates and Solubility Limits 
The acceptance criteria in this section are found in Section 2.2.1.3.4.3 of Yucca Mountain Review 
Plan, Final Report (NRC 2003 [DIRS 163274]). 
Acceptance Criterion 1System Description and Model Integration Are Adequate 
(2) The abstraction of radionuclide release rates and solubility limits uses 
assumptions, technical bases, data, and models that are appropriate and 
consistent with other related U.S. Department of Energy abstractions. For 
example, the assumptions used for this model abstraction are consistent with the 
abstractions of Degradation of Engineered Barriers (Section 2.2.1.3.1); 
Mechanical Disruption of Waste Packages (Section 2.2.1.3.2); 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 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 radionuclide 
release rates and solubility limits; 
A conservative upper-limit model is to be used in the TSPA-LA model to describe the 
DSNF degradation and associated radionuclide release rate for all DSNF groups other 
than Group 1. The basis for modeling the naval SNF is provided elsewhere. 
(3) The abstraction of radionuclide release rates and solubility limits provides 
sufficient, consistent design information on waste packages and engineered 
barrier systems. For example, inventory calculations and selected radionuclides 
are based on the detailed information provided on the distribution (both spatially 
and by compositional phase) of the radionuclide inventory, within the various 
types of high-level radioactive waste; 

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The DSNF degradation model reflects available information on DSNF types (Section 6). 
(4) The U.S. Department of Energy reasonably accounts for the range of 
environmental conditions expected inside breached waste packages and in the 
engineered barrier environment surrounding the waste package. For example, 
the U.S. Department of Energy should provide a description and sufficient 
technical bases for its abstraction of changes in hydrologic properties in the near 
field, caused by coupled thermal-hydrologicmechanical- chemical processes; 
Not applicable to this report; a conservative upper-limit model is to be used in the TSPA-LA 
model. 
(5) The description of process-level conceptual and mathematical models is 
sufficiently complete, with respect to thermal-hydrologic processes affecting 
radionuclide release from the emplacement drifts. For example, if the U.S. 
Department of Energy uncouples coupled processes, the demonstration that 
uncoupled model results bound predictions of fully coupled results is adequate; 
Not applicable to this report; a conservative upper-limit model is to be used in the TSPA-LA 
model. 
(6) Technical bases for inclusion of any thermal-hydrologic-mechanical-chemical 
couplings and features, events, and processes in the radionuclide release rates 
and solubility limits model abstraction are adequate. For example, technical 
bases may include activities, such as independent modeling, laboratory or field 
data, or sensitivity studies; 
Not applicable to this report; a conservative upper-limit model is to be used in the TSPA-LA 
model. 
Acceptance Criterion 2Data Are Sufficient for Model Justification 
(1) Geological, hydrological, and geochemical values used in the license 
application are adequately justified. Adequate description of how the data were 
used, interpreted, and appropriately synthesized into the parameters is provided; 
Not applicable to this report; a conservative upper-limit model is to be used in the TSPA-LA 
model. 
(2) Sufficient data have been collected on the characteristics of the natural system 
and engineered materials to establish initial and boundary conditions for 
conceptual models and simulations of thermal-hydrologic-chemical coupled 
processes. For example, sufficient data should be provided on design features, 
such as the type, quantity, and reactivity of materials, that may affect 
radionuclide release for this abstraction; 
Not applicable to this report; a conservative upper-limit model is to be used in the TSPA-LA 
model. 

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(4) The corrosion and radionuclide release testing program for high-level 
radioactive waste forms intended for disposal provides consistent, sufficient, and 
suitable data for the in-package and in-drift chemistry used in the abstraction of 
radionuclide release rates and solubility limits. For expected environmental 
conditions, the U.S. Department of Energy provides sufficient justification for the 
use of test results, not specifically collected from the Yucca Mountain site, for 
engineered barrier components, such as high-level radioactive waste forms, drip 
shield, and backfill. 
Not applicable to this report; a conservative upper-limit model is to be used in the TSPA-LA 
model. 
Acceptance Criterion 3Data Uncertainty Is Characterized and Propagated Through the 
Model Abstraction 
(1) Models use parameter values, assumed ranges, probability distributions, and 
bounding assumptions that are technically defensible, reasonably account for 
uncertainties and variabilities, and do not result in an under-representation of the 
risk estimate; 
Not applicable to this report; a conservative upper-limit model is to be used in the TSPA-LA 
model. 
(2) Parameter values, assumed ranges, probability distributions, and bounding 
assumptions used in the abstractions of radionuclide release rates and solubility 
limits in the total system performance assessment are technically defensible and 
reasonable based on data from the Yucca Mountain region, laboratory tests, and 
natural analogs. For example, parameter values, assumed ranges, probability 
distributions, and bounding assumptions adequately reflect the range of 
environmental conditions expected inside breached waste packages; 
Not applicable to this report; a conservative upper-limit model is to be used in the TSPA-LA 
model. 
(3) The U.S. Department of Energy uses reasonable or conservative ranges of 
parameters or functional relations to determine effects of coupled thermalhydrologic-
chemical processes on radionuclide release. These values are 
consistent with the initial and boundary conditions and the assumptions for the 
conceptual models and design concepts for natural and engineered barriers at the 
Yucca Mountain site. If any correlations between the input values exist, they are 
adequately established in the total system performance assessment. For example, 
estimations are based on a thermal loading and ventilation strategy; engineered 
barrier system design (including drift liner, backfill, and drip-shield); and natural 
system masses and fluxes that are consistent with those used in other 
abstractions; 
Not applicable to this report; a conservative upper-limit model is to be used in the TSPA-LA 
model. 

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(4) Uncertainty is adequately represented in parameter development for 
conceptual models, process models, and alternative conceptual models 
considered in developing the abstraction of radionuclide release rates and 
solubility limits, either through sensitivity analyses or use of bounding analyses; 
Uncertainty is addressed in Section 8.2. A conservative upper-limit model is to be used in the 
TSPA-LA model to account for uncertainties in the degradation performance of the DSNF 
inventory. As indicated in Section 8.2, there is no uncertainty associated with this model. 
(8) The U.S. Department of Energy adequately considers the uncertainties, in the 
characteristics of the natural system and engineered materials, such as the type, 
quantity, and reactivity of material, in establishing initial and boundary 
conditions for conceptual models and simulations of thermal-hydrologic-chemical 
coupled processes that affect radionuclide release; and 
Uncertainty is addressed in Section 8.2. A conservative upper-limit model is to be used in the 
TSPA-LA model to account for uncertainties in the degradation performance of the DSNF 
inventory. As indicated in Section 8.2, there is no uncertainty associated with this model. 
Acceptance Criterion 4Model Uncertainty Is Characterized and Propagated Through 
the Model Abstraction 
(1) Alternative modeling approaches of features, events, and processes are 
considered and are consistent with available data and current scientific 
understanding, and the results and limitations are appropriately considered in the 
abstraction; 
Uncertainty is addressed in Section 8.2. A conservative upper-limit model is to be used in the 
TSPA-LA model to account for uncertainties in the degradation performance of the DSNF 
inventory. As indicated in Section 8.2, there is no uncertainty associated with this model. 
(2) In considering alternative conceptual models for radionuclide release rates 
and solubility limits, the U.S. Department of Energy uses appropriate models, 
tests, and analyses that are sensitive to the processes modeled for both natural 
and engineering systems. Conceptual model uncertainties are adequately defined 
and documented, and effects on conclusions regarding performance are properly 
assessed. For example, in modeling flow and radionuclide release from the drifts, 
the U.S. Department of Energy represents significant discrete features, such as 
fault zones, separately, or demonstrates that their inclusion in the equivalent 
continuum model produces a conservative effect on calculated performance; and 
Alternative modeling approaches that are consistent with available data and current scientific 
understanding are considered and discussed in Section 6.3. It is shown that the base model is 
more appropriate than the alternative conceptual models considered. 
(3) Consideration of conceptual model uncertainty is consistent with available 
site characterization data, laboratory experiments, field measurements, natural 
analog information and process-level modeling studies; and the treatment of 

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conceptual model uncertainty does not result in an under-representation of the 
risk estimate; and 
Not applicable to this model report; a conservative upper-limit model is to be used in the TSPALA 
model. 
(4) The effects of thermal-hydrologic-chemical coupled processes that may occur 
in the natural setting, or from interactions with engineered materials, or their 
alteration products, on radionuclide release, are appropriately considered. 
Not applicable to this model report; a conservative upper-limit model is to be used in the TSPALA 
model. 
Acceptance Criterion 5Model Abstraction Output Is Supported by Objective 
Comparisons 
(1) The models implemented in this total system performance assessment 
abstraction provide results consistent with output from detailed process-level 
models and/or empirical observations (laboratory and field testings and/or 
natural analogs); 
Not applicable to this model report; a conservative upper-limit model is to be used in the TSPALA 
model. 
(3) The U.S. Department of Energy adopts well-documented procedures that have 
been accepted by the scientific community to construct and test the numerical 
models, used to simulate coupled thermal-hydrologic-chemical effects on 
radionuclide release. For example, the U.S. Department of Energy demonstrates 
that the numerical models used for high-level radioactive waste degradation and 
dissolution, and radionuclide release from the engineered barrier system, are 
adequate representations; include consideration of uncertainties; and are not 
likely to underestimate radiological exposures to the reasonably maximally 
exposed individual and releases of radionuclides into the accessible environment; 
and 
The DSNF degradation model is developed using ASTM C1174-97 [DIRS 105725]. 

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9. INPUTS AND REFERENCES 
9.1 DOCUMENTS CITED 
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National Laboratory. TIC: 246008. 
159159 Abrefah, J.; Buchanan, H.C.; and Marschman, S.C. 1998. Oxidation Rate of K-Basin 
Spent Nuclear Fuel in Moist Air. PNNL-11844. Richland, Washington: Pacific 
Northwest National Laboratory. TIC: 243092. 
151126 Abrefah, J.; Buchanan, H.C.; Gerry, W.M.; Gray, W.J.; and Marschman, S.C. 1998. 
Dry Air Oxidation Kinetics of K-Basin Spent Nuclear Fuel. PNNL-11786. Richland, 
Washington: Pacific Northwest National Laboratory. TIC: 243094. 
153644 Abrefah, J.; Gray, W.J.; and Shelton-Davis, C.V. 2000. Dissolution Rate of 
Unirradiated N-Reactor Fuel. PNNL-SA-33740. Richland, Washington: Pacific 
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T.A. 1995. K-Basin Spent Nuclear Fuel Characterization Data Report. PNL-10778. 
Richland, Washington: Pacific Northwest National Laboratory. TIC: 243197. 
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T.A. 1996. K-Basin Spent Nuclear Fuel Characterization Data Report II. PNNL- 
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T.A. 1999. Analysis of Ignition Testing on K-West Basin Fuel. PNNL-11816. 
Richland, Washington: Pacific Northwest National Laboratory. TIC: 248558. 
103753 ASM International. 1987. Corrosion. Volume 13 of Metals Handbook. 9th Edition. 
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168796 BSC 2003. Risk Information to Support Prioritization of Performance Assessment 
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171583 BSC 2004. Technical Work Plan For: Regulatory Integration Modeling and Analysis 
of the Waste Form and Waste Package. TWP-WIS-MD-000009 REV 00 ICN 01. Las 
Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20040910.0001. 
166275 Canori, G.F. and Leitner, M.M. 2003. Project Requirements Document. TER-MGRMD-
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ACC: DOC.20031222.0006. 
100362 CRWMS M&O 1998. Waste Form Degradation, Radionuclide Mobilization, and 
Transport Through the Engineered Barrier System. Chapter 6 of Total System 
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136060 CRWMS M&O 2000. CSNF Waste Form Degradation: Summary Abstraction. ANLEBS-
MD-000015 REV 00. Las Vegas, Nevada: CRWMS M&O. 
ACC: MOL.20000121.0161. 
125038 CRWMS M&O 2000. Performance Assessment Sensitivity Analyses of Selected U.S. 
Department of Energy Spent Fuels. ANL-WIS-MD-000001 REV 00. Las Vegas, 
Nevada: CRWMS M&O. ACC: MOL.20000710.0530. 
153595 CRWMS M&O 2000. Relative Contribution of Individual Radionuclides to 
Inhalation and Ingestion Dose. CAL-WIS-MD-000002 REV 00. Las Vegas, Nevada: 
CRWMS M&O. ACC: MOL.20010103.0211. 
153246 CRWMS M&O 2000. Total System Performance Assessment for the Site 
Recommendation. TDR-WIS-PA-000001 REV 00 ICN 01. Las Vegas, Nevada: 
CRWMS M&O. ACC: MOL.20001220.0045. 
107790 DOE (U.S. Department of Energy) 1999. DOE Spent Nuclear Fuel Information in 
Support of TSPA-SR. DOE/SNF/REP-0047, Rev. 0. Washington, D.C.: U.S. 
Department of Energy, Office of Environmental Management. 
ACC: DOC.20030905.0018. 
118968 DOE 2000. DOE Spent Nuclear Fuel Grouping in Support of Criticality, DBE, TSPALA. 
DOE/SNF/REP-046, Rev. 0. Idaho Falls, Idaho: U.S. Department of Energy, 
Idaho Operations Office. ACC: DOC.20030905.0021. 
158405 DOE 2002. DOE Spent Nuclear Fuel Information in Support of TSPA-SR. 
DOE/SNF/REP-047, Rev. 2. Idaho Falls, Idaho: U.S. Department of Energy, Idaho 
Operations Office. TIC: 252089. 
171539 DOE 2004. Quality Assurance Requirements and Description. DOE/RW-0333P, Rev. 
16. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive 
Waste Management. ACC: DOC.20040907.0002. 

DSNF and Other Waste Form Degradation Abstraction 
ANL-WIS-MD-000004 REV 04 9-3 November 2004 
152658 DOE 2000. Review of Oxidation Rates of DOE Spent Nuclear Fuel, Part 1: Metallic 
Fuel. DOE/SNF/REP-054, Rev. 0. Washington, D.C.: U.S. Department of Energy. 
TIC: 248978. 
154365 Freeze, G.A.; Brodsky, N.S.; and Swift, P.N. 2001. The Development of Information 
Catalogued in REV00 of the YMP FEP Database. TDR-WIS-MD-000003 REV 00 
ICN 01. Las Vegas, Nevada: Bechtel SAIC Company. ACC: MOL.20010301.0237. 
159482 Goldberg, M.M. 2000. Corrosion Testing of Spent Nuclear Fuel Performed at 
Argonne National Laboratory for Repository Acceptance. ANL/CMT/CP-102400. 
Argonne, Illinois: Argonne National Laboratory. TIC: 252837. 
151876 Gray, W.J. 2000. Dissolution Rates of Aluminum-Based Spent Fuels Relevant to 
Geologic Disposal. PNNL-11979. Richland, Washington: Pacific Northwest National 
Laboratory. TIC: 248761. 
109691 Gray, W.J. and Einziger, R.E. 1998. Initial Results from Dissolution Rate Testing of 
N-Reactor Spent Fuel Over a Range of Potential Geologic Repository Aqueous 
Conditions. DOE/SNF/REP-022, Rev. 0. Washington, D.C.: U.S. Department of 
Energy. TIC: 245380. 
161679 Kaminski, M. 2001. Batch Tests with Unirradiated Uranium Metal Fuel Program 
Report. ANL-01/33. Argonne, Illinois: Argonne National Laboratory. TIC: 252908. 
125091 LMITCO (Lockheed Martin Idaho Technologies Company) 1997. Uranium- 
Zirconium Hydride Fuels for TRIGA Reactors. UZR-28. San Diego, California: 
General Atomics. TIC: 245412. 
152077 Mowbray, G.E. 2000. Revised Source Term Data for Naval Spent Nuclear Fuel and 
Special Case Waste Packages. Letter from G.E. Mowbray (Department of the Navy) 
to Dr. J.R. Dyer (DOE/YMSCO), September 6, 2000, with attachments. 
ACC: MOL.20000911.0354. 
163274 NRC (U.S. Nuclear Regulatory Commission) 2003. Yucca Mountain Review Plan, 
Final Report. NUREG-1804, Rev. 2. Washington, D.C.: U.S. Nuclear Regulatory 
Commission, Office of Nuclear Material Safety and Safeguards. TIC: 254568. 
125082 Thornton, T.A. 1998. HPPP Issue 1; Preliminary TSPA for a Pyrophoric Event 
Involving N-Reactor SNF Waste Packages. Interoffice correspondence from T.A. 
Thornton (CRWMS M&O) to J.S. Clouet, S.S. Sareen, and D. Stahl, September 21, 
1998, LV.WP.TAT.09/98-179, with attachment. ACC: MOL.19981019.0001. 
107796 Thornton, T.A. 1998. HPPP Issue 1; Preliminary TSPA for N-Reactor SNF Waste 
Packages. Interoffice correspondence from T.A. Thornton (CRWMS M&O) to J.S. 
Clouet, S.S. Sareen, and D. Stahl, June 1, 1998, LV.WP.TAT.05/98-111, with 
attachment. ACC: MOL.19980730.0142. 

DSNF and Other Waste Form Degradation Abstraction 
ANL-WIS-MD-000004 REV 04 9-4 November 2004 
159408 Totemeier, T.C.; Pahl, R.G.; and Frank, S.M. 1999. Oxidation Kinetics of Hydride- 
Bearing Uranium Metal Corrosion Products. Journal of Nuclear Materials, 265, (3), 
308-320. New York, New York: Elsevier. TIC: 252838. 
159597 Tyfield, S.P. 1988. Corrosion of Reactor Grade Uranium in Aqueous Solutions 
Relevant to Storage and Transport. Nuclear Energy, 27, (2), 91-98. London, 
England: British Nuclear Energy Society. TIC: 252839. 
151225 Welsh, T.L.; Baker, R.B.; Hoppe, E.W.; Schmidt, A.J.; Abrefah, J.; Tingey, J.M.; 
Bredt, P.R.; Golcar, G.R.; and Makenas, B.J. 1997. Analysis of Sludge from Hanford 
K East Basin Canisters. HNF-SP-1201. Richland, Washington: Duke Engineering & 
Services Hanford. TIC: 248549. 
107984 Wiersma, B.J. and Mickalonis, J.I. 1998. Preliminary Report on the Dissolution Rate 
and Degradation of Aluminum Spent Nuclear Fuels in Repository Environments. 
WSRC-TR-98-00290. Aiken, South Carolina: Westinghouse Savannah River. 
TIC: 245382. 
9.2 CODES, STANDARDS, REGULATIONS, AND PROCEDURES 
156605 10 CFR 63. Energy: Disposal of High-Level Radioactive Wastes in a Geologic 
Repository at Yucca Mountain, Nevada. Readily available. 
AP-3.15Q, Rev. 4, ICN 5. Managing Technical Product Inputs. Washington, D.C.: 
U.S. Department of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20040812.0004. 
AP-SIII.10Q, Rev. 2, ICN 7. Models. Washington, D.C.: U.S. Department of Energy, 
Office of Civilian Radioactive Waste Management. ACC: DOC.20040920.0002. 
105725 ASTM C 1174-97. 1998. Standard Practice for Prediction of the Long-Term 
Behavior of Materials, Including Waste Forms, Used in Engineered Barrier Systems 
(EBS) for Geological Disposal of High-Level Radioactive Waste. West 
Conshohocken, Pennsylvania: American Society for Testing and Materials. 
TIC: 246015. 
9.3 SOURCE DATA, LISTED BY DATA TRACKING NUMBER 
160443 MO0210SPAOXIDA.001. Oxidation Rates for N-Reactor Spent Fuel and Uranium 
Metal in Oxygen, Water Vapor, and Aqueous Water. Submittal date: 10/03/2002.