ASTM C1431-99(2010)e1
(Guide)Standard Guide for Corrosion Testing of Aluminum-Based Spent Nuclear Fuel in Support of Repository Disposal
Standard Guide for Corrosion Testing of Aluminum-Based Spent Nuclear Fuel in Support of Repository Disposal
SIGNIFICANCE AND USE
Disposition of aluminum-based spent nuclear fuel will involve:
Removal from the existing storage or transfer facility,
Characterization or treatment, or both, of the fuel or the resulting waste form, or both,
Placement of the waste form in a canister,
Placement of the canister in a safe and environmentally sound interim storage facility,
Removal from the interim storage facility and transport to the repository,
placement in a waste container,
Emplacement in the repository, and
Repository closure and geologic disposal. Actions in each of these steps may significantly impact the success of any subsequent step.
Aluminum-based spent nuclear fuel and the aluminum-based waste forms display physical and chemical characteristics that differ significantly from the characteristics of commercial nuclear fuels and from high level radioactive waste glasses. For example, some are highly enriched and most have heterogeneous microstructures that include very small, uranium-rich particles. The impact of this difference on repository performance must be evaluated and understood.
The U.S. Nuclear Regulatory Commission has licensing authority over public domain transportation and repository disposal (and most of the interim dry storage) of spent nuclear fuels and high-level radioactive waste under the requirements established by 10 CFR 60, 10 CFR 71, and 10 CFR 72. These requirements outline specific information needs that should be met through test protocols, for example, those mentioned in this guide. The information developed from the tests described in this guide is not meant to be comprehensive. However, the tests discussed here will provide corrosion property data to support the following information needs.
A knowledge of the solubility, leaching, oxidation/reduction reactions, and corrosion of the waste form constituents in/by the repository environment (dry air, moist air, and repository relevant water) (see 10 CFR 60.112 and 135).
A knowledge of the eff...
SCOPE
1.1 This guide covers corrosion testing of aluminum-based spent nuclear fuel in support of geologic repository disposal (per the requirements in 10 CFR 60 and 40CFR191). The testing described in this document is designed to provide data for analysis of the chemical stability and radionuclide release behavior of aluminum-based waste forms produced from aluminum-based spent nuclear fuels. The data and analyses from the corrosion testing will support the technical basis for inclusion of aluminum-based spent nuclear fuels in the repository source term. Interim storage and transportation of the spent fuel will precede geologic disposal; therefore, reference is also made to the requirements for interim storage (per 10 CFR 72) and transportation (per 10 CFR 71). The analyses that will be based on the data developed are also necessary to support the safety analyses reports (SARs) and performance assessments (PAs) for disposal systems.
1.2 Spent nuclear fuel that is not reprocessed must be safely managed prior to transportation to, and disposal in, a geologic repository. Placement in an interim storage facility may include direct placement of the irradiated fuel or treatment of the fuel prior to placement, or both. The aluminum-based waste forms may be required to be ready for geologic disposal, or road ready, prior to placement in extended interim storage. Interim storage facilities, in the United States, handle fuel from civilian commercial power reactors, defense nuclear materials production reactors, and research reactors. The research reactors include both foreign and domestic reactors. The aluminum-based fuels in the spent fuel inventory in the United States are primarily from defense reactors and from foreign and domestic research reactors. The aluminum-based spent fuel inventory includes several different fuel forms and levels of 235U enrichment. Highly enriched fuels (235U enrichment levels >20 %) are part of this inv...
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Designation: C1431 − 99 (Reapproved 2010)
Standard Guide for
Corrosion Testing of Aluminum-Based Spent Nuclear Fuel in
Support of Repository Disposal
This standard is issued under the fixed designation C1431; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
´ NOTE—Editorial corrections were made throughout in June 2010.
1. Scope 1.3 Knowledge of the corrosion behavior of aluminum-
based spent nuclear fuels is required to ensure safety and to
1.1 This guide covers corrosion testing of aluminum-based
support licensing or other approval activities, or both, neces-
spent nuclear fuel in support of geologic repository disposal
sary for disposal in a geologic repository. The response of the
(per the requirements in 10 CFR 60 and 40CFR191). The
aluminum-based spent nuclear fuel waste form(s) to disposal
testing described in this document is designed to provide data
environments must be established for configuration-safety
for analysis of the chemical stability and radionuclide release
analyses, criticality analyses, PAs, and other analyses required
behavior of aluminum-based waste forms produced from
to assess storage, treatment, transportation, and disposal of
aluminum-based spent nuclear fuels. The data and analyses
spentnuclearfuels.Thisisparticularlyimportantforthehighly
from the corrosion testing will support the technical basis for
enriched, aluminum-based spent nuclear fuels. The test proto-
inclusion of aluminum-based spent nuclear fuels in the reposi-
cols described in this guide are designed to establish material
tory source term. Interim storage and transportation of the
response under the repository-relevant conditions.
spent fuel will precede geologic disposal; therefore, reference
is also made to the requirements for interim storage (per 10
1.4 The majority of the aluminum-based spent nuclear fuels
CFR 72) and transportation (per 10 CFR 71).The analyses that
are aluminum clad, aluminum-uranium alloys. The aluminum-
will be based on the data developed are also necessary to
uraniumalloytypicallyconsistsofuraniumaluminideparticles
support the safety analyses reports (SARs) and performance
dispersed in an aluminum matrix. Other aluminum-based fuels
assessments (PAs) for disposal systems.
include dispersions of uranium oxide, uranium silicide, or
1.2 Spent nuclear fuel that is not reprocessed must be safely uranium carbide particles in an aluminum matrix. These
managed prior to transportation to, and disposal in, a geologic particles, including the aluminides, are generally cathodic to
repository.Placementinaninterimstoragefacilitymayinclude the aluminum matrix. Selective leaching of the aluminum in
direct placement of the irradiated fuel or treatment of the fuel the exposure environment may provide a mechanism for
prior to placement, or both. The aluminum-based waste forms redistribution and relocation of the uranium-rich particles.
may be required to be ready for geologic disposal, or road Particle redistribution tendencies will depend on the nature of
ready, prior to placement in extended interim storage. Interim the aluminum corrosion processes and the size, shape, distri-
storagefacilities,intheUnitedStates,handlefuelfromcivilian bution and relative reactivity of the uranium-rich particles.
Interpretation of test data will require an understanding of the
commercial power reactors, defense nuclear materials produc-
tion reactors, and research reactors. The research reactors materialbehavior.Thisunderstandingwillenableevaluationof
the design and configuration of the waste package to ensure
include both foreign and domestic reactors. The aluminum-
based fuels in the spent fuel inventory in the United States are that unfilled regions in the waste package do not provide sites
for the relocation of the uranium-rich particles into nuclear
primarily from defense reactors and from foreign and domestic
research reactors. The aluminum-based spent fuel inventory criticalconfigurations.Testsamplesmustbeevaluated,priorto
testing, to ensure that the size and shape of the uranium-rich
includes several different fuel forms and levels of U
enrichment. Highly enriched fuels ( U enrichment levels particles in the test samples are representative of the particles
in the waste form being evaluated.
>20 %) are part of this inventory.
1.5 The use of the data obtained by the testing described in
this guide will be optimized to the extent the samples mimic
This guide is under the jurisdiction ofASTM Committee C26 on Nuclear Fuel
Cycle and is the direct responsibility of C26.13 on Spent Fuel and High Level
the condition of the waste form during actual repository
Waste.
exposure. The use of Practice C1174 is recommended for
Current edition approved June 1, 2010. Published June 2010. Originally
guidance. The selection of test samples, which may be unaged
approved in 1999. Last previous edition approved in 2005 as C1431 – 99 (2005).
DOI: 10.1520/C1431-99R10E01. or artificially aged, should ensure that the test samples and
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
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C1431 − 99 (2010)
conditions bound the waste form/repository conditions. The withitsstorageordisposalenvironment.Thecorrosionproduct
test procedures should carefully describe any artificial aging may be the result of aqueous corrosion, oxidation, reaction
treatment used in the test program and explain why that with moist air, or other types of chemical interaction.
treatment was selected.
3.2.8 interim storage installation—a facility designed to
store spent nuclear fuels for an extended period of time that
2. Referenced Documents
meets the intent of the requirements of an independent spent
2.1 ASTM Standards:
fuel storage installation or a monitored retrievable storage
C1174 Practice for Prediction of the Long-Term Behavior of
facility, as described in 10 CFR 72.
Materials, Including Waste Forms, Used in Engineered
3.2.9 melt-dilute process—a process to lower the fraction of
Barrier Systems (EBS) for Geological Disposal of High-
U in highly enriched, aluminum-based spent nuclear fuel by
Level Radioactive Waste
melting and adding depleted uranium to the waste from.
2.2 Government Documents
3.2.10 performance assessment—an analysis that identifies
10 CFR 60 US Code of Federal Regulations Title 10, Part
the processes and events that might affect a disposal system,
60, Disposal of High Level Radioactive Wastes in Geo-
examines the effects of those processes and events on the
logic Repositories
performance of the disposal system, and estimates the cumu-
10 CFR 71 US Code of Federal Regulations Title 10, Part
lative releases of radionuclides considering the associated
71, Packaging and Transport of Radioactive Materials
uncertainties caused by all significant processes and events.
10 CFR 72 US Code of Federal Regulations Title 10, Part
72, Licensing Requirements for the Independent Storage
3.2.11 safety analysis—an analysis to determine the risk to
of Spent Nuclear and High-Level Radioactive Waste
the public health and safety associated with the storage,
treatment, transportation, or disposal, or combination thereof,
3. Terminology
of aluminum-based spent nuclear fuel.
3.1 Definitions:
3.2.12 service condition test—a test of a material conducted
3.1.1 Terms used in this guide are defined in Practice
under conditions in which the values of the independent
C1174, by common usage, by Webster’s New World
variables characterizing the service environment are in the
Dictionary, or as described in 3.2, or combination thereof.
range expected in actual service.
3.2 Definitions:
3.2.1 aluminum-based spent nuclear fuel—irradiated 4. Significance and Use
nuclear fuel or target elements or assemblies, or both, that are
4.1 Disposition of aluminum-based spent nuclear fuel will
clad in aluminum or aluminum-rich alloys. The microstruc-
involve:
tures contain a continuous aluminum-rich matrix with
4.1.1 Removal from the existing storage or transfer facility,
uranium-rich particles dispersed in this matrix.
4.1.2 Characterization or treatment, or both, of the fuel or
3.2.2 aluminum-based spent nuclear fuel form or waste
the resulting waste form, or both,
form—anymetallicformproducedfromaluminum-basedspent
4.1.3 Placement of the waste form in a canister,
nuclear fuel and having a microstructure containing a continu-
4.1.4 Placement of the canister in a safe and environmen-
ous aluminum-rich matrix with uranium-rich particles dis-
tally sound interim storage facility,
persed in this matrix. This term may include the fuel itself.
4.1.5 Removal from the interim storage facility and trans-
port to the repository,
3.2.3 artificial aging—any short time treatment that is
designedtoduplicateorsimulatethematerial/propertychanges 4.1.6 placement in a waste container,
that normally occur after prolonged exposure and radioactive 4.1.7 Emplacement in the repository, and
decay.
4.1.8 Repository closure and geologic disposal. Actions in
each of these steps may significantly impact the success of any
3.2.4 attribute test—a test conducted to provide material
subsequent step.
properties that are required as input to behavior models, but
that are not themselves responses to the environment.
4.2 Aluminum-based spent nuclear fuel and the aluminum-
based waste forms display physical and chemical characteris-
3.2.5 bounding—a test, sample condition or calculation
designed to provide an evaluation of the limits to material ticsthatdiffersignificantlyfromthecharacteristicsofcommer-
cial nuclear fuels and from high level radioactive waste
behavior under relevant conditions.
glasses. For example, some are highly enriched and most have
3.2.6 characterization test—in high-level radioactive waste
heterogeneous microstructures that include very small,
management, any test conducted principally to furnish infor-
uranium-rich particles.The impact of this difference on reposi-
mation for a mechanistic understanding of alteration.
tory performance must be evaluated and understood.
3.2.7 corrosion product—an ion or compound formed dur-
4.3 The U.S. Nuclear Regulatory Commission has licensing
ing the interaction of the aluminum-based spent nuclear fuel
authority over public domain transportation and repository
disposal (and most of the interim dry storage) of spent nuclear
For referenced ASTM standards, visit the ASTM website, www.astm.org, or
fuels and high-level radioactive waste under the requirements
contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM
established by 10 CFR 60, 10 CFR 71, and 10 CFR 72. These
Standards volume information, refer to the standard’s Document Summary page on
the ASTM website. requirements outline specific information needs that should be
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C1431 − 99 (2010)
met through test protocols, for example, those mentioned in 5.1.1 The effective release rates (as solute or colloidal
this guide. The information developed from the tests described species) and dissolution rates of waste form constituents and
in this guide is not meant to be comprehensive. However, the corrosionproductsinwatercompositionsrelevanttorepository
tests discussed here will provide corrosion property data to disposal,
support the following information needs. 5.1.2 The temperature dependence of, and the effect of
radiolysis products on, waste form constituent solubility in
4.3.1 A knowledge of the solubility, leaching, oxidation/
repository relevant water compositions,
reduction reactions, and corrosion of the waste form constitu-
5.1.3 The corrosion rate or relative corrosion rates of the
ents in/by the repository environment (dry air, moist air, and
various constituents in the waste form, or both,
repository relevant water) (see 10 CFR 60.112 and 135).
5.1.4 An understanding of the effect of waste form micro-
4.3.2 Aknowledge of the effects of radiolysis and tempera-
structure (the size, shape, distribution, and volume fraction of
ture on the oxidation, corrosion, and leaching behavior (see 10
the uranium-rich particles, for example), corrosion products,
CFR 60.135).
and their formation sequence on corrosion and oxidation
4.3.3 A knowledge of the temperature dependence of the
behavior, and
solubility of waste form constituents plus oxidation and corro-
5.1.5 An understanding of the release of uranium-rich
sion products (see 10 CFR 60.135).
colloids or particles, or both, during storage and disposition.
4.3.4 Information from laboratory experiments or technical
analyses, or both, about time dependence of the internal 5.2 Tests conducted to supply the data needs described in
5.1 would ideally provide sufficient information to help estab-
condition of the waste package (see 10 CFR 60.143 and 10
CFR 72.76). lishmechanisticmodels,or,inanycase,empiricalcorrelations,
for:
4.3.5 Laboratory demonstrations of the effects of the elec-
5.2.1 Corrosion rates under the bounding or potential range
trochemical differences between the aluminum-based waste
of repository conditions,
form and the candidate packaging materials on galvanic
5.2.2 The effective solubility of waste form constituents,
corrosion (see 10 CFR 71.43) or the significance of electrical
including corrosion products, as a function of the temperature
contactbetweenthewasteformandthepackagingmaterialson
and chemistry of the water that may surround the waste form
items outlined in 4.3.1-4.3.4 (see 10 CFR 60.135), or both.
after a canister breach in the repository, and
4.3.6 Information on the risk involved in the receipt,
5.2.3 The selective leaching of the aluminum matrix from
handling, packaging, storage, and retrieval of the waste forms
the uranium-rich particles with anticipated waste package/
(see 10 CFR 72.3).
repository environments.
4.3.7 Information on the physical and chemical condition of
the waste form upon repository placement so that items 5.3 Theinformationneedsdescribedin5.1andcorrelations/
outlined in 4.3.1-4.3.4 can be evaluated (see 10 CFR 60.135). models described in 5.2 should enable the calculation of the
rateofreleaseofradionuclidesfromthealuminum-basedwaste
4.3.8 Knowledge of the degradation of the waste form
forms stored in the repository.
during interim storage so that operational safety problems with
respect to its removal from storage can be assessed, if such
5.4 The information needs described in 5.1 should provide
removal is necessary (see 10 CFR 72.123).
the necessary particle size and leach rate information for an
4.3.9 Knowledge o
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