ASTM C1682-21

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Designation: C1682 − 17 C1682 − 21
Standard Guide for
Characterization of Spent Nuclear Fuel in Support of Interim
Storage, Transportation and Geologic Repository Disposal
This standard is issued under the fixed designation C1682; 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.
1. Scope
1.1 This guide provides guidance for the types and extent of testing that would be involved in characterizing the physical and
chemical nature of spent nuclear fuel (SNF) in support of its interim storage, transport, and disposal in a geologic repository. This
guide applies primarily to commercial light water reactor (LWR) spent fuel and spent fuel from weapons production, although the
individual tests/analyses may be used as applicable to other spent fuels such as those from research and test reactors, test reactors,
molten salt reactors and mixed oxide (MOX) spent fuel. The testing is designed to provide information that supports the design,
safety analysis, and performance assessment of a geologic repository for the ultimate disposal of the SNF.
1.2 The testing described includes characterization of such physical attributes as physical appearance, weight, density,
shape/geometry, degree, and type of SNF cladding damage. The testing described also includes the measurement/examination of
such chemical attributes as radionuclide content, microstructure, and corrosion product content, and such environmental response
characteristics as drying rates, oxidation rates (in dry air, water vapor, and liquid water), ignition temperature, and dissolution/
degradation rates. Not all of the characterization tests described herein must necessarily be performed for any given analysis of
SNF performance for interim storage, transportation, or geological repository disposal, particularly in areas where an extensive
body of literature already exists for the parameter of interest in the specific service condition.
1.3 It is assumed in formulating the SNF characterization activities in this guide that the SNF has been stored in an interim storage
facility at some time between reactor discharge and dry transport to a repository. The SNF may have been stored either wet (for
example, a spent fuel pool), or dry (for example, an independent spent fuel storage installation (ISFSI)), or both, and that the
manner of interim storage may affect the SNF characteristics.
1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.
1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of
the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory
limitations prior to use.
1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility
of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of
regulatory limitations prior to use.
1.6 This international standard was developed in accordance with internationally recognized principles on standardization
established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued
by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
This guide is under the jurisdiction of ASTM Committee C26 on Nuclear Fuel Cycle and is the direct responsibility of Subcommittee C26.13 on Spent Fuel and High
Level Waste.
Current edition approved July 1, 2017Oct. 1, 2021. Published August 2017November 2021. Originally approved in 2009. Last previous edition approved in 20092017 as
C1682 – 09.C1682 – 17. DOI: 10.1520/C1682-17.10.1520/C1682-21.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
C1682 − 21
2. Referenced Documents
2.1 ASTM Standards:
C170/C170M Test Method for Compressive Strength of Dimension Stone
C696 Test Methods for Chemical, Mass Spectrometric, and Spectrochemical Analysis of Nuclear-Grade Uranium Dioxide
Powders and Pellets
C698 Test Methods for Chemical, Mass Spectrometric, and Spectrochemical Analysis of Nuclear-Grade Mixed Oxides ((U,
Pu)O )
C859 Terminology Relating to Nuclear Materials
C1174 Guide for Evaluation of Long-Term Behavior of Materials Used in Engineered Barrier Systems (EBS) for Geological
Disposal of High-Level Radioactive Waste
C1380 Test Method for the Determination of Uranium Content and Isotopic Composition by Isotope Dilution Mass
Spectrometry (Withdrawn 2018)
C1413 Test Method for Isotopic Analysis of Hydrolyzed Uranium Hexafluoride and Uranyl Nitrate Solutions by Thermal
Ionization Mass Spectrometry
C1454 Guide for Pyrophoricity/Combustibility Testing in Support of Pyrophoricity Analyses of Metallic Uranium Spent Nuclear
Fuel (Withdrawn 2016)
C1553 Guide for Drying of Spent Nuclear Fuel
E170 Terminology Relating to Radiation Measurements and Dosimetry
2.2 U.S. Government Documents
Code of Federal Regulations, Title 10, Part 60 Disposal of High-Level Radioactive Wastes in Geologic Repositories, U.S.
Nuclear Regulatory Commission
Code of Federal Regulations, Title 10, Part 63 Disposal of High-Level Radioactive Wastes in a Geologic Repository at Yucca
Mountain, Nevada, U.S. Nuclear Regulatory Commission
Code of Federal Regulations, Title 10, Part 71 Packaging and Transport of Radioactive Materials
Code of Federal Regulations, Title 10, Part 72 Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and
High-Level Radioactive Waste
Code of Federal Regulations, Title 10, Part 961 Standard contract for the Disposal of Spent Nuclear Fuel and/or High Level
Waste
Code of Federal Regulations, Title 40, Part 191 Environmental Radiation Protection Standards for Management and Disposal
of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes
Code of Federal Regulations Title 40, Part 197 Protection of Environment: Public Health and Environmental Radiation
Standards for Yucca Mountain, Nevada
3. Terminology
3.1 Definitions—Definitions used in this guide are as currently existing in Terminology C859 or Test Method C170/C170M, or as
commonly accepted in dictionaries of the English language, except for those terms defined below for the specific usage of this
standard. For consistency, many of the definitions are based on definitions from Federal Regulations in the United States.
3.2 Definitions of Terms Specific to This Standard:
3.2.1 alteration, n—any change to the form, state, or properties of a material.
3.2.2 attribute test, n—a test conducted to provide material properties that are required as input to materials behavior models, but
are not themselves responses to the materials environment (for example, thermal conductivity, mechanical properties, radionuclide
content of waste forms, etc).
3.2.3 characterization test, n—any test conducted principally to furnish information for a mechanistic understanding of alteration
(for example, electrochemical polarization tests, leach tests, solubility tests, etc).
3.2.4 combustible, adj—capable of burning or undergoing rapid chemical oxidation.
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volume information, refer to the standard’s Document Summary page on the ASTM website.
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www.access.gpo.gov.
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3.2.5 breached fuel, n—(per Code of Federal Regulations, Title 10, Part 72, Section 122(h)) any spent fuel with extreme
degradation or gross rupture, such that fuel particulates or pieces can be released from the fuel rod. (“The spent fuel cladding must
be protected during storage against degradation that leads to gross ruptures or the fuel must be otherwise confined such that
degradation of the fuel during storage will not pose operational safety problems with respect to its removal from storage,” Code
of Federal Regulations, Title 10, Part 72, Section 122(h)). It is not expected that minor cladding defects such as pinhole cracks
would permit significant release of particulate matter from the spent fuel rod.
3.2.5.1 Discussion—
“The spent fuel cladding must be protected during storage against degradation that leads to gross ruptures or the fuel must be
otherwise confined such that degradation of the fuel during storage will not pose operational safety problems with respect to its
removal from storage,” Code of Federal Regulations, Title 10, Part 72, Section 122(h). It is not expected that minor cladding
defects such as pinhole cracks would permit significant release of particulate matter from the spent fuel rod.
3.2.6 damaged fuel, n—spent nuclear fuel elements or assemblies that as a result of their irradiation or handling (or both) have
significantly altered dimensions or cladding through-wall cracks or penetrations such that it cannot fulfill its direct or indirect
regulatory or design function. For example any SNF assembly with rod(s) that are significantly displaced for purposes of criticality
calculations (application dependent and function of the stage in the nuclear fuel cycle).
3.2.6.1 Discussion—
For example any SNF assembly with rod(s) that are significantly displaced for purposes of criticality calculations (application
dependent and function of the stage in the nuclear fuel cycle).
3.2.7 degraded cladding, n—spent fuel cladding which has corroded or been physically altered in-reactor or during subsequent
interim storage (or both), to the extent that the alteration must be accounted for in the evaluation of its behavior during transport,
storage, or disposal (for example, cladding corrosion/thinning, hydride embrittlement, etc.).
3.2.8 failed fuel (geologic disposal), n—any significant alteration in the shape, dimensions, or configuration of a spent fuel
assembly or fuel element, or through-wall crack in the cladding, that could degrade or open further under long-term exposure to
the repository environment.
3.2.9 failed fuel (interim storage and transport), n—fuel rods/assemblies whose cladding has been perforated to the extent that
powder or pieces of the fuel can relocate or be released from the cladding.
3.2.9.1 Discussion—
Code of Federal Regulations, Title 10, Part 961, the Standard Contract between the USDOE and the US commercial nuclear
utilities defines categories of commercial LWR spent fuel as “Standard,” “Non-Standard,” and “Failed.” These categories are based
on the type of handling—normal or special—required for transport and storage of the SNF. The “Standard” classification includes
most normal and handle-able LWR (PWR and BWR) spent fuel. “Non-Standard” spent fuel includes non-LWR spent fuel,
consolidated fuel, older design fuel, etc. “Failed” fuel includes: Class F-1: (via visual examination) visual failure or
damage—“Assemblies which (i) are structurally deformed or have damaged cladding to the extent that special handling may be
required or (ii) for any reason cannot be handled with normal fuel handling equipment …” Class F-2: radioactive “ leakage” or
“any fuel that allows gaseous communication between the inside and the outside of the cladding.” Class F-3: Encapsulated—Note
that the terms used in this guide for failed fuel, damaged fuel, and degraded cladding can fit the “Failed Fuel” definition of Code
of Federal Regulations, Title 10, Part 961. Also, the Code of Federal Regulations, Title 10, Part 961 categories of spent fuel are
partially based on the fact that the repository is required by statute to accept all commercial LWR spent fuel, including
damaged/failed.) The classification of SNF within the country of interest may differ and the national guidelines for nomenclature
should be used. Documentation of the parallel definition given within this standard is recommended.
3.2.10 ignite, v—to cause to burn and reach a state of rapid oxidation, which is maintained without requiring an external heat
source.
3.2.11 model, n—a simplified representation of a system or phenomenon, often mathematical.
3.2.12 performance assessment (PA), n—an analysis that identifies the processes and events that might affect the disposal system;
examines the effects of these processes and events on the performance of the disposal system; and, estimates the cumulative
releases of radionuclides, considering the associated uncertainties, caused by all significant processes and events. These estimates
shall be incorporated into an overall probability distribution of cumulative release to the extent practicable (see Code of Federal
Regulations, Title 10, Part 63 Section 2) and Code of Federal Regulations, Title 40, Part 191 Section 15).
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3.2.12.1 Discussion—
These estimates shall be incorporated into an overall probability distribution of cumulative release to the extent practicable (see
Code of Federal Regulations, Title 10, Part 63 Section 2) and Code of Federal Regulations, Title 40, Part 191 Section 15).
3.2.13 pyrophoric, adj—capable of igniting spontaneously under temperature, chemical, or physical/mechanical conditions
specific to the storage, handling, or transportation environment.conditions.
3.2.14 sibling sample, n—one of two or more test samples that are nearly indistinguishable with respect to their chemical and
physical properties.
3.2.15 spent nuclear fuel (SNF), n—nuclear fuel that has been exposed to, and removed from, a nuclear reactor and not intended
to be reinserted in a reactor.
3.2.16 waste form (WF), n—(from Practice C1174) the radioactive waste materials and any encapsulating or stabilizing matrix in
which it is incorporated.
3.2.17 waste package (WP), n—(from Practice C1174) the waste form and any containers, shielding, packing and other absorbent
materials immediately surrounding an individual waste container.
4. Summary of Guide
4.1 The characterization of spent nuclear fuel (SNF)—in support of interim storage, transport, and disposal in a geologic
repository—described in this guide includes the examination/testing of such physical attributes as physical appearance, weight,
density, shape/geometry, degree and type of cladding damage, etc. It also includes the measurement/examination of such chemical
aspects as drying characteristics, water content, radionuclide content, microstructure, zirconium hydride content (of commercial
SNF cladding), uranium hydride content (of metallic uranium SNF), and such environmental response characteristics as oxidation
rate (in dry air, water vapor, and liquid water), ignition temperature, and dissolution/degradation rates.
4.2 The primary issues involved in the characterization of uranium dioxide-based commercial light water reactor (LWR) SNF are
the fraction of fuel rods with non-intact cladding (that is, the amount of “failed fuel” as defined in Section 3 above), the structural
integrity of the fuel assembly (that is, the amount of “damaged fuel” as defined in Section 3 above), the amount and structure of
zirconium hydride in the cladding (for example, “degraded cladding” as defined in Section 3 above), particularly with respect to
high burnup LWR SNF. Also, the radionuclide content of the fuel, the thickness of the zirconium oxide on the external surface of
the cladding, and the leaching/dissolution behavior characteristics when in contact with the (repository-relevant) air/water
environment are factors that could affect SNF behavior in repository disposal.
4.3 The primary issue involved in characterization of metallic uranium SNF is the extent of damage to the cladding (that is,
exposure of metallic uranium to air and water) and its consequently enhanced chemical activity and pyrophoricity/combustibility
characteristics. Metallic uranium SNF, largely from plutonium production reactors, has been temporarily stored in water basins in
several countries prior to reprocessing or ultimate direct disposal of unreprocessed fuel. SNF. In some cases the manner of
discharge (for example, those involving physical trauma to the fuel element) of the fuel elements from these reactors, and the type
of wet storage environment in which they were emplaced after discharge, has resulted in significant amounts of fuel cladding
damage and extensive corrosion of the consequently exposed uranium metal. This corrosion and damage has resulted in alteration
of the physical integrity/dimensions of the elements and the chemical reactivity of the material such that the physical and chemical
properties of the material no longer straightforwardly resemble, or can be represented by, the properties of the as-fabricated,
unirradiated fuel.
5. Significance and Use
5.1 In order to demonstrate conformance to regulatory requirements and support the post-closure repository performance
assessment information is required about the attributes, characteristics, and behavior of the SNF. These properties of the SNF in
turn support the transport, interim storage, and repository pre-closure safety analyses, and repository post-closure performance
assessment. In the United States, the interim dry storage of commercial LWR SNF is regulated per the Code of Federal
Regulations, Title 10, Part 72, which requires that the cladding must not sustain during the interim storage period any “gross”
damage sufficient to release fuel from the cladding into the container environment. In other countries, the appropriate governing
body will set regulations regarding interim dry storage of commercial LWR SNF. However, cladding damage insufficient to allow
the release of fuel during the interim storage period may still occur in the form of small cracks or pinholes. pinholes that can
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develop into much larger defects. These cracks/pinholes could be sufficient to classify the fuel as “ failed “failed fuel” or “breached
fuel” per the definitions given in Section 3 for repository disposal purposes, because they could allow contact of water vapor or
liquid with the spent fuel matrix and thus provide a pathway for radionuclide release from the waste form. Also, pinholes/cracks
in fuel rods in dry or wet interim storage can also develop into much larger defects (for example, the phenomenon of cladding
“unzipping”) under long-term repository conditions. Therefore SNF characterization should be adequate to determine the amount
of “failed fuel” for either usage as required. This could involve the examination of reactor operating records, ultrasonic testing,
sipping, and analysis of the residual water and drying kinetics of the spent fuel assemblies or canisters.
5.2 Regulations in each country may contain constraints and limitations on the chemical or physical (or both) properties and
long-term degradation behavior of the spent fuel and HLW in the repository. Evaluating the design and performance of the waste
form (WF), waste packaging (WP), and the rest of the engineered barrier system (EBS) with respect to these regulatory constraints
requires knowledge of the chemical/physical characteristics and degradation behavior of the SNF that could be provided by the
testing and data evaluation methods provided by this guide, using the United States as an example, as follows:
5.2.1 In the United States, for example, Code of Federal Regulations, Title 10, Part 60 Sections 135 and 113 require that the WF
be a material that is solid, non-particulate, non-pyrophoric, and non-chemically reactive, that the waste package contain no liquid,
particulates, or combustible materials and that the materials/components of the EBS be designed to provide—assuming anticipated
processes and events—substantially complete containment of the HLW for the NRC-designated regulatory period.
5.2.2 In the United States, for example, Code of Federal Regulations, Title 10, Part 63 Section 113 requires that the EBS be
designed such that, working in combination with the natural barriers, the performance assessment of the EBS demonstrates
conformance to the annual reasonably expected individual dose protection standard of Code of Federal Regulations, Title 10, Part
63 Section 311 and the reasonably maximally exposed individual standard of Code of Federal Regulations, Title 10, Part 63 Section
312, and shall not exceed EPA dose limits for protection of groundwater of Code of Federal Regulations, Title 10, Part 63 Section
331 during the NRC-designated regulatory compliance period after permanent closure.
5.2.3 In the United States, for example, Code of Federal Regulations, Title 10, Part 63 Section 114 (e), (f), and (g) and Code of
Federal Regulations, Title 10, Part 63 Section 115 (c) require that a technical basis be provided for the inclusion or exclusion of
degradation/alteration processes pertinent to the barriers of the EBS, and that likewise a technical basis be provided for the
degradation/alteration models used in the post-closure performance assessment of the capability of the EBS barriers to isolate
waste.
5.3 The enhanced chemical reactivity and degraded condition of corroded/damaged uranium metal-based SNF must be accounted
for in both the pre-closure safety analyses and the post-closure performance assessment of the geologic repository. An example
of this would be the potential for pyrophoric behavior in uranium metal-based SNF (see Guide C1454). Due to the combustibility
of the metallic uranium or uranium hydride (or both), and the enhanced aqueous dissolution rates for the exposed uranium metal,
the potential for enhanced chemical activity or pyrophoric behavior must be factored into the repository or interim storage facility
safety analyses, and assessments of the potential for radionuclide releases from the repository site boundary after repository
closure.
5.4 Characterization of several key properties of SNF may be required to support the design and performance analyses of both
repository above-ground SNF receipt and lag storage facilities, the WP into which the SNF is placed, and the subsurface permanent
emplacement drift EBS.
5.4.1 Repository waste package design must ensure that the waste to be placed in the repository can be accommodated within the
radionuclide and thermal loading ranges of the waste package drift emplacement licensing conditions. To do this the radionuclide
content and oxidation rate when exposed to oxygen/water environments should be determined.
5.4.2 The condition of the LWR spent fuel cladding (particularly with respect to hydride content and morphology) could
potentially influence the performance of the cladding in interim storage, transportation, and geologic repository disposal. The
corrosion and consequent failure rate of cladding with high hydride content may be greater than that of low or no hydride content.
If the performance assessment is found to be sensitive to the failure rate of the cladding, it may be necessary to perform zirconium
hydride content and orientation testing, particularly for high burnup LWR SNF.
5.4.3 Metallic uranium-based spent fuel introduces aspects of chemical reactivity, such as combustibility and pyrophoricity (see
C1454), that should be addressed in WP design and performance assessment, and in safety analyses associated with interim storage
and transportation prior to repository emplacement. Metallic uranium-based nuclear fuel has been widely used in nuclear reactors;
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sometimes for commercial reactors (for example, Magnox) but more often in plutonium and tritium production reactors. The
manner of discharge of metallic uranium SNF from these production reactors, and/or the manner of temporary wet storage of that
portion of the spent fuel that was not reprocessed has in many instances resulted in significant corrosion and mechanical damage
to the SNF assemblies. This damage has resulted in the direct exposure of the metallic uranium to the basin water. The relatively
high chemical reactivity of uranium in contact with water can result in significant physical damage to the assemblies as the result
of corrosion product buildup, and the creation in the exposed fuel surface and fuel matrix of uranium hydride inclusions which
in turn further increase the chemical activity of the material. The reaction of this spent fuel with air, water vapor, or liquid water
can introduce a significant heat source term into design basis events. In order to support the evaluation of these events, the physical
condition (that is, the degree of optically/visually observable damage), the chemical oxidation kinetics, the ignition characteristics,
and radionuclide release characteristics of the SNF should be investigated.
5.4.4 The thermal analysis of the waste package/engineered barrier system requires quantification of the potential chemical heat
source. To determine this, the amount of reactive uranium metal in the waste canisters
...