ASTM C1868-18
(Practice)Standard Practice for Ceramographic Preparation of UO2 and Mixed Oxide (U,Pu)O2 Pellets for Microstructural Analysis
Standard Practice for Ceramographic Preparation of UO<inf>2</inf> and Mixed Oxide (U,Pu)O<inf>2</inf> Pellets for Microstructural Analysis
SIGNIFICANCE AND USE
5.1 The ceramographic examination of the nuclear fuel pellet is mandatory to ensure that the microstructural characteristics are in compliance with the fuel specifications relative to performance in reactor, particularly concerning thermo-mechanical behavior and fission gas release.
5.2 This practice is applicable for sintered UO2 pellets with any 235U concentration and (U,Pu)O2 pellets containing up to 15 weight % PuO2 with less than 10 % porosity.
SCOPE
1.1 This practice describes the procedure for preparing nuclear-grade uranium dioxide (UO2) or mixed uranium-plutonium dioxide (MOX or (U,Pu)O2)), sintered and non-irradiated pellets for subsequent microstructural analysis (hereafter referred to as ceramographic examination).
1.2 The ceramographic examination is performed to confirm that the microstructure of the sintered pellet is in compliance with the fuel specification, for example as defined in Specifications C776 and C833, as a function of the initial raw material properties and manufacturing process parameters.
1.3 The microstructure of a ceramic pellet includes: grain size, porosity size and distribution, and phase distribution for (U,Pu)O2 pellets, that is, Pu-rich cluster size and distribution.2
1.4 The microstructural characteristics of the pellet are accessible after preparation which involves: sawing, mounting in a resin, surface polishing, and chemical etching, thermal etching, or both.
1.5 This practice describes the preparation processes mentioned in 1.4; it does not discuss the associated sampling practices (for example, Practice E105) or ceramographic examination methods (for example, the methods for determining average grain size are covered in Test Method E112).
1.6 Due to the radiotoxicity associated with these nuclear materials, all operations described in this practice should be performed in glovebox for (U,Pu)O2 pellets and in a hood for UO2 pellets.
1.7 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.
1.8 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.
1.9 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.
General Information
Relations
Standards Content (Sample)
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.
Designation: C1868 − 18
Standard Practice for
Ceramographic Preparation of UO and Mixed Oxide
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(U,Pu)O Pellets for Microstructural Analysis
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This standard is issued under the fixed designation C1868; 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 priate safety, health, and environmental practices and deter-
mine the applicability of regulatory limitations prior to use.
1.1 This practice describes the procedure for preparing
1.9 This international standard was developed in accor-
nuclear-grade uranium dioxide (UO ) or mixed uranium-
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dance with internationally recognized principles on standard-
plutonium dioxide (MOX or (U,Pu)O )), sintered and non-
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ization established in the Decision on Principles for the
irradiated pellets for subsequent microstructural analysis (here-
Development of International Standards, Guides and Recom-
after referred to as ceramographic examination).
mendations issued by the World Trade Organization Technical
1.2 Theceramographicexaminationisperformedtoconfirm
Barriers to Trade (TBT) Committee.
that the microstructure of the sintered pellet is in compliance
2. Referenced Documents
with the fuel specification, for example as defined in Specifi-
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cationsC776andC833,asafunctionoftheinitialrawmaterial
2.1 ASTM Standards:
properties and manufacturing process parameters.
C776 Specification for Sintered Uranium Dioxide Pellets for
1.3 The microstructure of a ceramic pellet includes: grain Light Water Reactors
C833 Specification for Sintered (Uranium-Plutonium) Diox-
size, porosity size and distribution, and phase distribution for
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(U,Pu)O pellets, that is, Pu-rich cluster size and distribution. ide Pellets for Light Water Reactors
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C859 Terminology Relating to Nuclear Materials
1.4 The microstructural characteristics of the pellet are
D1193 Specification for Reagent Water
accessible after preparation which involves: sawing, mounting
E105 Practice for Probability Sampling of Materials
in a resin, surface polishing, and chemical etching, thermal
E112 Test Methods for Determining Average Grain Size
etching, or both.
3. Terminology
1.5 This practice describes the preparation processes men-
tioned in 1.4; it does not discuss the associated sampling
3.1 Except as otherwise defined herein, definitions of terms
practices (for example, Practice E105) or ceramographic ex-
are as given in Terminology C859.
amination methods (for example, the methods for determining
3.2 Definitions of Terms Specific to This Standard:
average grain size are covered in Test Method E112).
3.2.1 grain—single crystal; region of space occupied by a
1.6 Due to the radiotoxicity associated with these nuclear
continuous crystal lattice.
materials, all operations described in this practice should be
3.2.2 microstructure—structure of a material as observed
performed in glovebox for (U,Pu)O pellets and in a hood for
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from a magnified view in the range from 0.1 to 100 µm
UO pellets.
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involving properties such as grains, grain boundaries, pores,
1.7 The values stated in SI units are to be regarded as
micro-cracks, and phases distribution of the sintered pellet.
standard. No other units of measurement are included in this
3.2.3 MOX—mixed oxide, that is, a blend of uranium and
standard.
plutonium dioxides.
1.8 This standard does not purport to address all of the
3.2.4 porosity—amount of pores (voids) in an object ex-
safety concerns, if any, associated with its use. It is the
pressed as percentage of the total volume.
responsibility of the user of this standard to establish appro-
3.2.5 sintered pellet—densified ceramic compact after heat
treatment at elevated temperatures but below the melting point
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This practice is under the jurisdiction of ASTM Committee C26 on Nuclear
of the main component.
Fuel Cycle and is the direct responsibility of Subcommittee C26.05 on Methods of
Test.
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Current edition approved Jan. 1, 2018. Published January 2018. DOI: 10.1520/ For referenced ASTM standards, visit the ASTM website, www.astm.org, or
C1868-18. contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM
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(U,Pu)O fuel pellets are characterized by fissile Pu-rich zones dispersed in a Standards volume information, refer to the standard’s Document Summary page on
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fertile depleted UO matrix. the ASTM website.
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