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ASTM C1817-15

(Test Method)

Standard Test Method for The Determination of the Oxygen to Metal (O/M) Ratio in Sintered Mixed Oxide ((U, Pu)O2) Pellets by Gravimetry

Standard Test Method for The Determination of the Oxygen to Metal (O/M) Ratio in Sintered Mixed Oxide ((U, Pu)O<inf>2</inf>) Pellets by Gravimetry

SIGNIFICANCE AND USE
5.1 MOX is used as a nuclear-reactor fuel. This test method is designed to determine whether the O/M ratio meets the requirements of the fuel specification. Examples for establishing a fuel specification are given in Specification C833.  
5.2 This method is suitable for all sintered MOX pellets containing up to 10 weight % PuO2 when the UO2 and PuO2 meet the requirements of Specifications C753 and C757.
SCOPE
1.1 This practice is an alternative method to Test Method C698 for the determination of the oxygen-to-metal atom ratio (O/M) in sintered mixed oxide fuel (MOX) pellets. The method presented in Test Method C698 is a one-step thermogravimetric method for determining O/M ratio in sintered MOX powders and pellets. As stated in Test Method C698, thermogravimetric methods using a two-step heating cycle are also satisfactory (1, 2).2 The method presented in this test method is a two-step heating cycle method. This test method is applicable to sintered MOX pellets containing up to 10 weight percent PuO2.  
1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.3 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.

General Information

Status
Historical
Publication Date
31-May-2015
ICS
27.120.30 - Fissile materials and nuclear fuel technology
Technical Committee
C26 - Nuclear Fuel Cycle
Drafting Committee
C26.05 - Methods of Test
Current Stage
Ref Project

Relations

Replaced By
ASTM C1817-16 - Standard Test Method for The Determination of the Oxygen to Metal (O/M) Ratio in Sintered Mixed Oxide ((U, Pu)O<inf>2</inf>) Pellets by Gravimetry
Effective Date
15-Jan-2016
Referred By
ASTM C698-16 - Standard Test Methods for Chemical, Mass Spectrometric, and Spectrochemical Analysis of Nuclear-Grade Mixed Oxides ((U, Pu)O<inf>2</inf>)
Effective Date
01-Jun-2015

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ASTM C1817-15 - Standard Test Method for The Determination of the Oxygen to Metal (O/M) Ratio in Sintered Mixed Oxide ((U, Pu)O<inf>2</inf>) Pellets by GravimetryASTM C1817-15 - Standard Test Method for The Determination of the Oxygen to Metal (O/M) Ratio in Sintered Mixed Oxide ((U, Pu)O<inf>2</inf>) Pellets by Gravimetry
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ASTM C1817-15 - Standard Test Method for The Determination of the Oxygen to Metal (O/M) Ratio in Sintered Mixed Oxide ((U, Pu)O<inf>2</inf>) Pellets by Gravimetry
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Standards Content (Sample)

NOTICE: This standard has either been superseded and replaced by a new version or withdrawn.
Contact ASTM International (www.astm.org) for the latest information
Designation: C1817 − 15
StandardTest Method for
The Determination of the Oxygen to Metal (O/M) Ratio in
1
Sintered Mixed Oxide ((U, Pu)O ) Pellets by Gravimetry
2
This standard is issued under the fixed designation C1817; 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 C859 Terminology Relating to Nuclear Materials
C1068 Guide for Qualification of Measurement Methods by
1.1 This practice is an alternative method to Test Method
a Laboratory Within the Nuclear Industry
C698 for the determination of the oxygen-to-metal atom ratio
C1672 Test Method for Determination of Uranium or Pluto-
(O/M)insinteredmixedoxidefuel(MOX)pellets.Themethod
nium Isotopic Composition or Concentration by the Total
presented in Test Method C698 is a one-step thermogravimet-
Evaporation Method Using a Thermal Ionization Mass
ric method for determining O/M ratio in sintered MOX
Spectrometer
powders and pellets. As stated in Test Method C698, thermo-
D1193 Specification for Reagent Water
gravimetric methods using a two-step heating cycle are also
2 D1356 Terminology Relating to Sampling and Analysis of
satisfactory (1, 2). The method presented in this test method is
Atmospheres
atwo-stepheatingcyclemethod.Thistestmethodisapplicable
4
2.2 ISO Standards:
to sintered MOX pellets containing up to 10 weight percent
ISO 21484 Nuclear Fuel Technology – Determination of the
PuO .
2
O/M Ratio in MOX Pellets – Gravimetric Method – First
1.2 The values stated in SI units are to be regarded as
Edition
standard. No other units of measurement are included in this
3. Terminology
standard.
1.3 This standard does not purport to address all of the
3.1 For definitions of terms used in this test method but not
safety concerns, if any, associated with its use. It is the defined herein, refer to Terminology C859.
responsibility of the user of this standard to establish appro-
3.2 Definitions of Terms Specific to This Standard:
priate safety and health practices and determine the applica-
3.2.1 average of the relative atomic mass—the weighted
bility of regulatory limitations prior to use.
average of the relative atomic mass of an element calculated as
a function of its isotopic composition.
2. Referenced Documents
3 3.2.2 mole fraction—the ratio of the number of molecules
2.1 ASTM Standards:
(or moles) of a compound or element to the total number of
C698 Test Methods for Chemical, Mass Spectrometric, and
molecules (or moles) present (Terminology D1356).
Spectrochemical Analysis of Nuclear-Grade Mixed Ox-
ides ((U, Pu)O ) 3.2.3 MOX—nuclearfuelcomposedofamixtureofuranium
2
and plutonium oxides ((U, Pu)O ).
C753 Specification for Nuclear-Grade, Sinterable Uranium
2
Dioxide Powder
3.2.4 O/M—ratio of the oxygen atoms divided by the metal
C757 Specification for Nuclear-Grade Plutonium Dioxide
atoms in the sample.
Powder, Sinterable
3.2.5 relative atomic mass—a dimensionless physical
C833 Specification for Sintered (Uranium-Plutonium) Diox-
quantity, the ratio of the average mass of atoms of an element
ide Pellets
1
(from a single given sample or source) to ⁄12 of the mass of an
atom of carbon-12 (known as the unified atomic mass unit).
1
This test method is under the jurisdiction ofASTM Committee C26 on Nuclear
3.2.6 scavenging—the process of pushing a gas out by
Fuel Cycle and is the direct responsibility of Subcommittee C26.05 on Methods of
Test. introducing a fresh flow in.
Current edition approved June 1, 2015. Published July 2015. DOI: 10.1520/
3.2.7 sintering—the process of forming a solid mass of
C1817-15.
2
material by heat or pressure, or both, without melting it to the
The boldface numbers in parentheses refer to a list of references at the end of
this standard.
point of liquefaction.
3
For referenced ASTM standards, visit the ASTM website, www.astm.org, or
contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM
4
Standards volume information, refer to the standard’s Document Summary page on Available from International Organization for Standardization (ISO), 1, ch. de
the ASTM website. la Voie-Creuse, CP 56, CH-1211 Geneva 20, Switzerland, http://www.iso.org.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
1

---------------------- Page: 1 ----------------------
C1817 − 15
4. Summary of Test Method gain due to redox reactions. Even inert impurities present in
sufficiently high amounts compromise the accuracy of O/M
4.1 The purpose of the analysis is to test the stoichiometry
ratios because the true mass of the ((U, Pu)O ) subjected to the
2
of the MOX pellet
...

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5.1 This test method allows the determination of 241Am in a plutonium solution without separation of the americium from the plutonium. It is generally applicable to any solution containing 241Am.  
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5.3 When the plutonium solution contains unacceptable levels of fission products or other materials, this method may be used following a tri-n-octylphosphine oxide (TOPO) extraction, ion exchange or other similar separation techniques (see Test Methods C758 and C759).  
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1.3 The values stated in SI units are to be regarded as standard. Additionally, the non-SI units of electron volts, kiloelectron volts, and liters are to be regarded as standard. No other units of measurement are included in this standard.  
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Standard Practice for Preparation and Dissolution of Plutonium Materials for Analysis

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5.1 Plutonium and uranium mixtures are used as nuclear reactor fuels. For use as a nuclear reactor fuel, the material must meet certain criteria for combined uranium and plutonium content, effective fissile content, and impurity content as described in Specifications C757 and C833. After dissolution using one of the procedures described in this practice, the material is assayed for plutonium and uranium to determine if the content is correct as specified by the purchaser.  
5.2 Unique plutonium materials, such as alloys, compounds, and scrap metals, are typically dissolved with various acid mixtures or by fusion with various fluxes. Many plutonium salts are soluble in hydrochloric acid. One or more of the procedures included in this practice may be effective for some of these materials; however their applicability to a particular plutonium material shall be verified by the user.
SCOPE
1.1 This practice is a compilation of dissolution techniques for plutonium materials that are applicable to the test methods used for characterizing these materials. Dissolution treatments for the major plutonium materials assayed for plutonium or analyzed for other components are listed. Aliquots of the dissolved materials are dispensed on a mass basis when one of the analyses must be of high precision, such as plutonium assay; otherwise they are dispensed on a volume basis.  
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1.3 One or more of the procedures in this practice may be applicable to unique plutonium materials, such as alloys, compounds, and scrap materials. The user must determine the applicability of this practice to such materials.  
1.4 The treatments, in order of presentation, are as follows:    
Procedure Number  
Procedure Title  
Section  
1  
Dissolution of Plutonium Metal with Hydrochloric Acid at Room Temperature  
9  
2  
Dissolution of Plutonium Metal with Hydrochloric Acid and Heating  
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3  
Dissolution of Plutonium Metal with Sulfuric Acid  
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4  
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12  
5  
Dissolution of Plutonium Oxide and Uranium-Plutonium Mixed Oxides by Sodium Bisulfate Fusion  
13  
6  
Dissolution of Uranium-Plutonium Mixed Oxides and Low-Fired Plutonium Oxide in Beakers  
14  
7  
Open-Vessel (with Reflux Condenser) Dissolution of Plutonium Oxide Powder  
15  
8  
Open-Vessel (with Reflux Condenser) Dissolution of Mixed Oxide Powder  
16  
9  
Closed-Vessel Hot Block Dissolution of Plutonium Oxide Powder  
17  
10  
Open-Vessel (with Reflux Condenser) Dissolution of Mixed Oxide Pellets  
18  
1.5 The values stated in SI units are to be regarded as standard. The non-SI unit of molarity (M) is also to be regarded as standard. Values in parentheses (non-SI units), where provided, are for information only and are not considered standard.  
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Standard Specification for Sintered (Uranium-Plutonium) Dioxide Pellets for Light Water Reactors

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This specification covers finished sintered and ground (uranium-plutonium) dioxide pellets for use in thermal reactors. It applies to uranium-plutonium dioxide pellets containing plutonium additions up to 15 % weight. The diversity of manufacturing methods shall be recognized by which uranium-plutonium dioxide pellets are produced and the many special requirements for chemical and physical characterization that may be imposed by the operating conditions to which the pellets will be subjected in specific reactor systems. The following are different chemical requirements that shall be determined: uranium content, plutonium content, impurity content, stoichiometry, moisture content, gas content, and americium-241 content. Nuclear requirements such as isotopic content, plutonium equivalent at a given date, equivalent boron content, and reactivity shall also be determined. Physical properties of the pellets like dimensions, density, grain size, pore morphology, plutonium-oxide homogeneity, plutonium-oxide particle size, plutonium-oxide particle distribution, integrity, and surface cracks shall be determined as well. The surfaces of finished pellets shall be visually free of loose chips, oil, macroscopic inclusions, and foreign materials. An estimate of the fuel pellet irradiation stability shall be obtained unless adequate allowance for such effects are factored into the fuel rod design. The estimate of the stability shall consist of either conformance to the thermal stability test as specified in the or by adequate correlation of manufacturing process or microstructure to in-reactor behavior, or both.
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1.1 This specification covers finished sintered and ground (U, Pu)O2 pellets for use in light water reactors. It applies to (U, Pu)O2 pellets containing a plutonium mass fraction up to 15 % (that is, mass of Pu divided by the sum of masses U, Pu, and Am yielding 0.15 or less).  
1.2 Pellets produced under this specification are available in four grades.  
1.2.1 Grade R—240Pu / (Pu + Am) isotope mass fraction is at least 19 %.  
1.2.2 Grade F—240Pu / (Pu + Am) isotope mass fraction is at least 7 % and less than 19 %.  
1.2.3 Grade N1—240Pu / (Pu + Am) isotope mass fraction is less than 7 %.  
1.2.4 Grade N2—240Pu /239Pu isotope mass fraction does not exceed 0.10 (10 %).  
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4.1 Materials handling equipment operability and long-term integrity are concerns that originate during the design and fabrication sequences. Such concerns are most efficiently addressed during one or the other of these stages. Equipment operability and integrity can be compromised during handling and installation sequences. For this reason, the subject equipment should be handled and installed under closely controlled and supervised conditions.  
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4.5.2 Materials handling equipment should be capable of withstanding rigorous chemical cleaning and decontamination procedures.  
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SCOPE
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SIGNIFICANCE AND USE
4.1 Factors governing selection of a method for the determination of uranium include available quantity of sample, sample purity, desired level of reliability, and equipment availability.  
4.2 This test method is suitable for samples between 20 mg to 300 mg of uranium, is applicable to fast breeder reactor (FBR)-mixed oxides having a uranium to plutonium ratio of 2.5 and greater, is tolerant towards most metallic impurity elements usually specified for FBR-mixed oxide fuel, and uses no special equipment.  
4.3 The ruggedness of the titration method has been studied for both the volumetric (6) and the weight (7) titration of uranium with dichromate.
SCOPE
1.1 This test method covers unirradiated uranium-plutonium mixed oxide having a uranium to plutonium ratio of 2.5 and greater. The presence of larger amounts of plutonium (Pu) that give lower uranium to plutonium ratios may give low analysis results for uranium (U) (1)2, if the amount of plutonium together with the uranium is sufficient to slow the reduction step and prevent complete reduction of the uranium in the allotted time. Use of this test method for lower uranium to plutonium ratios may be possible, especially when 20 mg to 50 mg quantities of uranium are being titrated rather than the 100 mg to 300 mg in the study cited in Ref (1). Confirmation of that information should be obtained before this test method is used for ratios of uranium to plutonium less than 2.5.  
1.2 The amount of uranium determined in the data presented in Section 12 was 20 mg to 50 mg. However, this test method, as stated, contains iron in excess of that needed to reduce the combined quantities of uranium and plutonium in a solution containing 300 mg of uranium with uranium to plutonium ratios greater than or equal to 2.5. Solutions containing up to 300 mg uranium with uranium to plutonium ratios greater than or equal to 2.5 have been analyzed (1) using the reagent volumes and conditions as described in Section 10.  
1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.4 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. For specific hazard statements, see Section 8.  
1.5 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.

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ASTM
ASTM C1347-08(2023)(Practice)
Standard Practice for Preparation and Dissolution of Uranium Materials for Analysis

SIGNIFICANCE AND USE
4.1 The materials covered that must meet ASTM specifications are uranium metal and uranium oxide.  
4.2 Uranium materials are used as nuclear reactor fuel. For this use, these materials must meet certain criteria for uranium content, uranium-235 enrichment, and impurity content, as described in Specifications C753 and C776. The material is assayed for uranium to determine whether the content is as specified.  
4.3 Uranium alloys, refractory uranium materials, and uranium containing scrap and ash are unique uranium materials for which the user must determine the applicability of this practice. In general, these unique uranium materials are dissolved with various acid mixtures or by fusion with various fluxes.
SCOPE
1.1 This practice covers dissolution treatments for uranium materials that are applicable to the test methods used for characterizing these materials for uranium elemental, isotopic, and impurities determinations. Dissolution treatments for the major uranium materials assayed for uranium or analyzed for other components are listed.  
1.2 The treatments, in order of presentation, are as follows:    
Procedure Title  
Section  
Dissolution of Uranium Metal and Oxide with Nitric Acid  
8.1  
Dissolution of Uranium Oxides with Nitric Acid and Residue
Treatment  
8.2  
Dissolution of Uranium-Aluminum Alloys in Hydrochloric Acid
with Residue Treatment  
8.3  
Dissolution of Uranium Scrap and Ash by Leaching with Nitric
Acid and Treatment of Residue by Carbonate Fusion  
8.4  
Dissolution of Refractory Uranium-Containing Material by
Carbonate Fusion  
8.5  
Dissolution of Uranium—Aluminum Alloys
Uranium Scrap and Ash, and Refractory
Uranium-Containing Materials by
Microwave Treatment  
8.6  
1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.4 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. Specific hazards statements are given in Section 7.  
1.5 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.

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