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ASTM C1843-16

(Test Method)

Standard Test Method for Determining Moisture Content in Uranium-Ore Concentrate

Standard Test Method for Determining Moisture Content in Uranium-Ore Concentrate

SIGNIFICANCE AND USE
5.1 The test method within this standard is used to demonstrate uranium ore-concentrate material meets the moisture specification defined in Specification C967 or other applicable requirements.
SCOPE
1.1 This test method is used to determine compliance with Specification C967 for the requirements of moisture in uranium ore concentrates (UOC). A procedure is given to determine the approximate temperature for drying the UOC; normally 110°C but possibly 165°C for uranyl peroxides. The dried uranium ore-concentrate resulting from this procedure is then used for performing additional analyses described in Specification C967.  
1.2 The values stated in SI units are to be regarded as the standard. The values given in parentheses are mathematical equivalents that are provided for information only and are not considered 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
14-Dec-2016
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

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ASTM C1843-16 - Standard Test Method for Determining Moisture Content in Uranium-Ore ConcentrateASTM C1843-16 - Standard Test Method for Determining Moisture Content in Uranium-Ore Concentrate
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ASTM C1843-16 - Standard Test Method for Determining Moisture Content in Uranium-Ore Concentrate
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NOTICE: This standard has either been superseded and replaced by a new version or withdrawn.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: C1843 − 16
Standard Test Method for
1
Determining Moisture Content in Uranium-Ore Concentrate
This standard is issued under the fixed designation C1843; 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 4. Summary of Test Method
4.1 The need to establish an appropriate temperature for
1.1 This test method is used to determine compliance with
drying may only occur once for a specific ore concentrate but
SpecificationC967fortherequirementsofmoistureinuranium
may be evaluated if the property of the ore is suspected to have
ore concentrates (UOC).Aprocedure is given to determine the
changed. An example of a UOC material that may require
approximate temperature for drying the UOC; normally 110°C
drying at 165°C is uranyl peroxide. Other concentrates, such as
but possibly 165°C for uranyl peroxides. The dried uranium
U O,UO , and UO are generally dried at 110°C.
ore-concentrate resulting from this procedure is then used for 3 8 2 3
performing additional analyses described in Specification
4.2 Aweighed portion of uranium ore concentrate sample is
C967.
placed into an oven that is capable of maintaining a tempera-
ture of 110 or 165°C. The sample is heated and weighed
1.2 The values stated in SI units are to be regarded as the
periodically to a constant weight or an insignificant change in
standard. The values given in parentheses are mathematical
weight.
equivalents that are provided for information only and are not
4.3 A weight change that is insignificant would be defined
considered standard.
by the sample size and the required precision.
1.3 This standard does not purport to address all of the
safety concerns, if any, associated with its use. It is the
5. Significance and Use
responsibility of the user of this standard to establish appro-
5.1 The test method within this standard is used to demon-
priate safety and health practices and determine the applica-
strate uranium ore-concentrate material meets the moisture
bility of regulatory limitations prior to use.
specification defined in Specification C967 or other applicable
requirements.
2. Referenced Documents
2 6. Safety Precautions
2.1 ASTM Standards:
C859 Terminology Relating to Nuclear Materials
6.1 Proper precautions should be taken to prevent inhalation
C967 Specification for Uranium Ore Concentrate or ingestion of uranium-ore concentrate during ore evaluation,
C1022 Test Methods for Chemical and Atomic Absorption
sample preparation, sample analysis, and sample packaging.
Analysis of Uranium-Ore Concentrate Precautions used to prevent inhalation or ingestion should
C1075 Practices for Sampling Uranium-Ore Concentrate
include a ventilation system and personal protective equip-
E11 Specification for Woven Wire Test Sieve Cloth and Test ment. Generally the ventilation system is in the form of
Sieves
laboratory hoods with a dust collection system. Personal
protective equipment used should be a respirator designed for
3. Terminology particulate matter.
3.1 Except as otherwise defined herein, definitions of terms
7. Apparatus
are as given in Terminology C859.
7.1 Drying Oven, capable of maintaining 110 or 165°C.
7.2 Desiccator.
1
7.3 Vacuum Pump, capable of reducing pressure by 34 kPa
This test method is under the jurisdiction ofASTM Committee C26 on Nuclear
Fuel Cycle and is the direct responsibility of Subcommittee C26.05 on Methods of
(5 psi) (25 cm of mercury).
Test.
7.4 Weighing trays.
Current edition approved Dec. 15, 2016. Published January 2017. DOI: 10.1520/
C1843-16.
7.5 Pulverizer.
2
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
7.6 Blender.
Standards volume information, refer to the standard’s Document Summary page on
the ASTM website. 7.7 Balance, 3000 g capacity, readable to at least 0.1 g.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
1

---------------------- Page: 1 ----------------------
C1843 − 16
7.8 Sample Jars, glass with metal lids, able to hold 150 g or 11.3 Divide the remaining blended material, about 2500 g,
as agreed upon. into two roughly equal portions and spread evenly in two
stainless steel drying pans.
7.9 Test Sieve, 0.177 mm (U.S. Sieve 80 mesh).
7.10 Test Sieve, 0.149 mm (U.S. Sieve 100 mesh).
12. Procedure
NOTE 1—See Specification E11.
12.1 Transfer directly two un-pulverized samples of about
7.11 Flat Iron, able to heat sample jar lids to ~82°C.
1300 g each, which are designated for the moisture
determination, and spread in two tared dryin
...

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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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5.6 This test method is applicable only to homogeneous solutions. This test method is not suitable for solutions containing solids.  
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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.
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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.  
1.2 Procedures in this practice are intended for the dissolution of plutonium metal, plutonium oxide, and uranium-plutonium mixed oxides. Aliquots of dissolved materials are analyzed using test methods, such as those developed by Subcommittee C26.05 on Methods of Test, to demonstrate compliance with applicable requirements. These may include product specifications such as Specifications C757 and C833.  
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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Open-Vessel (with Reflux Condenser) Dissolution of Mixed Oxide Powder  
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9  
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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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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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1.1 This terminology standard contains terms, definitions, descriptions of terms, nomenclature, and explanations of acronyms and symbols specifically associated with standards under the jurisdiction of Committee C26 on Nuclear Fuel Cycle. The content of this terminology standard may also be applicable to documents not under the jurisdiction of Committee C26, in which case this terminology standard may be referenced in those documents.  
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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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1.1.1 This guide covers materials handling equipment used in hot cells (shielded cells) for the processing and handling of nuclear and radioactive materials. The intent of this guide is to aid in the selection and design of materials handling equipment for hot cells in order to minimize equipment failures and maximize the equipment utility.  
1.1.2 It is intended that this guide record the principles and caveats that experience has shown to be essential to the design, fabrication, installation, maintenance, repair, replacement, and decontamination and decommissioning of materials handling equipment capable of meeting the stringent demands of operating, dependably and safely, in a hot cell environment where operator visibility is limited due to the radiation exposure hazards.  
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5.1 Factors governing selection of a method for the determination of plutonium include available quantity of sample, sample purity, desired level of reliability, and equipment.  
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1.1 This test method describes the determination of dissolved plutonium from unirradiated nuclear-grade (that is, high-purity) materials by controlled-potential coulometry. Controlled-potential coulometry may be performed in a choice of supporting electrolytes, such as 0.9 mol/L (0.9 M) HNO3, 1 mol/L (1 M) HClO4, 1 mol/L (1 M) HCl, 5 mol/L (5 M) HCl, and 0.5 mol/L (0.5 M) H2SO4. Limitations on the use of selected supporting electrolytes are discussed in Section 6. Optimum quantities of plutonium for this procedure are 5 mg to 20 mg.  
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1.3 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.  
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.  
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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.  
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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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    9 pages
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