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ASTM C776-06

(Specification)

Standard Specification for Sintered Uranium Dioxide Pellets

Standard Specification for Sintered Uranium Dioxide Pellets

ABSTRACT
This specification covers sintered uranium dioxide pellets containing 235U for use in nuclear reactors. Chemical requirements include uranium content, impurity content, stoichiometry, and moisture content. Maximum concentration limits are specified for impurity elements such as: aluminum, carbon, calcium+magnesium, chlorine, chromium, cobalt, fluorine, hydrogen, iron, nickel, nitrogen, silicon, and thorium. Chemical analyses shall be performed. Nuclear requirements include isotopic content and equivalent boron content. The following are physical characteristics of the material: dimensions, pellet density, grain size and pore morphology, pellet integrity–surface cracks and chips, and cleanliness and workmanship. Fuel pellet irradiation stability shall be estimated. Pellet lot shall be retained throughout processing without mixing with other established lots.
SCOPE
1.1 This specification is for finished sintered uranium dioxide pellets. It applies to uranium dioxide pellets containing uranium of any  235U concentration for use in nuclear reactors.
1.2 This specification recognizes the presence of reprocessed uranium in the fuel cycle and consequently defines isotopic limits for uranium dioxide pellets made from commercial grade UO2. Such commercial grade UO 2 is defined so that, regarding fuel design and manufacture, the product is essentially equivalent to that made from unirradiated uranium. UO 2 falling outside these limits cannot necessarily be regarded as equivalent and may thus need special provisions at the fuel fabrication plant or in the fuel design.
1.3 This specification does not include  (a) provisions for preventing criticality accidents or  (b) requirements for health and safety. Observance of this specification does not relieve the user of the obligation to be aware of and conform to all federal, state, and local regulations pertaining to possessing, shipping, processing, or using source or special nuclear material. Examples of U.S. Government documents are Code of Federal Regulations (Latest Edition), Title 10, Part 50, Title 10, Part 71, and Title 49, Part 173.
The following precautionary caveat pertains only to the technical requirements portion, Section , of this specification: 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 or regulatory limitations prior to use.

General Information

Status
Historical
Publication Date
31-Jul-2006
ICS
27.120.30 - Fissile materials and nuclear fuel technology
Technical Committee
C26 - Nuclear Fuel Cycle
Drafting Committee
C26.02 - Fuel and Fertile Material Specifications
Current Stage
Ref Project

Relations

Replaces
ASTM C776-00 - Standard Specification for Sintered Uranium Dioxide Pellets
Effective Date
01-Aug-2006
Replaced By
ASTM C776-06(2011) - Standard Specification for Sintered Uranium Dioxide Pellets
Effective Date
01-Jun-2011
Referred By
ASTM C1457-18 - Standard Test Method for Determination of Total Hydrogen Content of Uranium Oxide Powders and Pellets by Carrier Gas Extraction
Effective Date
01-Aug-2006
Referred By
ASTM C1413-18 - Standard Test Method for Isotopic Analysis of Hydrolyzed Uranium Hexafluoride and Uranyl Nitrate Solutions by Thermal Ionization Mass Spectrometry
Effective Date
01-Aug-2006
Referred By
ASTM C1430-18 - Standard Test Method for Determination of Uranium, Oxygen to Uranium (O/U), and Oxygen to Metal (O/M) in Sintered Uranium Dioxide and Gadolinia-Uranium Dioxide Pellets by Atmospheric Equilibration
Effective Date
01-Aug-2006
Referred By
ASTM C1647-20 - Standard Practice for Removal of Uranium or Plutonium, or both, for Impurity Assay in Uranium or Plutonium Materials
Effective Date
01-Aug-2006
Referred By
ASTM C1502-16 - Standard Test Method for Determination of Total Chlorine and Fluorine in Uranium Dioxide and Gadolinium Oxide
Effective Date
01-Aug-2006
Referred By
ASTM C1287-18 - Standard Test Method for Determination of Impurities in Nuclear Grade Uranium Compounds by Inductively Coupled Plasma Mass Spectrometry
Effective Date
01-Aug-2006
Referred By
ASTM C1832-23 - Standard Test Method for Determination of Uranium Isotopic Composition by Modified Total Evaporation (MTE) Method Using Thermal Ionization Mass Spectrometer
Effective Date
01-Aug-2006
Referred By
ASTM C696-19 - Standard Test Methods for Chemical, Mass Spectrometric, and Spectrochemical Analysis of Nuclear-Grade Uranium Dioxide Powders and Pellets
Effective Date
01-Aug-2006
Referred By
ASTM C1625-19 - Standard Test Method for Uranium and Plutonium Concentrations and Isotopic Abundances by Thermal Ionization Mass Spectrometry
Effective Date
01-Aug-2006
Referred By
ASTM C1453-19 - Standard Test Method for the Determination of Uranium by Ignition and the Oxygen to Uranium (O/U) Atomic Ratio of Nuclear Grade Uranium Dioxide Powders and Pellets
Effective Date
01-Aug-2006
Referred By
ASTM C1871-22 - Standard Test Method for Determination of Uranium Isotopic Composition by the Double Spike Method Using a Thermal Ionization Mass Spectrometer
Effective Date
01-Aug-2006
Referred By
ASTM C1408-16 - Standard Test Method for Carbon (Total) in Uranium Oxide Powders and Pellets By Direct Combustion-Infrared Detection Method
Effective Date
01-Aug-2006
Referred By
ASTM C1474-19 - Standard Test Method for Analysis of Isotopic Composition of Uranium in Nuclear-Grade Fuel Material by Quadrupole Inductively Coupled Plasma-Mass Spectrometry
Effective Date
01-Aug-2006

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ASTM C776-06 - Standard Specification for Sintered Uranium Dioxide Pellets
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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: C776 – 06
Standard Specification for
1
Sintered Uranium Dioxide Pellets
This standard is issued under the fixed designation C776; 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.
INTRODUCTION
This specification is intended to provide the nuclear industry with a general specification for
uranium dioxide pellets. It recognizes the diversity of manufacturing methods by which uranium
dioxide pellets are produced and the many special requirements for chemical and physical
characterization which may be imposed by the operating conditions to which the pellets will be
subjected in specific reactor systems. Therefore, it is anticipated that the purchaser may supplement
this specification with additional requirements for specific applications.
1. Scope 2. Referenced Documents
2
1.1 This specification is for finished sintered uranium diox- 2.1 ASTM Standards:
ide pellets. It applies to uranium dioxide pellets containing C696 Test Methods for Chemical, Mass Spectrometric, and
235
uranium of any U concentration for use in nuclear reactors. Spectrochemical Analysis of Nuclear-Grade Uranium Di-
1.2 This specification recognizes the presence of repro- oxide Powders and Pellets
cessed uranium in the fuel cycle and consequently defines C753 Specification for Nuclear-Grade, Sinterable Uranium
isotopic limits for uranium dioxide pellets made from commer- Dioxide Powder
cial grade UO . Such commercial grade UO is defined so that, C859 Terminology Relating to Nuclear Materials
2 2
regarding fuel design and manufacture, the product is essen- C996 Specification for Uranium Hexafluoride Enriched to
235
tially equivalent to that made from unirradiated uranium. UO Less Than 5 % U
2
falling outside these limits cannot necessarily be regarded as C1233 Practice for Determining Equivalent Boron Contents
equivalent and may thus need special provisions at the fuel of Nuclear Materials
fabrication plant or in the fuel design. E105 Practice for Probability Sampling of Materials
3
1.3 This specification does not include (a) provisions for 2.2 ANSI Standard:
preventing criticality accidents or (b) requirements for health ANSI/ASME NQA-1 Quality Assurance Requirements for
andsafety.Observanceofthisspecificationdoesnotrelievethe Nuclear Facility Applications
useroftheobligationtobeawareofandconformtoallfederal, 2.3 U.S. Government Documents:
state, and local regulations pertaining to possessing, shipping, Code of Federal Regulations (Latest Edition), Title 10, Part
processing, or using source or special nuclear material. Ex- 50, Energy (10 CFR 50) Domestic Licensing of Produc-
4
amples of U.S. Government documents are Code of Federal tion and Utilization Facilities
Regulations(LatestEdition),Title10,Part50,Title10,Part71, Code of Federal Regulations, Title 10, Part 71, Packaging
4
and Title 49, Part 173. and Transportation of Radioactive Material
1.4 The following precautionary caveat pertains only to the Code of Federal Regulations, Title 49, Part 173, General
4
4, of this specification: Requirements for Shipments and Packaging
technical requirements portion, Section
This standard does not purport to address all of the safety Regulatory Guide NUREG 1.126 AnAcceptable Model and
concerns, if any, associated with its use. It is the responsibility Related Statistical Methods for the Analysis of Fuel
of the user of this standard to establish appropriate safety and
health practices and determine the applicability or regulatory
limitations prior to use.
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
Standards volume information, refer to the standard’s Document Summary page on
1
This specification is under the jurisdiction of ASTM Committee C26 on the ASTM website.
3
Nuclear Fuel Cycle and is the direct responsibility of Subcommittee C26.02 on Fuel Available fromAmerican National Standards Institute (ANSI), 25 W. 43rd St.,
and Fertile Material Specifications. 4th Floor, New York, NY 10036, http://www.ansi.org.
4
Current edition approved Aug. 1, 2006. Published September 2006. Originally AvailablefromU.S.GovernmentPrintingOfficeSuperintendentofDocuments,
approved in 1976. Last previous edition approved in 2000 as C776 – 00. DOI: 732 N. Capitol St., NW, Mail Stop: SDE, Washington, DC 20401, http://
10.1520/C0776-06. www.access.gpo.gov.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.
1

---------------------- Page: 1 ----------------------
C776 – 06
5
Densification, Rev. 1 March 197
...

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SIGNIFICANCE AND USE
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Standard Practice for Preparation and Dissolution of Plutonium Materials for Analysis

SIGNIFICANCE AND USE
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.  
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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.  
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Procedure Number  
Procedure Title  
Section  
1  
Dissolution of Plutonium Metal with Hydrochloric Acid at Room Temperature  
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2  
Dissolution of Plutonium Metal with Hydrochloric Acid and Heating  
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6  
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7  
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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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ABSTRACT
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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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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4.5.1 Boots and similar protective covers should not restrict movement of the equipment, should be properly sealed to the equipment and should withstand the radiation, cell atmosphere, dust, cell temperatures, chemical exposures, and cleaning and decontamination reagents, and also resist snags and tearing.  
4.5.2 Materials handling equipment should be capable of withstanding rigorous chemical cleaning and decontamination procedures.  
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SCOPE
1.1 Intent:  
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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SIGNIFICANCE AND USE
5.1 Uranium hexafluoride is normally produced and handled in large (typically 1 to 14-ton) quantities and must, therefore, be characterized by reference to representative samples (see ISO 7195). The samples are used to determine compliance with the applicable commercial specification C787. The quantities involved, physical properties, chemical reactivity, and hazardous nature of UF6 are such that for representative sampling, specially designed equipment must be used and operated in accordance with the most carefully controlled and stringent procedures. This practice can be used by UF6 converters to review the effectiveness of existing procedures or as a guide to the design of equipment and procedures for future use.  
5.2 The intention of this practice is to avoid liquid UF6 sampling once the cylinder has been filled. For safety reasons, manipulation of large quantities of liquid UF6 should be avoided when possible.  
5.3 It is emphasized that this practice is not meant to address conventional or nuclear criticality safety issues.
SCOPE
1.1 This practice covers methods for withdrawing representative sample(s) of uranium hexafluoride (UF6) during a transfer occurring in the gas phase. Such transfer in the gas phase can take place during the filling of a cylinder during a continuous production process, for example the distillation column in a conversion facility. Such sample(s) may be used for determining compliance with the applicable commercial specification, for example Specification C787.  
1.2 Since UF6 sampling is taken during the filling process, this practice does not address any special additional arrangements that may be agreed upon between the buyer and the seller when the sampled bulk material is being added to residues already present in a container (“heels recycle”). Such arrangements will be based on QA procedures such as traceability of cylinder origin (to prevent for example contamination with irradiated material).  
1.3 If the receiving cylinder is purged after filling and sampling, special verifications must be performed by the user to verify the representativity of the sample(s). It is then expected that the results found on volatile impurities with gas phase sampling may be conservative.  
1.4 This practice is only applicable when the transfer occurs in the gas phase. When the transfer is performed in the liquid phase, Practice C1052 should apply. This practice does not apply to gas sampling after the cylinder has been filled since the sample taken will not be representative of the cylinder.  
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.

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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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ASTM
ASTM C1165-23(Test Method)
Standard Test Method for Determining Plutonium by Controlled-Potential Coulometry in H2SO4 at a Platinum Working Electrode

SIGNIFICANCE AND USE
5.1 This test method is used to ascertain whether or not materials meet specifications for plutonium concentration or plutonium mass fraction.  
5.1.1 The materials (nuclear grade plutonium nitrate solutions, plutonium metal, plutonium oxide powder, and mixed oxide and carbide powders and pellets) to which this test method applies are subject to nuclear safeguards regulations governing their possession and use. However, adherence to this test method does not automatically guarantee regulatory acceptance of the resulting safeguards measurements. It remains the sole responsibility of the user of this test method to ensure that its application to safeguards has the approval of the proper regulatory authorities.  
5.1.2 When used in conjunction with appropriate certified reference materials (CRMs), this test method can demonstrate traceability to the international measurements system (SI).  
5.2 Fitness for Purpose of Safeguards and Nuclear Safety Application—Methods intended for use in safeguards and nuclear safety applications shall meet the requirements specified by Guide C1068 for use in such applications.  
5.3 A chemical calibration of the coulometer is necessary for accurate results.
FIG. 1 Example of a Cell Design Used at Los Alamos National Laboratory (LANL)
SCOPE
1.1 This test method covers the determination of milligram quantities of plutonium in unirradiated uranium-plutonium mixed oxide having a U/Pu ratio range of 0.1 to 10. This test method is also applicable to plutonium metal, plutonium oxide, uranium-plutonium mixed carbide, various plutonium compounds including fluoride and chloride salts, and plutonium solutions.  
1.2 The recommended amount of plutonium for each aliquant in the coulometric analysis is 5 mg to 10 mg. Precision worsens for lower amounts of plutonium, and elapsed time of electrolysis becomes impractical for higher amounts of plutonium.  
1.3 The quantity values stated in SI units are to be regarded as standard. The quantity values with non-SI units are given in parentheses 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. Specific precautionary statements are given in Section 9.  
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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  • Standard
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    English language
    sale 15% off