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ASTM C1008-99(2008)

(Specification)

Standard Specification for Sintered (Uranium-Plutonium) Dioxide
Pellets—Fast Reactor Fuel (Withdrawn 2014)

Standard Specification for Sintered (Uranium-Plutonium) Dioxide<br> Pellets<span class='unicode'>&#x2014;</span>Fast Reactor Fuel (Withdrawn 2014)

ABSTRACT
This specification covers finished pellets composed of sintered uranium-plutonium dioxide for fast reactor fuel. Specimens shall be sampled and tested suitably, and shall conform accordingly to chemical (uranium and plutonium content, impurity content, stoichiometry, moisture, gas content, and americium-241 content), nuclear (isotopic content, and equivalent plutonium at a given date), and physical (dimensions, density, grain size an pore morphology, homogeneity, particle size and distribution, integrity, surface cracks, circumferential chips, pellet ends, cleanliness and workmanship, and identification) requirements.
SCOPE
1.1 This specification is for finished sintered (uranium-plutonium) dioxide pellets. It applies to (uranium-plutonium) dioxide pellets containing plutonium additions in the range from 10 to 40 weight % and uranium of any  235U content. The isotopic composition of the plutonium component will be as normally produced by in-reactor neutron irradiation of uranium having less than 5 %  235U or by in-reactor neutron irradiation of recycled plutonium mixed with uranium.
1.2 This specification does not include (1) provisions for preventing criticality accidents or (2) 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 applicable international, federal, state, and local regulations pertaining to possessing, processing, shipping, or using source or special nuclear material. Examples of U.S. government documents are Code of Federal Regulations Title 10, Part 50 — Domestic Licensing of Production and Utilization Facilities; Title 10, Part 71 — Packaging and Transportation of Radioactive Material; and Title 49, Part 173 — General Requirements for Shipments and Packaging.
1.3 The following safety hazards caveat pertains only to the technical requirements portion, Section 4, 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 of regulatory limitations prior to use.
WITHDRAWN RATIONALE
This specification was for finished sintered (uranium-plutonium) dioxide pellets. It applied to (uranium-plutonium) dioxide pellets containing plutonium additions in the range from 10 to 40 weight % and uranium of any  235U content. The isotopic composition of the plutonium component would have been as normally produced by in-reactor neutron irradiation of uranium having less than 5 %  235U or by in-reactor neutron irradiation of recycled plutonium mixed with uranium.
Formerly under the jurisdiction of Committee C26 on Nuclear Fuel Cycle, this test method was withdrawn in May 2014. This standard is being withdrawn without replacement due to its limited use by industry.

General Information

Status
Withdrawn
Publication Date
31-Dec-2007
Withdrawal Date
19-May-2014
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

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ASTM C1008-99(2008) - Standard Specification for Sintered (Uranium-Plutonium) Dioxide<br> Pellets<span class='unicode'>&#x2014;</span>Fast Reactor Fuel (Withdrawn 2014)ASTM C1008-99(2008) - Standard Specification for Sintered (Uranium-Plutonium) Dioxide<br> Pellets<span class='unicode'>&#x2014;</span>Fast Reactor Fuel (Withdrawn 2014)
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ASTM C1008-99(2008) - Standard Specification for Sintered (Uranium-Plutonium) Dioxide<br> Pellets<span class='unicode'>&#x2014;</span>Fast Reactor Fuel (Withdrawn 2014)
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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:C1008 −99(Reapproved2008)
Standard Specification for
Sintered (Uranium-Plutonium) Dioxide
1
Pellets—Fast Reactor Fuel
This standard is issued under the fixed designation C1008; 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 standard for
(uranium-plutonium) dioxide pellets for fast reactor use. It recognizes the diversity of manufacturing
methods by which (uranium-plutonium) 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. It does not recognize the
possible problems associated with the reprocessing of such pellets. It is therefore anticipated that the
buyer may supplement this specification with additional requirements for specific applications.
1. Scope of the user of this standard to establish appropriate safety and
health practices and determine the applicability of regulatory
1.1 This specification is for finished sintered (uranium-
limitations prior to use.
plutonium) dioxide pellets. It applies to (uranium-plutonium)
dioxide pellets containing plutonium additions in the range
2. Referenced Documents
235
from 10 to 40 weight % and uranium of any U content. The
2
2.1 ASTM Standards:
isotopic composition of the plutonium component will be as
C698 Test Methods for Chemical, Mass Spectrometric, and
normallyproducedbyin-reactorneutronirradiationofuranium
235
Spectrochemical Analysis of Nuclear-Grade Mixed Ox-
having less than 5 % U or by in-reactor neutron irradiation
ides ((U, Pu)O )
of recycled plutonium mixed with uranium. 2
C753 Specification for Nuclear-Grade, Sinterable Uranium
1.2 This specification does not include (1) provisions for
Dioxide Powder
preventing criticality accidents or (2) requirements for health
C757 Specification for Nuclear-Grade Plutonium Dioxide
andsafety.Observanceofthisspecificationdoesnotrelievethe
Powder, Sinterable
user of the obligation to be aware of and conform to all
C859 Terminology Relating to Nuclear Materials
applicable international, federal, state, and local regulations
C1165 Test Method for Determining Plutonium by
pertaining to possessing, processing, shipping, or using source
Controlled-Potential Coulometry in H SO at a Platinum
2 4
or special nuclear material. Examples of U.S. government
Working Electrode
documentsareCodeofFederalRegulationsTitle10,Part50—
C1204 Test Method for Uranium in Presence of Plutonium
Domestic Licensing of Production and Utilization Facilities;
by Iron(II) Reduction in Phosphoric Acid Followed by
Title 10, Part 71 — Packaging and Transportation of Radioac-
Chromium(VI) Titration
tive Material; and Title 49, Part 173 — General Requirements
C1206 Test Method for Plutonium by Iron (II)/Chromium
for Shipments and Packaging.
(VI) Amperometric Titration
1.3 The following safety hazards caveat pertains only to the
E105 Practice for Probability Sampling of Materials
technical requirements portion, Section 4, of this specification: 3
2.2 ANSI Standard:
This standard does not purport to address all of the safety
ANSI/ASME NQA-1 Quality Assurance Program Require-
concerns, if any, associated with its use. It is the responsibility
ments for Nuclear Facility Applications
1 2
This specification is under the jurisdiction of ASTM Committee C26 on For referenced ASTM standards, visit the ASTM website, www.astm.org, or
Nuclear Fuel Cycle and is the direct responsibility of Subcommittee C26.02 on Fuel contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM
and Fertile Material Specifications. Standards volume information, refer to the standard’s Document Summary page on
Current edition approved Jan. 1, 2008. Published February 2008. Originally the ASTM website.
3
approved in 1984. Last previous edition approved in 1999 as C1008 – 99. DOI: Available fromAmerican National Standards Institute (ANSI), 25 W. 43rd St.,
10.1520/C1008-99R08. 4th Floor, New York, NY 10036, http://www.ansi.org.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
1

---------------------- Page: 1 ----------------------
C1008−99(2008)
TABLE 1 Impurity Elements and Maximum Concentration Limits
4.1.3 Stoichiometry—The oxygen-to-heavy-metal ratio of
Maximum sinteredfuelpelletsshallbewithintherangefrom1.94to2.01.
Concentration
The nominal value and allowable tolerances shall be agreed
Elements
Limit
upon between the buyer and the selle
...

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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.
SCOPE
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.  
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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 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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SCOPE
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ASTM C1347-08(2023)(Practice)
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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.  
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Procedure Title  
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8.1  
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8.2  
Dissolution of Uranium-Aluminum Alloys in Hydrochloric Acid
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8.3  
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8.4  
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8.5  
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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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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 C1703-18(2023)(Practice)
Standard Practice for Sampling of Gaseous Uranium Hexafluoride for Enrichment

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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