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ASTM C1462-00(2008)

(Specification)

Standard Specification for Uranium Metal Enriched to More than 15 % and Less Than 20 % 235U

Standard Specification for Uranium Metal Enriched to More than 15 % and Less Than 20 % <sup> 235</sup>U

ABSTRACT
This specification covers nuclear grade uranium metal that has either been processed through an enrichment plant, or has been produced by the blending of highly enriched uranium with other uranium, to obtain uranium of any 235U concentration below 20 % (and greater than 15 %) and that is intended for research reactor fuel fabrication. The uranium content of commercial grade enriched uranium metal shall be greater than or equal to 99.85 weight percent. The isotopic requirements for enriched uranium metal are presented in details. The chemical and isotopic composition shall be tested to meet the requirements prescribed.
SCOPE
1.1 This specification covers nuclear grade uranium metal that has either been processed through an enrichment plant, or has been produced by the blending of highly enriched uranium with other uranium, to obtain uranium of any  235U concentration below 20 % (and greater than 15 %) and that is intended for research reactor fuel fabrication. The scope of this specification includes specifications for enriched uranium metal derived from commercial natural uranium, recovered uranium, or highly enriched uranium. Commercial natural uranium, recovered uranium and highly enriched uranium are defined in Section 3. The objectives of this specification are to define the impurity and uranium isotope limits for commercial grade enriched uranium metal.
1.2 This specification is intended to provide the nuclear industry with a standard for enriched uranium metal which is to be used in the production of research reactor fuel. In addition to this specification, the parties concerned may agree to other appropriate conditions.
1.3 The scope of this specification does not comprehensively cover all provisions for preventing criticality accidents or requirements for health and safety or for shipping. Observance of this standard does not relieve the user of the obligation to conform to all applicable international, federal, state, and local regulations for processing, shipping, or any other way of using uranium metal (see, for example, C 996 regarding references).

General Information

Status
Historical
Publication Date
31-Dec-2007
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 C1462-00(2005) - Standard Specification for Uranium Metal Enriched to More than 15% and Less Than 20%<sup>235</sup> U
Effective Date
01-Jan-2008

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ASTM C1462-00(2008) - Standard Specification for Uranium Metal Enriched to More than 15 % and Less Than 20 % <sup> 235</sup>UASTM C1462-00(2008) - Standard Specification for Uranium Metal Enriched to More than 15 % and Less Than 20 % <sup> 235</sup>U
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ASTM C1462-00(2008) - Standard Specification for Uranium Metal Enriched to More than 15 % and Less Than 20 % <sup> 235</sup>U
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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:C1462 −00(Reapproved 2008)
Standard Specification for
Uranium Metal Enriched to More than 15% and Less Than
235 1
20% U
This standard is issued under the fixed designation C1462; 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 oxide Powders and Pellets
C799 Test Methods for Chemical, Mass Spectrometric,
1.1 This specification covers nuclear grade uranium metal
Spectrochemical, Nuclear, and RadiochemicalAnalysis of
that has either been processed through an enrichment plant, or
Nuclear-Grade Uranyl Nitrate Solutions
has been produced by the blending of highly enriched uranium
235 C859 Terminology Relating to Nuclear Materials
with other uranium, to obtain uranium of any U concentra-
C996 Specification for Uranium Hexafluoride Enriched to
tion below 20 % (and greater than 15 %) and that is intended
235
Less Than 5 % U
for research reactor fuel fabrication. The scope of this speci-
C1233 Practice for Determining Equivalent Boron Contents
fication includes specifications for enriched uranium metal
of Nuclear Materials
derived from commercial natural uranium, recovered uranium,
C1295 Test Method for Gamma Energy Emission from
or highly enriched uranium. Commercial natural uranium,
Fission Products in Uranium Hexafluoride and Uranyl
recovered uranium and highly enriched uranium are defined in
Nitrate Solution
Section 3. The objectives of this specification are to define the
C1347 Practice for Preparation and Dissolution of Uranium
impurity and uranium isotope limits for commercial grade
Materials for Analysis
enriched uranium metal.
3
2.2 ANSI Standard
1.2 This specification is intended to provide the nuclear
ANSI-ASME NQA-1 Quality Assurance Program Require-
industrywithastandardforenricheduraniummetalwhichisto
ments for Nuclear Facility Applications
be used in the production of research reactor fuel. In addition
4
2.3 U.S. Government Documents
to this specification, the parties concerned may agree to other
CodeofFederalRegulations, Title10,Part50,(AppendixB)
appropriate conditions.
1.3 The scope of this specification does not comprehen-
3. Terminology
sively cover all provisions for preventing criticality accidents
3.1 Definitions of Terms Specific to This Standard:
or requirements for health and safety or for shipping. Obser-
vance of this standard does not relieve the user of the 3.1.1 Definitions of Terms Specific to This Standard—
Terms shall be defined in accordance with Terminology C859,
obligation to conform to all applicable international, federal,
state, and local regulations for processing, shipping, or any except for the following:
3.1.2 commercial grade enriched uranium metal— uranium
other way of using uranium metal (see, for example, C996
regarding references). metal derived from commercial natural uranium, recovered
uranium, or uranium obtained from the blending of highly
2. Referenced Documents
enriched uranium with commercial natural uranium, recovered
2
uranium, or depleted uranium.
2.1 ASTM Standards:
C696 Test Methods for Chemical, Mass Spectrometric, and 3.1.3 commercial natural uranium—any form of natural
235
Spectrochemical Analysis of Nuclear-Grade Uranium Di-
unirradiated uranium (containing 0.711 + 0.004 g U per 100
g U).
3.1.4 depleted uranium—any form of unirradiated uranium
1
This specification is under the jurisdiction of ASTM Committee C26 on
235
with a U content less than commercial natural uranium.
Nuclear Fuel Cycle and is the direct responsibility of Subcommittee C26.02 on Fuel
and Fertile Material Specifications.
Current edition approved Jan. 1, 2008. Published February 2008. Originally
approved in 2000. Last previous edition approved in 2005 as C1462 – 00 (2005).
3
DOI: 10.1520/C1462-00R08. Available fromAmerican National Standards Institute (ANSI), 25 W. 43rd St.,
2
For referenced ASTM standards, visit the ASTM website, www.astm.org, or 4th Floor, New York, NY 10036, http://www.ansi.org.
4
contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM AvailablefromU.S.GovernmentPrintingOfficeSuperintendentofDocuments,
Standards volume information, refer to the standard’s Document Summary page on 732 N. Capitol St., NW, Mail Stop: SDE, Washington, DC 20401, http://
the ASTM website. www.access.gpo.gov.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
1

---------------------- Page: 1 ----------------------
C1462−00 (2008)
3.1.5 highly enriched uranium—any form of uranium hav-
Rare Earths (Dy, Eu, Gd, Sm) 3.0
235
ing a U content equal to or in excess of 20 %.
The sum of the impurity ele
...

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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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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.  
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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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14  
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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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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4.5.1 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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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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SCOPE
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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4.3 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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