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ASTM C967-87(1996)

(Specification)

Standard Specification for Uranium Ore Concentrate

Standard Specification for Uranium Ore Concentrate

SCOPE
1.1 This specification covers uranium ore concentrate containing a minimum of 65 weight% uranium.  
1.2 This specification does not include requirements for health and safety. Observance of this standard does not relieve the user of the obligation to be aware of and conform to all applicable international, national, state, and local regulations pertaining to possessing, shipping, or using source nuclear material (see 2.2).  
1.3 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.

General Information

Status
Historical
Publication Date
31-Dec-1991
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

Replaced By
ASTM C967-02 - Standard Specification for Uranium Ore Concentrate
Effective Date
10-Feb-2002

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ASTM C967-87(1996) - Standard Specification for Uranium Ore ConcentrateASTM C967-87(1996) - Standard Specification for Uranium Ore Concentrate
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ASTM C967-87(1996) - Standard Specification for Uranium Ore Concentrate
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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NOTICE: This standard has either been superseded and replaced by a new version or
withdrawn.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: C 967 – 87 (Reapproved 1996)
AMERICAN SOCIETY FOR TESTING AND MATERIALS
100 Barr Harbor Dr., West Conshohocken, PA 19428
Reprinted from the Annual Book of ASTM Standards. Copyright ASTM
Standard Specification for
1
Uranium Ore Concentrate
This standard is issued under the fixed designation C 967; 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 (e) 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
ore concentrate. Material conforming to this specification will generally meet the requirements for
conversion to uranium hexafluoride. However, the converter may relax or supplement this specifica-
tion upon mutual agreement with the customer.
1. Scope 2.3 ANSI Standard:
ANSI/ASME NQA-1 Quality Assurance Requirements for
1.1 This specification covers uranium ore concentrate con-
5
Nuclear Facilities
taining a minimum of 65 weight % uranium.
1.2 This specification does not include requirements for
3. Terminology Definitions
health and safety. Observance of this specification does not
3.1 Except as otherwise defined herein, definitions of terms
relieve the user of the obligation to be aware of and conform to
are as given in Terminology C 859.
all applicable international, national, state, and local regula-
tions pertaining to possessing, shipping, or using source
4. Chemical Composition
nuclear material (see 2.2).
4.1 Uranium Content—The uranium content, as received,
1.3 The values stated in SI units are to be regarded as the
shall be a minimum of 65 weight %.
standard. The values given in parentheses are for information
4.2 Isotopic Content—The isotopic content shall be that of
only.
235
naturally occurring uranium (0.7105 to 0.7115 % U).
4.3 Insoluble Uranium—The uranium insoluble in nitric
2. Referenced Documents
acid shall be a maximum of 0.10 weight % on a uranium basis.
2.1 ASTM Standards:
4.4 Extractable Organic—The extractable organic shall be a
2
C 859 Terminology Relating to Nuclear Materials
maximum of 0.10 weight % on an as-received basis of an
C 1022 Test Methods for Chemical and Atomic Absorption
undried sample.
2
Analysis of Uranium-Ore Concentrate
4.5 Impurity Content—The impurity content shall be less
2
C 1075 Practices for Sampling Uranium-Ore Concentrate
than the maximum limits specified in Table 1.
2.2 U.S. Government Documents:
Nuclear Materials Licensing Code of Federal Regulations
5. Physical Properties
(latest edition) Title 10, Chapter 1, Nuclear Regulatory
5.1 Particle Size—All of a representative sample (Section
3
Commission
1
6) shall pass through a sieve with an aperature of 6.35 mm ( ⁄4
Nuclear Materials Licensing Code of Federal Regulations,
in.).
Title 49, Transportation Chapter 1, Materials Transporta-
5.2 Ability to Flow—Concentrate shall be sufficiently free-
4
tion Bureau
flowing to permit sampling.
Nuclear Materials Licensing Code of Federal Regulations,
5.3 Foreign Matter—Concentrate shall be free of all mate-
Energy Part 50 (10CFR 50) Licensing of Domestic Pro-
rials and objects that: (a) are not produced as a constituent of
3
duction and Utilization Facilities
concentrates in the milling of uranium ore, or, (b) would or
could be detrimental to the sampling of concentrates or to the
1 equipment used in such sampling.
This specification is under the jurisdiction of ASTM Committee C-26 on
Nuclear Fuel Cycle and is the direct responsibility of Subcommittee C26.02 on Fuel
6. Sampling
and Fertile Material Specifications.
Current edition approved May 29, 1987. Published July 1987. Originally
6.1 The lot size and number of tests (and retests when
published as C 967 – 81. Last previous edition C 967 – 81.
2
required) shall be as mutually agreed.
Annual Book of ASTM Standards, Vol 12.01.
3
Available from the Nuclear Regulatory Commission, 1717 H Street, N. W.,
Washington, DC 20555.
4 5
Available from the Materials Transportation Bureau, 400 Seventh St., Wash- Available from American National Standards Institute, 11 W. 42nd St., 13th
ington, DC, 20590. Floor, New York, NY 10036.
1

---------------------- Page: 1 ----------------------
C 967
TABLE 1 Impurities and Maximum Concentration Limits
7.2 Test Methods C 1022 is referenced as a guide.
Maximum Concentration Limit (Wt %,
A
Uranium Basis)
8. Certification
Impurity
Limit Without Limit Without
8.1 When specified in the purchase order or contract,
B C
Penalty Rejection
certification shall be furnished that the material was sampled
As 0.05 0.10
and tested in accordance with this specification and has been
B 0.005 0.10
Ca 0.05 1.00 found to meet the requirements. When specifi
...

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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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5.6 This test method is applicable only to homogeneous solutions. This test method is not suitable for solutions containing solids.  
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ASTM C1168-23(Practice)
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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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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4  
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12  
5  
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13  
6  
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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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Standard Test Method for Isotopic Analysis of Uranium Hexafluoride by Single–Standard Gas Source Multiple Collector Mass Spectrometer Method

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5.1 Uranium hexafluoride is a basic material used to produce nuclear reactor fuel. To be suitable for this purpose, the material must meet criteria for isotopic composition. This test method is designed to determine whether the material meets the requirements described in Specifications C787 and C996.
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ASTM C833-23(Specification)
Standard Specification for Sintered (Uranium-Plutonium) Dioxide Pellets for Light Water Reactors

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 The purpose of this guide is to provide information that will help to ensure that nuclear fuel dissolution facilities are conceived, designed, fabricated, constructed, and installed in an economic and efficient manner. This guide will help facilities meet the intended performance functions, eliminate or minimize the possibility of nuclear criticality and provide for the protection of both the operator personnel and the public at large under normal and abnormal (emergency) operating conditions as well as under credible failure or accident conditions.
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1.2 Applicability and Exclusions:  
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Standard Guide for Materials Handling Equipment for Hot Cells

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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 This service imposes stringent requirements on the quality and the integrity of the equipment, as follows:  
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.  
4.5.3 Materials handling equipment should be designed and fabricated to remain dimensionally stable throughout its life cycle.  
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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.  
1.1.3 This guide may apply to materials handling equipment in other radioactive remotely operated facilities such as suited entry repair areas and canyons, but does not apply to materials handling equipment used in commercial power reactors.  
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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.  
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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
    6 pages
    English language
    sale 15% off