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ASTM C787-96

(Specification)

Standard Specification for Uranium Hexafluoride for Enrichment

Standard Specification for Uranium Hexafluoride for Enrichment

SCOPE
1.1 This specification applies to uranium hexafluoride containing uranium of  235 U concentration not exceeding 5%  235 U intended for feed to an enrichment plant. This specification includes specifications for natural uranium hexafluoride (UF6) and UF6 that has been derived from irradiated uranium which has been reprocessed, and converted to UF6 for enrichment and subsequent reuse. The objectives of this specification are twofold: ( ) To define the impurity and uranium isotope limits for commercial natural UF6 feedstock so that the corresponding enriched product is essentially equivalent to enriched product made entirely from virgin natural UF6 and ( ) To define additional limits for reprocessed UF6 (or any mixture of reprocessed and commercial natural). For such UF6, special provisions may need to be made to ensure that no extra hazard arises to the employees, process equipment, or the environment.
1.2 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 international, federal, state, and local regulations for processing, shipping, or in any other way using UF6. (see, for example, TID-7016, DP-532, ORNL-NUREGCSD-6, and DOE 5633.3).  
1.3 The values stated in SI units are to be regarded as the standard. The values given in parentheses are given for information only.

General Information

Status
Historical
Publication Date
31-Dec-1995
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 C787-03e1 - Standard Specification for Uranium Hexafluoride for Enrichment
Effective Date
10-Jun-2003
Referred By
ASTM C1052-20 - Standard Practice for Bulk Sampling of Liquid Uranium Hexafluoride
Effective Date
01-Jan-1996
Referred By
ASTM C1842-16 - Standard Test Method for The Analysis of Boron and Silicon in Uranium Hexfluoride via Fourier-Transform Infrared (FTIR) Spectroscopy
Effective Date
01-Jan-1996
Referred By
ASTM C1429-21 - Standard Test Method for Isotopic Analysis of Uranium Hexafluoride by Double-Standard Multi-Collector Gas Mass Spectrometer
Effective Date
01-Jan-1996
Referred By
ASTM C1913-21 - Standard Practice for Sampling Gaseous Uranium Hexafluoride Using Zeolite in Single-Use Destructive Assay Sampler
Effective Date
01-Jan-1996
Referred By
ASTM C1295-15 - Standard Test Method for Gamma Energy Emission from Fission and Decay Products in Uranium Hexafluoride and Uranyl Nitrate Solution
Effective Date
01-Jan-1996
Referred By
ASTM C1477-19 - Standard Test Method for Isotopic Abundance Analysis of Uranium Hexafluoride and Uranyl Nitrate Solutions by Multi-Collector, Inductively Coupled Plasma-Mass Spectrometry
Effective Date
01-Jan-1996
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-Jan-1996
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-Jan-1996
Referred By
ASTM C1346-19 - Standard Practice for Dissolution of UF<inf>6</inf> from P-10 Tubes
Effective Date
01-Jan-1996
Referred By
ASTM C1703-18(2023) - Standard Practice for Sampling of Gaseous Uranium Hexafluoride for Enrichment
Effective Date
01-Jan-1996
Referred By
ASTM C1771-19 - Standard Test Method for Determination of Boron, Silicon, and Technetium in Hydrolyzed Uranium Hexafluoride by Inductively Coupled Plasma&#x2014;Mass Spectrometer After Removal of Uranium by Solid Phase Extraction
Effective Date
01-Jan-1996
Referred By
ASTM C1561-10(2016) - Standard Guide for Determination of Plutonium and Neptunium in Uranium Hexafluoride and U-Rich Matrix by Alpha Spectrometry
Effective Date
01-Jan-1996
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-Jan-1996
Referred By
ASTM C1883-19 - Standard Practice for Sampling of Gaseous Enriched Uranium Hexafluoride
Effective Date
01-Jan-1996

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ASTM C787-96 - Standard Specification for Uranium Hexafluoride for EnrichmentASTM C787-96 - Standard Specification for Uranium Hexafluoride for Enrichment
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ASTM C787-96 - Standard Specification for Uranium Hexafluoride for Enrichment
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NOTICE: This standard has either been superseded and replaced by a new version or withdrawn.
Please contact ASTM International (www.astm.org) for the latest information.
Designation: C 787 – 96
Standard Specification for
Uranium Hexafluoride for Enrichment
This standard is issued under the fixed designation C 787; 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 Fission Products in Uranium Hexafluoride
2.2 ANSI Standard:
1.1 This specification covers uranium hexafluoride (UF )
N14.1 Packaging of Uranium Hexafluoride for Transport
intended for feeding to an enrichment plant. Included are
2.3 U.S. Government Documents:
specifications for UF derived from unirradiated natural ura-
Inspection, Weighing, and Sampling of Uranium Hexafluo-
nium and UF derived from irradiated uranium that has been
ride Cylinders, Procedures for Handling and Analysis of
reprocessed and converted to UF for enrichment and subse-
Uranium Hexafluoride, Vol. 1, DOE Report ORO-671-1,
quent reuse. The objectives of this specification are twofold:
latest revision
(1) To define the impurity and uranium isotope limits for
Uranium Hexafluoride: A Manual of Good Handling Prac-
Commercial Natural UF feedstock so that the corresponding
tices, United States Enrichment Corporation Report
enriched uranium is essentially equivalent to enriched uranium
USEC-651, latest revision
made entirely from virgin natural UF ; and (2) To define
NuclearSafetyGuide,U.S.NuclearRegulatoryCommission
additional limits for Reprocessed UF (or any mixture of
Report TID-7016, Rev. 2, 1978, and ORNL-NUREG-
Reprocessed UF and Commercial Natural UF ). For such
6 6
CSD-6
UF , special provisions may be needed to ensure that no extra
Clarke, H. K., Handbook of Nuclear Safety, DOE Report
hazard arises to the work force, process equipment, or the
DP-532
environment.
Control and Accountability of Nuclear Materials, Basic
1.2 The scope of this specification does not comprehen-
Principles, U.S. DOE Order 5633.3
sively cover all provisions for preventing criticality accidents
or requirements for health and safety or for shipping. Obser-
3. Terminology
vance of this specification does not relieve the user of the
3.1 Definitions of Terms Specific to This Standard—Terms
obligation to conform to all international, federal, state, and
shall be defined in accordance with Terminology C 859C 859,
local regulations for processing, shipping, or in any other way
except for the following:
using UF (see, for example, TID-7016, DP-532, ORNL-
3.1.1 Commercial Natural UF —UF from natural unirra-
6 6
NUREG-CSD-6, and DOE 5633.3B).
diated uranium (containing 0.711 6 0.004 g U per 100 g U).
1.3 The values stated in SI units are to be regarded as the
3.1.1.1 Discussion—It is recognized that some contamina-
standard. The values given in parentheses are for information
tion with reprocessed uranium may occur during routine
only.
processing. This is acceptable provided that the UF meets the
2. Referenced Documents requirements for Commercial Natural UF .
3.1.2 Reprocessed UF —any UF made from uranium that
2.1 ASTM Standards: 6 6
has been exposed in a neutron irradiation facility and subse-
C 761 Test Methods for Chemical, Mass Spectrometric,
quently chemically separated from the fission products and
Spectrochemical, Nuclear, and Radiochemical Analysis of
transuranic isotopes so generated.
Uranium Hexafluoride
2 3.1.2.1 Discussion—The requirements for Reprocessed UF
C 859 Terminology Relating to Nuclear Materials
given in this specification are intended to be typical of
C 1052 Practice for Bulk Sampling of Liquid Uranium
reprocessed spent fuel that has achieved burnup levels of up to
Hexafluoride
50 000 MW days per tonne of uranium in light water reactors
C 1219 Test Methods forArsenic in Uranium Hexafluoride
and has been cooled for ten years after discharge. It is
C 1295 Test Method for Gamma Energy Emission from
1 3
This specification is under the jurisdiction of ASTM Committee C-26 on Available from American National Standards Institute, 11 W. 42nd St., 13th
Nuclear Fuel Cycle and is the direct responsibility of Subcommittee C26.02 on Fuel Floor, New York, NY 10036.
and Fertile Material Specifications. Available from Superintendent of Documents, U.S. Government Printing
Current edition approved July 10, 1996. Published September 1996. Originally Office, Washington, DC 20402.
published as C 787 – 76. Last previous edition C 787 – 90. Available from United States Enrichment Corporation, 6903 Rockledge Drive,
Annual Book of ASTM Standards, Vol 12.01. Bethesda, MD 20817.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.
NOTICE: This standard has either been superseded and replaced by a new version or withdrawn.
Please contact ASTM International (www.astm.org) for the latest information.
C787–96
recognized that different limits would be necessary to accom- 5. Chemical, Physical, and Isotopic Requirements
modate different fuel histories.
5.1 Plants preparing UF will have to control the purity of
processchemicalsandalsoemploylowcorrosionequipmentto
4. Safety, Health Physics, and Criticality Requirements
be successful in meeting the specifications for most impurities.
Both Commercial Natural UF and Reprocessed UF will have
4.1 The UF concentration shall be not less than 99.5 g UF 6 6
6 6
to meet the same specification criteria for most elements. In
per 100 g of sample in order to limit the potential hydrogen
addition, Reprocessed UF must meet additional specification
content for nuclear criticality safety. 6
limitsforartificiallycreatedradioactivespecies.Forevaluating
4.2 The total absolute vapor pressure shall not exceed the
Commercial Natural UF , the measured concentration of U
values given below:
will be used as an indicator for contamination with reprocessed
380 kPa at 80°C (55 psia at 176°F), or
uranium, on the assumption that there is no opportunity for
517 kPa at 93°C (75 psia at 200°F), or
862 kPa at 112°C (125 psia at 235°F) contamination with irradiated uranium that has not been
processed to remove the majority of fission products. Provided
Additionally, if a measurement is taken over solid UF , then
that this isotope does not exceed the concentration limit for
the vapor pressure shall not exceed the values given below:
Commercial Natural UF listed in 5.5, the expected concentra-
50 kPa at 20°C (7 psia at 68°F), or
tions of artificial isotopes would be so far below normal
69 kPa at 35°C (10 psia at 95°F)
detection limits that measurements to determine compliance
The purpose of the pressure check is to limit the hydrogen with the separate limits are not appropriate. Uranium hexafluo-
fluoride, air, or other volatile components that might cause ride that fails to meet Commercial Natural UF limits would
overpressure when heating the shipping container to obtain a require further testing to determine its acceptability as Repro-
liquid sample or withdraw the contents. cessed UF .
5.2 The UF content shall be reported as gUF /100 g
4.3 The total hydrocarbon, chlorocarbon, and partially sub- 6 6
sample.
stituted halohydrocarbon content shall not exceed 0.01 mol %
5.3 The total of all the following listed elements that form
of the UF . The reason for the exclusion
...

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ASTM C1295-24(Test Method)
Standard Test Method for Gamma Energy Emission from Fission and Decay Products in Uranium Hexafluoride and Uranyl Nitrate Solution

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5.1 Specific gamma-ray emitting radionuclides in UF6 are identified and quantified using a high-resolution gamma-ray energy analysis system, which includes a high-resolution germanium detector. This test method shall be used to meet the health and safety specifications of C787, C788, and C996 regarding applicable fission products in reprocessed uranium solutions. This test method may also be used to provide information to parties such as conversion facilities on the level of uranium decay products in such materials. Pa-231 is a specific uranium decay product that may be present in uranium ore concentrate and is amenable to analysis by gamma spectrometry.
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1.1 This test method covers the measurement of gamma energy emitted from fission products in uranium hexafluoride (UF6) and uranyl nitrate solution. This test method may also be used to measure the concentration of some uranium decay products. It is intended to provide a method for demonstrating compliance with UF6 Specifications C787 and C996, uranyl nitrate Specification C788, and uranium ore concentrate Specification C967.  
1.2 The lower limit of detection is estimated at 5000 MeV Bq/kg (MeV kg-1/s-1) of uranium and is the square root of the sum of the squares of the individual reporting limits of the nuclides to be measured. The limit of detection was determined on a pure, aged natural uranium (ANU) solution. The value is dependent upon the detector efficiency and background that can be achieved.  
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ASTM C859-24(Terminology)
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1.1 This terminology standard covers 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.  
1.2 While subcommittees within Committee C26 are free to only provide terms and definitions within individual standards, each subcommittee may request the addition of utilized terms and definitions to this terminology standard if it believes that such serves the broader interest of Committee C26 and the nuclear fuel cycle profession. Therefore, terms and definitions proposed for inclusion in Terminology C859 need not be used in more than one committee standard before being considered.  
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5.4 The total evaporation method allows for a wide range of sample loading with no significant change in precision or accuracy. The method is also suitable for trace-level loadings with some loss of precision and accuracy. The total evaporation method and modern instrumentation allow for the measurement of minor isotopes using ion counting detectors, while the major isotope(s) is(are) simultaneously measured using Faraday cup detectors.  
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1.1 This method describes the determination of the isotopic composition, or the concentration, or both, of uranium, plutonium, and americium as nitrate solutions by the total evaporation method using a thermal ionization mass spectrometer (TIMS) instrument. Purified uranium, plutonium, or americium nitrate solutions are deposited onto a metal filament and placed in the mass spectrometer. Under computer control, ion currents are generated by heating of the filament(s). The ion currents are continually measured until the whole deposited solution sample is exhausted. The measured ion currents are integrated over the course of the measurement and normalized to a reference isotope ion current to yield isotope ratios.  
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ASTM C1807-15(2023)(Guide)
Standard Guide for Nondestructive Assay of Special Nuclear Material (SNM) Holdup Using Passive Neutron Measurement Methods

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5.1 This guide assists in satisfying requirements in such areas as safeguards, SNM inventory control, nuclear criticality safety, waste disposal, and decontamination and decommissioning (D&D). This guide can apply to the measurement of holdup in process equipment or discrete items whose neutron production properties may be measured or estimated. These methods may meet target accuracy for items with complex distributions of SNM in the presence of moderators, absorbers, and neutron poisons; however, the results are subject to larger measurement uncertainties than measurements of less complex items.  
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1.1 This test method covers the determination of 25 elements in plutonium (Pu) materials. The Pu is dissolved in acid, the Pu matrix is separated from the target impurities by an ion exchange separation, and the concentrations of the impurities are determined by inductively coupled plasma-atomic emission spectroscopy (ICP-AES).  
1.2 This test method is specific for the determination of impurities in 8 M HNO3 solutions. Impurities in other plutonium materials, including plutonium oxide samples, may be determined if they are appropriately dissolved (see Practice C1168) and converted to 8 M HNO3  solutions.  
1.3 The values stated in SI units are to be regarded as standard. The values given in parentheses are mathematical conversions that are provided for information only and are not considered standard. Additionally, the non-SI units of molarity and centimeters of mercury are to be regarded as standard.  
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ASTM C1268-23(Test Method)
Standard Test Method for Quantitative Determination of 241Am in Plutonium by Gamma-Ray Spectrometry

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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.  
5.2 The 241Am in solid plutonium materials may be determined when these materials are dissolved (see Practice C1168).  
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5.5 The minimal sample preparation reduces the amount of sample handling and exposure to the analyst.  
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1.1 This test method covers the quantitative determination of 241Am by gamma-ray spectrometry in plutonium nitrate solution samples that do not contain significant amounts of radioactive fission products or other high specific activity gamma-ray emitters.  
1.2 This test method can be used to determine the 241Am in samples of plutonium metal, oxide and other solid forms, when the solid is appropriately sampled and dissolved.  
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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ASTM C1168-23(Practice)
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  
10  
3  
Dissolution of Plutonium Metal with Sulfuric Acid  
11  
4  
Dissolution of Plutonium Oxide and Uranium-Plutonium Mixed
Oxide by the Sealed-Reflux Technique  
12  
5  
Dissolution of Plutonium Oxide and Uranium-Plutonium Mixed Oxides by Sodium Bisulfate Fusion  
13  
6  
Dissolution of Uranium-Plutonium Mixed Oxides and Low-Fired Plutonium Oxide in Beakers  
14  
7  
Open-Vessel (with Reflux Condenser) Dissolution of Plutonium Oxide Powder  
15  
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.  
1.6 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.  
1.7 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 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.
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.  
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 %).  
1.3 There is no discussion of or provision for preventing criticality incidents, nor are health and safety requirements, the avoidance of hazards, or shipping precautions and controls discussed. 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; Code of Federal Regulations Title 10, Part 71—Packaging and Transportation of Radioactive Material; and Code of Federal Regulations Title 49, Part 173—General Requirements for Shipments and Packaging.  
1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.5 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, 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 Technic...

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ASTM
ASTM C1428-18(2023)(Test Method)
Standard Test Method for Isotopic Analysis of Uranium Hexafluoride by Single–Standard Gas Source Multiple Collector Mass Spectrometer Method

SIGNIFICANCE AND USE
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.
SCOPE
1.1 This test method is applicable to the isotopic analysis of uranium hexafluoride (UF6) with  235U concentrations less than or equal to 5 % and 234U,  236U concentrations of 0.0002 to 0.1 %.  
1.2 This test method may be applicable to the analysis of the entire range of 235U isotopic compositions providing that adequate Certified Reference Materials (CRMs or traceable standards) are available.  
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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.4 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 C1062-23(Guide)
Standard Guide for Design, Fabrication, and Installation of Nuclear Fuel Dissolution Facilities

SIGNIFICANCE AND USE
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.
SCOPE
1.1 It is the intent of this guide to set forth criteria and procedures for the design, fabrication and installation of nuclear fuel dissolution facilities. This guide applies to and encompasses all processing steps or operations beyond the fuel shearing operation (not covered), up to and including the dissolving accountability vessel.  
1.2 Applicability and Exclusions:  
1.2.1 Operations—This guide does not cover the operation of nuclear fuel dissolution facilities. Some operating considerations are noted to the extent that these impact upon or influence design.
1.2.1.1 Dissolution Procedures—Fuel compositions, fuel element geometry, and fuel manufacturing methods are subject to continuous change in response to the demands of new reactor designs and requirements. These changes preclude the inclusion of design considerations for dissolvers suitable for the processing of all possible fuel types. This guide will only address equipment associated with dissolution cycles for those fuels that have been used most extensively in reactors as of the time of issue (or revision) of this guide. (See Appendix X1.)  
1.2.2 Processes—This guide covers the design, fabrication and installation of nuclear fuel dissolution facilities for fuels of the type currently used in Pressurized Water Reactors (PWR). Boiling Water Reactors (BWR), Pressurized Heavy Water Reactors (PHWR) and Heavy Water Reactors (HWR) and the fuel dissolution processing technologies discussed herein. However, much of the information and criteria presented may be applicable to the equipment for other dissolution processes such as for enriched uranium-aluminum fuels from typical research reactors, as well as for dissolution processes for some thorium and plutonium-containing fuels and others. The guide does not address equipment design for the dissolution of high burn-up or mixed oxide fuels.
1.2.2.1 This guide does not address special dissolution processes that may require substantially different equipment or pose different hazards than those associated with the fuel types noted above. Examples of precluded cases are electrolytic dissolution and sodium-bonded fuels processing. The guide does not address the design and fabrication of continuous dissolvers.  
1.2.3 Ancillary or auxiliary facilities (for example, steam, cooling water, electrical services) are not covered. Cold chemical feed considerations are addressed briefly.  
1.2.4 Dissolution Pretreatment—Fuel pretreatment steps incidental to the preparation of spent fuel assemblies for dissolution reprocessing are not covered by this guide. This exclusion applies to thermal treatment steps such as “Voloxidation” to drive off gases prior to dissolution, to mechanical decladding operations or process steps associated with fuel elements disassembly and removal of end fittings, to chopping and shearing operations, and to any other pretreatment operations judged essential to an efficient nuclear fuels dissolution step.  
1.2.5 Fundamentals—This guide does not address specific chemical, physical or mechanical technology, fluid mechanics, stress analysis or other engineering fundamentals that are also applied in the creation of a safe design for nuclear fuel dissolution facilities.  
1.3 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI unit...

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