iTeh StandardsiTeh Standards
EURUSD
Your shopping cart is empty.
ASTM

ASTM C776-00

(Specification)

Standard Specification for Sintered Uranium Dioxide Pellets

Standard Specification for Sintered Uranium Dioxide Pellets

SCOPE
1.1 This specification is for finished sintered uranium dioxide pellets. It applies to uranium dioxide pellets containing uranium of any 234U concentration for use in nuclear reactors.  
1.2 This specification does not include (a) provisions for preventing criticality accidents or (b) requirements for health and safety. Observance of this specification does not relieve the user of the obligation to be aware of and conform to all federal, state, and local regulations pertaining to possessing, shipping, processing, or using source or special nuclear material. Examples of U.S. Government documents are Code of Federal Regulations (Latest Edition), Title 10, Part 50, Title 10, Part 71, and Title 49, Part 173.  
1.3 The following precautionary 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 or regulatory limitations prior to use.

General Information

Status
Historical
Publication Date
09-Jan-2000
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 C776-06 - Standard Specification for Sintered Uranium Dioxide Pellets
Effective Date
01-Aug-2006
Referred By
ASTM C1347-08(2023) - Standard Practice for Preparation and Dissolution of Uranium Materials for Analysis
Effective Date
10-Jan-2000
Referred By
ASTM C1502-16 - Standard Test Method for Determination of Total Chlorine and Fluorine in Uranium Dioxide and Gadolinium Oxide
Effective Date
10-Jan-2000
Referred By
ASTM C1457-18 - Standard Test Method for Determination of Total Hydrogen Content of Uranium Oxide Powders and Pellets by Carrier Gas Extraction
Effective Date
10-Jan-2000
Referred By
ASTM C1453-19 - Standard Test Method for the Determination of Uranium by Ignition and the Oxygen to Uranium (O/U) Atomic Ratio of Nuclear Grade Uranium Dioxide Powders and Pellets
Effective Date
10-Jan-2000
Referred By
ASTM C1625-19 - Standard Test Method for Uranium and Plutonium Concentrations and Isotopic Abundances by Thermal Ionization Mass Spectrometry
Effective Date
10-Jan-2000
Referred By
ASTM C1408-16 - Standard Test Method for Carbon (Total) in Uranium Oxide Powders and Pellets By Direct Combustion-Infrared Detection Method
Effective Date
10-Jan-2000
Referred By
ASTM C1672-17 - Standard Test Method for Determination of Uranium or Plutonium Isotopic Composition or Concentration by the Total Evaporation Method Using a Thermal Ionization Mass Spectrometer
Effective Date
10-Jan-2000
Referred By
ASTM C1832-23 - Standard Test Method for Determination of Uranium Isotopic Composition by Modified Total Evaporation (MTE) Method Using Thermal Ionization Mass Spectrometer
Effective Date
10-Jan-2000
Referred By
ASTM C1517-16 - Standard Test Method for Determination of Metallic Impurities in Uranium Metal or Compounds by DC-Arc Emission Spectroscopy
Effective Date
10-Jan-2000
Referred By
ASTM C1474-19 - Standard Test Method for Analysis of Isotopic Composition of Uranium in Nuclear-Grade Fuel Material by Quadrupole Inductively Coupled Plasma-Mass Spectrometry
Effective Date
10-Jan-2000
Referred By
ASTM C1868-18 - Standard Practice for Ceramographic Preparation of UO<inf>2</inf> and Mixed Oxide (U,Pu)O<inf>2</inf> Pellets for Microstructural Analysis
Effective Date
10-Jan-2000
Referred By
ASTM C696-19 - Standard Test Methods for Chemical, Mass Spectrometric, and Spectrochemical Analysis of Nuclear-Grade Uranium Dioxide Powders and Pellets
Effective Date
10-Jan-2000
Referred By
ASTM C1647-20 - Standard Practice for Removal of Uranium or Plutonium, or both, for Impurity Assay in Uranium or Plutonium Materials
Effective Date
10-Jan-2000
Referred By
ASTM C1430-18 - Standard Test Method for Determination of Uranium, Oxygen to Uranium (O/U), and Oxygen to Metal (O/M) in Sintered Uranium Dioxide and Gadolinia-Uranium Dioxide Pellets by Atmospheric Equilibration
Effective Date
10-Jan-2000

Buy Standard

ASTM C776-00 - Standard Specification for Sintered Uranium Dioxide PelletsASTM C776-00 - Standard Specification for Sintered Uranium Dioxide Pellets
PreviousNext
Technical specification
ASTM C776-00 - Standard Specification for Sintered Uranium Dioxide Pellets
English language
4 pages
sale 15% off
Preview
sale 15% off
Preview

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 776 – 00
Standard Specification for
Sintered Uranium Dioxide Pellets
This standard is issued under the fixed designation C 776; 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 specification for
uranium dioxide pellets. It recognizes the diversity of manufacturing methods by which uranium
dioxide pellets are produced and the many special requirements for chemical and physical
characterization which may be imposed by the operating conditions to which the pellets will be
subjected in specific reactor systems. Therefore, it is anticipated that the purchaser may supplement
this specification with additional requirements for specific applications.
1. Scope C 859 Terminology Relating to Nuclear Materials
C 996 Specification for Uranium Hexafluoride Enriched to
1.1 This specification is for finished sintered uranium diox-
Less than 5 % U
ide pellets. It applies to uranium dioxide pellets containing
C 1233 Practice for Determining the Equivalent Boron
uranium of any U concentration for use in nuclear reactors.
Content of Nuclear Materials
1.2 This specification does not include (a) provisions for
E 105 Practice for Probability Sampling of Materials
preventing criticality accidents or (b) requirements for health
2.2 ANSI Standard:
andsafety.Observanceofthisspecificationdoesnotrelievethe
NQA-1 Quality Assurance Program Requirements for
user of the obligation to be aware of and conform to all federal,
Nuclear Facilities
state, and local regulations pertaining to possessing, shipping,
2.3 U.S. Government Documents:
processing, or using source or special nuclear material. Ex-
Code of Federal Regulations, Title 10, Part 50, Energy (10
amples of U.S. Government documents are Code of Federal
CFR 50) Domestic Licensing of Production and Utiliza-
Regulations(LatestEdition),Title10,Part50,Title10,Part71,
tion Facilities
and Title 49, Part 173.
Code of Federal Regulations, Title 10, Part 71, Packaging
1.3 The following precautionary caveat pertains only to the
and Transportation of Radioactive Material
technical requirements portion, Section 4, of this specification:
Code of Federal Regulations, Title 49, Part 173, General
This standard does not purport to address all of the safety
Requirements for Shipments and Packaging
concerns, if any, associated with its use. It is the responsibility
of the user of this standard to establish appropriate safety and
3. Terminology
health practices and determine the applicability or regulatory
3.1 Definitions—For definitions of terms, refer to Terminol-
limitations prior to use.
ogy C 859.
2. Referenced Documents
4. Technical Requirements
2.1 ASTM Standards:
4.1 Chemical Requirements—Allchemicalanalysesshallbe
C 696 Test Methods for Chemical, Mass Spectrometric, and
performed on portions of the representative sample prepared in
Spectrochemical Analysis of Nuclear-Grade Uranium Di-
2 accordance with Section 6.Analytical chemistry methods used
oxide Powders and Pellets
shall be as stated in Test Methods C 696 (latest edition) or
C 753 Specification for Nuclear-Grade, Sinterable Uranium
2 demonstrated equivalent as mutually agreed upon between the
Dioxide Powder
seller and the buyer.
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 Annual Book of ASTM Standards, Vol 14.02.
and Fertile Material Specifications. Available from American National Standards Institute, 11 W. 42nd St., 13th
Current edition approved Jan. 10, 2000. Published March 2000. Originally Floor, New York, NY 10036.
published as C 776 – 76. Last previous edition C 776 – 94. Available from Superintendent of Documents, U. S. Government Printing
Annual Book of ASTM Standards, Vol 12.01. Office, Washington, DC 20402.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.
C 776
4.1.1 Uranium Content— The uranium content shall be a reactor use, the above limitation on EBC does not apply. The
minimum of 87.7 weight % on a dry weight basis. (Dry weight EBC of each element shall be calculated individually using the
is defined as the sample weight minus the moisture content.) following equation:
4.1.2 Impurity Content— The impurity content shall not
EBC of element5~EBC factor!3~µg element/g uranium! (1)
exceed the individual element limit specified in Table 1 on a
4.3 Physical Characteristics:
uranium weight basis. The summation of the contribution of
4.3.1 Dimensions—The dimensions of the pellet shall be
each of the impurity elements listed inTable 1 shall not exceed
specified by the buyer. These shall include diameter, length,
1500 µg/g. If an element analysis is reported as “less than” a
perpendicularity, and, as required, other geometric parameters
given concentration, this “less than” value shall be used in the
including surface finish.
determination of total impurities.
4.3.2 Pellet Density— The density of sintered pellets shall
4.1.3 Stoichiometry— The oxygen-to-uranium ratio of sin-
be as specified by the buyer.The theoretical density for UO of
tered fuel pellets shall be within the range from 1.99 to 2.02.
natural isotopic content shall be considered as 10.96 g/cm .
4.1.4 Moisture Content— The moisture content limit is
Density measurements shall be made by the geometric method
included in the total hydrogen limit (see Table 1).
stated in the Specification C 753Annex, an immersion method
4.2 Nuclear Requirements:
or by a demonstrated equivalent method as mutually agreed
4.2.1 Isotopic Content:
upon between the buyer and the seller.
4.2.1.1 For UO pellets with an isotopic content of U
4.3.3 Grain Size and Pore Morphology—The performance
between that of natural uranium and 5 %, the isotopic limits of
of UO fuel pellets is affected by their grain size and pore
Specification C 996 shall apply, unless otherwise agreed upon 2
morphology. These characteristics shall be mutually agreed
between the buyer and the seller. If the U content is greater
upon between the buyer and the seller.
thanenrichedcommercialgradeUF requirements,theisotopic
4.3.4 Pellet Integrity— Pellets shall be inspected to criteria
analysis requirements of Specification C 996 shall apply. The
which maintain adequate fuel performance and ensure that
specific isotopic measurements required by Specification
excessive breakage will not occur during fuel-rod loading.
C 996 may be waived, provided that the seller can demonstrate
Acceptable test methods include a visual (13) comparison
compliance with Specification C 996, for instance, through the
withpelletstandardsorothermethods,forexample,loadability
seller’s quality assurance records. U content greater than one
tests, approved by both the buyer and the seller.
specified in Specification C 996 for Commercial grade UF
4.3.4.1 Surface Cracks— The suggested limits for surface
may be agreed between the buyer and the seller since it is not
cracks are defined as follows:
a safety concern.
4.2.1.2 For UO pellets not having an assay in the range set (1) Axial Cracks, including those leading to the Pellet
Ends— ⁄2 the pellet length.
forth in 4.2.1.1, the isotopic requirements shall be as agreed
upon between
...

Questions, Comments and Discussion

Ask us and Technical Secretary will try to provide an answer. You can facilitate discussion about the standard in here.

This May Also Interest You

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

SIGNIFICANCE AND USE
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.
SCOPE
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.  
1.3 The fission product nuclides to be measured are 106Ru/106Rh, 103Ru, 137Cs, 144Ce, 144Pr, 141Ce, 95Zr, 95Nb, and 125Sb. Among the uranium decay product nuclides that may be measured is 231Pa. Other gamma energy-emitting fission and uranium decay nuclides present in the spectrum at detectable levels should be identified and quantified as required by the data quality objectives.  
1.4 The values stated in SI units are to be regarded as standard. Additionally, the non-SI units of kiloelectron volts and megaelectron volts are to be regarded as standard. No other units of measurement are included in this standard.  
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.

  • Standard
    5 pages
    English language
    sale 15% off
  • Standard
    5 pages
    English language
    sale 15% off
ASTM
ASTM C859-24(Terminology)
Standard Terminology Relating to Nuclear Materials

SCOPE
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.  
1.3 In general, technical terms that are defined in common dictionaries would not also be defined in this terminology standard unless there is a need to emphasize a specific definition in making appropriate use of a Committee C26 standard.  
1.4 Subcommittee C26.10 (Nondestructive Assay) also has a terminology standard applicable to its standards: Terminology C1673.  
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.

  • Standard
    6 pages
    English language
    sale 15% off
  • Standard
    6 pages
    English language
    sale 15% off
ASTM
ASTM C1672-23(Test Method)
Standard Test Method for Determination of the Uranium, Plutonium or Americium Isotopic Composition or Concentration by the Total Evaporation Method Using a Thermal Ionization Mass Spectrometer

SIGNIFICANCE AND USE
5.1 The total evaporation method is used to measure the isotopic composition of uranium, plutonium, and americium materials, and may be used to measure the elemental concentrations of these elements when employing the IDMS technique.  
5.2 Uranium and plutonium compounds are used as nuclear reactor fuels. In order to be suitable for use as a nuclear fuel the starting material must meet certain criteria, such as found in Specifications C757, C833, C753, C776, C787, C967, C996, or as specified by the purchaser. The uranium concentration, plutonium concentration, or both, and isotope abundances are measured by TIMS following this method.  
5.3 Americium-241 is the decay product of 241Pu isotope. The abundance of the 241Am isotope together with the abundance of the 241Pu parent isotope can be used to estimate radio-chronometric age of the Pu material for nuclear forensic applications Ref (6). The americium concentration and isotope abundances are measured by TIMS following this method.  
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.  
5.5 The new generation of miniaturized ion counters allow extremely small samples, in the picogram range, to be measured via the total evaporation method. The method may be employed for measuring environmental or safeguards inspection samples containing nanogram quantities of uranium or plutonium. Very small loadings require special sample handling and careful evaluation of measurement uncertainties.  
5.6 Typical uranium analyses are conducted using sample loadings between 50 nanograms and 800 nanograms. For uranium isotope ratios the total evapo...
SCOPE
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.  
1.2 In principle, the total evaporation method should yield isotope ratios that do not require mass bias correction. In practice, samples may require this bias correction. Compared to the conventional TIMS method described in Test Method C1625, the total evaporation method is approximately two times faster, improves precision of the isotope ratio measurements by a factor of two to four, and utilizes smaller sample sizes. Compared to the C1625 method, the total evaporation method provides “major” isotope ratios 235U/238U, 240Pu/239Pu, and 241Am/243Am with improved accuracy.  
1.3 The total evaporation method is prone to biases in the “minor” isotope ratios (233U/238U, 234U/238U, and 236U/238U ratios for uranium materials and 238Pu/239Pu, 241Pu/239Pu, 242Pu/239Pu, and 244Pu/239Pu ratios for plutonium materials) due to peak tailing from adjacent major isotopes. The magnitude of the absolute bias is dependent on measurement and instrumental characteristics. The relative bias, however, depends on the relative isotopic abundances of the sample. The use of an electron multiplier equipped with an energy filter may eliminate or diminish peak tailing effects. Measurement of the abundance sensitivity of the instrument m...

  • Standard
    22 pages
    English language
    sale 15% off
  • Standard
    22 pages
    English language
    sale 15% off
ASTM
ASTM C1432-23(Test Method)
Standard Test Method for Determination of Impurities in Plutonium: Acid Dissolution, Ion Exchange Matrix Separation, and Inductively Coupled Plasma-Atomic Emission Spectroscopic (ICP/AES) Analysis

SIGNIFICANCE AND USE
5.1 This test method can be used on plutonium matrices in nitrate solutions.  
5.2 This test method has been validated for all elements listed in Test Methods C757 except sulfur (S) and tantalum (Ta).  
5.3 This test method has been validated for all of the cation elements measured in Table 1. Phosphorus (P) requires a vacuum or an inert gas purged optical path instrument.
SCOPE
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.  
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. Some specific hazards statements are given in Section 9 on Hazards.  
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.

  • Standard
    5 pages
    English language
    sale 15% off
  • Standard
    5 pages
    English language
    sale 15% off
ASTM
ASTM C1807-15(2023)(Guide)
Standard Guide for Nondestructive Assay of Special Nuclear Material (SNM) Holdup Using Passive Neutron Measurement Methods

SIGNIFICANCE AND USE
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.  
5.2 Quantitative Measurements—These measurements result in quantification of the mass of SNM in the holdup. They include all the corrections and descriptive information, such as isotopic composition, that are available.  
5.2.1 High-quality results require detailed knowledge of radiation sources and detectors, radiation transport, calibration, facility operations, and error analysis. Consultation with qualified NDA personnel is recommended (Guide C1490).  
5.2.2 Holdup estimates for a single piece of process equipment or piping often include some compilation of multiple measurements. The holdup estimate must appropriately combine the results of each individual measurement. In addition, uncertainty estimates for each individual measurement must be made and appropriately combined.  
5.3 Scan—Radiation scanning, typically gamma, may be used to provide a qualitative description of the extent, location, and the relative quantity of holdup. It can be used to plan or supplement the quantitative neutron measurements. Other indicators (for example, visual) may also indicate a need for a holdup measurement.  
5.4 Nuclide Mapping—To appropriately interpret the neutron data, the specific neutron yield is needed. Isotopic measurements to determine the relative isotopic composition of the holdup at specific locations may be required, depending on the facility.  
5.5 Spot Check an...
SCOPE
1.1 This guide describes passive neutron measurement methods used to nondestructively estimate the amount of neutron-emitting special nuclear material compounds remaining as holdup in nuclear facilities. Holdup occurs in all facilities in which nuclear material is processed. Material may exist, for example, in process equipment, in exhaust ventilation systems, and in building walls and floors.  
1.1.1 The most frequent uses of passive neutron holdup techniques are for the measurement of uranium or plutonium deposits in processing facilities.  
1.2 This guide includes information useful for management, planning, selection of equipment, consideration of interferences, measurement program definition, and the utilization of resources.  
1.3 Counting modes include both singles (totals) or gross counting and neutron coincidence techniques.  
1.3.1 Neutron holdup measurements of uranium are typically performed on neutrons emitted during (α, n) reactions and spontaneous fission using singles (totals) or gross counting. While the method does not preclude measurement using coincidence or multiplicity counting for uranium, measurement efficiency is generally not sufficient to permit assays in reasonable counting times.  
1.3.2 For measurement of plutonium in gloveboxes, installed measurement equipment may provide sufficient efficiency for performing counting using neutron coincidence techniques in reasonable counting times.  
1.4 The measurement of nuclear material holdup in process equipment requires a scientific knowledge of radiation sources and detectors, radiation transport, modeling methods, calibration, facility operations, and uncertainty analysis. It is subject to the constraints of the facility, management, budget, and schedule, plus health and safety requirements, as well as the laws of physics. This guide does not purport to instruct the NDA practitioner on these principles.  
1.5 The measurement process includes...

  • Guide
    12 pages
    English language
    sale 15% off
ASTM
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.  
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.
SCOPE
1.1 This practice is a compilation of dissolution techniques for plutonium materials that are applicable to the test methods used for characterizing these materials. Dissolution treatments for the major plutonium materials assayed for plutonium or analyzed for other components are listed. Aliquots of the dissolved materials are dispensed on a mass basis when one of the analyses must be of high precision, such as plutonium assay; otherwise they are dispensed on a volume basis.  
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.

  • Standard
    10 pages
    English language
    sale 15% off
  • Standard
    10 pages
    English language
    sale 15% off
ASTM
ASTM C1268-23(Test Method)
Standard Test Method for Quantitative Determination of 241Am in Plutonium by Gamma-Ray Spectrometry

SIGNIFICANCE AND USE
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).  
5.3 When the plutonium solution contains unacceptable levels of fission products or other materials, this method may be used following a tri-n-octylphosphine oxide (TOPO) extraction, ion exchange or other similar separation techniques (see Test Methods C758 and C759).  
5.4 This test method is less subject to interferences from plutonium than alpha counting since the energy of the gamma ray used for the analysis is better resolved from other gamma rays than the alpha particle energies used for alpha counting.  
5.5 The minimal sample preparation reduces the amount of sample handling and exposure to the analyst.  
5.6 This test method is applicable only to homogeneous solutions. This test method is not suitable for solutions containing solids.  
5.7 Solutions containing 241Am at concentrations as little as 1 × 10−5 g/L may be analyzed using this method. The lower limit depends on the detector used and the counting geometry. Solutions containing high concentrations may be analyzed following an appropriate dilution.
SCOPE
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.  
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.  
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.

  • Standard
    6 pages
    English language
    sale 15% off
  • Standard
    6 pages
    English language
    sale 15% off
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.

  • Standard
    4 pages
    English language
    sale 15% off
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...

  • Technical specification
    5 pages
    English language
    sale 15% off
  • Technical specification
    5 pages
    English language
    sale 15% off
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...

  • Guide
    17 pages
    English language
    sale 15% off
  • Guide
    17 pages
    English language
    sale 15% off