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ASTM C1854-17

(Test Method)

Standard Test Method for Determination of Hydrogen (total from all sources) in Mixed Oxide ((U, Pu)O2) Sintered Pellets by the Inert Gas Fusion Technique Followed by Thermal Conductivity Measurement

Standard Test Method for Determination of Hydrogen (total from all sources) in Mixed Oxide ((U, Pu)O<inf>2</inf>) Sintered Pellets by the Inert Gas Fusion Technique Followed by Thermal Conductivity Measurement

SIGNIFICANCE AND USE
5.1 MOX is used as a nuclear-reactor fuel. This test method is designed to determine whether the hydrogen content of the pellets meet the requirements of fuel specification. Examples of these requirements are given in Specification C833. Other requirements may apply based on agreements between the supplier and the customer.  
5.2 This method is suitable for all sintered MOX pellets containing up to 15 weight % PuO2 when the UO2 and PuO2 meet the requirements of Specifications C753 and C757. The method uncertainty is related to the concentration of the hydrogen in the sample. At lower concentrations, the relative uncertainty increases. At hydrogen contents close to the typical hydrogen content specification limit (1.3 μg hydrogen/g U + Pu metal); the combined relative uncertainty at the 95 % confidence level (k = 2) is approximately 30 %.
SCOPE
1.1 This test method covers the determination of hydrogen in nuclear-grade mixed oxides of uranium and plutonium ((U, Pu)O2) sintered fuel pellets. This test method is an alternative to Test Method C698 for the determination of moisture in nuclear-grade sintered mixed oxide (MOX) fuel pellets. Test Method C698 describes the detection of moisture in mixed oxides using a coulometric, electrolytic moisture analyzer. Although the main source of H2 in the fuel pellets is moisture, there could be other sources. The MOX pellet Specification C833 specifies a limit for hydrogen from all sources, not only moisture. The inert gas fusion followed by thermal conductivity detector specified in this test method allows for detection of hydrogen from all sources. Therefore, this test method can be used to determine the limit specified in C833.  
1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
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 and health 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.

General Information

Status
Published
Publication Date
31-May-2017
ICS
27.120.30 - Fissile materials and nuclear fuel technology
Technical Committee
C26 - Nuclear Fuel Cycle
Drafting Committee
C26.05 - Methods of Test
Current Stage
Ref Project

Relations

Refers
ASTM C753-16 - Standard Specification for Nuclear-Grade, Sinterable Uranium Dioxide Powder
Effective Date
01-Feb-2016
Refers
ASTM C757-16 - Standard Specification for Nuclear-Grade Plutonium Dioxide Powder for Light Water Reactors
Effective Date
01-Apr-2016
Refers
ASTM C698-16 - Standard Test Methods for Chemical, Mass Spectrometric, and Spectrochemical Analysis of Nuclear-Grade Mixed Oxides ((U, Pu)O<inf>2</inf>)
Effective Date
01-Jun-2016
Refers
ASTM C859-14a - Standard Terminology Relating to Nuclear Materials
Effective Date
15-Jun-2014

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ASTM C1854-17 - Standard Test Method for Determination of Hydrogen (total from all sources) in Mixed Oxide ((U, Pu)O<inf>2</inf>) Sintered Pellets by the Inert Gas Fusion Technique Followed by Thermal Conductivity MeasurementASTM C1854-17 - Standard Test Method for Determination of Hydrogen (total from all sources) in Mixed Oxide ((U, Pu)O<inf>2</inf>) Sintered Pellets by the Inert Gas Fusion Technique Followed by Thermal Conductivity Measurement
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ASTM C1854-17 - Standard Test Method for Determination of Hydrogen (total from all sources) in Mixed Oxide ((U, Pu)O<inf>2</inf>) Sintered Pellets by the Inert Gas Fusion Technique Followed by Thermal Conductivity Measurement
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Standards Content (Sample)

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.
Designation: C1854 − 17
Standard Test Method for
Determination of Hydrogen (total from all sources) in Mixed
Oxide ((U, Pu)O ) Sintered Pellets by the Inert Gas Fusion
2
1
Technique Followed by Thermal Conductivity Measurement
This standard is issued under the fixed designation C1854; 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 2. Referenced Documents
2
2.1 ASTM Standards:
1.1 This test method covers the determination of hydrogen
C698 Test Methods for Chemical, Mass Spectrometric, and
in nuclear-grade mixed oxides of uranium and plutonium ((U,
Spectrochemical Analysis of Nuclear-Grade Mixed Ox-
Pu)O ) sintered fuel pellets. This test method is an alternative
2
ides ((U, Pu)O )
2
to Test Method C698 for the determination of moisture in
C753 Specification for Nuclear-Grade, Sinterable Uranium
nuclear-grade sintered mixed oxide (MOX) fuel pellets. Test
Dioxide Powder
Method C698 describes the detection of moisture in mixed
C757 Specification for Nuclear-Grade Plutonium Dioxide
oxides using a coulometric, electrolytic moisture analyzer.
Powder for Light Water Reactors
Although the main source of H in the fuel pellets is moisture,
2
C833 Specification for Sintered (Uranium-Plutonium) Diox-
there could be other sources. The MOX pellet Specification
ide Pellets for Light Water Reactors
C833 specifies a limit for hydrogen from all sources, not only
C859 Terminology Relating to Nuclear Materials
moisture. The inert gas fusion followed by thermal conductiv-
C1068 Guide for Qualification of Measurement Methods by
ity detector specified in this test method allows for detection of
a Laboratory Within the Nuclear Industry
hydrogen from all sources. Therefore, this test method can be
3. Terminology
used to determine the limit specified in C833.
3.1 For definitions of terms used in this test method but not
1.2 The values stated in SI units are to be regarded as
defined herein, refer to Terminology C859.
standard. No other units of measurement are included in this
3.2 Definitions of Terms Specific to This Standard:
standard.
3.2.1 MOX—nuclearfuelcomposedofamixtureofuranium
1.3 This standard does not purport to address all of the
and plutonium oxides ((U, Pu)O ).
2
safety concerns, if any, associated with its use. It is the
3.2.2 reference material—material traceable to a reference
responsibility of the user of this standard to establish appro-
material from a national standards body such as the U.S.
priate safety and health practices and determine the applica-
National Institute for Standards and Technology (NIST) or
bility of regulatory limitations prior to use.
equivalent.
1.4 This international standard was developed in accor-
3.2.3 sintering—to increase the bonding in a mass of pow-
dance with internationally recognized principles on standard-
der or a compact by heating below the melting point of the
ization established in the Decision on Principles for the
main constituent.
Development of International Standards, Guides and Recom-
3.3 Acronyms:
mendations issued by the World Trade Organization Technical
3.3.1 LIMS—Laboratory Information Management System
Barriers to Trade (TBT) Committee.
3.3.2 TCD—Thermal Conductivity Detector
4. Summary of Test Method
4.1 Themethodforthedeterminationoftotalhydrogen(H )
2
fromallsourcespresentedinthistestmethodconsistsoffusion
1
This test method is under the jurisdiction ofASTM Committee C26 on Nuclear
2
Fuel Cycle and is the direct responsibility of Subcommittee C26.05 on Methods of For referenced ASTM standards, visit the ASTM website, www.astm.org, or
Test. contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM
Current edition approved June 1, 2017. Published June 2017. DOI: 10.1520/ Standards volume information, refer to the standard’s Document Summary page on
C1854-17. the ASTM website.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
1

---------------------- Page: 1 ----------------------
C1854 − 17
of the sample in an electrode impulse furnace in a stream of uncertainty increases.At hydrogen contents close to the typical
nitrogen (N ) or argon (Ar) gas at a temperature sufficient to hydrogencontentspecificationlimit(1.3µghydrogen/gU+Pu
2
release all hydrogen in the sample. The stream of gas carries metal); the combined relative uncertainty at the 95 % confi-
the released hydrogen through a series of filters to remove dence lev
...

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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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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.  
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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.
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:  
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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.
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1.1 This terminology standard contains terms, definitions, descriptions of terms, nomenclature, and explanations of acronyms and symbols specifically associated with standards under the jurisdiction of Committee C26 on Nuclear Fuel Cycle. The content of this terminology standard may also be applicable to documents not under the jurisdiction of Committee C26, in which case this terminology standard may be referenced in those documents.  
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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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.  
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1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
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ASTM C1204-14(2023)(Test Method)
Standard Test Method for Uranium in Presence of Plutonium by Iron(II) Reduction in Phosphoric Acid Followed by Chromium(VI) Titration

SIGNIFICANCE AND USE
4.1 Factors governing selection of a method for the determination of uranium include available quantity of sample, sample purity, desired level of reliability, and equipment availability.  
4.2 This test method is suitable for samples between 20 mg to 300 mg of uranium, is applicable to fast breeder reactor (FBR)-mixed oxides having a uranium to plutonium ratio of 2.5 and greater, is tolerant towards most metallic impurity elements usually specified for FBR-mixed oxide fuel, and uses no special equipment.  
4.3 The ruggedness of the titration method has been studied for both the volumetric (6) and the weight (7) titration of uranium with dichromate.
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1.1 This test method covers unirradiated uranium-plutonium mixed oxide having a uranium to plutonium ratio of 2.5 and greater. The presence of larger amounts of plutonium (Pu) that give lower uranium to plutonium ratios may give low analysis results for uranium (U) (1)2, if the amount of plutonium together with the uranium is sufficient to slow the reduction step and prevent complete reduction of the uranium in the allotted time. Use of this test method for lower uranium to plutonium ratios may be possible, especially when 20 mg to 50 mg quantities of uranium are being titrated rather than the 100 mg to 300 mg in the study cited in Ref (1). Confirmation of that information should be obtained before this test method is used for ratios of uranium to plutonium less than 2.5.  
1.2 The amount of uranium determined in the data presented in Section 12 was 20 mg to 50 mg. However, this test method, as stated, contains iron in excess of that needed to reduce the combined quantities of uranium and plutonium in a solution containing 300 mg of uranium with uranium to plutonium ratios greater than or equal to 2.5. Solutions containing up to 300 mg uranium with uranium to plutonium ratios greater than or equal to 2.5 have been analyzed (1) using the reagent volumes and conditions as described in Section 10.  
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1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use. For specific hazard statements, see Section 8.  
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ASTM C1455-14(2023)(Test Method)
Standard Test Method for Nondestructive Assay of Special Nuclear Material Holdup Using Gamma-Ray Spectroscopic Methods

SIGNIFICANCE AND USE
5.1 Measurement results from this test method assists in demonstrating regulatory compliance in such areas as safeguards SNM inventory control, criticality control, waste disposal, and decontamination and decommissioning (D&D). This test method can apply to the measurement of holdup in process equipment or discrete items whose gamma-ray absorption properties may be measured or estimated. This method may be adequate to accurately measure items with complex distributions of radioactive and attenuating material, however, the results are subject to larger measurement uncertainties than measurements of less complex distributions of radioactive material.  
5.2 Scan—A scan is used to provide a qualitative indication of the extent, location, and the relative quantity of holdup. It can be used to plan or supplement the quantitative measurements.  
5.3 Nuclide Mapping—Nuclide mapping measures the relative isotopic composition of the holdup at specific locations. It can also be used to detect the presence of radionuclides that emit radiation which could interfere with the assay. Nuclide mapping is best performed using a high resolution detector (such as HPGe) for best nuclide and interference detection. If the holdup is not isotopically homogeneous at the measurement location, that measured isotopic composition will not be a reliable estimate of the bulk isotopic composition.  
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1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
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SIGNIFICANCE AND USE
5.1 Factors governing selection of a method for the determination of plutonium include available quantity of sample, sample purity, desired level of reliability, and equipment.  
5.1.1 This test method determines 5 mg to 20 mg of plutonium with prior dissolution using Practice C1168.  
5.1.2 This test method calculates plutonium mass fraction in solutions and solids using an electrical calibration based upon Ohm’s Law and the Faraday Constant.  
5.1.3 Chemical standards are used for quality control. When prior chemical separation of plutonium is necessary to remove interferences, the quality control standards should be included with each chemical separation batch (9).  
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.
SCOPE
1.1 This test method describes the determination of dissolved plutonium from unirradiated nuclear-grade (that is, high-purity) materials by controlled-potential coulometry. Controlled-potential coulometry may be performed in a choice of supporting electrolytes, such as 0.9 mol/L (0.9 M) HNO3, 1 mol/L (1 M) HClO4, 1 mol/L (1 M) HCl, 5 mol/L (5 M) HCl, and 0.5 mol/L (0.5 M) H2SO4. Limitations on the use of selected supporting electrolytes are discussed in Section 6. Optimum quantities of plutonium for this procedure are 5 mg to 20 mg.  
1.2 Plutonium-bearing materials are radioactive and toxic. Adequate laboratory facilities, such as gloved boxes, fume hoods, controlled ventilation, etc., along with safe techniques must be used in handling specimens containing these materials.  
1.3 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.  
1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
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 C1128-23(Guide)
Standard Guide for Preparation of Working Reference Materials for Use in Analysis of Nuclear Fuel Cycle Materials

SIGNIFICANCE AND USE
5.1 Certified reference materials (CRMs) prepared from nuclear materials are well characterized, traceable, and sufficiently homogenous and stable for their intended use. Usually they are certified using the most unbiased and precise measurement methods available, often with more than one laboratory being used on a national or international level. CRMs are at the top of the metrological hierarchy of reference materials. A graphical representation of a typical national nuclear measurement system is shown in Fig. 3.
FIG. 3 Typical National Nuclear Measurement System  
5.2 Working reference materials (WRMs) need to have quality characteristics that are similar to CRMs, although the rigor used to achieve those characteristics is not usually as stringent as for CRMs. Similarly, production of WRMs should be in accordance with applicable requirements of ISO 17034. Where possible, CRMs are typically used to calibrate the methods used for establishing reference values assigned to WRMs, thus providing traceability to CRMs as required by ISO/IEC 17025. A WRM is normally prepared for a specific application.  
5.3 Because of the importance of having highly reliable measurement data from nuclear material analysis, particularly for material control and accountability purposes, CRMs are used for calibration when available. However, CRMs prepared from nuclear materials are not always available for specific applications. Thus, there may be a need for a laboratory to prepare nuclear material WRMs to meet specific needs; for example, to match the matrix in process samples. In such cases, a WRM can be tailored to meet specific needs of a process or laboratory. Also, CRM supply may be too limited for use in the quantities needed for long-term, routine use. When properly prepared, WRMs will serve equally well as CRMs for most applications, and using WRMs will help preserve supplies of CRMs.  
5.4 Difficulties may be encountered in the preparation of RMs from nuclear materials becaus...
SCOPE
1.1 This guide covers the preparation and characterization of working reference materials (WRM) that are produced by a laboratory for its own use in the analysis of nuclear fuel cycle materials. Guidance is provided for proper planning, preparation, packaging, and storage; requirements for characterization; homogeneity and stability considerations; and value assignment. When traceability to SI is desired for a WRM, it will be achieved by a defined, statistically sound characterization process that is traceable to a certified value on a certified reference materials. While the guidance provided is generic for nuclear fuel cycle materials, detailed examples for some materials are provided in the appendixes.  
1.2 This guide does not apply to the production and characterization of certified reference materials (CRM). Refer to ISO 17034 and ISO Guide 35 for guidance on reference material production, characterization, certification, sale, and distribution requirements.  
1.3 The information provided by this guide is found in the following sections:    
Section  
Perform WRM Planning  
6  
Prepare and Process Materials  
7  
Packaging and Storage of Materials  
8  
Perform Homogeneity Study  
9  
Perform Stability Studies  
10  
Characterize Materials  
11  
Perform Uncertainty Analysis  
12  
Produce Documentation  
13  
Carry Out WRM Utilization and Monitoring  
14  
1.4 The values stated in SI units are to be regarded as standard. The non-SI units of molar, M, and normal, N, are also regarded as standard. Any non-SI units of measurement shown in parentheses are for information only.  
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...

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