C26 - Nuclear Fuel Cycle

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.
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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.  
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.

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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.  
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.

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...
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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...

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.
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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.  
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.

SIGNIFICANCE AND USE
5.1 This practice may be used to determine concentrations of elements leached from nuclear waste materials (glasses, ceramics, cements) using an aqueous leachant. If the nuclear waste material is radioactive, a suitably contained and shielded ICP-AES spectrometer system with a filtered exit-gas system must be used, but no other changes in the practice are required. The leachant may be deionized water or any aqueous solution containing less than 1 % total solids.  
5.2 This practice as written is for the analysis of solutions containing 1 % nitric acid. It can be modified to specify the use of the same or another mineral acid at the same or higher concentration. In such cases, the only change needed in this practice is to substitute the preferred acid and concentration value whenever 1 % nitric acid appears here. It is important that the acid type and content of the reference and check solutions closely match the leachate solutions to be analyzed.  
5.3 This practice can be used to analyze leachates from static leach testing of waste forms using Test Method C1220.
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1.1 This practice is applicable to the determination of low concentration and trace elements in aqueous leachate solutions produced by the leaching of nuclear waste materials, using inductively coupled plasma-atomic emission spectroscopy (ICP-AES).  
1.2 The nuclear waste material may be a simulated (non-radioactive) solid waste form or an actual solid radioactive waste material.  
1.3 The leachate may be deionized water or any natural or simulated leachate solution containing less than 1 % total dissolved solids.  
1.4 This practice should be used by analysts experienced in the use of ICP-AES, the interpretation of spectral and non-spectral interferences, and procedures for their correction.  
1.5 No detailed operating instructions are provided because of differences among various makes and models of suitable ICP-AES instruments. Instead, the analyst shall follow the instructions provided by the manufacturer of the particular instrument. This test method does not address comparative accuracy of different devices or the precision between instruments of the same make and model.  
1.6 This practice contains notes that are explanatory and are not part of the mandatory requirements of the method.  
1.7 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.8 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.  
1.9 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.

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...
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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.  
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...

SIGNIFICANCE AND USE
4.1 Remote Viewing Components:  
4.2 The long-term applicability of a remotely operated radiological facility will be greatly affected by the provisions for remote viewing of normal and off-normal operations within the facility. The deployment of remote viewing systems can most efficiently be addressed during the design and construction phases.  
4.2.1 The purpose of this guide is to provide general guidelines for the design and operation of remote viewing equipment to ensure longevity and reliability throughout the period of service.  
4.2.2 It is intended that this guide record the general conditions and practices that experience has shown are necessary to minimize equipment failures and maximize the effectiveness and utility of remote viewing equipment. It is also intended to inform designers and engineers of those features that are highly desirable for the selection of equipment that has proven reliable in high radiation environments.  
4.2.3 This guide is intended as a supplement to other standards, and to federal and state regulations, codes, and criteria applicable to the design of equipment intended for hot cell use.  
4.2.4 This guide is intended to be generic and applies to a wide range of types and configurations of hot cell equipment and remote viewing systems.
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1.1 Intent:  
1.1.1 This guide establishes the minimum requirements for viewing systems for remotely operated facilities, including hot cells (shielded cells), used for the processing and handling of nuclear and radioactive materials. The intent of this guide is to aid in the design, selection, installation, modification, fabrication, and quality assurance of remote viewing systems to maximize their usefulness and to minimize equipment failures.  
1.1.2 It is intended that this guide record the principles and caveats that experience has shown to be essential to the design, fabrication, installation, maintenance, repair, replacement, and, decontamination and decommissioning of remote viewing equipment capable of meeting the stringent demands of operating, dependably and safely, in a hot cell environment where operator visibility is limited due to the radiation exposure hazards.  
1.1.3 This guide is intended to apply to methods of remote viewing for nuclear applications but may be applicable to any environment where remote operational viewing is desirable.  
1.2 Applicability:  
1.2.1 This guide applies to, but is not limited to, radiation hardened and non-radiation hardened cameras (black-and-white and color), lenses, camera housings and positioners, periscopes, through wall/roof viewing, remotely deployable cameras, crane/robot mounted cameras, endoscope cameras, borescopes, video probes, flexible probes, mirrors, lighting, fiber lighting, and support equipment.  
1.2.2 This guide is intended to be applicable to equipment used under one or more of the following conditions:
1.2.2.1 The remote operation facility that contains a significant radiation hazard to man or the environment.
1.2.2.2 The facility equipment can neither be accessed directly for purposes of operation or maintenance, nor can the equipment be viewed directly, for example, without shielding viewing windows, periscopes, or a video monitoring system.
1.2.2.3 The facility can be viewed directly but portions of the views are restricted (for example, the back or underside of objects) or where higher magnification or specialized viewing is beneficial.  
1.2.3 The remote viewing equipment may be intended for either long-term application (commonly, in excess of several years) or for short-term usage (for example, troubleshooting). Both types of applications are addressed in sections that follow.  
1.2.4 This guide is not intended to cover the detailed design and application of remote handling connectors for services (for example, electrical, instrumentation, video, etc.).  
1.2.5 The system of units employed in this guide is the metric ...

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.
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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.

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.

SIGNIFICANCE AND USE
4.1 The determination of uranium isotopic composition by gamma-ray spectrometry is a nondestructive technique and when used with other nondestructive techniques that quantify a single isotope, such as Test Methods C1133 (Segmented Gamma Scanning), C1221 (Solution Assay), C1455 (Holdup),and C1718 (Tomographic Gamma Scanning), can provide a wholly nondestructive assay of uranium mass necessary for material accountancy and safeguards needs. This method can be used with calorimetry (Test Method C1458) for kilogram quantities of high-enriched uranium and is also used to convert an Active-Well Coincidence Counter (4) measurement of 235U mass to total uranium mass.  
4.2 Because gamma-ray spectrometry systems are typically automated, the routine use of the test method is fast, reliable, and is not labor intensive. The test method is nondestructive, requires no sample preparation, and does not create waste disposal problems.  
4.3 The test method does not require that the system be calibrated to a specific geometry.  
4.4 The test method assumes that all uranium in the measured item has the same isotopic distribution. This is often termed isotopic homogeneity.  
4.5 The application of the test method does not depend upon the physical or chemical form of the material being analyzed.  
4.6 The 236U abundance is not measured by this test method and must be estimated from isotopic correlation techniques, stream averages, historical information, or other measurement techniques.  
4.7 The isotopic composition of a given item of uranium is an attribute of that item and, once determined, can be used in subsequent inventory measurements to verify the identity of an item within the measurement uncertainties.  
4.8 The method can also measure the ratio of other gamma-emitting isotopes in the measured item to uranium assuming they have the same spatial distribution as the uranium in the item. Some of these “other” gamma-emitting isotopes include daughter isotopes of uranium, ces...
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1.1 This test method applies to the nondestructive determination of the isotopic abundances of uranium, typically 234U, 235U, 236U, and 238U, in isotopically homogeneous uranium-bearing materials using gamma spectrometry. The material is commonly inside a container and is measured without specimen preparation.  
1.2 This test method is applicable to items containing sub-gram quantities of uranium to the maximum uranium mass allowed by criticality considerations.  
1.3 Measurable gamma ray emissions from uranium cover the energy range from below 80 keV to above 1000 keV. K-X-ray emissions from the isotopes of uranium and their daughters are found in the energy region around 100 keV. This test method has been applied to all portions of this energy range.  
1.4 The isotopic abundance of 236U is usually not directly determined because its low-energy gamma rays are too weak (1)2 to be detected under normal measurement conditions. Isotopic correlation techniques have been used to estimate its relative abundance (2).  
1.5 This test method has been demonstrated in routine use for isotopic amount fraction (atom %) of 235U from 0.2 % to 97 %.  
1.6 This test method requires decay equilibrium (160 days for 99 %) between 238U and its 24.1 d half-life 234Th daughter. Corrections can be made if the date of chemical separation of the 234Th daughter is known.  
1.7 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.8 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.  
1.9 This international standard was developed in accordance with internationally recognized principles on standardization established in the Dec...

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...

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.

SIGNIFICANCE AND USE
4.1 Radiation Shielding Window Components:  
4.1.1 Radiation shielding window components operability and long-term integrity are concerns that originate during the design and fabrication sequences. Such concerns can only be addressed, or are most efficiently addressed, during one or the other of these stages. The operability and integrity can be compromised during handling and installation sequences. For this reason, the subject equipment should be handled and installed under closely controlled and supervised conditions.  
4.1.2 This standard is intended as a supplement to other standards and to federal and state regulations, codes, and criteria applicable to the design of radiation shielding window components.
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1.1 Intent:  
1.1.1 The intent of this standard is to provide guidance for the design, fabrication, quality assurance, inspection, testing, packaging, shipping, installation, and maintenance of radiation shielding window components. These window components include wall liner embedments, dry lead glass radiation shielding window assemblies, oil-filled lead glass radiation shielding window assemblies, shielding wall plugs, barrier shields, view ports, and the installation/extraction table/device required for the installation and removal of the window components.  
1.2 Applicability:  
1.2.1 This standard is intended for those persons who are tasked with the planning, design, procurement, fabrication, installation, and operation of the radiation shielding window components that may be used in the operation of hot cells, high level caves, mini-cells, canyon facilities, and very high level radiation areas.  
1.2.2 This standard applies to radiation shielding window assemblies used in normal concrete walls, high-density concrete walls, steel walls and lead walls.  
1.2.3 The values stated in SI units or inch-pound units are to be regarded separately as standard. The values stated in each system may not be exact equivalents; therefore, each system shall be used independently of the other. Combining values from the two systems may result in nonconformance with the standard. Common nomenclature for specifying some terms; specifically shielding, uses a combination of metric units and inch-pound units.  
1.2.4 This standard identifies the special information required by the Manufacturer for the design of window components. Table A1.1 shows a sample list of the radiation source spectra and geometry information, typically required for shielding analysis. Table A2.1 shows a detailed sample list of specific data typically required to determine the physical size, glass types, and viewing characteristics of the shielding window, or view port. Annex A3 shows general window configuration sketches. Blank copies of Table A1.2 and Table A2.1 are found in the respective Annexes for the Owner–Operator's use.  
1.2.5 This standard is intended to be generic and to apply to a wide range of configurations and types of lead glass radiation shielding window components used in hot cells. It does not address glovebox, water, X-ray glass, or zinc bromide windows.  
1.2.6 Supplementary information on viewing systems in hot cells may be found in Guides C1533 and C1661.  
1.3 Caveats:  
1.3.1 Consideration shall be given when preparing the shielding window designs for the safety related issues discussed in the Hazard Sources and Failure Modes, Section 11; such as dielectric discharge, over-pressurization, radiation exposure, contamination, and overturning of the installation/extraction table/device.  
1.3.2 In many cases, the use of the word “shall” has been purposely used in lieu of “should” to stress the importance of the statements that have been made in this standard.  
1.3.3 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 prac...

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...

SIGNIFICANCE AND USE
4.1 Materials handling equipment operability and long-term integrity are concerns that originate during the design and fabrication sequences. Such concerns are most efficiently addressed during one or the other of these stages. Equipment operability and integrity can be compromised during handling and installation sequences. For this reason, the subject equipment should be handled and installed under closely controlled and supervised conditions.  
4.2 This guide is intended as a supplement to other standards (Section 2, Referenced Documents), and to federal and state regulations, codes, and criteria applicable to the design of equipment intended for this use.  
4.3 This guide is intended to be generic and to apply to a wide range of types and configurations of materials handling equipment.  
4.4 The term materials handling equipment is used herein in a generic sense. It includes manipulators, cranes, carts or bogies, and special equipment for handling tools and material in hot cells.  
4.5 This service imposes stringent requirements on the quality and the integrity of the equipment, as follows:  
4.5.1 Boots and similar protective covers should not restrict movement of the equipment, should be properly sealed to the equipment and should withstand the radiation, cell atmosphere, dust, cell temperatures, chemical exposures, and cleaning and decontamination reagents, and also resist snags and tearing.  
4.5.2 Materials handling equipment should be capable of withstanding rigorous chemical cleaning and decontamination procedures.  
4.5.3 Materials handling equipment should be designed and fabricated to remain dimensionally stable throughout its life cycle.  
4.5.4 Attention to fabrication tolerances is necessary to allow the proper fit-up between components for the proper installation and mounting of materials handling equipment in hot cells, for example, when parts or components are being replaced. Fabrication tolerances should be controlled to provide sufficie...
SCOPE
1.1 Intent:  
1.1.1 This guide covers materials handling equipment used in hot cells (shielded cells) for the processing and handling of nuclear and radioactive materials. The intent of this guide is to aid in the selection and design of materials handling equipment for hot cells in order to minimize equipment failures and maximize the equipment utility.  
1.1.2 It is intended that this guide record the principles and caveats that experience has shown to be essential to the design, fabrication, installation, maintenance, repair, replacement, and decontamination and decommissioning of materials handling equipment capable of meeting the stringent demands of operating, dependably and safely, in a hot cell environment where operator visibility is limited due to the radiation exposure hazards.  
1.1.3 This guide may apply to materials handling equipment in other radioactive remotely operated facilities such as suited entry repair areas and canyons, but does not apply to materials handling equipment used in commercial power reactors.  
1.1.4 This guide covers mechanical master-slave manipulators and electro-mechanical manipulators, but does not cover electro-hydraulic manipulators.  
1.2 Applicability:  
1.2.1 This guide is intended to be applicable to equipment used under one or more of the following conditions:
1.2.1.1 The materials handled or processed constitute a significant radiation hazard to man or to the environment.
1.2.1.2 The equipment will generally be used over a long-term life cycle (for example, in excess of two years), but equipment intended for use over a shorter life cycle is not excluded.
1.2.1.3 The equipment can neither be accessed directly for purposes of operation or maintenance, nor can the equipment be viewed directly, for example, without shielded viewing windows, periscopes, or a video monitoring system.  
1.3 User Caveats:  
1.3.1 This standard is not a substitute for applied engin...

SIGNIFICANCE AND USE
5.1 This test method provides a description of the design of the Stirred Reactor Coupon Analysis (SRCA) apparatus and identifies aspects of the performance of the SRCA tests and interpretation of the test results that must be addressed by the experimenter to provide confidence in the measured dissolution rate.  
5.2 The SRCA methods described in this test method can be used to characterize several aspects of glass corrosion that can be included in mechanistic models of long-term durability of glasses, including nuclear waste glasses.  
5.3 Depending on the test parameters investigated, the SRCA results can be used to measure the intrinsic dilute glass dissolution rate, as well as the effects of conditions such as temperature, pH, and solution chemistry on the dissolution rate.  
5.4 Due to the scalable nature of the method, it is particularly applicable to studies of the impact of glass composition on dilute-condition corrosion. Models of glass behavior can be parameterized by testing glass composition matrices and establishing quantitative structure-property relationships.  
5.5 The step heights present on the corroded sample can be measured by a variety of techniques including profilometry (optical or stylus), atomic force microscopy, interferometry or other techniques capable of determining relative depths on a sample surface. The sample can also be interrogated with other techniques such as scanning electron microscopy to characterize the corrosion behavior. These further analyses can determine if the sample corroded homogenously and possible formation of secondary phases or leached layers. Occurrence of these features may impact the accuracy of glass dissolution. This test method does not address these solid-state characterizations.
SCOPE
1.1 This test method describes a test method in which the dissolution rate of a homogenous silicate glass is measured through corrosion of monolithic samples in stirred dilute conditions.  
1.2 Although the test method was designed for simulated nuclear waste glass compositions per Guide C1174, the method is applicable to glass compositions for other applications including, but not limited to, display glass, pharmaceutical glass, bioglass, and container glass compositions.  
1.3 Various test solutions can be used at temperatures less than 100 °C. While the durability of the glass can be impacted by dissolving species from the glass, and thus the test can be conducted in dilute conditions or concentrated condition to determine the impact of such species, care must be taken to avoid, acknowledge, or account for the production of alteration layers which may confound the step height measurements.  
1.4 The dissolution rate measured by this test is, by design, an average of all corrosion that occurs during the test. In dilute conditions, glass is assumed to dissolve congruently and the dissolution rate is assumed to be constant.  
1.5 Tests are carried out via the placement of the monolithic samples in a large well-mixed volume of solution, achieving a high volume to surface area ratio resulting in dilute conditions with agitation of the solution.  
1.6 This test method excludes test methods using powdered glass samples, or in which the reactor solution saturates with time. Glass fibers may be used without a mask if the diameter is known to high accuracy before the test.  
1.7 Tests may be conducted with ASTM Type I water (see Specification D1193 and Terminology D1129), buffered water or other chemical solutions, simulated or actual groundwaters, biofluids, or other dissolving solutions.  
1.8 Tests are conducted with monolithic glass samples with at least a single flat face. Although having two plane-parallel faces is helpful for certain step height measurements, it is not required. The geometric dimensions of the monolith are not required to be known. The reacted monolithic sample is to be analyzed following the reaction to measure a corroded d...

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.  
5.4 Quantitative Measurements—These measurements result in quantification of the mass of the measured nuclides in the holdup. They include all the corrections, such as attenuation, and descriptive information, such as isotopic composition, that are available  
5.4.1 High quality results require detailed knowledge of radiation sources and detectors, transmission of radiation, calibration, facility operations and error analysis. Judicious use of subject matter experts is required (Guide C1490).  
5.5 Holdup Monitoring—Periodic re-measurement of holdup at a defined point using the same technique and a...
SCOPE
1.1 This test method describes gamma-ray methods used to nondestructively measure the quantity of 235U or  239Pu present as holdup in nuclear facilities. Holdup may occur in any facility where nuclear material is processed, in process equipment, in exhaust ventilation systems and in building walls and floors.  
1.2 This test method includes information useful for management, planning, selection of equipment, consideration of interferences, measurement program definition, and the utilization of resources  (1, 2, 3, 4) .2  
1.3 The measurement of nuclear material hold up in process equipment requires a scientific knowledge of radiation sources and detectors, transmission of radiation, calibration, facility operations and uncertainty analysis. It is subject to the constraints of the facility, management, budget, and schedule; plus health and safety requirements. The measurement process includes defining measurement uncertainties and is sensitive to the form and distribution of the material, various backgrounds, and interferences. The work includes investigation of material distributions within a facility, which could include potentially large holdup surface areas. Nuclear material held up in pipes, ductwork, gloveboxes, and heavy equipment, is usually distributed in a diffuse and irregular manner. It is difficult to define the measurement geometry, to identify the form of the material, and to measure it without interference from adjacent sources of radiation.  
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...

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.

SIGNIFICANCE AND USE
4.1 The materials covered that must meet ASTM specifications are uranium metal and uranium oxide.  
4.2 Uranium materials are used as nuclear reactor fuel. For this use, these materials must meet certain criteria for uranium content, uranium-235 enrichment, and impurity content, as described in Specifications C753 and C776. The material is assayed for uranium to determine whether the content is as specified.  
4.3 Uranium alloys, refractory uranium materials, and uranium containing scrap and ash are unique uranium materials for which the user must determine the applicability of this practice. In general, these unique uranium materials are dissolved with various acid mixtures or by fusion with various fluxes.
SCOPE
1.1 This practice covers dissolution treatments for uranium materials that are applicable to the test methods used for characterizing these materials for uranium elemental, isotopic, and impurities determinations. Dissolution treatments for the major uranium materials assayed for uranium or analyzed for other components are listed.  
1.2 The treatments, in order of presentation, are as follows:    
Procedure Title  
Section  
Dissolution of Uranium Metal and Oxide with Nitric Acid  
8.1  
Dissolution of Uranium Oxides with Nitric Acid and Residue
Treatment  
8.2  
Dissolution of Uranium-Aluminum Alloys in Hydrochloric Acid
with Residue Treatment  
8.3  
Dissolution of Uranium Scrap and Ash by Leaching with Nitric
Acid and Treatment of Residue by Carbonate Fusion  
8.4  
Dissolution of Refractory Uranium-Containing Material by
Carbonate Fusion  
8.5  
Dissolution of Uranium—Aluminum Alloys
Uranium Scrap and Ash, and Refractory
Uranium-Containing Materials by
Microwave Treatment  
8.6  
1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use. Specific hazards statements are given in Section 7.  
1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

SIGNIFICANCE AND USE
5.1 This guide offers several options for the determination of 99Tc in soil samples. Sample sizes of up to 200 g are possible, depending on the method chosen to extract Tc from the soil matrix. It is up to the user to determine if it is appropriate for the intended use of the final data.
SCOPE
1.1 This guide is intended to serve as a reference for laboratories wishing to perform 99Tc analyses in soil. Several options are given for selection of a tracer and for the method of extracting the Tc from the soil matrix. Separation of Tc from the sample matrix is performed using an extraction chromatography resin. Options are then given for the determination of the 99Tc activity in the original sample. It is up to the user to determine which options are appropriate for use, and to generate acceptance data to support the chosen procedure.  
1.2 Due to the various extraction methods available, various tracers used, variable detection methods used, and lack of certified reference materials for 99Tc in soil, there is insufficient data to support a single method written as a standard method.  
1.3 The values stated in SI units are to be regarded as standard, except where the non-SI unit of molar, M, is used for the concentration of chemicals and reagents. The values given in parentheses after SI units are provided for information only and are not considered 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.

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.
SCOPE
1.1 This test method covers unirradiated uranium-plutonium mixed oxide having a uranium to plutonium ratio of 2.5 and greater. The presence of larger amounts of plutonium (Pu) that give lower uranium to plutonium ratios may give low analysis results for uranium (U) (1)2, if the amount of plutonium together with the uranium is sufficient to slow the reduction step and prevent complete reduction of the uranium in the allotted time. Use of this test method for lower uranium to plutonium ratios may be possible, especially when 20 mg to 50 mg quantities of uranium are being titrated rather than the 100 mg to 300 mg in the study cited in Ref (1). Confirmation of that information should be obtained before this test method is used for ratios of uranium to plutonium less than 2.5.  
1.2 The amount of uranium determined in the data presented in Section 12 was 20 mg to 50 mg. However, this test method, as stated, contains iron in excess of that needed to reduce the combined quantities of uranium and plutonium in a solution containing 300 mg of uranium with uranium to plutonium ratios greater than or equal to 2.5. Solutions containing up to 300 mg uranium with uranium to plutonium ratios greater than or equal to 2.5 have been analyzed (1) using the reagent volumes and conditions as described in Section 10.  
1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use. For specific hazard statements, see Section 8.  
1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

SIGNIFICANCE AND USE
5.1 The determination of actinides by alpha spectrometry is an essential function of many environmental and other programs. Alpha spectrometry allows the identification and quantification of most alpha-emitting actinides. Although numerous separation methods are used, the final sample preparation technique has historically been by electrodeposition (Practice C1284). However, electrodeposition may have some drawbacks, such as time required, incompatibility with prior chemistry, thick deposits, and low recoveries. These problems may be minimized by using the neodymium fluoride coprecipitation method whose performance is well documented (1-6).4 To a lesser extent cerium fluoride has been used (7) but is not addressed in this practice.  
5.2 The sample mounting technique described in this practice is rapid, adds an additional purification step, since only those elements that form insoluble fluorides are mounted, and the sample and filter media can be dissolved and remounted if problems occur. The recoveries are better and resolution approaches normal in electrodeposited samples. Recoveries are sufficiently high that for survey work, if quantitative recoveries are not necessary, tracers can be omitted. Drawbacks to this technique include use of very hazardous hydrofluoric acid and the possibility of a non-reproducible and ill-defined counting geometry from filters that are not flat and may not be suitable for long retention. Also, although the total turn around time for coprecipitation may be less than for electrodeposition, coprecipitation requires more time and attention from the analyst.
SCOPE
1.1 This practice covers the preparation of separated fractions of actinides for alpha spectrometry. It is applicable to any of the actinides that can be dissolved in dilute hydrochloric acid. Examples of applicable samples would be the final elution from an ion exchange separation or the final strip from a solvent extraction separation.2  
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, health, and environmental practices and determine the applicability of regulatory limitations prior to use. For a specific hazard statement, see Section 9.  
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.

SIGNIFICANCE AND USE
5.1 Uranium material is used as a fuel in certain types of nuclear reactors. To be suitable for use as nuclear fuel, the starting material shall meet certain specifications such as those described in Specifications C753, C776, C787, C833, C967, C996, and C1008, or as specified by the purchaser. The isotope amount ratios of uranium material can be measured by mass spectrometry following this test method to ensure that they meet the specification.  
5.2 The MTE method can be used for a wide range of sample sizes even in samples containing as low as 20 µg of uranium. If the uranium sample is in the form of uranium hexafluoride, it has to be converted into a uranium nitrate solution for measurement by the MTE method. The concentration of the loading solution for MTE has to be in the range of 1 mg/g to 6 mg/g to allow a sample loading of 2 µg to 6 µg of uranium. A minimum loading of 3 µg uranium per filament is strongly recommended. This is needed to have a sufficient and stable ion signal especially for the two minor isotopes (234U and 236U) thus enabling the internal calibration of SEM versus the Faraday cups using the 234U ion beam signal during the measurement.  
5.3 Until now, the instrument capabilities for the MTE method have only been implemented on the TRITON™ TIMS instrument.5 Therefore, all recommendations for measurement parameters in this test method are specified for the TRITON instrument. The manufacturers of other TIMS instruments (for example, IsotopX and Nu Instruments) have indicated plans to implement the modifications needed in their instruments to use the MTE method.  
5.4 The MTE method described here can also be extended to measurement of elements other than uranium. Note that the MTE method has already been implemented for plutonium and calcium.
SCOPE
1.1 This test method describes the determination of the isotope amount ratios of uranium material as nitrate solutions by the modified total evaporation (MTE) method using a thermal ionization mass spectrometer (TIMS) instrument.  
1.2 The analytical performance in the determination of the 235U/238U major isotope amount ratio by MTE is similar to the (“classical”) total evaporation (TE) method as described in C1672. However, in the MTE method, the evaporation process is interrupted on a regular basis to allow measurements and subsequent corrections for background from peak tailing, perform internal calibration of a secondary electron multiplier (SEM) detector versus the Faraday cups, peak centering, and ion source refocusing. Performing these calibrations and corrections on a regular basis during the measurement, improves precision, and significantly reduces uncertainties for the minor isotope amount ratios 234U/238U and 236U/238U as compared to the TE method.  
1.3 In principle, the MTE method may yield major isotope amount ratios without the need for mass fractionation correction. However, depending on the measurement conditions, small variations are observed between sample turrets. Therefore, a small correction based on measurements of a certified reference material is recommended to improve consistency. The uncertainty around the mass fractionation correction factor usually includes unity.  
1.4 Units—The values stated in SI units are to be regarded as standard. The values given in parentheses after SI units are provided for information only and are not considered 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 issu...

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...

SIGNIFICANCE AND USE
5.1 This test method is used to ascertain whether or not materials meet specifications for plutonium concentration or plutonium mass fraction.  
5.1.1 The materials (nuclear grade plutonium nitrate solutions, plutonium metal, plutonium oxide powder, and mixed oxide and carbide powders and pellets) to which this test method applies are subject to nuclear safeguards regulations governing their possession and use. However, adherence to this test method does not automatically guarantee regulatory acceptance of the resulting safeguards measurements. It remains the sole responsibility of the user of this test method to ensure that its application to safeguards has the approval of the proper regulatory authorities.  
5.1.2 When used in conjunction with appropriate certified reference materials (CRMs), this test method can demonstrate traceability to the international measurements system (SI).  
5.2 Fitness for Purpose of Safeguards and Nuclear Safety Application—Methods intended for use in safeguards and nuclear safety applications shall meet the requirements specified by Guide C1068 for use in such applications.  
5.3 A chemical calibration of the coulometer is necessary for accurate results.
FIG. 1 Example of a Cell Design Used at Los Alamos National Laboratory (LANL)
SCOPE
1.1 This test method covers the determination of milligram quantities of plutonium in unirradiated uranium-plutonium mixed oxide having a U/Pu ratio range of 0.1 to 10. This test method is also applicable to plutonium metal, plutonium oxide, uranium-plutonium mixed carbide, various plutonium compounds including fluoride and chloride salts, and plutonium solutions.  
1.2 The recommended amount of plutonium for each aliquant in the coulometric analysis is 5 mg to 10 mg. Precision worsens for lower amounts of plutonium, and elapsed time of electrolysis becomes impractical for higher amounts of plutonium.  
1.3 The quantity values stated in SI units are to be regarded as standard. The quantity values with non-SI units are given in parentheses for information only.  
1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use. Specific precautionary statements are given in Section 9.  
1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

SIGNIFICANCE AND USE
4.1 The process of selection, training and qualification of personnel involved with NDA measurements is one of the quality assurance elements for an overall quality NDA measurement program.  
4.2 This guide describes an approach to selection, qualification, and training of personnel that is to be used in conjunction with other NDA Quality Assurance (QA) program elements. The selection, qualification and training processes can vary and this guide provides one such approach.  
4.3 The qualification activities described in this guide assume that NDA personnel are already proficient in general facility operations and safety procedures. The training and activities that developed this proficiency are not covered in this guide.  
4.4 This guide describes a basic approach and principles for the qualification of NDA professionals and technical specialists and operators. A different approach may be adopted by the management organization based on its particular organization and facility specifics. However, if a variation of the approach of this guide is applied, the resulting selection, training, and qualification programs must meet the requirements of the facility quality assurance program and should provide all the applicable functions of Section 5.  
4.5 This guide may be used as an aid in the preparation of a Training Implementation Plan (TIP) for the Transuranic Waste Characterization Program (TWCP).  
4.6 This guide describes education and expertise guidance for NDA auditors due to the importance and complexity of proper oversight of NDA activities.
SCOPE
1.1 This guide contains good practices for the selection, training, qualification, and professional development of personnel performing analysis, calibration, physical measurements, or data review using nondestructive assay equipment, methods, results, or techniques. The guide also covers NDA personnel involved with NDA equipment setup, selection, diagnosis, troubleshooting, or repair. General guidelines for the selection, training, and qualification of NDA auditors are included as well, but at a lower level of detail due to the variability of the personnel’s responsibilities performing this functions. Selection, training, and qualification programs based on this guide are intended to provide assurance that NDA personnel are suitably qualified and experienced personnel (SQEP) to perform their jobs competently. This guide presents a series of options but does not recommend a specific course of action.
This standard guide does not address the qualifications per se of an NDA Manager. However, it is expected that the NDA Manager is familiar with NDA techniques, and can make informed decisions on the acceptability of the assay results. If an NDA Manager does not have adequate technical qualifications in the NDA field, they are recommended to undergo training to gain familiarity in this area.
An NDA Manager with no relevant NDA experience should have access to a Senior NDA Professional who will give guidance for all technical decisions such as applicability and limitation of methods, reasonableness of results, needed upgrades and advantageous development investments.  
1.2 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.

SIGNIFICANCE AND USE
5.1 Uranium hexafluoride is normally produced and handled in large (typically 1 to 14-ton) quantities and must, therefore, be characterized by reference to representative samples (see ISO 7195). The samples are used to determine compliance with the applicable commercial specification C787. The quantities involved, physical properties, chemical reactivity, and hazardous nature of UF6 are such that for representative sampling, specially designed equipment must be used and operated in accordance with the most carefully controlled and stringent procedures. This practice can be used by UF6 converters to review the effectiveness of existing procedures or as a guide to the design of equipment and procedures for future use.  
5.2 The intention of this practice is to avoid liquid UF6 sampling once the cylinder has been filled. For safety reasons, manipulation of large quantities of liquid UF6 should be avoided when possible.  
5.3 It is emphasized that this practice is not meant to address conventional or nuclear criticality safety issues.
SCOPE
1.1 This practice covers methods for withdrawing representative sample(s) of uranium hexafluoride (UF6) during a transfer occurring in the gas phase. Such transfer in the gas phase can take place during the filling of a cylinder during a continuous production process, for example the distillation column in a conversion facility. Such sample(s) may be used for determining compliance with the applicable commercial specification, for example Specification C787.  
1.2 Since UF6 sampling is taken during the filling process, this practice does not address any special additional arrangements that may be agreed upon between the buyer and the seller when the sampled bulk material is being added to residues already present in a container (“heels recycle”). Such arrangements will be based on QA procedures such as traceability of cylinder origin (to prevent for example contamination with irradiated material).  
1.3 If the receiving cylinder is purged after filling and sampling, special verifications must be performed by the user to verify the representativity of the sample(s). It is then expected that the results found on volatile impurities with gas phase sampling may be conservative.  
1.4 This practice is only applicable when the transfer occurs in the gas phase. When the transfer is performed in the liquid phase, Practice C1052 should apply. This practice does not apply to gas sampling after the cylinder has been filled since the sample taken will not be representative of the cylinder.  
1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
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.

ABSTRACT
This specification provides the chemical and physical properties and requirements for nuclear-grade aluminum oxide powder intended for fabrication into shapes for nuclear applications. The materials shall conform to physical requirements as to particle size distribution, and specific surface area, and chemical requirements as to loss-on-ignition, and total and elemental concentrations of all impurities. Impurities may include silicaon, iron-chromium-nickel, magnesium, sodium, calcium, hafnium, fluorine, fluorine-chlorine-iodine-bromine, gadolinium, samarium, europium, and dysprosium.
SCOPE
1.1 This specification provides the chemical and physical requirements for nuclear-grade aluminum oxide powder intended for fabrication into shapes for nuclear applications. Two specific uses for which this powder is intended are Al2O3  pellets and Al2O3 − B4C composite pellets for use as thermal insulator or burnable neutron absorbers, respectively.  
1.2 The material described herein shall be particulate in nature.  
1.3 Units—The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.4 This 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.

ABSTRACT
This specification covers the properties and requirements for pellets of stabilized cubic hafnium oxide used in nuclear reactors. Hafnium oxide should consist of a stabilizing agent, the recommended of which is yttrium oxide, though others such as calcium oxide and magnesium oxide may also be used as agreed upon by the buyer and seller. The material shall meet specified values of the following requirements: physical dimensions; density; mechanical properties; phase stabilization; impurity concentration limits; moisture concentration limit; visual appearance; end and circumferential chips; cracks; and fissures and other defects.
SCOPE
1.1 This specification applies to pellets of stabilized cubic hafnium oxide used in nuclear reactors.  
1.2 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.  
1.3 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.

SCOPE
1.1 This specification applies to pellets of aluminum oxide that may be ultimately used in a reactor core, for example, as filler or spacers within fuel, burnable poison, or control rods. In order to distinguish between the subject pellets and “burnable poison” pellets, it is established that the subject pellets are not intended to be used as neutron-absorbing material.  
1.2 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard.  
1.3 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.

SCOPE
1.1 This specification defines the physical and chemical requirements for hafnium oxide powder intended for fabrication into shapes for use in a nuclear reactor core.  
1.2 The material described herein shall be particulate in nature.  
1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.4 This 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.

SIGNIFICANCE AND USE
5.1 Shielding performance testing should be performed to verify analytical predictions of shielding effectiveness, compliance with design requirements, and determine the location(s) of shielding deficiencies that may require either supplemental shielding, design modification, or changes to operating methods and procedures to resolve.  
5.2 Dose rates higher than adjacent shielding may be expected in the vicinity of master-slave manipulator penetrations as a result of cable or tape clearance requirements within the mechanisms where they pass through the shield walls. This is true even with supplemental shielding installed in the manipulator through tube.  
5.3 Similar to manipulator penetrations, when sources are placed directly in front of penetrations and gaps resulting from access doors or panels, radiation levels directly on the other side of the gaps or penetrations may be higher than levels in adjacent shielding or analytical predictions. If these test configurations are representative of how the shielded enclosure will actually be operated (that is, the test configuration is representative of engineered source and normally occupied work locations) than the additional expense of designing or modifying the shielded enclosure should be considered if that location is normally occupied and will result in a significant increase of dose rates to personnel.  
5.4 Frames around shield windows may have a higher potential for shielding deficiencies.  
5.5 Shield walls constructed of concrete may be subject to the formation of void spaces that could result in diminished shielding performance.  
5.6 Background radiation levels in the area where the dose measurement data are being gathered shall be monitored and their contribution to measured dose rates accounted for in the data analysis as part of the test report.
SCOPE
1.1 This guide identifies appropriate test methods for determining the sufficiency of radiological shielding for hot cells and shielded enclosures.  
1.2 After constructing or modifying radiological shielding, it is necessary to verify that shielding performance meets or exceeds the shielding performance requirements. This is typically accomplished using sealed test sources of much less activity than the design basis. This allows for modifications or correction of any discrepancies identified before the commissioning of the hot cell.  
1.3 The guidance and practices recommended by this guide are applicable to both new and existing shielded facilities and enclosures for evaluating shielding suitability and locating the existence of shine paths or other shielding anomalies that result from design, manufacture, or construction.  
1.4 Two types of testing may be performed.  
1.4.1 Shielding performance verification testing provides evidence that the shielding configuration is sufficient for meeting established performance criteria and for identifying deficiencies in the shielding configuration or components that may not have been addressed during design. Test results are expected to demonstrate that shielding performance meets or exceeds design criteria, not match the dose rates predicted analytically.  
1.4.2 Shielding performance verification testing identifies shielding deficiencies (hot spots) in the installed configuration relative to adjacent shielding but does not demonstrate compliance with any quantitative shielding performance requirement.  
1.5 Performance testing should be specified and performed to assess shielding adequacy with sources in all critical locations.  
1.6 Requirements for shielding performance testing should be clearly defined in design basis or procurement documentation.  
1.7 This guide is not applicable to neutron radiation shielding performance evaluations.  
1.8 Units—The values stated in either SI units or inch-pound units are to be regarded separately as standard. The values stated in each system may not be exact equivalents; therefore, ea...

ABSTRACT
These practices are intended to provide the nuclear industry with procedures for obtaining representative bulk samples from uranium-ore concentrates (UOC) and for obtaining a series of representative secondary samples from the original bulk sample for the determination of moisture and other test purposes, and for the preparation of pulverized analytical samples. These practices consist of a number of alternative procedures for (1) primary sampling such as one-stage falling stream, two-stage falling stream, and Auger sampling; (2) secondary sampling such as straight-path (reciprocating) cutter sampling and rotating (Vezin) cutter multi-sampling; (3) sample preparation such as concurrent-drying, natural moisture, and calcination; and (4) sample packaging such as wax sealing and vacuum sealing. These procedures do not include requirements for health, safety, and accountability. The material and sampling equipment requirements are detailed. Schematic diagrams of the primary and secondary samplers are provided.
SCOPE
1.1 These practices are intended to provide the nuclear industry with procedures for obtaining representative bulk samples from uranium-ore concentrates (UOC) (see Specification C967).  
1.2 These practices also provide for obtaining a series of representative secondary samples from the original bulk sample for the determination of moisture and other test purposes, and for the preparation of pulverized analytical samples (see Test Methods C1022 and C1843).  
1.3 These practices consist of a number of alternative procedures for sampling and sample preparation which have been shown to be satisfactory through long experience in the nuclear industry. These procedures are described in the following order.    
Stage  
Procedure  
Section  
Primary Sampling  
One-stage falling stream  
5    
Two-stage falling stream  
6    
Auger  
7    
Secondary Sampling  
Straight-path (reciprocating)  
8    
Rotating (Vezin)  
9, 10  
1.3.1 The primary and secondary sampling stages can be organized in the following way:    
1.3.2 It is possible to combine the various elements of these stages in different ways to give satisfactory results depending on the agreed requirements of the contracting parties. For a given stage, however, each procedure must be regarded as a whole. It is highly inadvisable to mix elements belonging to different procedures.  
1.4 These procedures do not include requirements for health, safety, and accountability. The observance of these practices does not relieve the user of the obligation to be aware of and to conform to all applicable international, federal, state, and local regulations pertaining to processing, shipping, or using uranium-ore concentrates. (Guidance is provided in CFR 10, Chapter 1.)  
1.5 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.  
1.6 This standard does not purport to address all of the safety problems, 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.

SCOPE
1.1 This specification defines the physical and chemical requirements for zirconium oxide powder intended for fabrication into shapes, either entirely or partially of zirconia, for use in a nuclear reactor core.  
1.2 The material described herein shall be particulate in nature.  
1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.4 This 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.

SCOPE
1.1 This specification covers uranium oxides, including processed byproducts or scrap material (powder, pellets, or pieces), that are intended for dissolution into uranyl nitrate solution meeting the requirements of Specification C788 prior to conversion into nuclear grade UO2 powder with a 235U content of less than 5 %. This specification defines the impurity and uranium isotope limits for such urania powders that are to be dissolved prior to processing to nuclear grade UO2 as defined in Specification C753.  
1.2 This specification provides the nuclear industry with a general standard for such uranium oxide powders. It recognizes the diversity of conversion processes and the processes to which such powders are subsequently to be subjected (for instance, by solvent extraction). It is therefore anticipated that it may be necessary to include supplementary specification limits by agreement between the buyer and seller.  
1.3 The scope of this specification does not comprehensively cover all provisions for preventing criticality accidents, for health and safety, or for shipping. Observance of this specification does not relieve the user of the obligation to conform to all international, national, state and local regulations for processing, shipping, or any other way of using urania powders (see 2.2 and 2.3).  
1.4 Units—The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
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.

SIGNIFICANCE AND USE
4.1 The purpose of this guide is to provide general guidelines for the design and operation of hot cell equipment to ensure longevity and reliability throughout the period of service.  
4.2 It is intended that this guide record the general conditions and practices that experience has shown is necessary to minimize equipment failures and maximize the effectiveness and utility of hot cell equipment. It is also intended to alert designers to those features that are highly desirable for the selection of equipment that has proven reliable in high radiation environments.  
4.3 This guide is intended as a supplement to other standards, and to federal and state regulations, codes, and criteria applicable to the design of equipment intended for hot cell use.  
4.4 This guide is intended to be generic and to apply to a wide range of types and configurations of hot cell equipment.
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1.1 Intent:  
1.1.1 The intent of this guide is to provide general design and operating considerations for the safe and dependable operation of remotely operated hot cell equipment. Hot cell equipment is hardware used to handle, process, or analyze nuclear or radioactive material in a shielded room. The equipment is placed behind radiation shield walls and cannot be directly accessed by the operators or by maintenance personnel because of the radiation exposure hazards. Therefore, the equipment is operated remotely, either with or without the aid of viewing.  
1.1.2 This guide may apply to equipment in other radioactive remotely operated facilities such as suited entry repair areas, canyons or caves, but does not apply to equipment used in commercial power reactors.  
1.1.3 This guide does not apply to equipment used in gloveboxes.  
1.2 Applicability:  
1.2.1 This guide is intended for persons who are tasked with the planning, design, procurement, fabrication, installation, or testing of equipment used in remote hot cell environments.  
1.2.2 The equipment will generally be used over a long-term life cycle (for example, in excess of two years), but equipment intended for use over a shorter life cycle is not excluded.  
1.2.3 The system of units employed in this standard is the metric unit, also known as SI Units, which are commonly used for International Systems, and defined by IEEE/ASTM SI 10: American National Standard for Use of the International System of Units (SI): The Modern Metric System.  
1.3 Caveats:  
1.3.1 This guide does not address considerations relating to the design, construction, operation, or safety of hot cells, caves, canyons, or other similar remote facilities. This guide deals only with equipment intended for use in hot cells.  
1.3.2 Specific design and operating considerations are found in other ASTM documents.  
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.

SIGNIFICANCE AND USE
4.1 This guide is relevant to the design of specialized support equipment and tools that are remotely operated, maintained, or viewed through shielding windows, or combinations thereof, or by other remote viewing systems.  
4.2 Hot cells contain substances and processes that may be extremely hazardous to personnel or the external environment, or both. Process safety and reliability are improved with successful design, installation, and operation of specialized mechanical and support equipment.  
4.3 Use of this guide in the design of specialized mechanical and support equipment can reduce costs, improve productivity, reduce failed hardware replacement time, and provide a standardized design approach.
SCOPE
1.1 Intent:  
1.1.1 This guide presents practices and guidelines for the design and implementation of equipment and tools to assist assembly, disassembly, alignment, fastening, maintenance, or general handling of equipment in a hot cell. Operating in a remote hot cell environment significantly increases the difficulty and time required to perform a task compared to completing a similar task directly by hand. Successful specialized support equipment and tools minimize the required effort, reduce risks, and increase operating efficiencies.  
1.2 Applicability:  
1.2.1 This guide may apply to the design of specialized support equipment and tools anywhere it is remotely operated, maintained, and viewed through shielding windows or by other remote viewing systems.  
1.2.2 Consideration should be given to the need for specialized support equipment and tools early in the design process.  
1.2.3 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard.  
1.3 Caveats:  
1.3.1 This guide is generic in nature and addresses a wide range of remote working configurations. Other acceptable and proven international configurations exist and provide options for engineer and designer consideration. Specific designs are not a substitute for applied engineering skills, proven practices, or experience gained in any specific situation.  
1.3.2 This guide does not supersede federal or state regulations, or both, or codes applicable to equipment under any conditions.  
1.3.3 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.

SIGNIFICANCE AND USE
3.1 The purpose of this guide is to establish criteria for the design of gloveboxes as primary confinement systems to ensure the safety of the workers and the protection of the environment when storing, handling, processing, and disposing of both combustible and non-combustible forms of plutonium. The use of this guide will provide the user with guidance to design a successfully performing glovebox system.
SCOPE
1.1 This guide defines criteria for the design of glovebox systems to be used for the handling of plutonium in any chemical or physical form or isotopic composition or when mixed with other elements or compounds. Not included in the criteria are systems auxiliary to the glovebox systems such as utilities, ventilation, alarm, and waste disposal. Also not addressed are hot cells or open-face hoods.  
1.2 The scope of this guide excludes specific license requirements relating to provisions for criticality prevention, hazards control, safeguards, packaging, and material handling. Observance of this guide does not relieve the user of the obligation to conform to all federal, state, and local regulations for design and construction of glovebox systems.  
1.3 Units—The values stated in either SI units or inch-pound units are to be regarded separately as standard. The values stated in each system may not be exact equivalents; therefore, each system shall be used independently of the other. Combining values from the two systems may result in nonconformance with the 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.

SIGNIFICANCE AND USE
5.1 Factors governing selection of a method for the determination of uranium include available quantity of sample, homogeneity of material sampled, sample purity, desired level of reliability, and facility available equipment.  
5.2 This uranium assay method is referenced in the Test Methods for Chemical, Mass Spectrometric, and Spectrochemical Analysis of Nuclear-Grade Uranium Dioxide Powders and Pellets (Test Methods C696) and in the Test Methods for Chemical, Mass Spectrometric, and Spectrochemical, Nuclear, and Radiochemical Analysis of Nuclear-Grade Uranyl Nitrate Solutions (Test Methods C799). This uranium assay method may also be used for uranium hexafluoride and uranium ore concentrate. This test method determines 20 mg to 200 mg of uranium; is applicable to product, fuel, and scrap material after the material is dissolved; is tolerant towards most metallic impurity elements usually specified in product and fuel; and uses no special equipment.  
5.3 The ruggedness of the titration method has been studied for both the volumetric (6) and the weight (7)  titration of uranium with dichromate.  
5.4 Fitness for Purpose of Safeguards and Nuclear Safety Application—Methods intended for use in safeguards and nuclear safety applications shall meet the requirements specified by Guide C1068 for use in such applications.  
5.4.1 When used in conjunction with the appropriate certified reference materials (SRM6 or CRM), this procedure can demonstrate traceability to the national measurement base. However, use of the test method does not automatically guarantee regulatory acceptance of the resulting safeguards measurements. It remains the sole responsibility of the user of this test method to assure that its application to safeguards has the approval of the proper regulatory authorities.
SCOPE
1.1 This test method, commonly referred to as the Modified Davies and Gray technique, covers the titration of uranium in product, fuel, and scrap materials after the material is dissolved. The test method is versatile and has been ruggedness tested. With appropriate sample preparation, this test method can give precise and unbiased uranium assays over a wide variety of material types (1, 2).2 Details of the titration procedure in the presence of plutonium with appropriate modifications are given in Test Method C1204.  
1.2 Uranium levels titrated are usually 20 mg to 50 mg, but up to 200 mg uranium can be titrated using the reagent volumes stated in this test method.  
1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use. For specific safeguard and safety precaution statements, see Section 5.  
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.

ABSTRACT
This specification applies to pellets of stabilized zirconium oxide used in nuclear reactors. The chemical composition requirements such as the stabilizing additive (calcium oxide or yttrium oxide), analytical chemistry methods, impurity concentration (including hafnium, boron, gadolinium, samarium, europium, dysprosium, cobalt, silicon, iron, calcium, magnesium, aluminum, titanium, thorium, fluorine, chlorine, bromine, iodine, and hydrogen), and moisture concentration are prescribed. The nuclear grade pellets shall conform to the specified physical requirements which includes the following: physical dimensions, density, mechanical properties and test methods such as compressive test and thermal cycling test, and visual appearance such as end chips, circumferential chips, cracks, and fissures. The requirements for cleanliness before and after sampling and packaging are given.
SCOPE
1.1 This specification applies to pellets of stabilized zirconium oxide used in nuclear reactors.  
1.2 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.  
1.3 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.

ABSTRACT
This specification covers sintered uranium dioxide pellets containing 235U for use in nuclear reactors. Chemical requirements include uranium content, impurity content, stoichiometry, and moisture content. Maximum concentration limits are specified for impurity elements such as: aluminum, carbon, calcium+magnesium, chlorine, chromium, cobalt, fluorine, hydrogen, iron, nickel, nitrogen, silicon, and thorium. Chemical analyses shall be performed. Nuclear requirements include isotopic content and equivalent boron content. The following are physical characteristics of the material: dimensions, pellet density, grain size and pore morphology, pellet integrity –
SCOPE
1.1 This specification is for finished sintered UO2 pellets. It applies to UO2 pellets containing uranium (U) of any  235U concentration for use in nuclear reactors.  
1.2 This specification recognizes the presence of reprocessed U in the fuel cycle and consequently defines isotopic limits for UO2 pellets made from commercial grade UO2. Such commercial grade UO2 is defined so that, regarding fuel design and manufacture, the product is essentially equivalent to that made from unirradiated U. UO2 falling outside these limits cannot necessarily be regarded as equivalent and may thus need special provisions at the fuel fabrication plant or in the fuel design.  
1.3 This specification does not include (a) provisions for preventing criticality accidents, (b) requirements for health and safety, (c) avoidance of hazards, or (d) shipping precautions and controls. 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 70, Title 10, Part 71, and Title 49, Part 173.  
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 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, health, and environmental practices and determine the applicability or 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.

SIGNIFICANCE AND USE
4.1 Mechanical drive systems operability and long-term integrity are concerns that should be addressed primarily during the design phase; however, problems identified during fabrication and testing should be resolved and the changes in the design documented. Equipment operability and integrity can be compromised during handling and installation sequences. For this reason, the subject equipment should be handled and installed under closely controlled and supervised conditions.  
4.2 This standard is intended as a supplement to other standards, and to federal and state regulations, codes, and criteria applicable to the design of equipment intended for this use.  
4.3 This standard is intended to be generic and to apply to a wide range of types and configurations of mechanical drive systems.
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1.1 Intent:  
1.1.1 The intent of this standard is to provide general guidelines for the design, selection, quality assurance, installation, operation, and maintenance of mechanical drive systems used in remote hot cell environments. The term mechanical drive systems used herein, encompasses all individual components used for imparting motion to equipment systems, subsystems, assemblies, and other components. It also includes complete positioning systems and individual units that provide motive power and any position indicators necessary to monitor the motion.  
1.2 Applicability:  
1.2.1 This standard is intended to be applicable to equipment used under one or more of the following conditions:  
1.2.1.1 The materials handled or processed constitute a significant radiation hazard to man or to the environment.
1.2.1.2 The equipment will generally be used over a long-term life cycle (for example, in excess of two years), but equipment intended for use over a shorter life cycle is not excluded.
1.2.1.3 The equipment can neither be accessed directly for purposes of operation or maintenance, nor can the equipment be viewed directly, for example, without radiation shielding windows, periscopes, or a video monitoring system (Guides C1572 and C1661).  
1.2.2 The values stated in either SI units or inch-pound units are to be regarded separately as standard. The values stated in each system may not be exact equivalents; therefore, each system shall be used independently of the other. Combining values from the two systems may result in non-conformance with the standard.  
1.3 User Caveats:  
1.3.1 This standard is not a substitute for applied engineering skills, proven practices and experience. Its purpose is to provide guidance.
1.3.1.1 The guidance set forth in this standard relating to design of equipment is intended only to alert designers and engineers to those features, conditions, and procedures that have been found necessary or highly desirable to the design, selection, operation and maintenance of mechanical drive systems for the subject service conditions.
1.3.1.2 The guidance set forth results from discoveries of conditions, practices, features, or lack of features that were found to be sources of operational or maintenance problems, or causes of failure.  
1.3.2 This standard does not supersede federal or state regulations, or both, and codes applicable to equipment under any conditions.  
1.3.3 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.

SIGNIFICANCE AND USE
4.1 The test methods in this standard are designed to show whether a given material meets the specifications prescribed in Specification C967.  
4.2 Because of the variability of matrices of uranium-ore concentrate and the limited availability of suitable reference or calibration materials, the precision and bias of these test methods should be established by each individual laboratory that will use them. The precision and bias statements given for each test method are those reported by various laboratories and can be used as a guideline.  
4.3 Instrumental test methods such as X-ray fluorescence and emission spectroscopy can be used for the determination of some impurities where such equipment is available.
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1.1 These test methods cover procedures for the chemical and atomic absorption analysis of uranium-ore concentrates to determine compliance with the requirements prescribed in Specification C967.  
1.2 The analytical procedures appear in the following order:    
Sections    
Uranium by Ferrous Sulfate Reduction—Potassium Dichromate
Titrimetry  
9    
Nitric Acid-Insoluble Uranium  
10 to 18  
Extractable Organic Material  
19 to 26  
Determination of Arsenic  
27  
Carbonate by CO2 Gravimetry  
28 to 34  
Fluoride by Ion-Selective Electrode  
35 to 42  
Halides by Volhard Titration  
43 to 50  
Phosphorus by Spectrophotometry  
52 to 60  
Determination of Silicon  
61  
Determination of Thorium  
62  
Calcium, Iron, Magnesium, Molybdenum, Titanium, and Vana-
dium by Atomic Absorption Spectrophotometry  
63 to 72  
Potassium and Sodium by Atomic Absorption
Spectrophotometry  
73 to 82  
Boron by Spectrophotometry  
83 to 92  
1.3 The values stated in SI units are to be regarded as 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. A specific precautionary statement is given in Section 7.  
1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

SIGNIFICANCE AND USE
5.1 The guide is applicable to the analysis of materials to demonstrate compliance with the specifications set forth in Specifications C787 and C996. Some other specifications may be expressed in terms of mass of  232U per mass of only  238U (see ISO 21847–3:2007).
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1.1 This guide covers the determination of  232U in uranium hexafluoride by alpha spectrometry.  
1.2 The values stated in SI units are to be regarded as standard, except where the non-SI unit of molar, M, is used for the concentration of chemicals and reagents. The unit of electronvolt (eV) is outside the SI but its use with the SI is accepted by the International Committee for Weights and Measures (CIPM, Comité International des Poids et Mesures) and the U. S. National Institute of Science and Technology (NIST). 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, 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.

SIGNIFICANCE AND USE
5.1 Uranium material is used as a fuel in certain types of nuclear reactors. To be suitable for use as nuclear fuel, the starting material shall meet certain specifications such as those described in Specifications C753, C776, C787, C833, C967, C996, and C1008, or as specified by the purchaser. The 235U/238U isotope amount ratios and the amount content of uranium material can be measured by mass spectrometry following this test method to ensure that they meet the specification.  
5.2 The double spike method has been used for studies of uranium fractionation effects in isotope geochemistry and cosmochemistry, for uranium source attribution in nuclear forensics and for investigation of conversion or sampling processes in nuclear industry and nuclear safeguards (7-11). Most recently, the double spike method has been used for the validation of the Cristallini sampling method of UF6   (12 and 13). The double spike method can be used for a wide range of sample sizes even in samples containing as low as 50 μg of uranium. The concentration of the loading solution for the DS method has to be in the range of 1 to 6 mg/g to allow a sample loading of 4 to 6 μg of uranium. A minimum loading of 4 μg uranium per filament is recommended.  
5.3 The measurement of 236U/238U ratios using this method is not possible due to the large isobaric interference from the 236U ion beam of the double spike onto the 236U ion beam from the sample (>50.000 times for close to natural material, for example, like IRMM-184).  
5.4 The application of the double spike method for measurements of 235U/238U ratio is limited by the isobaric interference between the 236U from the double spike material and the 236U contained in the sample. As a consequence, the method is not suitable for samples which contain significant amounts of 236U due to prior neutron capture from 235U in the predecessor materials. For samples with 236U/238U ratios higher than about 10–6, the double spike method should be applied with...
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1.1 This test method describes the determination of the isotope amount ratios of uranium material as nitrate solutions by the double spike (DS) method using a thermal ionization mass spectrometer (TIMS) instrument.  
1.2 The analytical performance in the determination of the 235U/238U major isotope amount ratio by the DS method is five to ten times better in terms of the internal and external reproducibility compared to the (“classical”) total evaporation (TE) method as described in Test Method C1672 and the “modified total evaporation” (MTE) as described in Test Method C1832. This is due to the use of an internal rather than external mass fractionation correction by using a double spike material with a known or certified 233U/236U isotope ratio, which is mixed with the sample prior to the measurement, either during the sample preparation or directly on the TIMS filament.  
1.3 The DS method cannot be applied for the determination of the 236U/238U minor isotope amount ratio, and is also not recommended for the determination of the 234U/238U minor isotope amount ratio.  
1.4 In case the uranium amount concentration of the double spike is known or certified, the uranium amount concentration of the sample can be determined using the isotope dilution mass spectrometry (IDMS) method as described in Test Method C1672, by blending the sample gravimetrically with the double spike and performing a DS measurement.  
1.5 An external mass fractionation correction by measurements of a certified reference material loaded on different filaments and measured in the same measurement sequence, as recommended for TE and required for MTE measurements, is not necessary for the DS method. However, for quality control (QC) purposes it is recommended to perform DS measurements of low enriched or natural uranium isotopic reference materials on a regular basis.  
1.6 The DS method can only be applied to uranium samples with relative isotope ...

SIGNIFICANCE AND USE
5.1 The test method within this standard is used to demonstrate uranium ore-concentrate material meets the moisture specification defined in Specification C967 or other applicable requirements.
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1.1 This test method is used to determine compliance with Specification C967 for the requirements of moisture in uranium ore concentrates (UOC). A procedure is given to determine the approximate temperature for drying the UOC; normally 110 °C but possibly 165 °C for uranyl peroxides. The dried uranium ore-concentrate resulting from this procedure is then used for performing additional analyses described in Specification C967.  
1.2 The values stated in SI units are to be regarded as the standard. The values given in parentheses are mathematical equivalents that are provided for information only and are not considered 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, 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.

SIGNIFICANCE AND USE
5.1 The major objective of this practice is to provide a common reference document for both applicants and certification authorities on the accepted practices for accomplishing package thermal qualification. Details and methods for accomplishing qualification are described in this document in more specific detail than available in the regulations. Methods that have been shown by experience to lead to successful qualification are emphasized. Possible problems and pitfalls that lead to unsatisfactory results are also described.  
5.2 The work described in this standard practice shall be done under a quality assurance program that is accepted by the regulatory authority that certifies the package for use. For packages certified in the United States, 10 CFR 71 Subpart H shall be used as the basis for the quality assurance (QA) program, while for international certification, ISO 9000 usually defines the appropriate program. The quality assurance program shall be in place and functioning prior to the initiation of any physical or analytical testing activities and prior to submittal of any information to the certifying authority.
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1.1 This practice defines detailed methods for thermal qualification of “Type B” radioactive materials packages under Title 10, Code of Federal Regulations, Part 71 (10CFR71) in the United States or, under International Atomic Energy Agency Regulation SSR-6. Under these regulations, packages transporting what are designated to be Type B quantities of radioactive material shall be demonstrated to be capable of withstanding a sequence of hypothetical accidents without significant release of contents.  
1.2 The unit system (SI metric or English) used for thermal qualification shall be agreed upon prior to submission of information to the certification authority. If SI units are to be standard, then use IEEE/ASTM SI-10. Additional units given in parentheses are for information purposes only.  
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 standard is used to measure and describe the response of materials, products, or assemblies to heat and flame under controlled conditions, but does not by itself incorporate all factors required for fire hazard or fire risk assessment of the materials, products, or assemblies under actual fire conditions.  
1.5 Fire testing is inherently hazardous. Adequate safeguards for personnel and property shall be employed in conducting these tests.  
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.

SIGNIFICANCE AND USE
4.1 The uranium hexfluoride (UF6), as described in Specifications C787 and C996, has to meet different requirements: one set of requirements being safety, health physics, and criticality and the other set being chemical, physical, and isotopic. To ensure the UF6 is in compliance with all requirements, sampling and analysis shall be performed. Therefore, packaging may have a significant impact on the quality of UF6.  
4.2 After sampling, the bottle will contain residues. There is contamination because of the equipment, other contamination caused by nonvolatile elements, and isotopic contamination as a result of UF6 hydrolysis.  
4.3 Cleaning shall be efficient. Special emphasis should be given to decontaminate the bottles without leaving any trace of cleaning products, make the bottles inert in UF6 medium (passivation bottle), and minimize waste. The cleaning process should be easy, safe, and environmentally friendly.  
4.4 This practice describes different protocols for cleaning bottles by gas and liquid.
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1.1 This practice provides a description of the different ways to clean uranium hexafluoride (UF6) bottles.  
1.2 This practice describes two kinds of sample bottles: 1S and 2S bottles.  
1.3 Units—The values stated in SI units are to be regarded as the 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.

SIGNIFICANCE AND USE
5.1 In order to demonstrate conformance to regulatory requirements and support the post-closure repository performance assessment information is required about the attributes, characteristics, and behavior of the SNF. These properties of the SNF in turn support the transport, interim storage, and repository pre-closure safety analyses, and repository post-closure performance assessment. In the United States, the interim dry storage of commercial LWR SNF is regulated per the Code of Federal Regulations, Title 10, Part 72, which requires that the cladding must not sustain during the interim storage period any “gross” damage sufficient to release fuel from the cladding into the container environment. In other countries, the appropriate governing body will set regulations regarding interim dry storage of commercial LWR SNF. However, cladding damage insufficient to allow the release of fuel during the interim storage period may still occur in the form of small cracks or pinholes that can develop into much larger defects. These cracks/pinholes could be sufficient to classify the fuel as “failed fuel” or “breached fuel” per the definitions given in Section 3 for repository disposal purposes, because they could allow contact of water vapor or liquid with the spent fuel matrix and thus provide a pathway for radionuclide release from the waste form. Therefore SNF characterization should be adequate to determine the amount of “failed fuel” for either usage as required. This could involve the examination of reactor operating records, ultrasonic testing, sipping, and analysis of the residual water and drying kinetics of the spent fuel assemblies or canisters.  
5.2 Regulations in each country may contain constraints and limitations on the chemical or physical (or both) properties and long-term degradation behavior of the spent fuel and HLW in the repository. Evaluating the design and performance of the waste form (WF), waste packaging (WP), and the rest of the engineered barrie...
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1.1 This guide provides guidance for the types and extent of testing that would be involved in characterizing the physical and chemical nature of spent nuclear fuel (SNF) in support of its interim storage, transport, and disposal in a geologic repository. This guide applies primarily to commercial light water reactor (LWR) spent fuel and spent fuel from weapons production, although the individual tests/analyses may be used as applicable to other spent fuels such as those from research reactors, test reactors, molten salt reactors and mixed oxide (MOX) spent fuel. The testing is designed to provide information that supports the design, safety analysis, and performance assessment of a geologic repository for the ultimate disposal of the SNF.  
1.2 The testing described includes characterization of such physical attributes as physical appearance, weight, density, shape/geometry, degree, and type of SNF cladding damage. The testing described also includes the measurement/examination of such chemical attributes as radionuclide content, microstructure, and corrosion product content, and such environmental response characteristics as drying rates, oxidation rates (in dry air, water vapor, and liquid water), ignition temperature, and dissolution/degradation rates. Not all of the characterization tests described herein must necessarily be performed for any given analysis of SNF performance for interim storage, transportation, or geological repository disposal, particularly in areas where an extensive body of literature already exists for the parameter of interest in the specific service condition.  
1.3 It is assumed in formulating the SNF characterization activities in this guide that the SNF has been stored in an interim storage facility at some time between reactor discharge and dry transport to a repository. The SNF may have been stored either wet (for example, a spent fuel pool), or dry (for example, an independent spent f...

SIGNIFICANCE AND USE
5.1 This procedure can be used for (but is not limited to) the following applications:
(1) support glass formulation development to make sure that processing criteria are met,
(2) support production (for example, processing or troubleshooting), and
(3) support model validation.
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1.1 These test methods cover procedures for determining the liquidus temperature (TL) of nuclear waste, mixed nuclear waste, simulated nuclear waste, or hazardous waste glass in the temperature range from 600 °C to 1600 °C. This test method differs from Practice C829 in that it employs additional methods to determine TL. TL is useful in waste glass plant operation, glass formulation, and melter design to determine the minimum temperature that must be maintained in a waste glass melt to make sure that crystallization does not occur or is below a particular constraint, for example, 1 volume % crystallinity or T1%. As of now, many institutions studying waste and simulated waste vitrification are not in agreement regarding this constraint (1).2  
1.2 Three methods are included, differing in (1) the type of equipment available to the analyst (that is, type of furnace and characterization equipment), (2) the quantity of glass available to the analyst, (3) the precision and accuracy desired for the measurement, and (4) candidate glass properties. The glass properties, for example, glass volatility and estimated TL, will dictate the required method for making the most precise measurement. The three different approaches to measuring TL described here include the following: Gradient Temperature Furnace Method (GT), Uniform Temperature Furnace Method (UT), and Crystal Fraction Extrapolation Method (CF). This procedure is intended to provide specific work processes, but may be supplemented by test instructions as deemed appropriate by the project manager or principle investigator. The methods defined here are not applicable to glasses that form multiple immiscible liquid phases. Immiscibility may be detected in the initial examination of glass during sample preparation (see 9.3). However, immiscibility may not become apparent until after testing is underway.  
1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
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.