4.1 Nuclear grade graphite is a composite material made from petroleum or a coal-tar-based coke and a pitch binder. Manufacturing graphite is an iterative process of baking and pitch impregnation of a formed billet prior to final graphitization, which occurs at temperatures greater than 2500 °C. The impregnation and rebake step is repeated several times until the desired product density is obtained. Integral to this process is the use of isotropic cokes and a forming process (that is, isostatically molded, vibrationally molded, or extruded) that is intended to obtain an isotropic or near isotropic material. However, the source, size, and blend of the starting materials as well as the forming process of the green billet will impart unique material properties as well as variations within the final product. There will be density variations from the billet surface inward and different physical properties with and transverse to the grain direction. Material variations are expected within individual billets as well as billet-to-billet and lot-to-lot. Other manufacturing defects of interest include large pores, inclusions, and cracks. In addition to the material variation inherent to the manufacturing process, graphite will experience changes in volume, mechanical strength, and thermal properties while in service in a nuclear reactor along with the possibility of cracking due to stress and oxidation resulting from constituents in the gas coolant or oxygen ingress. Therefore, there is the recognized need to be able to nondestructively characterize a variety of material attributes such as uniformity, isotropy, and porosity distributions as a means to assure consistent stock material. This need also includes the ability to detect isolated defects such as cracks, large pores and inclusions, or distributed material damage such as material loss due to oxidation. The use of this guide is to acquire a basic understanding of the unique attributes of nuclear grade graphite and its application that either permits or hinders the use of conventional eddy current, ultrasonic, or X-ray inspection technologies.1.1 This guide provides general tutorial information regarding the application of conventional nondestructive evaluation technologies (NDE) to nuclear grade graphite. An introduction will be provided to the characteristics of graphite that defines the inspection technologies that can be applied and the limitations imposed by the microstructure. This guide does not provide specific techniques or acceptance criteria for end-user examinations but is intended to provide information that will assist in identifying and developing suitable approaches.1.2 The values stated in SI units are to be regarded as the standard.1.2.1 Exception—Alternative units provided in parentheses are for information 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 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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This specification covers blended uranium oxides with a 235U content of less than 5% for direct hydrogen reduction to nuclear grade uranium dioxide. For commercial-grade uranium oxide with an isotopic content of 235U between that of natural uranium and 5%, the isotopic limits shall apply. Physical and chemical requirements include: uranium content, oxygen-to-uranium ratio, impurity content, equivalent boron content, bulk density, moisture content, ability to flow, particle size, and reduction and sinterability. Maximum concentration limit is specified for impurity elements such as: aluminum, barium, beryllium, bismuth, calcium+magnesium, carbon, chlorine, chromium, cobalt, copper, fluorine, iron, lead, manganese, molybdenum, nickel, phosphorus, silicon, sodium, tantalum, thorium, tin, titanium, tungsten, vanadium, and zinc. The identity of a lot shall be retained throughout.1.1 This specification covers blended uranium trioxide (UO3), U3O8, or mixtures of the two, powders that are intended for conversion into a sinterable uranium dioxide (UO2) powder by means of a direct reduction process. The UO2 powder product of the reduction process must meet the requirements of Specification C 753 and be suitable for subsequent UO2 pellet fabrication by pressing and sintering methods. This specification applies to uranium oxides with a 235U enrichment less than 5 %.1.2 This specification includes chemical, physical, and test method requirements for uranium oxide powders as they relate to the suitability of the powder for storage, transportation, and direct reduction to UO2 powder. This specification is applicable to uranium oxide powders for such use from any source.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 uranium oxide powders (see 2.2 and 2.3).1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.1.5 The following safety hazards caveat pertains only to the test methods portion of the annexes in this specification: This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

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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.

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5.1 The results of this test method give a conservative estimate of the performance of nuclear-grade activated carbon used in all nuclear power plant HVAC systems for the removal of radioiodine.1.1 This test method is a very stringent procedure for establishing the capability of new and used activated carbon to remove radio-labeled methyl iodide from air and gas streams. The single test method described is for application to both new and used carbons, and should give test results comparable to those obtained from similar tests required and performed throughout the world. The conditions employed were selected to approximate operating or accident conditions of a nuclear reactor which would severely reduce the performance of activated carbons. Increasing the temperature at which this test is performed generally increases the removal efficiency of the carbon by increasing the rate of chemical and physical absorption and isotopic exchange, that is, increasing the kinetics of the radioiodine removal mechanisms. Decreasing the relative humidity of the test generally increases the efficiency of methyl iodide removal by activated carbon. The water vapor competes with the methyl iodide for adsorption sites on the carbon, and as the amount of water vapor decreases with lower specified relative humidities, the easier it is for the methyl iodide to be adsorbed. Therefore, this test method is a very stringent test of nuclear-grade activated carbon because of the low temperature and high relative humidity specified. This test method is recommended for the qualification of new carbons and the quantification of the degradation of used carbons.1.1.1 Guidance for testing new and used carbons using conditions different from this test method is offered in Annex A1.1.2 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.1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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1.1 This terminology standard contains terms, definitions, descriptions of terms, nomenclature, and explanations of acronyms and symbols specifically associated with standards under the jurisdiction of Committee C26 on Nuclear Fuel Cycle. The content of this terminology standard may also be applicable to documents not under the jurisdiction of Committee C26, in which case this terminology standard may be referenced in those documents.1.2 While subcommittees within Committee C26 are free to only provide terms and definitions within individual standards, each subcommittee may request the addition of utilized terms and definitions to this terminology standard if it believes that such serves the broader interest of Committee C26 and the nuclear fuel cycle profession. Therefore, terms and definitions proposed for inclusion in Terminology C859 need not be used in more than one committee standard before being considered.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.

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4.1 This test method is designed to provide a uniform test to assess the suitability of coatings, used in nuclear power facilities, under radiation exposure for the life of the facilities, including radiation during a DBA (Coating Service Level I areas only). Specific plant radiation exposure may exceed or be less than the amount specified in 7.2 of this standard. If required by the licensee design basis, the gamma dose used may exceed the actual anticipated plant gamma dose to account for beta dose. Coatings in Level II and III areas (outside primary containment) are expected to be exposed to lower accumulated radiation doses.1.1 This test method covers a standard procedure for evaluating the lifetime radiation tolerance of coatings to be used in nuclear power plants. This test method is applicable to Coating Service Levels I, II, and III.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 standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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5.1 To establish a proper calibration area for nuclear surface gauges.5.2 To reduce the chance of improper calibration.NOTE 1: The quality of the results produced by this standard is dependent on the competence of the personnel performing it, and the suitability of the equipment and facilities used. Agencies that meet the criteria of practice D3740 are generally considered capable of competent and objective testing/inspection/etc. Users of this standard are cautioned that compliance with practice D3740 does not in itself assure a means of evaluating some of those factors.1.1 This guide outlines procedures for setup of a nuclear gauge calibration facility in either a shielded bay or an unshielded area—Guide A and Guide B, respectively.1.2 This guide does not attempt to describe the calibration techniques or methods. It is assumed that this guide will be used by persons familiar with the operations of the gauge and in performing proper calibration, service and maintenance.1.3 This guide does not attempt to address maintenance or service procedures related to the gauge.1.4 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.1.5 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 guide offers an organized collection of information or a series of options and does not recommend a specific course of action. This document cannot replace education or experience and should be used in conjunction with professional judgment. Not all aspects of this guide may be applicable in all circumstances. This ASTM standard is not intended to represent or replace the standard of care by which the adequacy of a given professional service must be judged, nor should this document be applied without consideration of a project’s many unique aspects. The word “Standard” in the title of this document has been approved through ASTM consensus process.1.7 All observed and calculated values shall conform to the guidelines for significant digits and rounding established in practice D6026.1.7.1 The method used to specify how data are collected, calculated, or recorded in this standard is not directly related to the accuracy to which the data can be applied in the design or other uses, or both. How one applies the results obtained using this standard is beyond its scope.1.8 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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4.1 Gadolinium oxide powder is used, with subsequent processing, in nuclear fuel applications, such as an addition to uranium dioxide. These test methods are designed to determine whether the material meets the requirements described in Specification C888.4.1.1 The material is analyzed to determine whether it contains the minimum gadolinium oxide content specified.4.1.2 The loss on ignition and impurity content are determined to ensure that the weight loss and the maximum concentration limit of specified impurity elements are not exceeded.1.1 These test methods cover procedures for the chemical and mass spectrometric analysis of nuclear-grade gadolinium oxide powders to determine compliance with specifications.1.2 The analytical procedures appear in the following order:  SectionsCarbon by Direct Combustion—Thermal Conductivity 2C1408 Test Method for Carbon (Total) in Uranium Oxide Powders and Pellets By Direct Combustion-Infrared Detection Method 3Total Chlorine and Fluorine by Pyrohydrolysis Ion— Selective Electrode 4C1502 Test Method for Determination of Total Chlorine and Fluorine in Uranium Dioxide and Gadolinium Oxide 3Loss of Weight on Ignition 8 – 14Sulfur by Combustion—Iodometric Titration 5Impurity Elements by a Spark-Source Mass Spectrographic Method   C761 Test Methods for Chemical, Mass Spectrometric, Spectrochemical, Nuclear, and Radiochemical Analysis of Uranium Hexafluoride 3 C1287 Test Method for Determination of Impurities in Nuclear Grade Uranium Compounds by Inductively Coupled Plasma Mass Spectrometry 3Gadolinium Content in Gadolinium Oxide by Impurity Correction Method 15 – 181.3 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 6.1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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4.1 Quality assurance, as covered by this practice, comprises all those planned and systematic actions necessary to provide adequate confidence that safety-related coating work in nuclear facilities as defined in Guide D5144, will perform satisfactorily in service.4.2 It is not practical to impose all the requirements of this practice on certain specific items that require only a small quantity of coating material. The licensee, consistent with his formal Quality Assurance Program, may accept affidavits of compliance or certification attesting to the quality of a shop or field coating for such items. If required by licensing commitment; safety-related coatings that are not qualified or for which the quantification basis is indeterminate as defined in Guide D5144, shall be identified, quantified, and documented.4.3 This practice may be incorporated in a project specification by direct reference or may be used to provide guidelines for the quality assurance program for coatings, on the basis of the licensee’s requirements. Effective use of this practice may also require the incorporation of applicable sections in project specifications for coatings on concrete, steel, equipment, and other related items.1.1 This standard replaces ANSI N101.4 and provides a common basis for, and specifically comprises quality assurance requirements applicable to, safety-related protective coating work in Coating Service Level I areas of nuclear facilities as defined in Guide D5144.1.2 This standard meets the requirements of ANSI N101.4 while also recognizing advancements in technology and industry practices since transfer to ASTM responsibility for updating, rewriting, and issuing replacement standards to ANSI N101.4.1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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5.1 Segmented gamma-ray scanning provides a nondestructive means of measuring the nuclide content of scrap and waste where the specific nature of the matrix and the chemical form and relationship between the nuclide and matrix may be unknown.5.2 The procedure can serve as a diagnostic tool that provides a vertical profile of transmission and nuclide concentration within the item.5.3 Item preparation is generally limited to good waste/scrap segregation practices that produce relatively homogeneous items that are required for any successful waste/inventory management and assay scheme, regardless of the measurement method used. Also, process knowledge should be used, when available, as part of a waste management program to complement information on item parameters, container properties, and the appropriateness of calibration factors.5.4 To obtain the lowest detection levels, a two-pass assay should be used. The two-pass assay also reduces problems related to potential interferences between transmission peaks and assay peaks. For items with higher activities, a single-pass assay may be used to increase throughput.1.1 This test method covers the transmission-corrected nondestructive assay (NDA) of gamma-ray emitting special nuclear materials (SNMs), most commonly 235U, 239Pu, and 241Am, in low-density scrap or waste, packaged in cylindrical containers. The method can also be applied to NDA of other gamma-emitting nuclides including fission products. High-resolution gamma-ray spectroscopy is used to detect and measure the nuclides of interest and to measure and correct for gamma-ray attenuation in a series of horizontal segments (collimated gamma detector views) of the container. Corrections are also made for counting losses occasioned by signal processing limitations (1-3).21.2 There are currently several systems in use or under development for determining the attenuation corrections for NDA of radioisotopic materials (4-8). A related technique, tomographic gamma-ray scanning (TGS), is not included in this test method (9, 10, 11).1.2.1 This test method will cover two implementations of the Segmented Gamma Scanning (SGS) procedure: (1) Isotope Specific (Mass) Calibration, the original SGS procedure, uses standards of known radionuclide masses to determine detector response in a mass versus corrected count rate calibration that applies only to those specific radionuclides for which it is calibrated, and (2) Efficiency Curve Calibration, an alternative method, typically uses non-SNM radionuclide sources to determine system detection efficiency vs. gamma energy and thereby calibrate for all gamma-emitting radionuclides of interest (12).1.2.1.1 Efficiency Curve Calibration, over the energy range for which the efficiency is defined, has the advantage of providing calibration for many gamma-emitting nuclides for which half-life and gamma emission intensity data are available.1.3 The assay technique may be applicable to loadings up to several hundred grams of nuclide in a 208-L [55-gal] drum, with more restricted ranges to be applicable depending on specific packaging and counting equipment considerations.1.4 Measured transmission values must be available for use in calculation of segment-specific attenuation corrections at the energies of analysis.1.5 A related method, SGS with calculated correction factors based on item content and density, is not included in this standard.1.6 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.1.7 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 10.1.8 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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3.1 These test methods are useful as a rapid, nondestructive technique for determination of asphalt content of asphalt mixtures.3.2 These test methods are suitable for quality control and acceptance testing for construction and for research and development applications. The test method is used for determination of asphalt content only as it does not provide extracted aggregate for gradation analysis.3.3 The nondestructive nature of the test allows repetitive measurements to be made on a single test sample for statistical analysis of test data.3.4 These test methods determine the asphalt content of a test sample by comparing the measured asphalt content with previously established calibration data.3.4.1 The asphalt content of a material expressed as a percentage is the ratio of the mass of asphalt in a given mass of material to the total mass of the sample or to the mass of the solid material particles.1.1 These test methods cover the procedures for determining the asphalt content of samples of uncompacted asphalt mixtures (Test Method A), and of laboratory compacted specimens of asphalt mixtures (Test Method B) by examining a test sample with an apparatus that utilizes neutron thermalization techniques.1.2 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 nonconformance with the standard.1.3 A precision and bias statement for Method B in this standard has not been developed at this time. Therefore, Method B should not be used for acceptance or rejection of a material for purchasing purposes.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. See Section 6 and 8.4.2, 8.5.6, and Note 4 for specific 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.

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1.1 This specification establishes the requirements for plate, sheet, strip, and rolled bar copper alloy UNS No. C15815, dispersion strengthened copper (DSC), which has been modified by the addition of natural or enriched boron for absorption of neutrons. 1.1.1 The products made to this specification are under consideration for use in storage containers for spent nuclear fuel. Note 1- Dispersion strengthened copper is thermally stable high-strength copper which retains a high percentage of its strength at elevated temperature. In addition, it does not recrystallize or soften after exposure to high temperatures almost to the melting point of copper. Since dispersion strengthened copper uses inert aluminum oxide particles to strengthen the copper matrix, the thermal conductivity of the copper is not significantly decreased. Note 2- Borated dispersion strengthened coppers covered by this specification, because of their particular structure and specialized properties, may require special care in their fabrication and welding. 1.2 The values stated in inch-pound units are the standard. SI units are provided for information only.

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4.1 For criticality control of nuclear fuel in dry storage and transportation, the most commonly used neutron absorber materials are borated stainless steel alloys, borated aluminum alloys, and boron carbide aluminum alloy composites. The boron used in these neutron absorber materials may be natural or enriched in the nuclide 10B. The boron is usually incorporated either as an intermetallic phase (for example, AlB2, TiB2, CrB2, etc.) in an aluminum alloy or stainless steel, or as a stable chemical compound particulate such as boron carbide (B4C), typically in an aluminum MMC or cermet.4.2 While other neutron absorbers continue to be investigated, 10B has been most widely used in these applications, and it is the only thermal neutron absorber addressed in this standard.4.3 In service, many neutron absorber materials are inaccessible and not amenable to a surveillance program. These neutron absorber materials are often expected to perform over an extended period.4.4 Qualification and acceptance procedures demonstrate that the neutron absorber material has the necessary characteristics to perform its design functions during the service lifetime.4.5 The criticality control function of neutron absorber materials in dry cask storage systems and transportation packagings is only significant in the presence of a moderator, such as during loading of fuel under water, or water ingress resulting from hypothetical accident conditions.4.6 The expected users of this standard include designers, neutron absorber material suppliers and purchasers, government agencies, consultants and utility owners. Typical use of the practice is to summarize practices which provide input for design specification, material qualification, and production acceptance. Adherence to this standard does not guarantee regulatory approval; a government regulatory authority may require different tests or additional tests, and may impose limits or restrictions on the use of a neutron absorber material.1.1 This practice provides procedures for qualification and acceptance of neutron absorber materials used to provide criticality control by absorbing thermal neutrons in systems designed for nuclear fuel storage, transportation, or both.1.2 This practice is limited to neutron absorber materials consisting of metal alloys, metal matrix composites (MMCs), and cermets, clad or unclad, containing the neutron absorber boron-10 (10B).1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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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 C1220.1.1 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.

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SNM monitors are an efficient and sensitive means of unobtrusively (without a body search) meeting the requirements of 10 CFR (Code of Federal Regulations) Part 73 or DOE Order 5632.4 (May 1986) that individuals exiting nuclear material access areas (MAAs) be searched for concealed SNM. The monitors sense radiation emitted by SNM, which is an excellent but otherwise imperceptible clue to the presence of the material. Because the monitors operate in a natural radiation environment and must detect small intensity increases as clues, the monitors must be well designed and maintained to operate without unnecessary nuisance alarms. This guide provides information on different types of monitors for searching pedestrians and vehicles. Each monitor has an inherent sensitivity at a particular nuisance alarm rate that must be low enough to maintain the monitor’credibility. Sensitivity and nuisance alarm rates are both governed by the alarm threshold so it is very important that corresponding values for both be known when measured, estimated, or specified values are discussed. Fitting SNM monitors into a facility physical protection plan must not only consider adequate sensitivity but also a sufficiently low nuisance alarm rate.1.1 This guide briefly describes the state-of-the-art of radiation monitors for detecting special nuclear material (SNM) (see 3.1.11) in order to establish the context in which to write performance standards for the monitors. This guide extracts information from technical documentation to provide information for selecting, calibrating, testing, and operating such radiation monitors when they are used for the control and protection of SNM. This guide offers an unobtrusive means of searching pedestrians, packages, and motor vehicles for concealed SNM as one part of a nuclear material control or security plan for nuclear materials. The radiation monitors can provide an efficient, sensitive, and reliable means of detecting the theft of small quantities of SNM while maintaining a low likelihood of nuisance alarms. 1.2 Dependable operation of SNM radiation monitors rests on selecting appropriate monitors for the task, operating them in a hospitable environment, and conducting an effective program to test, calibrate, and maintain them. Effective operation also requires training in the use of monitors for the security inspectors who attend them. Training is particularly important for hand-held monitoring where the inspector plays an important role in the search by scanning the instrument over pedestrians and packages or throughout a motor vehicle. 1.3 SNM radiation monitors are commercially available in three forms: 1.3.1 Small Hand-Held Monitors—These monitors may be used by an inspector to manually search pedestrians and vehicles that stop for inspection. 1.3.2 Automatic Pedestrian Monitors—These monitors are doorway or portal monitors that search pedestrians in motion as they pass between radiation detectors, or wait-in monitoring booths that make extended measurements to search pedestrians while they stop to obtain exit clearance. 1.3.3 Automatic Vehicle Monitors—These monitors are portals that monitor vehicles as they pass between radiation detectors, or vehicle monitoring stations that make extended measurements to search vehicles while they stop to obtain exit clearance. 1.4 Guidance for applying SNM monitors is available as Atomic Energy Commission/Nuclear Regulatory Commission (AEC/NRC) regulatory guides, AEC/ERDA/DOE performance standards, and more recently as handbooks and applications guides published by national laboratories under DOE sponsorship. This broad information base covering the pertinent physics, engineering practice, and equipment available for monitoring has had no automatic mechanism for periodic review and revision. This ASTM series of guides and standards will consolidate the information in a form that is reexamined and updated on a fixed schedule. 1.5 Up-to-date information on monitoring allows both nuclear facilities and regulatory agencies to be aware of the current range of monitoring alternatives. Up-to-date information also allows manufacturers to be aware of the current goals of facilities and regulators, for example, to obtain particular sensitivities at a low nuisance alarm rate with instrumentation that is dependable and easy to maintain. 1.6 This guide updates and expands the scope of NRC regulatory guides and AEC/ERDA/DOE SNM monitor performance standards using the listed publications as a technical basis. 1.7 The values stated in SI units are to be regarded as the standard. 1.8 This standard may involve hazardous materials, operations, and equipment. This standard does not purport to address all of the safety problems associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

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