4.1 This test method is designed to provide a uniform test to determine the suitability of Coating Service Level 1 coatings used inside primary containment of light-water nuclear facilities under simulated DBA conditions. This test method is intended only to demonstrate that under DBA conditions, the coatings will remain intact and not form debris which could unacceptably compromise the operability of engineered safety systems. Deviations in actual surface preparation and in application and curing of the coating materials from qualification test parameters require an engineering evaluation to determine if additional testing is required.4.2 Since different plants have different tolerance levels for coating conditions, the definition of appropriate acceptance criteria is to be developed by the license holder based on individual plant engineered safety systems operability considerations.4.3 Use of this standard is predicated on the testing facility having a quality assurance program acceptable to the licensee.1.1 This test method establishes procedures for evaluating protective coating systems test specimens under simulated DBA conditions. Included are a description of conditions and apparatus for temperature-pressure testing, and requirements for preparing, irradiating, testing, examining, evaluating, and documenting the samples.1.2 Consideration should be given to testing using worst case conditions (for example, surface preparation, temperature and pressure profile, irradiation, spray chemistry, chemical resistance, etc.) in an effort to reduce the number of tests required by changing plant accident calculations, changes in coating selection, etc.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 units that 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.

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5.1 Performance properties are dependent on the number and type of short chain branches. This test method permits measurement of these branches for ethylene copolymers with propylene, butene-1, hexene-1, octene-1, and 4-methylpentene-1.1.1 This test method determines the molar composition of copolymers prepared from ethylene (ethene) and a second alkene-1 monomer. This second monomer can include propene, butene-1, hexene-1, octene-1, and 4-methylpentene-1.1.2 Calculations of this test method are valid for products containing units EEXEE, EXEXE, EXXE, EXXXE, and of course EEE where E equals ethene and X equals alkene-1. Copolymers containing a considerable number of alkene-1 blocks (such as, longer blocks than XXX) are outside the scope of this test method.1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. See Section 8 for a specific hazard statement.NOTE 1: There is no known ISO equivalent to this standard.

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5.1 The test method is designed to show whether or not a material meets the specifications as given in Specifications C753 or C776.5.2 The powder’s stoichiometry is useful for predicting the oxide's sintering behavior in the pellet production process.1.1 This test method covers the determination of uranium and the oxygen to uranium atomic ratio in nuclear grade uranium dioxide powder and pellets.1.2 This test method does not include provisions for preventing criticality accidents or requirements for health and safety. Observance of this test method does not relieve the user of the obligation to be aware of and conform to all international, national, or federal, state and local regulations pertaining to possessing, shipping, processing, or using source or special nuclear material.1.3 This test method also is applicable to UO3 and U3O8 powder.1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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3.1 The lining test described in 6.2 may be used to evaluate the chemical resistance characteristics of coating systems for lining surfaces of tanks, vessels and similar facilities used in Coating Service Level I and II applications in a nuclear power plant. For the evaluation of linings in Coating Service Level III water immersion applications in nuclear power plants use the test methods and guidance found in Guide D7230.3.2 The specific chemical resistance tests described in 6.1 are dependent upon the relative severity of the service conditions. The specific chemical reagents to be used shall be specified to reflect the intended service conditions.3.3 At the discretion of the user, the methods presented may also be used to evaluate coatings and linings for applications in other types of power plants or other industrial services.1.1 This test method establishes procedures for the evaluation of the chemical resistance of coatings and linings for use in Coating Service Level I and II applications in nuclear power plants.1.2 This test method is intended to be used as a screening test to evaluate coatings and linings on steel and concrete substrates.1.3 This test method addresses two exposure intervals:(1) Short Term (Typically 5 days): Such exposures are primarily applicable for coatings exposed to chemical splash or spill.(2) Long Term (Typically 180 days): Such exposures are primarily applicable for linings exposed to continuous or near-continuous chemical immersion.1.4 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.1.5 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 guide covers testing protocols for testing the pyrophoricity/combustibility characteristics of metallic uranium-based spent nuclear fuel (SNF). The testing will provide basic data for input into more detailed computer codes or analyses of thermal, chemical, and mechanical SNF responses. These analyses would support the engineered barrier system (EBS) design bases and safety assessment of extended interim storage facilities and final disposal in a geologic repository. The testing also could provide data related to licensing requirements for the design and operation of a monitored retrievable storage facility (MRS) or independent spent fuel storage installation (ISFSI).1.2 This guide describes testing of metallic uranium and metallic uranium-based SNF in support of transportation (in accordance with the requirements of 10CFR71), interim storage (in accordance with the requirements of 10CFR72), and geologic repository disposal (in accordance with the requirements of 10CFR60/63). The testing described herein is designed to provide basic data related to the evaluation of the pyrophoricity/combustibility characteristics of containers or waste packages containing metallic uranium SNF in support of safety analyses (SAR), or performance assessments (PA) of transport, storage, or disposal systems, or a combination thereof.1.3 Spent nuclear fuel that is not reprocessed must be emplaced in secure temporary interim storage as a step towards its final disposal in a geologic repository. In the United States, SNF, from both civilian commercial power reactors and defense nuclear materials production reactors, will be sent to interim storage, and subsequently, to deep geologic disposal. U.S. commercial SNF comes predominantly from light water reactors (LWRs) and is uranium dioxide-based, whereas U.S. Department of Energy (DOE) owned defense reactor SNF is in several different chemical forms, but predominantly (approximately 80 % by weight of uranium) consists of metallic uranium.1.4 Knowledge of the pyrophoricity/combustibility characteristics of the SNF is required to support licensing activities for extended interim storage and ultimate disposition in a geologic repository. These activities could include interim storage configuration safety analyses, conditioning treatment development, preclosure design basis event (DBE) analyses of the repository controlled area, and postclosure performance assessment of the EBS.1.5 Metallic uranium fuels are clad, generally with zirconium, aluminum, stainless steel, or magnesium alloy, to prevent corrosion of the fuel and to contain fission products. If the cladding is damaged and the metallic SNF is stored in water the consequent corrosion and swelling of the exposed uranium enhances the chemical reactivity of the SNF by further rupturing the cladding and creating uranium hydride particulates and/or inclusions in the uranium metal matrix. The condition of the metallic SNF will affect its behavior in transport, interim storage or repository emplacement, or both, and therefore, influence the engineering decisions in designing the pathway to disposal.1.6 Zircaloy spent fuel cladding has occasionally demonstrated pyrophoric behavior. This behavior often occurred on cladding pieces or particulate residues left after the chemical dissolution of metallic uranium or uranium dioxide during fuel reprocessing of commercial spent fuel and/or extraction of plutonium from defense reactor spent fuel. Although it is generally believed that zirconium is not as intrinsically prone to pyrophoric behavior as uranium or plutonium, it has in the past ignited after being sensitized during the chemical extraction process. Although this guide primarily addresses the pyrophoricity of the metallic uranium component of the spent fuel, some of the general principles involved could also apply to zirconium alloy spent fuel cladding.1.7 The interpretation of the test data depends on the characteristics of the sample tested and/or the usage to which the test results are put. For example, usage could include simple comparison of the relative ignition temperature of different sample configurations or as inputs to more complex computer simulations of spontaneous ignition. The type and the size of the SNF sample must be chosen carefully and accounted for in the usage of the data. The use of the data obtained by the testing described herein may require that samples be used which mimic the condition of the SNF at times far into the future, for example, the repository postcontainment period. This guide does not specifically address methods for `aging' samples for this purpose. The section in Practice C 1174 concerning the accelerated testing of waste package materials is recommended for guidance on this subject.

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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 available5.4.1 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 assumptions can be used to detect or track relative changes in the holdup quantity at that point over time. Either a qualitative or a quantitative method can be used.5.6 Indirect Measurements—Quantity of a radionuclide can be determined by measurement of a daughter radionuclide or of a second radionuclide if the ratio of the abundances of the two radionuclides is known and secular equilibrium (Terminology C1673) is present. This can be used when there are interfering gamma rays or when the parent radionuclide does not have a sufficiently strong gamma-ray signal to be readily measured. If this method is employed, it is important that the ratio of the two radionuclides be known with sufficient accuracy to meet assay uncertainty goals.5.7 Mathematical Modeling—Modeling is an aid in the evaluation of complex measurement situations. Measurement data are used with a mathematical model describing the physical location of equipment and materials. (3, 5, 6, 7, 8) .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) .21.3 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 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 the requirements for sheathed, Type K and N thermocouples for nuclear service. This specification can be used for sheathed thermocouples which are required for laboratory or general commercial applications where the environmental conditions exceed normal service requirements. The measuring junction styles for thermocouples are as follows: Style G2 (grounded) in which measuring junction is electrically connected to conductive sheaths and Style U2 (ungrounded) in which measuring junctions are electrically isolated from conductive sheaths and from reference ground. Different properties of the sheath such as integrity, cracks, voids, inclusions, surface finish, surface defect, and metallurgical structure shall be determined by performing different tests. Insulation resistance between thermoelements and the sheath shall be measured as well.1.1 This specification covers the requirements for simplex, compacted mineral-insulated, metal-sheathed (MIMS), Type K and N thermocouples for nuclear or other high reliability service. Depending on size, these thermocouples are normally suitable for operating temperatures to 1652 °F [900 °C]; special conditions of environment and life expectancy may permit their use at temperatures in excess of 2012 °F [1100 °C]. This specification was prepared to detail requirements for this type of MIMS thermocouple for use in nuclear environments, but they can also be used for laboratory or general commercial applications where the environmental conditions exceed normal service requirements. The intended use of a MIMS thermocouple in a specific nuclear application will require evaluation of the compatibility of the thermocouple, including the effect of the temperature, atmosphere, and integrated neutron flux on the materials and accuracy of the thermoelements in the proposed application by the purchaser.1.2 This specification does not attempt to include all possible specifications, standards, etc., for materials that may be used as sheathing, insulation, and thermocouple wires for sheathed-type construction. The requirements of this specification include only the austenitic stainless steels and other alloys as allowed by Specification E585/E585M for sheathing, magnesium oxide or aluminum oxide as insulation, and Type K and N thermocouple wires for thermoelements (see Note 1).1.3 General Design—Nominal sizes of the finished thermocouples shall be 0.0400 in., 0.0625 in., 0.125 in., 0.1875 in., or 0.250 in. [1.000 mm, 1.500 mm, 3.000 mm, 4.500 mm, or 6.000 mm]. Sheath dimensions and tolerances for each nominal size shall be in accordance with Table 1 and Figs. 1 and 2. The measuring junction styles for thermocouples covered by this specification are as follows:FIG. 1 Grounded Measuring Junction, Style GFIG. 2 Ungrounded Measuring Junction, Style U1.3.1 Style G2 (grounded)—The measuring junction is electrically connected to its conductive sheath, and1.3.2 Style U2 (ungrounded)—The measuring junction is electrically isolated from its conductive sheath and from reference ground.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 are not exact equivalents or conversions; 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 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 Use of this guide will ensure that the potential impact on the surrounding environment from planned decommissioning activities has been properly assessed.5.2 Use of this guide will ensure that the adequacy of environmental sampling has been assessed for location, frequency, analytical techniques, and media type to monitor the environment and to detect site-related releases and their impact.1.1 This guide covers the development or assessment of environmental monitoring plans for decommissioning nuclear facilities. This guide addresses: (1) development of an environmental baseline prior to commencement of decommissioning activities; (2) determination of release paths from site activities and their associated exposure pathways in the environment; and (3) selection of appropriate sampling locations and media to ensure that all exposure pathways in the environment are monitored appropriately. This guide also addresses the interfaces between the environmental monitoring plan and other planning documents for site decommissioning, such as radiation protection, site characterization, and waste management plans, and federal, state, and local environmental protection laws and guidance. This guide is applicable up to the point of completing D&D activities and the reuse of the facility or area for other purposes.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.

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5.1 The mathematical and statistical techniques described in this guide support implementation of the calibration requirements of Practice D7282 and the guidance for uncertainty analysis given in Guide D8293. The guidance is intended for use either by qualified specialists at a radioanalytical laboratory or by developers of software for calibration of nuclear instruments.5.2 Applications for single-point calibrations might include:5.2.1 Alpha-particle spectrometry,5.2.2 Gas proportional counters used for thin sources with negligible attenuation, and5.2.3 Gamma-ray spectrometers used for single nuclides.5.3 Applications for calibration curves determined by LLS might include:5.3.1 Mass attenuation curves for gas proportional counters (polynomial), and5.3.2 Quench calibration curves for liquid scintillation counters (polynomial).5.4 Applications for calibration curves determined by NLLS might include:5.4.1 Gamma-ray spectrometry across a range of gamma-ray energies,5.4.2 Mass attenuation curves for gas proportional counters, and5.4.3 Quench calibration curves for liquid scintillation counters.5.5 Although this guide focuses on efficiency calibrations for nuclear instruments, the same general principles and paradigms should apply to other types of calibrations and to other instruments, as long as there are valid uncertainty models for the calibration data.1.1 This guide describes data analysis for efficiency calibrations of nuclear instruments using radioactive sources. It includes the calculation of the calibration parameters, evaluation and use of their uncertainties and covariances, and testing of the calibration data for outliers and overall lack of fit. It also provides guidelines for summarizing and reporting the results of a calibration.1.2 The instrument counting efficiency is assumed to be independent of the radiation emission rate.1.3 Guidance is provided for both single-point calibrations and calibration curves.1.4 The guidance presumes the existence of measurement uncertainty models to provide statistical weighting factors for the calibration data.1.5 This guide does not cover calibrations involving physically-based computer simulations.1.6 The system of units for this guide is not specified. Dimensional quantities in the guide are presented only as illustrations of calculation methods. The examples are not binding on products or test methods treated.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.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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1. Scope 1.1 This Standard outlines the requirements for seismic instrumentation systems for nuclear power plants where site- specific seismic responses are required to be determined and recorded.

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1. Scope 1.1 This Standard covers the design, fabrication, qualification, installation, and inspection of CANDU nuclear power plant electrical and instrument air support power systems. These support power systems are part of the category of systems ca

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This PDF includes GI #2. 1. Scope This Standard provides requirements and guidelines for protection against fires in CANDU nuclear power plants. It applies to the design, construction, commissioning, all operational phases, and decommissioning of the

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1.1 These test methods cover the determination of the total or wet density of soil and soil-rock mixtures by the attenuation of gamma radiation where the source and detector(s) remain on the surface (Backscatter Method) or the source or detector is placed at a known depth up to 300 mm (12 in.) while the detector(s) or source remains on the surface (Direct Transmission Method).1.2 The density in mass per unit volume of the material under test is determined by comparing the detected rate of gamma radiation with previously established calibration data.1.3 The values tested in SI units are to be regarded as the standard. The inch-pound equivalents may be approximate.1.4 It is common practice in the engineering profession to concurrently use pounds to represent both a unit of mass (lbm) and a unit of force (lbf). This implicitly combines two separate systems of units; that is, the absolute system and the gravitational system. It is scientifically undesirable to combine the use of two separate sets of inch-pound units within a single standard. These test methods have been written using the gravitational system of units when dealing with the inch-pound system. In this system the pound (lbf) represents a unit of force (weight). However, the use of balances or scales recording pounds of mass (lbm), or the recording of density in lbm/ft 3 should not be regarded as nonconformance with these test methods.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. For specific Hazard statements, see Section .

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4.1 Disposition of aluminum-based spent nuclear fuel will involve:4.1.1 Removal from the existing storage or transfer facility,4.1.2 Characterization or treatment, or both, of the fuel or the resulting waste form, or both,4.1.3 Placement of the waste form in a canister,4.1.4 Placement of the canister in a safe and environmentally sound interim storage facility,4.1.5 Removal from the interim storage facility and transport to the repository,4.1.6 Placement in a waste container,4.1.7 Emplacement in the repository, and4.1.8 Repository closure and geologic disposal. Actions in each of these steps may significantly impact the success of any subsequent step.4.2 Aluminum-based spent nuclear fuel and the aluminum-based waste forms display physical and chemical characteristics that differ significantly from the characteristics of commercial nuclear fuels and from high level radioactive waste glasses. For example, some are highly enriched and most have heterogeneous microstructures that include very small, uranium-rich particles. The impact of this difference on repository performance must be evaluated and understood.4.3 The U.S. Nuclear Regulatory Commission has licensing authority over public domain transportation and repository disposal (and most of the interim dry storage) of spent nuclear fuels and high-level radioactive waste under the requirements established by 10 CFR 60, 10 CFR 71, and 10 CFR 72. These requirements outline specific information needs that should be met through test protocols, for example, those mentioned in this guide. The information developed from the tests described in this guide is not meant to be comprehensive. However, the tests discussed here will provide corrosion property data to support the following information needs.4.3.1 A knowledge of the solubility, leaching, oxidation/reduction reactions, and corrosion of the waste form constituents in/by the repository environment (dry air, moist air, and repository relevant water) (see 10 CFR 60.112 and 135).4.3.2 A knowledge of the effects of radiolysis and temperature on the oxidation, corrosion, and leaching behavior (see 10 CFR 60.135).4.3.3 A knowledge of the temperature dependence of the solubility of waste form constituents plus oxidation and corrosion products (see 10 CFR 60.135).4.3.4 Information from laboratory experiments or technical analyses, or both, about time dependence of the internal condition of the waste package (see 10 CFR 60.143 and 10 CFR 72.76).4.3.5 Laboratory demonstrations of the effects of the electrochemical differences between the aluminum-based waste form and the candidate packaging materials on galvanic corrosion (see 10 CFR 71.43) or the significance of electrical contact between the waste form and the packaging materials on items outlined in 4.3.1 – 4.3.4 (see 10 CFR 60.135), or both.4.3.6 Information on the risk involved in the receipt, handling, packaging, storage, and retrieval of the waste forms (see 10 CFR 72.3).4.3.7 Information on the physical and chemical condition of the waste form upon repository placement so that items outlined in 4.3.1 – 4.3.4 can be evaluated (see 10 CFR 60.135).4.3.8 Knowledge of the degradation of the waste form during interim storage so that operational safety problems with respect to its removal from storage can be assessed, if such removal is necessary (see 10 CFR 72.123).4.3.9 Knowledge of the condition of the waste form prior to repository placement so that items outlined in 4.3.1 – 4.3.4 are properly addressed (see 10 CFR 60.135).4.4 Conditions expected during each stage of the disposition process must be addressed. Exposure conditions anticipated over the interim storage through geologic disposition periods include dry and moist air, and aqueous environments. The air environments are associated with interim storage and the early stages of repository storage while the aqueous environments arise after water intrusion into the repository has caused corrosion-induced failure of the waste package.1.1 This guide covers corrosion testing of aluminum-based spent nuclear fuel in support of geologic repository disposal (per the requirements in 10 CFR 60 and 40CFR191). The testing described in this document is designed to provide data for analysis of the chemical stability and radionuclide release behavior of aluminum-based waste forms produced from aluminum-based spent nuclear fuels. The data and analyses from the corrosion testing will support the technical basis for inclusion of aluminum-based spent nuclear fuels in the repository source term. Interim storage and transportation of the spent fuel will precede geologic disposal; therefore, reference is also made to the requirements for interim storage (per 10 CFR 72) and transportation (per 10 CFR 71). The analyses that will be based on the data developed are also necessary to support the safety analyses reports (SARs) and performance assessments (PAs) for disposal systems.1.2 Spent nuclear fuel that is not reprocessed must be safely managed prior to transportation to, and disposal in, a geologic repository. Placement in an interim storage facility may include direct placement of the irradiated fuel or treatment of the fuel prior to placement, or both. The aluminum-based waste forms may be required to be ready for geologic disposal, or road ready, prior to placement in extended interim storage. Interim storage facilities, in the United States, handle fuel from civilian commercial power reactors, defense nuclear materials production reactors, and research reactors. The research reactors include both foreign and domestic reactors. The aluminum-based fuels in the spent fuel inventory in the United States are primarily from defense reactors and from foreign and domestic research reactors. The aluminum-based spent fuel inventory includes several different fuel forms and levels of 235U enrichment. Highly enriched fuels (235U enrichment levels >20 %) are part of this inventory.1.3 Knowledge of the corrosion behavior of aluminum-based spent nuclear fuels is required to ensure safety and to support licensing or other approval activities, or both, necessary for disposal in a geologic repository. The response of the aluminum-based spent nuclear fuel waste form(s) to disposal environments must be established for configuration-safety analyses, criticality analyses, PAs, and other analyses required to assess storage, treatment, transportation, and disposal of spent nuclear fuels. This is particularly important for the highly enriched, aluminum-based spent nuclear fuels. The test protocols described in this guide are designed to establish material response under the repository-relevant conditions.1.4 The majority of the aluminum-based spent nuclear fuels are aluminum clad, aluminum-uranium alloys. The aluminum-uranium alloy typically consists of uranium aluminide particles dispersed in an aluminum matrix. Other aluminum-based fuels include dispersions of uranium oxide, uranium silicide, or uranium carbide particles in an aluminum matrix. These particles, including the aluminides, are generally cathodic to the aluminum matrix. Selective leaching of the aluminum in the exposure environment may provide a mechanism for redistribution and relocation of the uranium-rich particles. Particle redistribution tendencies will depend on the nature of the aluminum corrosion processes and the size, shape, distribution and relative reactivity of the uranium-rich particles. Interpretation of test data will require an understanding of the material behavior. This understanding will enable evaluation of the design and configuration of the waste package to ensure that unfilled regions in the waste package do not provide sites for the relocation of the uranium-rich particles into nuclear critical configurations. Test samples must be evaluated, prior to testing, to ensure that the size and shape of the uranium-rich particles in the test samples are representative of the particles in the waste form being evaluated.1.5 The use of the data obtained by the testing described in this guide will be optimized to the extent the samples mimic the condition of the waste form during actual repository exposure. The use of Practice C1174 is recommended for guidance. The selection of test samples, which may be unaged or artificially aged, should ensure that the test samples and conditions bound the waste form/repository conditions. The test procedures should carefully describe any artificial aging treatment used in the test program and explain why that treatment was selected.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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4.1 Integral Mode Dosimetry—As shown in 3.2, two different integral relationships can be established using proton-recoil emulsion data. These two integral reactions can be obtained with roughly an order of magnitude reduction in scanning effort. Consequently, this integral mode is an important complementary alternative to the customary differential mode of NRE spectrometry. The integral mode can be applied over extended spatial regions, for example, perhaps up to as many as ten in-situ locations can be covered for the same scanning effort that is expended for a single differential measurement. Hence the integral mode is especially advantageous for dosimetry applications which require extensive spatial mapping, such as exist in Light Water Reactor-Pressure Vessel (LWR-PV) benchmark fields (see Test Method E1005). In low power benchmark fields, NRE can be used as integral dosimeters in a manner similar to RM, solid state track recorders (SSTR) and helium accumulation monitors (HAFM) neutron dosimeters (see Test Methods E854 and E910). In addition to spatial mapping advantages of these other dosimetry methods, NRE offer fine spatial resolution and can therefore be used in-situ for fine structure measurements. In integral mode scanning, both absolute reaction rates, that is I(ET) and J(Emin), are determined simultaneously. Separate software codes need to be used to permit operation of a computer based interactive system in the integral mode (see Section 9). It should be noted that the integrals I(ET) and J(Emin) possess different units, namely proton-recoil tracks/MeV per hydrogen atom and proton-recoil tracks per hydrogen atom, respectively. 4.2 Applicability for Spectral Adjustment Codes—In the integral mode, NRE provide absolute integral reaction rates that can be used in neutron spectrum least squares adjustment codes (see Guide E944). In the past, such adjustment codes could not utilize NRE integral reaction rates because of the non-existence of NRE data. NRE integral reaction rates provide unique benchmark data for use in least squares spectral adjustment codes. The unique significance of NRE integral data arises from a number of attributes, which are described separately below. Thus, inclusion of NRE integral reaction rate data in the spectral adjustment calculations can result in a significant improvement in the determination of neutron spectra in low power benchmark fields. 4.3 The Neutron Scattering Cross Section of Hydrogen—Integral NRE reaction rates are based on the standard neutron scattering cross section of hydrogen. For fast neutron spectrometry and dosimetry applications, the accuracy of this (n,p) cross section over extended energy regions is essentially unmatched. A semi-empirical representation of the energy-dependence of the (n,p) cross section is given in Eq 13. where: E is in MeV and σnp(E) is in barns. This energy-dependent representation of the (n,p) cross section possesses an uncertainty of approximately 1 % at the (1σ) level (19). 4.4 Threshold Energy Definition—In contrast with all other fast neutron dosimetry cross sections, the threshold energy of the I and J integral reaction rates can be varied. NRE integral reaction threshold variability extends down to approximately 0.3 to 0.4 MeV, which is the lower limit of applicability of the NRE method. Threshold variation is readily accomplished by using different lower bounds of proton track length to analyze NRE proton-recoil track length distributions. Furthermore, these NRE thresholds are more accurately defined than the corresponding thresholds of all other fast neutron dosimetry cross sections. NRE therefore provide a response with an extremely sharp energy cutoff that is not only unmatched by other cross sections, but an energy threshold that is independent of the in-situ neutron spectrum. No other fast neutron dosimetry cross sections possess a threshold response with these significant attributes. The behavior of the I-integral and J-integral response for different threshold energies is shown in Figs. 2 and 3, respectively, in comparison to the threshold 237Np(n,f) reaction used in RM dosimetry. FIG. 2 Comparison of the I-Integral Response with the 237Np (n,f) Threshold Reaction FIG. 3 Comparison of the J-Integral Response for ET = 0.404, 0.484, 0.554 and 0.620 MeV with the 237Np (n,f) Threshold Reaction 4.5 Complimentary Energy Response—It is of interest to compare the differential energy responses available from these two integral relations. From Eq 4 and 11, one finds responses of the form σ(E)/ E and (1 –Emin/E)σ(E) for the I and J integral relations, respectively. These two responses are compared in Fig. 4 using a common cut-off of 0.5 MeV for both ET and Emin. Since these two responses are substantially different, simultaneous application of these two integral relations would be highly advantageous. As shown in Fig. 4, the energy response of the I and J integral reaction rates complement each other. The J-integral response increases with increasing neutron energy above the threshold value and therefore possesses an energy dependence qualitatively similar to most fast neutron dosimetry cross sections. However, significant quantitative differences exist. As discussed above, the J-integral response is more accurately defined in terms of both the energy-dependent cross section and threshold energy definition. The I-integral possesses a maximum value at the threshold energy and decreases rapidly from this maximum value as neutron energy increases above the threshold value. As can be seen in Fig. 4, the I-integral possesses a much more narrowly defined energy response than the J-integral. While the J-integral response is broadly distributed, most of the I-integral response is concentrated in the neutron energy just above threshold. As a consequence, the I-integral reaction rate data generally provides a more rigorous test of the ability of neutron transport calculations to describe the complex spatial and energy variations that exist in benchmark fields than does the J-integral data. This conclusion is supported by the calculation to experiment ratios (C/E) obtained from NRE experiments in the VENUS-1 LWR-PV benchmark field. For these VENUS-1 NRE experiments, the C/E values for the I integral possessed larger variation and deviated more widely from unity than the corresponding C/E values for the J-integral (20). FIG. 4 Energy Dependent Response for the Integral Reactions I(ET) and J(Emin) 1.1 Nuclear Research Emulsions (NRE) have a long and illustrious history of applications in the physical sciences, earth sciences and biological sciences (1, 2)2. In the physical sciences, NRE experiments have led to many fundamental discoveries in such diverse disciplines as nuclear physics, cosmic ray physics and high energy physics. In the applied physical sciences, NRE have been used in neutron physics experiments in both fission and fusion reactor environments (3-6). Numerous NRE neutron experiments can be found in other applied disciplines, such as nuclear engineering, environmental monitoring and health physics. Given the breadth of NRE applications, there exist many textbooks and handbooks that provide considerable detail on the techniques used in the NRE method (1-4, 6). As a consequence, this practice will be restricted to the application of the NRE method for neutron measurements in reactor physics and nuclear engineering with particular emphasis on neutron dosimetry in benchmark fields (see Matrix E706). 1.2 NRE are passive detectors and provide time integrated reaction rates. As a consequence, NRE provide fluence measurements without the need for time-dependent corrections, such as arise with radiometric (RM) dosimeters (see Test Method E1005). NRE provide permanent records, so that optical microscopy observations can be carried out any time after exposure. If necessary, NRE measurements can be repeated at any time to examine questionable data or to obtain refined results. 1.3 Since NRE measurements are conducted with optical microscopes, high spatial resolution is afforded for fine structure experiments. The attribute of high spatial resolution can also be used to determine information on the angular anisotropy of the in-situ neutron field (4, 5, 7). It is not possible for active detectors to provide such data because of in-situ perturbations and finite-size effects (see Section 11). 1.4 The existence of hydrogen as a major constituent of NRE affords neutron detection through neutron scattering on hydrogen, that is, the well known (n,p) reaction. NRE measurements in low power reactor environments have been predominantly based on this (n,p) reaction. NRE have also been used to measure the 6Li (n,t) 4He and the 10B (n,α) 7Li reactions by including 6Li and 10B in glass specks near the mid-plane of the NRE (8, 9). Use of these two reactions does not provide the general advantages of the (n,p) reaction for neutron dosimetry in low power reactor environments (see Section 4). As a consequence, this standard will be restricted to the use of the (n,p) reaction for neutron dosimetry in low power reactor environments. 1.5 Limitations—The NRE method possesses four major limitations for applicability in low power reactor environments. 1.5.1 Gamma-Ray Sensitivity—Gamma-rays create a significant limitation for NRE measurements. Above a gamma-ray exposure of approximately 0.025 Gy, NRE can become fogged by gamma-ray induced electron events. At this level of gamma-ray exposure, neutron induced proton-recoil tracks can no longer be accurately measured. As a consequence, NRE experiments are limited to low power environments such as found in critical assemblies and benchmark fields. Moreover, applications are only possible in environments where the buildup of radioactivity, for example, fission products, is limited. 1.5.2 Low Energy Limit—In the measurement of track length for proton recoil events, track length decreases as proton-recoil energy decreases. Proton-recoil track length below approximately 3μm in NRE cannot be adequately measured with optical microscopy techniques. As proton-recoil track length decreases below approximately 3 μm, it becomes very difficult to measure track length accurately. This 3-μm track length limit corresponds to a low energy limit of applicability in the range of approximately 0.3 to 0.4 MeV for neutron induced proton-recoil measurements in NRE. 1.5.3 High-Energy Limits—As a consequence of finite-size limitations, fast-neutron spectrometry measurements are limited to ≤15 MeV. The limit for in-situ spectrometry in reactor environments is ≤8MeV. 1.5.4 Track Density Limit—The ability to measure proton recoil track length with optical microscopy techniques depends on track density. Above a certain track density, a maze or labyrinth of overlapping tracks is created, which precludes the use of optical microscopy techniques. For manual scanning, this limitation arises above approximately 104 tracks/cm2, whereas interactive computer-based scanning systems can extend this limit up to approximately 105 tracks/cm2. These limits correspond to neutron fluences of 106 − 10 7 cm−2, respectively. 1.6 Neutron Spectrometry (Differential Measurements)—For differential neutron spectrometry measurements in low-power reactor environments, NRE experiments can be conducted in two different modes. In the more general mode, NRE are irradiated in-situ in the low power reactor environment. This mode of NRE experiments is called the 4π mode, since the in-situ irradiation creates tracks in all directions (see 3.1.1). In special circumstances, where the direction of the neutron flux is known, NRE are oriented parallel to the direction of the neutron flux. In this orientation, one edge of the NRE faces the incident neutron flux, so that this measurement mode is called the end-on mode. Scanning of proton-recoil tracks is different for these two different modes. Subsequent data analysis is also different for these two modes (see 3.1.1 and 3.1.2). 1.7 Neutron Dosimetry (Integral Measurements)—NRE also afford integral neutron dosimetry through use of the (n,p) reaction in low power reactor environments. Two different types of (n,p) integral mode dosimetry reactions are possible, namely the I-integral (see 3.2.1) and the J-integral (see 3.2.2) (10, 11). Proton-recoil track scanning for these integral reactions is conducted in a different mode than scanning for differential neutron spectrometry (see 3.2). Integral mode data analysis is also different than the analysis required for differential neutron spectrometry (see 3.2). This practice will emphasize NRE (n,p) integral neutron dosimetry, because of the utility and advantages of integral mode measurements in low power benchmark fields. 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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