C26.13 - Spent Fuel and High Level Waste

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.
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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 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
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 test method can be used to provide a measure of the reactivity of a material in a dilute solution in which the test response is dominated by the dissolution or leaching of the test specimen. It can be used to compare the dissolution or leaching behaviors of candidate radioactive waste forms and to study the reactions during static exposure to dilute solutions in which solution feed-back effects can be maintained negligible, depending on the test conditions.  
5.2 The test is suitable for application to natural minerals, simulated waste form materials, and radioactive waste form material specimens.  
5.3 Data from this test may form part of the larger body of data that is necessary in the logical approach to long-term prediction of waste form behavior, as described in Practice C1174. In particular, measured solution concentrations and characterizations of altered surfaces may be used in the validation of geochemical modeling codes.  
5.4 This test method excludes the use of crushed or powdered specimens and organic materials.  
5.5 Several reference test parameter values and reference leachant solutions are specified to facilitate the comparison of results of tests conducted with different materials and at different laboratories. However, other test parameter values and leachant solution compositions can be used to characterize the specimen reactivity.  
5.5.1 Tests can be conducted with different leachant compositions to simulate groundwaters, buffer the leachate pH as the specimen dissolves, or measure the common ion effect of particular solutes.  
5.5.2 Tests can be conducted to measure the effects of various test parameter values on the specimen response, including time, temperature, and S/V ratio. Tests conducted for different durations and at various temperatures provide insight into the reaction kinetics. Tests conducted at different S/V ratio provide insight into chemical affinity (solution feed-back effects) and the approach to saturation.  
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1.1 This test method provides a measure of the chemical durability of a simulated or radioactive monolithic waste form, such as a glass, ceramic, cement (grout), or cermet, in a test solution at temperatures  
1.2 This test method can be used to characterize the dissolution or leaching behaviors of various simulated or radioactive waste forms in various leachants under the specific conditions of the test based on analysis of the test solution. Data from this test are used to calculate normalized elemental mass loss values from specimens exposed to aqueous solutions at temperatures  
1.3 The test is conducted under static conditions in a constant solution volume and at a constant temperature. The reactivity of the test specimen is determined from the amounts of components released and accumulated in the solution over the test duration. A wide range of test conditions can be used to study material behavior, including various leachant composition, specimen surface area-to-leachant volume ratios, temperatures, and test durations.  
1.4 Three leachant compositions and four reference test matrices of test conditions are recommended to characterize materials behavior and facilitate interlaboratory comparisons of tests results.  
1.5 Specimen surfaces may become altered during this test. Although not part of the test method, it is recommended that these altered surface regions be examined to characterize chemical and physical changes due to the reaction of waste forms during static exposure to solutions.  
1.6 This test method is not recommended for evaluating metallic materials, the degradation of which includes oxidation reactions that are not controlled by this test method.  
1.7 This test method must be performed in accordance with all applicable quality assurance requirements for acceptance of the data.  
1.8 The values stated in SI units are to be regarded as standard. Other units of measurement are included for r...

SIGNIFICANCE AND USE
4.1 Drying of the SNF and fuel cavity of the SNF container and its internals is needed to prepare for sealed dry storage, transportation, or permanent disposal at a repository. This guide provides technical information for use in determining the forms of water that need to be considered when choosing a drying process. This guide provides information to aid in (a) selecting a drying system, (b) selecting a drying method, and (c) demonstrating that adequate dryness was achieved (see Annex A2).  
4.2 The considerations affecting drying processes include:  
4.2.1 Water remaining on and in commercial, research, and production reactor spent nuclear fuels after removal from wet storage may become an issue when the fuel is sealed in a dry storage system or transport cask. The movement to a dry storage environment typically results in an increase in fuel temperature, which may be sufficient to cause the release of water from the fuel. The water release coupled with the temperature increase in a sealed container may result in container pressurization, corrosion of fuel or assembly structures, or both, that could affect retrieval of the fuel, and container corrosion.  
4.2.2 Removal of the water associated with the SNF may be accomplished by a variety of technologies including heating, imposing a vacuum over the system, flushing the system with dry gases, and combinations of these and other similar processes.  
4.2.3 Water removal processes are time, temperature, and pressure-dependent. Residual water in some form(s) should be anticipated.  
4.2.4 Drying processes may not readily remove the water that was retained in porous materials, capillaries, sludge, CRUD, physical features that retain water and as thin wetted surface films. Water trapped within breached SNF may be especially difficult to remove.  
4.2.5 Drying processes may be even less successful in removing bound water from the SNF and associated materials because removal of bound water will only occur when the t...
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1.1 This guide discusses three steps in preparing spent nuclear fuel (SNF) for placement in a sealed dry storage system: (1) evaluating the needs for drying the SNF after removal from a water storage pool and prior to placement in dry storage, (2) drying the SNF, and (3) demonstrating that adequate dryness has been achieved.  
1.1.1 The scope of SNF includes nuclear fuel of any design (fuel core, clad materials, and geometric configuration) discharged from power reactors and research reactors and its condition as impacted by reactor operation, handling, and water storage.  
1.1.2 The guide addresses drying methods and their limitations when applied to the drying of SNF that has been stored in water pools. The guide discusses sources and forms of water that may remain in the SNF, the container, or both after the drying process has been completed. It also discusses the important and potential effects of the drying process and any residual water on fuel integrity and container materials during the dry storage period. The effects of residual water are discussed mechanistically as a function of the container thermal and radiological environment to provide guidance on situations that may require extraordinary drying methods, specialized handling, or other treatments.  
1.1.3 The basic issues in drying are: (1) to determine how dry the SNF must be in order to prevent problems with fuel retrievability, container pressurization, or container corrosion during storage, handling, and transfer, and (2) to demonstrate that adequate dryness has been achieved. Achieving adequate dryness may be straightforward for intact commercial fuel but complex for any SNF where the cladding is breached prior to or during placement and storage at the spent fuel pools. Challenges in achieving adequate dryness may also result from the presence of sludge, CRUD, and any other hydrated compounds. These may be transferred with the SNF to the storage con...

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.

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1.1 Intent:  
1.1.1 The intent of this guide is to provide general considerations for the development, verification, validation, and documentation of tank simulants for hazardous materials (for example, radioactive wastes) and process streams. Due to the expense and hazards associated with obtaining and working with actual hazardous materials, especially radioactive wastes, simulants are used in a wide variety of applications including process and equipment development and testing, equipment acceptance testing, and plant commissioning. This standard guide facilitates a consistent methodology for development, preparation, verification, validation, and documentation of simulants.  
1.2 This guide provides direction on (1) defining simulant use, (2) defining simulant-design requirements, (3) developing a simulant preparation procedure, (4) verifying and validating that the simulant meets design requirements, and (5) documenting simulant-development activities and simulant preparation procedures.  
1.3 Applicability:  
1.3.1 This guide is intended for persons and organizations tasked with developing simulants to either mimic certain characteristics and properties of hazardous materials or provide representative performance for the phenomenon being evaluated. The process for simulant development, verification, validation, and documentation is shown schematically in Fig. 1. Specific approval requirements for the simulant developed under this guide are not provided. This topic is left to the performing organization. Approval requirements are associated with the design of the simulant, makeup procedures, and final simulant produced.
FIG. 1 Simulant Development, Verification, Validation, and Documentation Flowsheet  
1.3.2 While this guide is directed at simulants for radioactive materials (for example, nuclear waste), the guidance is also applicable to other aqueous based solutions and slurries.  
1.3.3 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 guide is not a substitute for sound chemistry and chemical engineering skills, proven practices and experience. It is not intended to be prescriptive but rather to provide considerations for the development and use of simulants.  
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.

SIGNIFICANCE AND USE
5.1 Measurements performed in this guide are limited to radioactive solutions, slurries, and sludges as well as simulants designed to model the properties of these radioactive materials.  
5.2 Data obtained from the measurement and calculation of physical and rheological properties of radioactive solutions, slurries, and sludges are essential in developing appropriate simulants for design and testing of retrieval, transport, mixing, and storage systems for treatment of radioactive materials. Details on methods to develop representative simulants are provided in the Guide C1750. These data also provide input parameters for modeling the flow behavior, processing, transport, safety, and storage of these radioactive materials.  
5.3 Consistency in the handling of samples, measurement methods, and calculations is essential in obtaining reproducible results of rheological and physical property measurements.  
5.4 This guide will be used to measure or calculate the physical properties listed below:  
5.4.1 Settled solids density.  
5.4.2 Bulk slurry density.  
5.4.3 Centrifuged solids density.  
5.4.4 Supernatant density.  
5.4.5 Settling rate.  
5.4.6 Volume percent centrifuged solids.  
5.4.7 Volume percent settled solids after settling.  
5.4.8 Undissolved solids content.  
5.4.9 Dissolved solids content.  
5.4.10 Weight percent centrifuged solids.  
5.4.11 Weight percent total oxides.  
5.4.12 Solids content of the centrifuged solids.  
5.4.13 Total solids content.  
5.5 This guide describes the process of performing measurement of the rheological properties. The rheological measurements and calculations described in this guide are limited to shear strength, shear stress versus shear rate, apparent viscosity, consistency, and yield stress.  
5.6 Due to the nature of some solutions, slurries, and sludges, not all of the measurements described in this standard may be applicable to all samples. For example, some sludges do not settle; therefore, settlin...
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1.1 Intent:  
1.1.1 The intent of this guide is to provide guidance for the measurement and calculation of physical and rheological properties of radioactive solutions, slurries, and sludges as well as simulants designed to model the properties of these radioactive materials.  
1.2 Applicability:  
1.2.1 This guide is intended for measurement of mass and volume of the solution, slurries, and sludges as well as dissolved solids content in the liquid fraction and solids content associated with the slurries and sludges. Particle size distribution is also measured.  
1.2.2 This guide identifies the data required and the equations recommended for calculation of density (bulk, settled solids, supernatant, and centrifuged solids), settling rate, volume and weight percent of the centrifuged solids and settled solids, and the weight percent undissolved solids, dissolved solids, and total oxides.  
1.2.3 This guide is intended for measurement of shear strength and shear stress as a function of shear rate.  
1.2.4 Rheological property measurement guidelines in this guide are limited to rotational rheometers.  
1.2.5 This guide is limited to measurements of viscous and incipient flow and does not include oscillatory rheometry.  
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 and health 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...

SIGNIFICANCE AND USE
4.1 The distribution coefficient, Kd, is an experimentally determined ratio quantifying the distribution of a chemical species between a given fluid and solid material sample under certain conditions, including the attainment of constant aqueous concentrations of the species of interest. The Kd concept is used in mass transport modeling, for example, to assess the degree to which the movement of a species will be delayed by interactions with the local geomedium as the solution migrates through the geosphere under a given set of underground geochemical conditions (pH, temperature, ionic strength, etc.). The retardation factor (Rf) is the ratio of the velocity of the groundwater divided by the velocity of the contaminant, which can be expressed as:
   where:  
  ρb  =  bulk density of the porous medium (mass/length3), and    ηe  =  effective porosity of the medium (unitless) expressed as a decimal.    
4.2 Because of the sensitivity of Kd to site specific conditions and materials, the use of literature derived  Kd values is strongly discouraged. For applications other than transport modeling, batch Kd measurements also may be used, for example, for parametric studies of the effects of changing chemical conditions and of mechanisms related to the interactions of fluids with solid material.
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1.1 This test method covers the determination of distribution coefficients, Kd, of chemical species to quantify uptake onto solid materials by a batch sorption technique. It is a laboratory method primarily intended to assess sorption of dissolved ionic species subject to migration through pores and interstices of site specific geomedia, or other solid material. It may also be applied to other materials such as manufactured adsorption media and construction materials. Application of the results to long-term field behavior is not addressed in this method. Kd for radionuclides in selected geomedia or other solid materials are commonly determined for the purpose of assessing potential migratory behavior of contaminants in the subsurface of contaminated sites and out of a waste form and in the surface of waste disposal facilities. This test method is also applicable to studies for parametric studies of the variables and mechanisms which contribute to the measured Kd.  
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.

SIGNIFICANCE AND USE
5.1 These test methods provide data useful for evaluating the chemical durability (see 3.1.5) of glass waste forms as measured by elemental release. Accordingly, it may be applicable throughout manufacturing, research, and development.  
5.1.1 Test Method A can specifically be used to obtain data to evaluate whether the chemical durability of glass waste forms have been consistently controlled during production (see Table 1).  
5.1.2 Test Method B can specifically be used to measure the chemical durability of glass waste forms under various test conditions, for example, varying test durations, test temperatures, sample surface area (SA)-to-leachant volume (V) ratios (see Appendix X1), and leachant types (see Table 1). Data from this test may form part of the larger body of data that are necessary in the logical approach to long-term prediction of waste form behavior (see Practice C1174).
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1.1 These product consistency Test Methods A and B provide a measure of the chemical durability of homogeneous glasses, phase separated glasses, devitrified glasses, glass ceramics, multiphase glass ceramic waste forms, or combinations thereof, hereafter collectively referred to as “glass waste forms” by measuring the concentrations of the chemical species released to a test solution under carefully controlled conditions.  
1.1.1 Test Method A is a seven-day chemical durability test performed at 90 ± 2 °C in a leachant of ASTM-Type I water. The test method is static and conducted in stainless steel vessels. The stainless steel vessels require a gasket to remain leak-tight (see Note 1) The stainless steel vessels are considered to be “closed system” tests. Test Method A can specifically be used to evaluate whether the chemical durability and elemental release characteristics of nuclear, hazardous, and mixed glass waste forms have been consistently controlled during production. This test method is applicable to radioactive and simulated glass waste forms as defined above.  
Note 1: TFE-fluorocarbon gaskets, available commercially, are acceptable and chemically inert up to radiation doses of 1 × 105 R of beta or gamma radiation which have been shown not to damage TFE-fluorocarbon. If higher radiation doses are anticipated, special gaskets fabricated from metals such as copper, gold, lead, or indium are recommended.  
1.1.2 Test Method B is a durability test that allows testing at various test durations, test temperatures, particle size and masses of glass sample, leachant volumes, and leachant compositions. This test method is static and can be conducted in stainless steel or PFA TFE-fluorocarbon vessels. The stainless steel vessels are considered to be “closed system” while the PFA TFE-fluorocarbon vessels are considered to be “open system” tests. Test Method B can specifically be used to evaluate the relative chemical durability characteristics of homogeneous glasses, phase separated glasses, devitrified glasses, glass ceramics, or multiphase glass ceramic waste forms, or combinations thereof. This test method is applicable to radioactive (nuclear) and mixed, hazardous, and simulated glass waste forms as defined above. Test Method B cannot be used as a consistency test for production of high level radioactive glass waste forms.  
1.2 These test methods must be performed in accordance with all quality assurance requirements for acceptance of the data.  
1.3 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.  
1.4 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 ...

SIGNIFICANCE AND USE
5.1 This test method can be used to measure the release of a component from a solidified waste form into water at the reference temperature near 20 °C and at elevated temperatures that accelerate the rate and extent of leaching relative to the values measured at the reference temperature. Results of this test method can be used to quantify an intrinsic property of a material, but should not be presumed to represent releases in specific disposal environments. Tests can be conducted under conditions that represent a specific disposal environment (for example, by using a representative groundwater) to determine effective parameter values for those conditions.  
5.2 This test method can be used to:  
5.2.1 Compare releases of waste components from various types of solidification agents and formulations.  
5.2.2 Determine the model parameter values quantifying the release of contaminants from a waste or waste form at a specific temperature.  
5.2.3 Promote greater extents of reaction than can be achieved under expected service conditions within a laboratory time frame to provide greater confidence in modeled contaminant releases.  
5.2.4 Determine the temperature dependence of contaminant release.  
5.3 Fitting the experimental results with a mechanistic model allows release behaviors to be extrapolated to long times and to full-scale waste forms under the following constraints:  
5.3.1 The same model must be used to represent the results of tests conducted at elevated temperatures and at the reference temperature because the mechanism must be the same.  
5.3.2 Projections of releases over long times require that the waste form matrix remain stable, which may be demonstrated by the physical robustness of specimens recovered from tests conducted at elevated temperatures.  
5.3.3 Extrapolations in time at any temperature within the range tested are limited to values that correspond to the maximum CFL value that was measured.
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1.1 This test method provides a procedure for measuring the release rates of elements from a solidified matrix material under conditions that mitigate solution feedback effects. Results can be analyzed by using different models to determine if the elemental releases are controlled by mass transport though the matrix (that is, by diffusion), by a surface dissolution process, or by a combination of processes. This test method is applicable to any material that does not deform during the test.  
1.1.1 If mass diffusion is the dominant process in the release mechanism, then the results of this test can be used to derive diffusion coefficients for use in diffusion-based mathematical models.  
1.1.2 If surface dissolution is the dominant process, then the results of this test can be used to derive the kinetic dissolution rate in the absence of reaction affinity effects for use in dissolution-based mathematical models.  
1.1.3 If release is controlled by coupled or combined dissolution and mass transport processes, then the results of this test can be used to derive effective coefficient values for a mechanistic or empirical model.  
1.2 Tests at elevated temperatures are used to accelerate the release process to determine the temperature range over which the release mechanism does not change and to generate results that can be used for calculating releases at lower temperatures over long times, provided that the release mechanism does not change with temperature.  
1.2.1 Tests conducted at high temperatures can be used to determine the temperature dependence of model coefficients.  
1.2.2 The mechanism is considered to remain unchanged over a range of temperatures if the model coefficients show Arrhenius behavior over that range.  
1.2.3 Releases at any temperature within that range can be projected in time up to the highest cumulative fractional release value that has been measured for that material.  
1.3 The values stated in SI units are to be regarded as s...

SIGNIFICANCE AND USE
4.1 Obtaining samples of high-level waste created during the reprocessing of spent nuclear fuels presents unique challenges. Generally, high-level waste is stored in tanks with limited access to decrease the potential for radiation exposure to personnel. Samples must be obtained remotely because of the high radiation dose from the bulk material and the samples, samples require shielding for handling, transport, and storage. The quantity of sample that can be obtained and transported is small due to the hazardous nature of the samples as well as their high radiation dose.  
4.2 Many high-level wastes have been treated to remove strontium (Sr) or cesium (Cs), or both, have undergone liquid volume reductions through pumping and forced evaporation or have been pH modified, or both, to decrease corrosion of the tanks. These processes, as well as waste streams added from multiple process plant operations, often resulted in precipitation, and produced multiphase wastes that are heterogeneous. Evaporation of water from waste with significant dissolved salts concentrations has occurred in some tanks due to the high heat load associated with the high-level waste and by pumping and intentional evaporative processing, resulting in the formation of a saltcake or crusts, or both. Organic layers exist in some waste tanks, creating additional heterogeneity in the wastes.  
4.3 Many of the sampling systems have limitations including the ability to sample varying depths in the tank and the depth of sampling. Sampling in Hanford tanks is constrained by riser diameter, riser location and riser availability.  
4.4 Due to these extraordinary challenges, substantial effort in research and development has been expended to develop techniques to provide grab samples of the contents of the high-level waste tanks. A summary of the primary techniques used to obtain samples from high-level waste tanks is provided in Table 1. These techniques will be summarized in this guideline with the assum...
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1.1 This guide addresses techniques used to obtain samples from tanks containing high-level radioactive waste created during the reprocessing of spent nuclear fuels. Guidance on selecting appropriate sampling devices for waste covered by the Resource Conservation and Recovery Act (RCRA) is also provided by the United States Environmental Protection Agency (EPA) (1).2 Vapor sampling of the head-space is not included in this guide because it does not significantly affect slurry retrieval, pipeline transport, plugging, or mixing.  
1.2 The values stated in inch-pound units are to be regarded as standard. No other units of measurement are included in this standard.  
1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.  
1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

SIGNIFICANCE AND USE
5.1 This guide supports the development of material behavior models that can be used to estimate performance of the EBS materials during the post-closure period of a high-level nuclear waste repository for times much longer than can be tested directly. This guide is intended for modeling the degradation behaviors of materials proposed for use in an EBS designed to contain radionuclides over tens of thousands of years and more. There is both national and international recognition of the importance of the use and long-term performance of engineered materials in geologic repository design. Use of the models developed following the approaches described in this guide is intended to address established regulations, such as:  
5.1.1 U.S. Public Law 97–425, the Nuclear Waste Policy Act of 1982, provides for the deep geologic disposal of high-level radioactive waste through a system of multiple barriers. These barriers include engineered barriers designed to prevent the migration of radionuclides out of the engineered system, and the geologic host medium that provides an additional transport barrier between the engineered system and biosphere. The regulations of the U.S. Nuclear Regulatory Commission for geologic disposal require a performance confirmation program to provide data through tests and analyses, where practicable, that demonstrate engineered systems and components that are designed or assumed to act as barriers after permanent closure are functioning as intended and anticipated.  
5.1.2 IAEA Safety Requirements specify that engineered barriers shall be designed and the host environment shall be selected to provide containment of the radionuclides associated with the wastes.  
5.1.3 The Swedish Regulatory Authority has provided general advice to the repository developer that the application of best available technique be followed in connection with disposal, which means that the siting, design, construction, and operation of the repository and appurtenant syste...
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1.1 This guide addresses how various test methods and data analyses can be used to develop models for the evaluation of the long-term alteration behavior of materials used in an engineered barrier system (EBS) for the disposal of spent nuclear fuel (SNF) and other high-level nuclear waste in a geologic repository. The alteration behavior of waste forms and EBS materials is important because it affects the retention of radionuclides within the disposal system either directly, as in the case of waste forms in which the radionuclides are initially immobilized, or indirectly, as in the case of EBS containment materials that restrict the ingress of groundwater or the egress of radionuclides that are released as the waste forms degrade.  
1.2 The purpose of this guide is to provide a scientifically-based strategy for developing models that can be used to estimate material alteration behavior after a repository is permanently closed (that is, in the post-closure period). Because the timescale involved with geological disposal precludes direct validation of predictions, mechanistic understanding of the processes based on detailed data and models and consideration of the range of uncertainty are recommended.  
1.3 This guide addresses the scientific bases and uncertainties in material behavior models and the impact on the confidence in the EBS design criteria and repository performance assessments using those models. This includes the identification and use of conservative assumptions to address uncertainty in the long-term performance of materials.  
1.3.1 Steps involved in evaluating the performance of waste forms and EBS materials include problem definition, laboratory and field testing, modeling of individual and coupled processes, and model confirmation.  
1.3.2 The estimates of waste form and EBS material performance are based on models derived from theoretical considerations, expert judgments, and interpretations of d...

SIGNIFICANCE AND USE
5.1 The vapor hydration test can be used to study the corrosion of glass and glass ceramic waste forms under conditions of high temperature and contact by water vapor or thin films of water. This method may serve as an accelerated test for some materials, since the high temperatures will accelerate thermally activated processes. A wide range of test temperatures have been reported in the literature –40°C (Ebert et al, 2005 (3), for example) to 300°C (Vienna et al, 2001 (4), for example). It should be noted that with increased test temperature comes the possibility of changing the corrosion rate determining mechanism and the types of phases formed upon alteration from those that occur in the disposal environment (Vienna et al, 2001 (4)).  
5.2 The vapor hydration test can be used as a screening test to determine the propensity of waste forms to alter and for relative comparisons in alteration rates between waste forms.
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1.1 The vapor hydration test method can be used to study the corrosion of a waste forms such as glasses and glass ceramics2 upon exposure to water vapor at elevated temperatures. In addition, the alteration phases that form can be used as indicators of those phases that may form under repository conditions. These tests; which allow altering of glass at high surface area to solution volume ratio; provide useful information regarding the alteration phases that are formed, the disposition of radioactive and hazardous components, and the alteration kinetics under the specific test conditions. This information may be used in performance assessment (McGrail et al, 2002 (1)3 for example).  
1.2 This test method must be performed in accordance with all quality assurance requirements for acceptance of the data.  
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 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, and  
4.1.8 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 rep...
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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 in...

SIGNIFICANCE AND USE
5.1 Information is provided in this document and other referenced documents to assist the licensee and the licensor in analyzing the materials aspects of performance of SNF and DCSS components during extended storage. The effects of the service conditions of the first licensing period are reviewed in the license renewal process. These service conditions are highlighted and discussed in Annex A1 as factors that affect materials performance in an ISFSI. Emphasis is on the effects of time, temperature, radiation, and the environment on the condition of the SNF and the performance of components of ISFSI storage systems.  
5.2 The storage of SNF that is irradiated under the regulations of 10 CFR Part 50 is governed by regulations in 10 CFR Part 72. Regulatory requirements for the subsequent geologic disposal of this SNF are presently given in 10 CFR Part 60, with specific requirements for the use of Yucca Mountain as a repository being given in the regulatory requirements of 10 CFR Part 63. Between the life-cycle phases of storage and disposal, SNF may be transported under the requirements of 10 CFR Part 71. Therefore, in storage, it is important to acknowledge the transport and disposal phases of the life cycle. In doing this, the materials properties that are important to these subsequent phases are to be considered in order to promote successful completion of these subsequent phases in the life cycle of SNF. Retrievability of SNF (or high-level radioactive waste) is set as a requirement in 10 CFR Part 72.122(g)(5) and 10 CFR Part 72.122(l). Care should be taken in operations conducted prior to disposal, for example, storage, transfer, and transport, to ensure that the SNF is not abused and that SNF assemblies will be retrievable, the protective value of the cladding is not degraded and remains capable of serving as an active barrier to radionuclide release during transfer and transport operations. It is possible that cladding could be altered during dry storage. The ...
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1.1 Part of the total inventory of commercial spent nuclear fuel (SNF) is stored in dry cask storage systems (DCSS) under licenses granted by the U.S. Nuclear Regulatory Commission (NRC). The purpose of this guide is to provide information to assist in supporting the renewal of these licenses, safely and without removal of the SNF from its licensed confinement, for periods beyond those governed by the term of the original license. This guide provides information on materials behavior under conditions that may be important to safety evaluations for the extended service of the renewal period. This guide is written for DCSS containing light water reactor (LWR) fuel that is clad in zirconium alloy material and stored in accordance with the Code of Federal Regulations (CFR), at an independent spent-fuel storage installation (ISFSI).2 The components of an ISFSI, addressed in this document, include the commercial SNF, canister, cask, and all parts of the storage installation including the ISFSI pad. The language of this guide is based, in part, on the requirements for a dry SNF storage license that is granted, by the U.S. Nuclear Regulatory Commission (NRC), for up to 20 years. Although government regulations may differ for various nations, the guidance on materials properties and behavior given here is expected to have broad applicability.  
1.2 This guide addresses many of the factors affecting the time-dependent behavior of materials under ISFSI service [10 CFR Part 72.42]. These factors are those regarded to be important to performance, in license extension, beyond the currently licensed 20-year period. Examples of these factors are given in this guide and they include materials alterations or environmental conditions for components of an ISFSI system that, over time, could have significance related to safety. For purposes of this guide, a license period of an additional 20 to 80 years is assumed.  
1.3 This guide address...

SIGNIFICANCE AND USE
5.1 This practice provides a prescriptive description of the design of a SPFT test apparatus and identifies aspects of the performance of SPFT tests and interpretation of test results that must be addressed by the experimenter to provide confidence in the measured dissolution rate.  
5.2 The SPFT test method described in this practice can be used to characterize various aspects of glass corrosion behavior that can be utilized in a mechanistic model for calculating long-term behavior of a nuclear waste glass.  
5.3 Depending on the values of test parameters that are used, the results of SPFT tests can be used to measure the intrinsic dissolution rate of a glass, the temperature and pH dependencies of the rate, and the effects of various dissolved species on the dissolution rate.  
5.4 The reacted sample recovered from a test may be examined with surface analytical techniques, such as scanning electron microscopy, to further characterize the corrosion behavior. Such examinations may provide evidence regarding whether the glass is dissolving stoichiometrically, if particular leached layers and secondary phases were formed on the specimen surface, and so forth. These occurrences may impact the accuracy of the glass dissolution rate that is measured using this method. This practice does not address the analysis of solid reaction materials.
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1.1 This practice describes a single-pass flow-through (SPFT) test method that can be used to measure the dissolution rate of a homogeneous silicate glass, including nuclear waste glasses, in various test solutions at temperatures less than 100°C. Tests may be conducted under conditions in which the effects from dissolved species on the dissolution rate are minimized to measure the forward dissolution rate at specific values of temperature and pH, or to measure the dependence of the dissolution rate on the concentrations of various solute species.  
1.2 Tests are conducted by pumping solutions in either a continuous or pulsed flow mode through a reaction cell that contains the test specimen. Tests must be conducted at several solution flow rates to evaluate the effect of the flow rate on the glass dissolution rate.  
1.3 This practice excludes static test methods in which flow is simulated by manually removing solution from the reaction cell and replacing it with fresh solution.  
1.4 Tests may be conducted with demineralized water, chemical solutions (such as pH buffer solutions, simulated groundwater solutions, and brines), or actual groundwater.  
1.5 Tests may be conducted with crushed glass of a known size fraction or monolithic specimens having known geometric surface area. The reacted solids may be examined to provide additional information regarding the behavior of the material in the test and the reaction mechanism.  
1.6 Tests may be conducted with glasses containing radionuclides. However, this test method does not address safety issues for radioactive samples.  
1.7 Data from these tests can be used to determine the values of kinetic model parameters needed to calculate the glass corrosion behavior in a disposal system over long periods (for example, see Practice C1174).  
1.8 This practice must be performed in accordance with all quality assurance requirements for acceptance of the data.  
1.9 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.  
1.10 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.  
1.11 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, Guid...

SIGNIFICANCE AND USE
4.1 The distribution coefficient, Kd, is an experimentally determined ratio quantifying the distribution of a chemical species between a given fluid and solid material sample under certain conditions, including the attainment of constant aqueous concentrations of the species of interest. The Kd concept is used in mass transport modeling, for example, to assess the degree to which the movement of a species will be delayed by interactions with the local geomedium as the solution migrates through the geosphere under a given set of underground geochemical conditions (pH, temperature, ionic strength, etc.). The retardation factor (Rf) is the ratio of the velocity of the groundwater divided by the velocity of the contaminant, which can be expressed as:
   where:  
  ρb  =  bulk density of the porous medium (mass/length3), and    ηe  =  effective porosity of the medium (unitless) expressed as a decimal.    
4.2 Because of the sensitivity of Kd to site specific conditions and materials, the use of literature derived  Kd values is strongly discouraged. For applications other than transport modeling, batch Kd measurements also may be used, for example, for parametric studies of the effects of changing chemical conditions and of mechanisms related to the interactions of fluids with solid material.
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1.1 This test method covers the determination of distribution coefficients, Kd, of chemical species to quantify uptake onto solid materials by a batch sorption technique. It is a laboratory method primarily intended to assess sorption of dissolved ionic species subject to migration through pores and interstices of site specific geomedia, or other solid material. It may also be applied to other materials such as manufactured adsorption media and construction materials. Application of the results to long-term field behavior is not addressed in this method. Kd for radionuclides in selected geomedia or other solid materials are commonly determined for the purpose of assessing potential migratory behavior of contaminants in the subsurface of contaminated sites and out of a waste form and in the surface of waste disposal facilities. This test method is also applicable to studies for parametric studies of the variables and mechanisms which contribute to the measured Kd.  
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.

SIGNIFICANCE AND USE
4.1 Obtaining samples of high-level waste created during the reprocessing of spent nuclear fuels presents unique challenges. Generally, high-level waste is stored in tanks with limited access to decrease the potential for radiation exposure to personnel. Samples must be obtained remotely because of the high radiation dose from the bulk material and the samples; samples require shielding for handling, transport, and storage. The quantity of sample that can be obtained and transported is small due to the hazardous nature of the samples as well as their high radiation dose.  
4.2 Many high-level wastes have been treated to remove strontium (Sr) or cesium (Cs), or both, have undergone liquid volume reductions through pumping and forced evaporation or have been pH modified, or both, to decrease corrosion of the tanks. These processes, as well as waste streams added from multiple process plant operations, often resulted in precipitation, and produced multiphase wastes that are heterogeneous. Evaporation of water from waste with significant dissolved salts concentrations has occurred in some tanks due to the high heat load associated with the high-level waste and by pumping and intentional evaporative processing, resulting in the formation of a saltcake or crusts, or both. Organic layers exist in some waste tanks, creating additional heterogeneity in the wastes.  
4.3 Many of the sampling systems have limitations including the ability to sample varying depths in the tank and the depth of sampling. Sampling in Hanford tanks is constrained by riser diameter, riser location and riser availability.  
4.4 Due to these extraordinary challenges, substantial effort in research and development has been expended to develop techniques to provide grab samples of the contents of the high-level waste tanks. A summary of the primary techniques used to obtain samples from high-level waste tanks is provided in Table 1. These techniques will be summarized in this guideline with the assum...
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1.1 This guide addresses techniques used to obtain samples from tanks containing high-level radioactive waste created during the reprocessing of spent nuclear fuels. Guidance on selecting appropriate sampling devices for waste covered by the Resource Conservation and Recovery Act (RCRA) is also provided by the United States Environmental Protection Agency (EPA) (1).2 Vapor sampling of the head-space is not included in this guide because it does not significantly affect slurry retrieval, pipeline transport, plugging, or mixing.  
1.2 The values stated in inch-pound units are to be regarded as standard. No other units of measurement are included in this standard.  
1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.  
1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

SIGNIFICANCE AND USE
5.1 This test method can be used to provide a measure of the reactivity of a material in a dilute solution in which the test response is dominated by the dissolution or leaching of the test specimen. It can be used to compare the dissolution or leaching behaviors of candidate radioactive waste forms and to study the reactions during static exposure to dilute solutions in which solution feed-back effects can be maintained negligible, depending on the test conditions.  
5.2 The test is suitable for application to natural minerals, simulated waste form materials, and radioactive waste form material specimens.  
5.3 Data from this test may form part of the larger body of data that is necessary in the logical approach to long-term prediction of waste form behavior, as described in Practice C1174. In particular, measured solution concentrations and characterizations of altered surfaces may be used in the validation of geochemical modeling codes.  
5.4 This test method excludes the use of crushed or powdered specimens and organic materials.  
5.5 Several reference test parameter values and reference leachant solutions are specified to facilitate the comparison of results of tests conducted with different materials and at different laboratories. However, other test parameter values and leachant solution compositions can be used to characterize the specimen reactivity.  
5.5.1 Tests can be conducted with different leachant compositions to simulate groundwaters, buffer the leachate pH as the specimen dissolves, or measure the common ion effect of particular solutes.  
5.5.2 Tests can be conducted to measure the effects of various test parameter values on the specimen response, including time, temperature, and S/V ratio. Tests conducted for different durations and at various temperatures provide insight into the reaction kinetics. Tests conducted at different S/V ratio provide insight into chemical affinity (solution feed-back effects) and the approach to saturation.  
...
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1.1 This test method provides a measure of the chemical durability of a simulated or radioactive monolithic waste form, such as a glass, ceramic, cement (grout), or cermet, in a test solution at temperatures  
1.2 This test method can be used to characterize the dissolution or leaching behaviors of various simulated or radioactive waste forms in various leachants under the specific conditions of the test based on analysis of the test solution. Data from this test are used to calculate normalized elemental mass loss values from specimens exposed to aqueous solutions at temperatures  
1.3 The test is conducted under static conditions in a constant solution volume and at a constant temperature. The reactivity of the test specimen is determined from the amounts of components released and accumulated in the solution over the test duration. A wide range of test conditions can be used to study material behavior, including various leachant composition, specimen surface area-to-leachant volume ratios, temperatures, and test durations.  
1.4 Three leachant compositions and four reference test matrices of test conditions are recommended to characterize materials behavior and facilitate interlaboratory comparisons of tests results.  
1.5 Specimen surfaces may become altered during this test. Although not part of the test method, it is recommended that these altered surface regions be examined to characterize chemical and physical changes due to the reaction of waste forms during static exposure to solutions.  
1.6 This test method is not recommended for evaluating metallic materials, the degradation of which includes oxidation reactions that are not controlled by this test method.  
1.7 This test method must be performed in accordance with all applicable quality assurance requirements for acceptance of the data.  
1.8 The values stated in SI units are to be regarded as standard. No other units of measurement are included in...

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1.1 Intent:  
1.1.1 The intent of this guideline is to provide general considerations for the development, verification, validation, and documentation of high-level waste (HLW) tank simulants. Due to the expense and hazards associated with obtaining and working with actual wastes, especially radioactive wastes, simulants are used in a wide variety of applications including process and equipment development and testing, equipment acceptance testing, and plant commissioning. This standard guide facilitates a consistent methodology for development, preparation, verification, validation, and documentation of waste simulants.  
1.2 This guideline provides direction on (1) defining simulant use, (2) defining simulant-design requirements, (3) developing a simulant preparation procedure, (4) verifying and validating that the simulant meets design requirements, and (5) documenting simulant-development activities and simulant preparation procedures.  
1.3 Applicability:  
1.3.1 This guide is intended for persons and organizations tasked with developing HLW simulants to mimic certain characteristics and properties of actual wastes. The process for simulant development, verification, validation, and documentation is shown schematically in Fig. 1. Specific approval requirements for the simulants developed under this guideline are not provided. This topic is left to the performing organization.
FIG. 1 Simulant Development, Verification, Validation, and Documentation Flowsheet  
1.3.2 While this guide is directed at HLW simulants, much of the guidance may also be applicable to other aqueous based solutions and slurries.  
1.3.3 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.  
1.4 User Caveats:  
1.4.1 This guideline is not a substitute for sound chemistry and chemical engineering skills, proven practices and experience. It is not intended to be prescriptive but rather to provide considerations for the development and use of waste simulants.  
1.4.2 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 This practice provides a prescriptive description of the design of a SPFT test apparatus and identifies aspects of the performance of SPFT tests and interpretation of test results that must be addressed by the experimenter to provide confidence in the measured dissolution rate.  
5.2 The SPFT test method described in this practice can be used to characterize various aspects of glass corrosion behavior that can be utilized in a mechanistic model for calculating long-term behavior of a nuclear waste glass.  
5.3 Depending on the values of test parameters that are used, the results of SPFT tests can be used to measure the intrinsic dissolution rate of a glass, the temperature and pH dependencies of the rate, and the effects of various dissolved species on the dissolution rate.  
5.4 The reacted sample recovered from a test may be examined with surface analytical techniques, such as scanning electron microscopy, to further characterize the corrosion behavior. Such examinations may provide evidence regarding whether the glass is dissolving stoichiometrically, if particular leached layers and secondary phases were formed on the specimen surface, and so forth. These occurrences may impact the accuracy of the glass dissolution rate that is measured using this method. This practice does not address the analysis of solid reaction materials.
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1.1 This practice describes a single-pass flow-through (SPFT) test method that can be used to measure the dissolution rate of a homogeneous silicate glass, including nuclear waste glasses, in various test solutions at temperatures less than 100°C. Tests may be conducted under conditions in which the effects from dissolved species on the dissolution rate are minimized to measure the forward dissolution rate at specific values of temperature and pH, or to measure the dependence of the dissolution rate on the concentrations of various solute species.  
1.2 Tests are conducted by pumping solutions in either a continuous or pulsed flow mode through a reaction cell that contains the test specimen. Tests must be conducted at several solution flow rates to evaluate the effect of the flow rate on the glass dissolution rate.  
1.3 This practice excludes static test methods in which flow is simulated by manually removing solution from the reaction cell and replacing it with fresh solution.  
1.4 Tests may be conducted with demineralized water, chemical solutions (such as pH buffer solutions, simulated groundwater solutions, and brines), or actual groundwater.  
1.5 Tests may be conducted with crushed glass of a known size fraction or monolithic specimens having known geometric surface area. The reacted solids may be examined to provide additional information regarding the behavior of the material in the test and the reaction mechanism.  
1.6 Tests may be conducted with glasses containing radionuclides. However, this test method does not address safety issues for radioactive samples.  
1.7 Data from these tests can be used to determine the values of kinetic model parameters needed to calculate the glass corrosion behavior in a disposal system over long periods (for example, see Practice C1174).  
1.8 This practice must be performed in accordance with all quality assurance requirements for acceptance of the data.  
1.9 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.  
1.10 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.  
1.11 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, Guid...

SIGNIFICANCE AND USE
5.1 The vapor hydration test can be used to study the corrosion of glass and glass ceramic waste forms under conditions of high temperature and contact by water vapor or thin films of water. This method may serve as an accelerated test for some materials, since the high temperatures will accelerate thermally activated processes. A wide range of test temperatures have been reported in the literature –40°C (Ebert et al, 2005 (3), for example) to 300°C (Vienna et al, 2001 (4), for example). It should be noted that with increased test temperature comes the possibility of changing the corrosion rate determining mechanism and the types of phases formed upon alteration from those that occur in the disposal environment (Vienna et al, 2001 (4)).  
5.2 The vapor hydration test can be used as a screening test to determine the propensity of waste forms to alter and for relative comparisons in alteration rates between waste forms.
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1.1 The vapor hydration test method can be used to study the corrosion of a waste forms such as glasses and glass ceramics2 upon exposure to water vapor at elevated temperatures. In addition, the alteration phases that form can be used as indicators of those phases that may form under repository conditions. These tests; which allow altering of glass at high surface area to solution volume ratio; provide useful information regarding the alteration phases that are formed, the disposition of radioactive and hazardous components, and the alteration kinetics under the specific test conditions. This information may be used in performance assessment (McGrail et al, 2002 (1)3 for example).  
1.2 This test method must be performed in accordance with all quality assurance requirements for acceptance of the data.  
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 This procedure can be used for (but is 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.
SCOPE
1.1 These practices 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 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).  
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: (A) Gradient Temperature Furnace Method (GT), (B) Uniform Temperature Furnace Method (UT), and (C) 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 and health 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 distribution coefficient, Kd, is an experimentally determined ratio quantifying the distribution of a chemical species between a given fluid and solid material sample under certain conditions, including the attainment of constant aqueous concentrations of the species of interest. The Kd concept is used in mass transport modeling, for example, to assess the degree to which the movement of a species will be delayed by interactions with the local geomedium as the solution migrates through the geosphere under a given set of underground geochemical conditions (pH, temperature, ionic strength, etc.). The retardation factor (Rf) is the ratio of the velocity of the groundwater divided by the velocity of the contaminant, which can be expressed as:
   where:
  ρb  =  bulk density of the porous medium (mass/length 3) and    ηe  =  effective porosity of the medium (unitless) expressed as a decimal.    
4.2 Because of the sensitivity of Kd to site specific conditions and materials, the use of literature derived  Kd values is strongly discouraged. For applications other than transport modeling, batch Kd measurements also may be used, for example, for parametric studies of the effects of changing chemical conditions and of mechanisms related to the interactions of fluids with solid material.
SCOPE
1.1 This test method covers the determination of distribution coefficients, Kd, of chemical species to quantify uptake onto solid materials by a batch sorption technique. It is a laboratory method primarily intended to assess sorption of dissolved ionic species subject to migration through pores and interstices of site specific geomedia, or other solid material. It may also be applied to other materials such as manufactured adsorption media and construction materials. Application of the results to long-term field behavior is not addressed in this method. Kd for radionuclides in selected geomedia or other solid materials are commonly determined for the purpose of assessing potential migratory behavior of contaminants in the subsurface of contaminated sites and out of a waste form and in the surface of waste disposal facilities. This test method is also applicable to studies for parametric studies of the variables and mechanisms which contribute to the measured Kd.  
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.

SIGNIFICANCE AND USE
5.1 This procedure can be used for (but is 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.
SCOPE
1.1 These practices 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 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).  
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: (A) Gradient Temperature Furnace Method (GT), (B) Uniform Temperature Furnace Method (UT), and (C) 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 and health 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 distribution coefficient, Kd, is an experimentally determined ratio quantifying the distribution of a chemical species between a given fluid and solid material sample under certain conditions, including the attainment of constant aqueous concentrations of the species of interest. The Kd concept is used in mass transport modeling, for example, to assess the degree to which the movement of a species will be delayed by interactions with the local geomedium as the solution migrates through the geosphere under a given set of underground geochemical conditions (pH, temperature, ionic strength, etc.). The retardation factor (Rf) is the ratio of the velocity of the groundwater divided by the velocity of the contaminant, which can be expressed as:
   where:
  ρb  =  bulk density of the porous medium (mass/length 3) and    ηe  =  effective porosity of the medium (unitless) expressed as a decimal.    
4.2 Because of the sensitivity of Kd to site specific conditions and materials, the use of literature derived  Kd values is strongly discouraged. For applications other than transport modeling, batch Kd measurements also may be used, for example, for parametric studies of the effects of changing chemical conditions and of mechanisms related to the interactions of fluids with solid material.
SCOPE
1.1 This test method covers the determination of distribution coefficients, Kd, of chemical species to quantify uptake onto solid materials by a batch sorption technique. It is a laboratory method primarily intended to assess sorption of dissolved ionic species subject to migration through pores and interstices of site specific geomedia, or other solid material. It may also be applied to other materials such as manufactured adsorption media and construction materials. Application of the results to long-term field behavior is not addressed in this method. Kd for radionuclides in selected geomedia or other solid materials are commonly determined for the purpose of assessing potential migratory behavior of contaminants in the subsurface of contaminated sites and out of a waste form and in the surface of waste disposal facilities. This test method is also applicable to studies for parametric studies of the variables and mechanisms which contribute to the measured Kd.  
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.

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. 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. Also, pinholes/cracks in fuel rods in dry or wet interim storage can also develop into much larger defects (for example, the phenomenon of cladding “unzipping”) under long-term repository conditions. 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 ...
SCOPE
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 and test 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 fuel storage installation (IS...

SIGNIFICANCE AND USE
5.1 Measurements performed in this guide are limited to radioactive solutions, slurries, and sludges as well as simulants designed to model the properties of these radioactive materials.  
5.2 Data obtained from the measurement and calculation of physical and rheological properties of radioactive solutions, slurries, and sludges are essential in developing appropriate simulants for design and testing of retrieval, transport, mixing, and storage systems for treatment of radioactive materials. These data also provide input parameters for modeling the flow behavior, processing, transport, safety, and storage of these radioactive materials.  
5.3 Consistency in the handling of samples, measurement methods, and calculations is essential in obtaining reproducible results of rheological and physical property measurements.  
5.4 This guide will be used to measure or calculate the physical properties listed below.  
5.4.1 Bulk slurry density  
5.4.2 Settled solids density  
5.4.3 Centrifuged solids density  
5.4.4 Supernatant density  
5.4.5 Settling rate  
5.4.6 Volume percent centrifuged solids  
5.4.7 Volume percent settled solids after settling  
5.4.8 Undissolved solids content  
5.4.9 Dissolved solids content  
5.4.10 Weight percent centrifuged solids  
5.4.11 Weight percent total oxides  
5.4.12 Solids content of the centrifuged solids  
5.4.13 Total solids content  
5.5 This guide describes the process of performing measurement of the rheological properties. The rheological measurements and calculations described in this guide are limited to shear strength, shear stress versus shear rate, apparent viscosity, consistency, and yield stress.  
5.6 Due to the nature of some solutions, slurries, and sludges, not all of the measurements described in this standard may be applicable to all samples. For example, some sludges do not settle; therefore, settling rate measurements are not applicable for these samples.
SCOPE
1.1 Intent:  
1.1.1 The intent of this guide is to provide guidance for the measurement and calculation of physical and rheological properties of radioactive solutions, slurries, and sludges as well as simulants designed to model the properties of these radioactive materials.  
1.2 Applicability:  
1.2.1 This guide is intended for measurement of mass and volume of the solution, slurries, and sludges as well as dissolved solids content in the liquid fraction and solids content associated with the slurries and sludges. Particle size distribution is also measured.  
1.2.2 This guide identifies the data required and the equations recommended for calculation of density (bulk, settled solids, supernatant, and centrifuged solids), settling rate, volume and weight percent of the centrifuged solids and settled solids, and the weight percent undissolved solids, dissolved solids, and total oxides.  
1.2.3 This guide is intended for measurement of shear strength and shear stress as a function of shear rate.  
1.2.4 Rheological property measurement guidelines in this guide are limited to rotational rheometers.  
1.2.5 This guide is limited to measurements of viscous and incipient flow and does not include oscillatory rheometry.  
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 and health 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 This practice supports the development of material behavior models that can be used to estimate performance of the EBS materials during the post-closure period of a high-level nuclear waste repository for times much longer than can be tested directly. This practice is intended for modeling the degradation behaviors of materials proposed for use in an EBS designed to contain radionuclides over tens of thousands of years and more. There is both national and international recognition of the importance of the use and long-term performance of engineered materials in geologic repository design. Use of the models developed following the approaches described in this practice is intended to address established regulations, such as:  
5.1.1 U.S. Public Law 97–425, the Nuclear Waste Policy Act of 1982, provides for the deep geologic disposal of high-level radioactive waste through a system of multiple barriers. These barriers include engineered barriers designed to prevent the migration of radionuclides out of the engineered system, and the geologic host medium that provides an additional transport barrier between the engineered system and biosphere. The regulations of the U.S. Nuclear Regulatory Commission for geologic disposal require a performance confirmation program to provide data through tests and analyses, where practicable, that demonstrate engineered systems and components that are designed or assumed to act as barriers after permanent closure are functioning as intended and anticipated.  
5.1.2 IAEA Safety Requirements specify that engineered barriers shall be designed and the host environment shall be selected to provide containment of the radionuclides associated with the wastes.  
5.1.3 The Swedish Regulatory Authority has provided general advice to the repository developer that the application of best available technique be followed in connection with disposal, which means that the siting, design, construction, and operation of the repository and appurten...
SCOPE
1.1 This practice addresses how various test methods and data analyses can be used to develop models for the evaluation of the long-term alteration behavior of materials used in engineered barrier system (EBS) for the disposal of spent nuclear fuel (SNF) and other high-level nuclear waste in a geologic repository. The alteration behavior of waste forms and EBS materials is important because it affects the retention of radionuclides within the disposal system either directly, as in the case of waste forms in which the radionuclides are initially immobilized, or indirectly, as in the case of EBS containment materials that restrict the ingress of groundwater or the egress of radionuclides that are released as the waste forms degrade.  
1.2 The purpose of this practice is to provide a scientifically-based strategy for developing models that can be used to estimate material alteration behavior after a repository is permanently closed (that is, the post-closure period) because the timescales involved with geological disposal preclude direct validation of predictions.  
1.3 This practice also addresses uncertainties in materials behavior models and the impact on the confidence in the EBS design criteria, the scientific bases of alteration models, and repository performance assessments using those models. This includes the identification and use of conservative assumptions to address uncertainty in the long-term performance of materials.  
1.3.1 Steps involved in evaluating the performance of waste forms and EBS materials include problem definition, laboratory and field testing, modeling of individual and coupled processes, and model confirmation.  
1.3.2 The estimates of waste form and EBS material performance are based on models derived from theoretical considerations, expert judgments, and interpretations of data obtained from tests and analyses of appropriate analogs.  
1.3.3 For the purpose of this practice, tests are ...

SIGNIFICANCE AND USE
5.1 This test method can be used to measure the release of a component from a cylindrical solidified waste form into water at the reference temperature of 20°C and at elevated temperatures that accelerate the rate and extent of leaching relative to the values measured at 20°C.  
5.2 This test method can be used to:  
5.2.1 Compare releases of waste components from various types of solidification agents and formulations.  
5.2.2 Determine the diffusion coefficients for the release of waste components from waste forms at a specific temperature.  
5.2.3 Promote greater extents of reaction than can be achieved under expected service conditions within a laboratory time frame to provide greater confidence in modeled diffusive releases.  
5.2.4 Determine the temperature dependence of diffusive release.  
5.3 Fitting the experimental results with a mechanistic model allows diffusive releases to be extrapolated to long times and to full-scale waste forms under the following constraints:  
5.3.1 Results of this test method address an intrinsic property of a material and should not be presumed to represent releases in specific disposal environments. Tests can be conducted under conditions that represent a specific disposal environment (for example, by using a representative groundwater) to determine an effective diffusion coefficient for those conditions.  
5.3.2 Projections of releases over long times requires that the waste form matrix remain stable, which may be demonstrated by the behavior of the specimen in ALTs at elevated temperatures.  
5.3.3 Extrapolations in time and scale are limited to values that correspond to the maximum CFL value obtained in an accelerated test.  
5.3.4 The mechanism must be the same at all temperatures used in the extrapolation. The same model that describes the results of tests conducted at elevated temperatures must also describe the results of tests run at the reference temperature of 20°C.
SCOPE
1.1 This test method provides procedures for measuring the leach rates of elements from a solidified matrix material, determining if the releases are controlled by mass diffusion, computing values of diffusion constants based on models, and verifying projected long-term diffusive releases. This test method is applicable to any material that does not degrade or deform during the test.  
1.1.1 If mass diffusion is the dominant step in the leaching mechanism, then the results of this test can be used to calculate diffusion coefficients using mathematical diffusion models. A computer program developed for that purpose is available as a companion to this test method (Note 1).  
1.1.2 It should be verified that leaching is controlled by diffusion by a means other than analysis of the leach test solution data. Analysis of concentration profiles of species of interest near the surface of the solid waste form after the test is recommended for this purpose.  
1.1.3 Potential effects of partitioning on the test results can be identified through modeling, although further testing and analyses are required to determine the cause of partitioning (for example, if it occurs during production of the material or as a result of leaching).  
1.2 The method is a modification of other semi-dynamic tests such as the IAEA test (1)2 and the ANS 16.1 Leach Test wherein elevated temperatures are used to accelerate diffusive release to an extent that would only be reached after very long times at lower temperatures. This approach provides a mechanistic basis for calculating diffusive releases at repository-relevant temperatures over long times, provided that the leaching mechanism does not change with temperature.  
1.2.1 Tests can be conducted at elevated temperatures to accelerate diffusive release and provide a mechanistic basis for calculating diffusive releases that would occur at lower temperatures over long times. Tests conducted at high temperatures allow the temperature dependen...

SIGNIFICANCE AND USE
4.1 Drying of the SNF and fuel cavity of the SNF container and its internals is needed to prepare for sealed dry storage, transportation, or permanent disposal at a repository. This guide provides technical information for use in determining the forms of water that need to be considered when choosing a drying process. This guide provides information to aid in (a) selecting a drying system, (b) selecting a drying method, and (c) demonstrating that adequate dryness was achieved.  
4.2 The considerations affecting drying processes include:  
4.2.1 Water remaining on and in commercial, research, and production reactor spent nuclear fuels after removal from wet storage may become an issue when the fuel is sealed in a dry storage system or transport cask. The movement to a dry storage environment typically results in an increase in fuel temperature, which may be sufficient to cause the release of water from the fuel. The water release coupled with the temperature increase in a sealed container may result in container pressurization, corrosion of fuel or assembly structures, or both, that could affect retrieval of the fuel, and container corrosion.  
4.2.2 Removal of the water associated with the SNF may be accomplished by a variety of technologies including heating, imposing a vacuum over the system, flushing the system with dry gases, and combinations of these and other similar processes.  
4.2.3 Water removal processes are time, temperature, and pressure-dependent. Residual water in some form(s) should be anticipated.  
4.2.4 Drying processes may not readily remove the water that was retained in porous materials, capillaries, sludge, CRUD, and as thin wetted surface films. Water trapped within damaged SNF may be especially difficult to remove.  
4.2.5 Drying processes may be even less successful in removing bound water from the SNF and associated materials because removal of bound water will only occur when the threshold energy required to break the specific water...
SCOPE
1.1 This guide discusses three steps in preparing spent nuclear fuel (SNF) for placement in a sealed dry storage system: (1) evaluating the needs for drying the SNF after removal from a water storage pool and prior to placement in dry storage, (2) drying the SNF, and (3) demonstrating that adequate dryness has been achieved.  
1.1.1 The guide addresses drying methods and their limitations when applied to the drying of SNF that has been stored in water pools. The guide discusses sources and forms of water that may remain in the SNF, the container, or both after the drying process has been completed. It also discusses the important and potential effects of the drying process and any residual water on fuel integrity and container materials during the dry storage period. The effects of residual water are discussed mechanistically as a function of the container thermal and radiological environment to provide guidance on situations that may require extraordinary drying methods, specialized handling, or other treatments.  
1.1.2 The basic issues in drying are: (1) to determine how dry the SNF must be in order to prevent problems with fuel retrievability, container pressurization, or container corrosion during storage, handling, and transfer, and (2) to demonstrate that adequate dryness has been achieved. Achieving adequate dryness may be straightforward for undamaged commercial fuel but complex for any SNF where cladding damage has occurred prior to or during placement and storage at the spent fuel pools. Challenges in achieving adequate dryness may also result from the presence of sludge, CRUD, and any other hydrated compounds. These may be transferred with the SNF to the storage container and may hold water and resist drying.  
1.1.3 Units are given in both SI and non-SI units as is industry standard. In some cases, mathematical equivalents are given in parentheses.  
1.2 This standard only discusses SNF drying and does n...

SIGNIFICANCE AND USE
5.1 These test methods provide data useful for evaluating the chemical durability (see 3.1.5) of glass waste forms as measured by elemental release. Accordingly, it may be applicable throughout manufacturing, research, and development.  
5.1.1 Test Method A can specifically be used to obtain data to evaluate whether the chemical durability of glass waste forms have been consistently controlled during production (see Table 1).  
5.1.2 Test Method B can specifically be used to measure the chemical durability of glass waste forms under various test conditions, for example, varying test durations, test temperatures, ratios of sample-surface area (S) to leachant volume (V) (see Appendix X1), and leachant types (see Table 1). Data from this test may form part of the larger body of data that are necessary in the logical approach to long-term prediction of waste form behavior (see Practice C1174).
SCOPE
1.1 These product consistency test methods A and B provide a measure of the chemical durability of homogeneous glasses, phase separated glasses, devitrified glasses, glass ceramics, multiphase glass ceramic waste forms, or combinations thereof, hereafter collectively referred to as “glass waste forms” by measuring the concentrations of the chemical species released to a test solution under carefully controlled conditions.  
1.1.1 Test Method A is a seven-day chemical durability test performed at 90 ± 2°C in a leachant of ASTM-Type I water. The test method is static and conducted in stainless steel vessels. The stainless steel vessels require a gasket to remain leak-tight.2 The stainless steel vessels are considered to be “closed system” tests. Test Method A can specifically be used to evaluate whether the chemical durability and elemental release characteristics of nuclear, hazardous, and mixed glass waste forms have been consistently controlled during production. This test method is applicable to radioactive and simulated glass waste forms as defined above.  
1.1.2 Test Method B is a durability test that allows testing at various test durations, test temperatures, particle size and masses of glass sample, leachant volumes, and leachant compositions. This test method is static and can be conducted in stainless steel or PFA TFE-fluorocarbon vessels. The stainless steel vessels are considered to be “closed system” while the PFA TFE-fluorocarbon vessels are considered to be “open system” tests. Test Method B can specifically be used to evaluate the relative chemical durability characteristics of homogeneous glasses, phase separated glasses, devitrified glasses, glass ceramics, or multiphase glass ceramic waste forms, or combinations thereof. This test method is applicable to radioactive (nuclear) and mixed, hazardous, and simulated glass waste forms as defined above. Test Method B cannot be used as a consistency test for production of high level radioactive glass waste forms.  
1.2 These test methods must be performed in accordance with all quality assurance requirements for acceptance of the data.  
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 and health practices and determine the applicability of regulatory limitations prior to use.

SIGNIFICANCE AND USE
5.1 This practice supports the development of materials behavior models that can be used to predict alterations in materials over the very long time periods pertinent to the operation of a high-level nuclear waste repository; periods of time much longer than can be tested directly. Under the very extended service periods relevant to geological disposal—much longer periods than those encountered in normal engineering practice—equilibrium or steady state conditions may be achieved and models for reaction kinetics may be replaced by models, if justified, describing equilibrium extents of alteration. This practice is intended for use for waste form materials and materials proposed for use in an EBS that is designed to contain radionuclides released from high-level nuclear waste forms as they degrade over tens of thousands of years and more. Various U.S. Government regulations pertinent to repository disposal in the United States are as follows:  
5.1.1 Public Law 97–425, the Nuclear Waste Policy Act of 1982, provides for the deep geologic disposal of high-level radioactive waste through a system of multiple barriers. The radiation release limits are to be set by the U.S. Environmental Protection Agency (EPA) (40 CFR 191). Licensing of such disposal will be done by the U.S. Nuclear Regulatory Commission (NRC).  
5.1.2 The analyses described in this Standard Guide can be used to support the demonstration of compliance of the EBS components and design to the applicable requirements of 10 CFR 60 (pertaining to any HLW repository in the U.S.) and 10 CFR 63 (pertaining to the planned HLW repository at Yucca Mountain, NV).
5.1.2.1 10 CFR 60.135 and 60.113 require that the waste form be a material that is solid, non-particulate, non-pyrophoric, and non-chemically reactive, and that the waste package contain no liquid, particulates, chemically reactive or combustible materials and that the materials/components of the EBS be designed to provide – assuming anticipated processe...
SCOPE
1.1 This practice describes test methods and data analyses used to develop models for the prediction of the long-term behavior of materials, such as engineered barrier system (EBS) materials and waste forms, used in the geologic disposal of spent nuclear fuel (SNF) and other high-level nuclear waste in a geologic repository. The alteration behavior of waste form and EBS materials is important because it affects the retention of radionuclides by the disposal system. The waste form and EBS materials provide a barrier to release either directly (as in the case of waste forms in which the radionuclides are initially immobilized), or indirectly (as in the case of containment materials that restrict the ingress of groundwater or the egress of radionuclides that are released as the waste forms and EBS materials degrade).  
1.1.1 Steps involved in making such predictions include problem definition, testing, modeling, and model confirmation.  
1.1.2 The predictions are based on models derived from theoretical considerations, expert judgment, interpretation of data obtained from tests, and appropriate analogs. 1.1.3 For the purpose of this practice, tests  
1.1.3 For the purpose of this practice, tests are categorized according to the information they provide and how it is used for model development and use. These tests may include but are not limited to the following:
1.1.3.1 Attribute tests to measure intrinsic materials properties,
1.1.3.2 Characterization tests to measure the effects of material and environmental variables on behavior,
1.1.3.3 Accelerated tests to accelerate alteration and determine important mechanisms and processes that can affect the performance of waste form and EBS materials,
1.1.3.4 Service condition tests to confirm the appropriateness of the model and variables for anticipated disposal conditions,
1.1.3.5 Confirmation tests to verify the predictive capacity of the model, and
1.1.3.6 Tes...

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.  
5.3 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.
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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 TS-R-1. 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 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.  
1.3 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.  
1.4 Fire testing is inherently hazardous. Adequate safeguards for personnel and property shall be employed in conducting these tests.

SIGNIFICANCE AND USE
Measurements performed in this guide are limited to radioactive solutions, slurries, and sludges as well as simulants designed to model the properties of these radioactive materials.
Data obtained from the measurement and calculation of physical and rheological properties of radioactive solutions, slurries, and sludges are essential in developing appropriate simulants for design and testing of retrieval, transport, mixing, and storage systems for treatment of radioactive materials. These data also provide input parameters for modeling the flow behavior, processing, transport, safety, and storage of these radioactive materials.
Consistency in the handling of samples, measurement methods, and calculations is essential in obtaining reproducible results of rheological and physical property measurements.
This guide will be used to measure or calculate the physical properties listed below.  
Bulk slurry density
Settled solids density
Centrifuged solids density
Supernatant density
Settling rate
Volume percent centrifuged solids
Volume percent settled solids after settling
Undissolved solids content
Dissolved solids content  
Weight percent centrifuged solids
Weight percent total oxides
Solids content of the centrifuged solids
Total solids content
This guide describes the process of performing measurement of the rheological properties. The rheological measurements and calculations described in this guide are limited to shear strength, shear stress versus shear rate, apparent viscosity, consistency, and yield stress.
Due to the nature of some solutions, slurries, and sludges, not all of the measurements described in this standard may be applicable to all samples. For example, some sludges do not settle; therefore, settling rate measurements are not applicable for these samples.
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1.1 Intent:
1.1.1 The intent of this guide is to provide guidance for the measurement and calculation of physical and rheological properties of radioactive solutions, slurries, and sludges as well as simulants designed to model the properties of these radioactive materials.
1.2 Applicability:
1.2.1 This guide is intended for measurement of mass and volume of the solution, slurries, and sludges as well as dissolved solids content in the liquid fraction and solids content associated with the slurries and sludges. Particle size distribution is also measured.
1.2.2 This guide identifies the data required and the equations recommended for calculation of density (bulk, settled solids, supernatant, and centrifuged solids), settling rate, volume and weight percent of the centrifuged solids and settled solids, and the weight percent undissolved solids, dissolved solids, and total oxides.  
1.2.3 This guide is intended for measurement of shear strength and shear stress as a function of shear rate.
1.2.4 Rheological property measurement guidelines in this guide are limited to rotational rheometers.
1.2.5 This guide is limited to measurements of viscous and incipient flow and does not include oscillatory rheometry.  
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 and health practices and determine the applicability of regulatory limitations prior to use.

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1.1 Intent:
1.1.1 The intent of this guideline is to provide general considerations for the development, verification, validation, and documentation of high-level waste (HLW) tank simulants. Due to the expense and hazards associated with obtaining and working with actual wastes, especially radioactive wastes, simulants are used in a wide variety of applications including process and equipment development and testing, equipment acceptance testing, and plant commissioning. This standard guide facilitates a consistent methodology for development, preparation, verification, validation, and documentation of waste simulants.
1.2 This guideline provides direction on (1) defining simulant use, (2) defining simulant-design requirements, (3) developing a simulant preparation procedure, (4) verifying and validating that the simulant meets design requirements, and (5) documenting simulant-development activities and simulant preparation procedures.
1.3 Applicability:
1.3.1 This guide is intended for persons and organizations tasked with developing HLW simulants to mimic certain characteristics and properties of actual wastes. The process for simulant development, verification, validation, and documentation is shown schematically in Fig. 1. Specific approval requirements for the simulants developed under this guideline are not provided. This topic is left to the performing organization.
1.3.2 While this guide is directed at HLW simulants, much of the guidance may also be applicable to other aqueous based solutions and slurries.
1.3.3 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.
1.4 User Caveats:
1.4.1 This guideline is not a substitute for sound chemistry and chemical engineering skills, proven practices and experience. It is not intended to be prescriptive but rather to provide considerations for the development and use of waste simulants.
1.4.2 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.
FIG. 1 Simulant Development, Verification, Validation, and Documentation Flowsheet

SIGNIFICANCE AND USE
Obtaining samples of high-level waste created during the reprocessing of spent nuclear fuels presents unique challenges. Generally, high-level waste is stored in tanks with limited access to decrease the potential for radiation exposure to personnel. Samples must be obtained remotely because of the high radiation dose from the bulk material and the samples; samples require shielding for handling, transport, and storage. The quantity of sample that can be obtained and transported is small due to the hazardous nature of the samples as well as their high radiation dose.
Many high-level wastes have been treated to remove strontium (Sr) or cesium (Cs), or both, underwent liquid volume reductions through forced evaporation or have been pH modified, or both, to decrease corrosion of the tanks. These processes, as well as waste streams added from multiple process plant operations, often resulted in precipitation, and produced multiphase wastes that are heterogeneous. Evaporation of water from waste with significant dissolved salts concentrations has occurred in some tanks due to the high heat load associated with the high-level waste and by intentional evaporative processing, resulting in the formation of a saltcake or crusts, or both. Organic layers exist in some waste tanks, creating additional heterogeneity in the wastes.  
Due to these extraordinary challenges, substantial effort in research and development has been expended to develop techniques to provide grab samples of the contents of the high-level waste tanks. A summary of the primary techniques used to obtain samples from high-level waste tanks is provided in Table 1. These techniques will be summarized in this guideline with the assumption that the tank headspace is adequately ventilated during sampling.  
TABLE 1 High-Level Waste Tank Sampling Methods  TechniqueMaterial TypeNotes SolidSlurryLiquid HAST in-tank
needle orifice
X
Orifice as part of Reverse-Flow
Diverter (RFD)
  B...
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1.1 This guide addresses techniques used to obtain grab samples from tanks containing high-level radioactive waste created during the reprocessing of spent nuclear fuels. Guidance on selecting appropriate sampling devices for waste covered by the Resource Conservation and Recovery Act (RCRA) is also provided by the United States Environmental Protection Agency (EPA) (1). Vapor sampling of the head-space is not included in this guide because it does not significantly affect slurry retrieval, pipeline transport, plugging, or mixing.
1.2 The values stated in inch-pound units are to be regarded as standard. No other units of measurement are included in this standard.
1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

SIGNIFICANCE AND USE
This procedure can be used for (but is 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 practices 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 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).  
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: (A) Gradient Temperature Furnace Method (GT), (B) Uniform Temperature Furnace Method (UT), and (C) 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 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.
1.4 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 and health practices and determine the applicability of regulatory limitations prior to use.

SIGNIFICANCE AND USE
5.1 This procedure can be used for (but is 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.
SCOPE
1.1 These practices 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 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).  
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: (A) Gradient Temperature Furnace Method (GT), (B) Uniform Temperature Furnace Method (UT), and (C) 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 and health practices and determine the applicability of regulatory limitations prior to use.

SIGNIFICANCE AND USE
The distribution coefficient, Kd, is an experimentally determined ratio quantifying the distribution of a chemical species between a given fluid and geomedium sample under certain conditions, including the attainment of constant aqueous concentrations of the species of interest. The Kd concept is used in mass transport modeling, for example, to assess the degree to which the movement of a species will be delayed by interactions with the geomedium as the solution migrates through the geosphere under a given set of underground geochemical conditions (pH, temperature, ionic strength, etc.). The retardation factor (Rf) is the ratio of the velocity of the groundwater divided by the velocity of the contaminant, which can be expressed as:
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1.1 This test method covers the determination of distribution coefficients of chemical species to quantify uptake onto solid materials by a batch sorption technique. It is a laboratory method primarily intended to assess sorption of dissolved ionic species subject to migration through pores and interstices of site specific geomedia. It may also be applied to other materials such as manufactured adsorption media and construction materials. Application of the results to long-term field behavior is not addressed in this method. Distribution coefficients for radionuclides in selected geomedia are commonly determined for the purpose of assessing potential migratory behavior of contaminants in the subsurface of contaminated sites and waste disposal facilities. This test method is also applicable to studies for parametric studies of the variables and mechanisms which contribute to the measured distribution coefficient.
1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.
1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

SIGNIFICANCE AND USE
This practice provides a prescriptive description of the design of a SPFT test apparatus and identifies aspects of the performance of SPFT tests and interpretation of test results that must be addressed by the experimenter to provide confidence in the measured dissolution rate.
The SPFT test method described in this practice can be used to characterize various aspects of glass corrosion behavior that can be utilized in a mechanistic model for calculating long-term behavior of a nuclear waste glass.
Depending on the values of test parameters that are used, the results of SPFT tests can be used to measure the intrinsic dissolution rate of a glass, the temperature and pH dependencies of the rate, and the effects of various dissolved species on the dissolution rate.
The reacted sample recovered from a test may be examined with surface analytical techniques, such as scanning electron microscopy, to further characterize the corrosion behavior. Such examinations may provide evidence regarding whether the glass is dissolving stoichiometrically, if particular leached layers and secondary phases were formed on the specimen surface, and so forth. These occurrences may impact the accuracy of the glass dissolution rate that is measured using this method. This practice does not address the analysis of solid reaction materials.
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1.1 This practice describes a single-pass flow-through (SPFT) test method that can be used to measure the dissolution rate of a homogeneous silicate glass, including nuclear waste glasses, in various test solutions at temperatures less than 100°C. Tests may be conducted under conditions in which the effects from dissolved species on the dissolution rate are minimized to measure the forward dissolution rate at specific values of temperature and pH, or to measure the dependence of the dissolution rate on the concentrations of various solute species.
1.2 Tests are conducted by pumping solutions in either a continuous or pulsed flow mode through a reaction cell that contains the test specimen. Tests must be conducted at several solution flow rates to evaluate the effect of the flow rate on the glass dissolution rate.
1.3 This practice excludes static test methods in which flow is simulated by manually removing solution from the reaction cell and replacing it with fresh solution.
1.4 Tests may be conducted with demineralized water, chemical solutions (such as pH buffer solutions, simulated groundwater solutions, and brines), or actual groundwater.
1.5 Tests may be conducted with crushed glass of a known size fraction or monolithic specimens having known geometric surface area. The reacted solids may be examined to provide additional information regarding the behavior of the material in the test and the reaction mechanism.
1.6 Tests may be conducted with glasses containing radionuclides. However, this test method does not address safety issues for radioactive samples.
1.7 Data from these tests can be used to determine the values of kinetic model parameters needed to calculate the glass corrosion behavior in a disposal system over long periods (for example, see Practice C1174).
1.8 This practice must be performed in accordance with all quality assurance requirements for acceptance of the data.
1.9 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.
1.10 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.

SIGNIFICANCE AND USE
This test method can be used to provide a measure of the reactivity of a material in a dilute solution in which the test response is dominated by the dissolution or leaching of the test specimen. It can be used to compare the dissolution or leaching behaviors of candidate radioactive waste forms and to study the reactions during static exposure to dilute solutions in which solution feed-back effects can be maintained negligible, depending on the test conditions.
The test is suitable for application to natural minerals, simulated waste form materials, and radioactive waste form material specimens.
Data from this test may form part of the larger body of data that is necessary in the logical approach to long-term prediction of waste form behavior, as described in Practice C1174. In particular, measured solution concentrations and characterizations of altered surfaces may be used in the validation of geochemical modeling codes.
This test method excludes the use of crushed or powdered specimens and organic materials.
Several reference test parameter values and reference leachant solutions are specified to facilitate the comparison of results of tests conducted with different materials and at different laboratories. However, other test parameter values and leachant solution compositions can be used to characterize the specimen reactivity.
Tests can be conducted with different leachant compositions to simulate groundwaters, buffer the leachate pH as the specimen dissolves, or measure the common ion effect of particular solutes.
Tests can be conducted to measure the effects of various test parameter values on the specimen response, including time, temperature, and S/V ratio. Tests conducted for different durations and at various temperatures provide insight into the reaction kinetics. Tests conducted at different S/V ratio provide insight into chemical affinity (solution feed-back effects) and the approach to saturation.  
Either aerated or deaerated solutions ...
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1.1 This test method provides a measure of the chemical durability of a simulated or radioactive monolithic waste form, such as a glass, ceramic, cement (grout), or cermet, in a test solution at temperatures 100°C under low specimen surface- area-to-leachant volume (S/V) ratio conditions.
1.2 This test method can be used to characterize the dissolution or leaching behaviors of various simulated or radioactive waste forms in various leachants under the specific conditions of the test based on analysis of the test solution. Data from this test are used to calculate normalized elemental mass loss values from specimens exposed to aqueous solutions at temperatures 100°C.
1.3 The test is conducted under static conditions in a constant solution volume and at a constant temperature. The reactivity of the test specimen is determined from the amounts of components released and accumulated in the solution over the test duration. A wide range of test conditions can be used to study material behavior, including various leachant composition, specimen surface area-to-leachant volume ratios, temperatures, and test durations.
1.4 Three leachant compositions and four reference test matrices of test conditions are recommended to characterize materials behavior and facilitate interlaboratory comparisons of tests results.
1.5 Specimen surfaces may become altered during this test. Although not part of the test method, it is recommended that these altered surface regions be examined to characterize chemical and physical changes due to the reaction of waste forms during static exposure to solutions.
1.6 This test method is not recommended for evaluating metallic materials, the degradation of which includes oxidation reactions that are not controlled by this test method.
1.7 This test method must be performed in accordance with all applicable quality assurance requirements for acceptance of the data.
1.8 The values stated in SI uni...

SIGNIFICANCE AND USE
Disposition of aluminum-based spent nuclear fuel will involve:
Removal from the existing storage or transfer facility,
Characterization or treatment, or both, of the fuel or the resulting waste form, or both,
Placement of the waste form in a canister,
Placement of the canister in a safe and environmentally sound interim storage facility,
Removal from the interim storage facility and transport to the repository,
placement in a waste container,
Emplacement in the repository, and
Repository closure and geologic disposal. Actions in each of these steps may significantly impact the success of any subsequent step.
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.
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.
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).
A knowledge of the eff...
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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 inv...

SIGNIFICANCE AND USE
Information is provided in this document and other referenced documents to assist the licensee and the licensor in analyzing the materials aspects of performance of SNF and DCSS components during extended storage. The effects of the service conditions of the first licensing period are reviewed in the license renewal process. These service conditions are highlighted and discussed in Annex A1 as factors that affect materials performance in an ISFSI. Emphasis is on the effects of time, temperature, radiation, and the environment on the condition of the SNF and the performance of components of ISFSI storage systems.
The storage of SNF that is irradiated under the regulations of 10 CFR Part 50 is governed by regulations in 10 CFR Part 72. Regulatory requirements for the subsequent geologic disposal of this SNF are presently given in 10 CFR Part 60, with specific requirements for the use of Yucca Mountain as a repository being given in the regulatory requirements of 10 CFR Part 63. Between the life-cycle phases of storage and disposal, SNF may be transported under the requirements of 10 CFR Part 71. Therefore, in storage, it is important to acknowledge the transport and disposal phases of the life cycle. In doing this, the materials properties that are important to these subsequent phases are to be considered in order to promote successful completion of these subsequent phases in the life cycle of SNF. Retrievability of SNF (or high-level radioactive waste) is set as a requirement in 10 CFR Part 72.122(g)(5) and 10 CFR Part 72.122(l). Care should be taken in operations conducted prior to disposal, for example, storage, transfer, and transport, to ensure that the SNF is not abused and that SNF assemblies will be retrievable, the protective value of the cladding is not degraded and remains capable of serving as an active barrier to radionuclide release during transfer and transport operations. It is possible that cladding could be altered during dry storage. The hydrogen ...
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1.1 Part of the total inventory of commercial spent nuclear fuel (SNF) is stored in dry cask storage systems (DCSS) under licenses granted by the U.S. Nuclear Regulatory Commission (NRC). The purpose of this guide is to provide information to assist in supporting the renewal of these licenses, safely and without removal of the SNF from its licensed confinement, for periods beyond those governed by the term of the original license. This guide provides information on materials behavior under conditions that may be important to safety evaluations for the extended service of the renewal period. This guide is written for DCSS containing light water reactor (LWR) fuel that is clad in zirconium alloy material and stored in accordance with the Code of Federal Regulations (CFR), at an independent spent-fuel storage installation (ISFSI). The components of an ISFSI, addressed in this document, include the commercial SNF, canister, cask, and all parts of the storage installation including the ISFSI pad. The language of this guide is based, in part, on the requirements for a dry SNF storage license that is granted, by the U.S. Nuclear Regulatory Commission (NRC), for up to 20 years. Although government regulations may differ for various nations, the guidance on materials properties and behavior given here is expected to have broad applicability.
1.2 This guide addresses many of the factors affecting the time-dependent behavior of materials under ISFSI service [10 CFR Part 72.42]. These factors are those regarded to be important to performance, in license extension, beyond the currently licensed 20-year period. Examples of these factors are given in this guide and they include materials alterations or environmental conditions for components of an ISFSI system that, over time, could have significance related to safety. For purposes of this guide, a license period of an additional 20 to 80 years is assumed.
1.3 This guide addresses ...

SIGNIFICANCE AND USE
In order to demonstrate conformance to regulatory requirements and support the postclosure repository performance assessment information is required about the attributes, characteristics, and behavior of the SNF need to be determined. These properties of the SNF in turn support the transport, interim storage, and repository preclosure safety analyses, and repository postclosure performance assessment. The interim dry storage of commercial LWR SNF is regulated per 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. 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. 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. Also, pinholes/cracks in fuel rods in dry or wet interim storage can also develop into much larger defects (for example, the phenomenon of cladding “unzipping”) under long-term repository conditions. 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.
In the U.S. the disposal of spent nuclear fuel (SNF) and high level waste (HLW) in a geologic repository is regulated under the Code of Federal Regulations, Title 10, Part 60 (which pertains to any geologic repository built in the U.S.), Code of Federal Regulations, Title 10, Part 63 (which pertains...
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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 and test reactors. 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 repository performance, particularly in areas where an extensive body of literature already exists for the parameter of interest.
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 the geologic repository. The SNF may have been stored either wet (for example, a spent fuel pool), or dry (for example, an independent spent fuel storage installation (ISFSI)), or both, and that the manner of interim storage may affect the SNF characteristics.
1.4 The values stated in ...

SIGNIFICANCE AND USE
The vapor hydration test can be used to study the corrosion of glass and glass ceramic waste forms under conditions of high temperature and contact by water vapor or thin films of water. This method may serve as an accelerated test for some materials, since the high temperatures will accelerate thermally activated processes. A wide range of test temperatures have been reported in the literature –40°C (Ebert et al, 2005 (3), for example) to 300°C (Vienna et al, 2001 (4), for example). It should be noted that with increased test temperature comes the possibility of changing the corrosion rate determining mechanism and the types of phases formed upon alteration from those that occur in the disposal environment (Vienna et al, 2001 (4)).
The vapor hydration test can be used as a screening test to determine the propensity of waste forms to alter and for relative comparisons in alteration rates between waste forms.
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1.1 The vapor hydration test method can be used to study the corrosion of a waste forms such as glasses and glass ceramics upon exposure to water vapor at elevated temperatures. In addition, the alteration phases that form can be used as indicators of those phases that may form under repository conditions. These tests; which allow altering of glass at high surface area to solution volume ratio; provide useful information regarding the alteration phases that are formed, the disposition of radioactive and hazardous components, and the alteration kinetics under the specific test conditions. This information may be used in performance assessment (McGrail et al, 2002 (1) for example).
1.2 This test method must be performed in accordance with all quality assurance requirements for acceptance of the data.
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.

SIGNIFICANCE AND USE
This test method can be used to measure the release of a component from a cylindrical solidified waste form into water at the reference temperature of 20°C and at elevated temperatures that accelerate the rate and extent of leaching relative to the values measured at 20°C.  
This test method can be used to:
Compare releases of waste components from various types of solidification agents and formulations.
Determine the diffusion coefficients for the release of waste components from waste forms at a specific temperature.
Promote greater extents of reaction than can be achieved under expected service conditions within a laboratory time frame to provide greater confidence in modeled diffusive releases.
Determine the temperature dependence of diffusive release.
Fitting the experimental results with a mechanistic model allows diffusive releases to be extrapolated to long times and to full-scale waste forms under the following constraints:
Results of this test method address an intrinsic property of a material and should not be presumed to represent releases in specific disposal environments. Tests can be conducted under conditions that represent a specific disposal environment (for example, by using a representative groundwater) to determine an effective diffusion coefficient for those conditions.
Projections of releases over long times requires that the waste form matrix remain stable, which may be demonstrated by the behavior of the specimen in ALTs at elevated temperatures.
Extrapolations in time and scale are limited to values that correspond to the maximum CFL value obtained in an accelerated test.
The mechanism must be the same at all temperatures used in the extrapolation. The same model that describes the results of tests conducted at elevated temperatures must also describe the results of tests run at the reference temperature of 20°C.
SCOPE
1.1 This test method provides procedures for measuring the leach rates of elements from a solidified matrix material, determining if the releases are controlled by mass diffusion, computing values of diffusion constants based on models, and verifying projected long-term diffusive releases. This test method is applicable to any material that does not degrade or deform during the test.
1.1.1 If mass diffusion is the dominant step in the leaching mechanism, then the results of this test can be used to calculate diffusion coefficients using mathematical diffusion models. A computer program developed for that purpose is available as a companion to this test method (Note 1).
1.1.2 It should be verified that leaching is controlled by diffusion by a means other than analysis of the leach test solution data. Analysis of concentration profiles of species of interest near the surface of the solid waste form after the test is recommended for this purpose.
1.1.3 Potential effects of partitioning on the test results can be identified through modeling, although further testing and analyses are required to determine the cause of partitioning (for example, if it occurs during production of the material or as a result of leaching).
1.2 The method is a modification of other semi-dynamic tests such as the IAEA test (1) and the ANS 16.1 Leach Test wherein elevated temperatures are used to accelerate diffusive release to an extent that would only be reached after very long times at lower temperatures. This approach provides a mechanistic basis for calculating diffusive releases at repository-relevant temperatures over long times, provided that the leaching mechanism does not change with temperature.
1.2.1 Tests can be conducted at elevated temperatures to accelerate diffusive release and provide a mechanistic basis for calculating diffusive releases that would occur at lower temperatures over long times. Tests conducted at high temperatures allow the temperature dependence of the diffusion coefficient to be determined. They also demonstrate tha...

SIGNIFICANCE AND USE
These test methods provide data useful for evaluating the chemical durability (see 3.1.4) of glass waste forms as measured by elemental release. Accordingly, it may be applicable throughout manufacturing, research, and development.  
Test Method A can specifically be used to obtain data to evaluate whether the chemical durability of glass waste forms have been consistently controlled during production (see Table 1).
Test Method B can specifically be used to measure the chemical durability of glass waste forms under various leaching conditions, for example, varying test durations, test temperatures, ratio of sample-surface area (S) to leachant volume (V) (see Appendix X1), and leachant types (see Table 1). Data from this test may form part of the larger body of data that are necessary in the logical approach to long-term prediction of waste form behavior (see Practice C 1174).
SCOPE
1.1 These product consistency test methods A and B evaluate the chemical durability of homogeneous glasses, phase separated glasses, devitrified glasses, glass ceramics, and/or multiphase glass ceramic waste forms hereafter collectively referred to as “glass waste forms” by measuring the concentrations of the chemical species released to a test solution.
1.1.1 Test Method A is a seven-day chemical durability test performed at 90 ± 2°C in a leachant of ASTM-Type I water. The test method is static and conducted in stainless steel vessels. Test Method A can specifically be used to evaluate whether the chemical durability and elemental release characteristics of nuclear, hazardous, and mixed glass waste forms have been consistently controlled during production. This test method is applicable to radioactive and simulated glass waste forms as defined above.
1.1.2 Test Method B is a durability test that allows testing at various test durations, test temperatures, mesh size, mass of sample, leachant volume, and leachant compositions. This test method is static and can be conducted in stainless steel or PFA TFE-fluorocarbon vessels, or both. Test Method B can specifically be used to evaluate the relative chemical durability characteristics of homogeneous glasses, phase separated glasses, devitrified glasses, glass ceramics, and/or multiphase glass ceramic waste forms. This test method is applicable to radioactive (nuclear) and mixed, hazardous, and simulated waste forms as defined above. Test Method B cannot be used as a consistency test for production of high level radioactive glass waste forms.
1.2 These test methods must be performed in accordance with all quality assurance requirements for acceptance of the data.
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 and health practices and determine the applicability of regulatory limitations prior to use.

SIGNIFICANCE AND USE
This guide provides technical information for use by SNF owners to determine the forms of water usually associated with spent nuclear fuel due to corrosion damage of the fuel, cladding and storage materials during irradiation and in storage pools. Drying may be needed to prepare the SNF for sealed dry storage, transportation, and/or permanent disposal at a repository. This guide provides information for:
4.1.1 Evaluating what drying system should be used,
4.1.2 Drying methods, and
4.1.3 Methods to confirm that adequate dryness was achieved.
The guide can be used to determine:
4.2.1 Drying technologies that are designed to remove most of the unbound water but will not remove all forms of water. Water remaining on and in commercial and research reactor spent nuclear fuels coming from water basin storage may become an issue when the fuel is sealed in a dry storage system or transport cask. The movement to a dry storage environment typically results in an increase in fuel temperature due to the decay heat. This temperature change could be significant to cause the release of water remaining in a sealed dry package that may result in container pressurization, fuel retrievability issues, and container corrosion.
4.2.2 A methodology for evaluating drying processes that may not readily remove all forms of water that may be retained in pores in fuel cladding, capillaries, sludge, crud, and thin wetted surface films. Drying techniques are even less successful in removing bound water. Removal of bound water will only occur when the specific threshold energy is applied to break the bonds involved and release the water. For spent nuclear fuel this threshold energy may come from the combination of thermal input and ionizing radiation.
4.2.3 How the residual water retained with the SNF, CRUD and sludge inside a sealed package may become available to react with the internal environment, the fuel, and the package materials as a result of extended time at equilibrium dry...
SCOPE
1.1 This guide is organized to discuss the three major components of significance in the drying behavior of spent nuclear fuel: evaluating the need for drying, drying spent nuclear fuel, and confirmation of adequate dryness.
1.1.1 The guide addresses drying methods and their limitations in drying spent nuclear fuels that have been in storage at water pools. The guide discusses sources and forms of water that remain in SNF, its container, or both, after the drying process and discusses the importance and potential effects they may have on fuel integrity, and container materials. The effects of residual water are discussed mechanistically as a function of the container thermal and radiological environment to provide guidance on situations that may require extraordinary drying methods, specialized handling, or other treatments.
1.1.2 The basic issue in drying is to determine how dry the SNF must be in order to prevent issues with fuel retrievability, container pressurization, or container corrosion. Adequate dryness may be readily achieved for undamaged commercial fuel but may become a complex issue for any SNF where cladding damage has occurred during fuel irradiation, storage, or both, at the spent fuel pools. Dryness issues may also result from the presence of sludge, crud, and other hydrated compounds connected to the SNF that hold water and resist drying efforts.
1.2 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.