4.1 The purpose of this guide is to provide information that will help to ensure that nuclear fuel dissolution facilities are conceived, designed, fabricated, constructed, and installed in an economic and efficient manner. This guide will help facilities meet the intended performance functions, eliminate or minimize the possibility of nuclear criticality and provide for the protection of both the operator personnel and the public at large under normal and abnormal (emergency) operating conditions as well as under credible failure or accident conditions.1.1 It is the intent of this guide to set forth criteria and procedures for the design, fabrication and installation of nuclear fuel dissolution facilities. This guide applies to and encompasses all processing steps or operations beyond the fuel shearing operation (not covered), up to and including the dissolving accountability vessel.1.2 Applicability and Exclusions: 1.2.1 Operations—This guide does not cover the operation of nuclear fuel dissolution facilities. Some operating considerations are noted to the extent that these impact upon or influence design.1.2.1.1 Dissolution Procedures—Fuel compositions, fuel element geometry, and fuel manufacturing methods are subject to continuous change in response to the demands of new reactor designs and requirements. These changes preclude the inclusion of design considerations for dissolvers suitable for the processing of all possible fuel types. This guide will only address equipment associated with dissolution cycles for those fuels that have been used most extensively in reactors as of the time of issue (or revision) of this guide. (See Appendix X1.)1.2.2 Processes—This guide covers the design, fabrication and installation of nuclear fuel dissolution facilities for fuels of the type currently used in Pressurized Water Reactors (PWR). Boiling Water Reactors (BWR), Pressurized Heavy Water Reactors (PHWR) and Heavy Water Reactors (HWR) and the fuel dissolution processing technologies discussed herein. However, much of the information and criteria presented may be applicable to the equipment for other dissolution processes such as for enriched uranium-aluminum fuels from typical research reactors, as well as for dissolution processes for some thorium and plutonium-containing fuels and others. The guide does not address equipment design for the dissolution of high burn-up or mixed oxide fuels.1.2.2.1 This guide does not address special dissolution processes that may require substantially different equipment or pose different hazards than those associated with the fuel types noted above. Examples of precluded cases are electrolytic dissolution and sodium-bonded fuels processing. The guide does not address the design and fabrication of continuous dissolvers.1.2.3 Ancillary or auxiliary facilities (for example, steam, cooling water, electrical services) are not covered. Cold chemical feed considerations are addressed briefly.1.2.4 Dissolution Pretreatment—Fuel pretreatment steps incidental to the preparation of spent fuel assemblies for dissolution reprocessing are not covered by this guide. This exclusion applies to thermal treatment steps such as “Voloxidation” to drive off gases prior to dissolution, to mechanical decladding operations or process steps associated with fuel elements disassembly and removal of end fittings, to chopping and shearing operations, and to any other pretreatment operations judged essential to an efficient nuclear fuels dissolution step.1.2.5 Fundamentals—This guide does not address specific chemical, physical or mechanical technology, fluid mechanics, stress analysis or other engineering fundamentals that are also applied in the creation of a safe design for nuclear fuel dissolution facilities.1.3 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard.1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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4.1 This specification provides uniform requirements for the preparation of test samples used for testing of coatings and linings to be used in nuclear power plants.4.2 At the users discretion, this standard may also be used when preparing samples to be tested for the purpose of assessing performance attributes for coating and lining systems that may be applied in other types of power plants or for other industrial facilities.4.3 Users of this guide must ensure that coatings work complies not only with this guide, but also with the licensee’s plant-specific quality assurance program and licensing commitments.AbstractThis specification defines the size composition and surface preparation requirements for the preparation of test samples used for qualification testing of coatings utilized in nuclear power plant construction and maintenance. All panels should be carbon steel. Materials shall be tested for abrasion, and shall conform to specified requirements of steel samples, and concrete blocks.1.1 This specification defines the size, composition, surface preparation, and coating application variables for preparing samples for evaluating coatings and linings over various substrates.1.2 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard.1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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1.1 This specification defines the physical and chemical requirements for zirconium oxide powder intended for fabrication into shapes, either entirely or partially of zirconia, for use in a nuclear reactor core.1.2 The material described herein shall be particulate in nature.1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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5.1 This test method is used to determine the U and Pu content of scrap and waste in containers. Active measurement times have typically been 100 to 1000 s. Passive measurement times have typically been 400 s to several hours. The following limits may be further restricted depending upon specific matrix, calibration material, criticality safety, or counting equipment considerations.5.1.1 The passive measurement has been applied to benign matrices in 208 L drums with Pu content ranging from 30 mg to 1 kg.5.1.2 The active measurement has been applied to waste drums with 235U content ranging from about 100 mg to 1 kg.5.2 This test method can be used to demonstrate compliance with the radioactivity levels specified in safeguards, waste, disposal, and environmental regulations (for example, see NRC regulatory guides 5.11, 5.53, DOE Order 5820.2a, and 10CFR61 sections 61.55 and sections 61.56, 40CFR191, and DOE/WIPP-069).5.3 This test method could be used to detect diversion attempts that use shielding to encapsulate nuclear material.5.4 The bias of the measurement results is related to the item size and density, the homogeneity and composition of the matrix, and the quantity and distribution of the nuclear material. The precision of the measurement results is related to the quantity of nuclear material and the count time of the measurement.5.4.1 For both the matrix-specific and the matrix-correction approaches, the method assumes the calibration materials match the items to be measured with respect to the homogeneity and composition of the matrix, the neutron moderator and absorber content, and the quantity of nuclear material, to the extent they affect the measurement.5.4.2 It is recommended that measurements be made on small containers of scrap and waste before they are combined in large containers. Special arrangement may be required to assay small containers to best effect in a large cavity general purpose shuffer.5.4.3 It is recommended that measurements be made on containers with homogeneous contents. In general, heterogeneity in the distribution of nuclear material, neutron moderators, and neutron absorbers has the potential to cause biased results.5.5 This test method requires that the relative isotopic compositions of the contributing elements are known.5.6 This test method assumes that the distribution of the contributing isotopes is uniform throughout the container when the matrix affects neutron transport.5.7 This test method assumes that lump affects are unimportant—that is to say that large quantities of special nuclear material are not concentrated in a small portion of the container.5.8 For best results from the application of this test method, appropriate packaging of the items is required. Suitable training of the personnel who package the scrap and waste prior to measurement should be provided (for example, see ANSI 15.20, Guide C1009, Guide C1490, and Guide C1068 for training guidance). Sometimes site specific conditions and requirements may have greater bearing.1.1 This test method covers the nondestructive assay of scrap and waste items for U, Pu, or both, using a 252 Cf shuffler. Shuffler measurements have been applied to a variety of matrix materials in containers of up to several 100 L. Corrections are made for the effects of matrix material. Applications of this test method include measurements for safeguards, accountability, TRU, and U waste segregation, disposal, and process control purposes (1, 2, 3).21.1.1 This test method uses passive neutron coincidence counting (4) to measure the 240Pu-effective mass. It has been used to assay items with total Pu contents between 0.03 g and 1000 g. It could be used to measure other spontaneously fissioning isotopes such as Cm and Cf. It specifically describes the approach used with shift register electronics; however, it can be adapted to other electronics.1.1.2 This test method uses neutron irradiation with a moveable Cf source and counting of the delayed neutrons from the induced fissions to measure the 235U equivalent fissile mass. It has been used to assay items with 235U contents between 0.1 g and 1000 g. It could be used to assay other fissile and fissionable isotopes.1.2 This test method requires knowledge of the relative isotopic composition (See Test Method C1030) of the special nuclear material to determine the mass of the different elements from the measurable quantities.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 The techniques described in this test method have been applied to materials other than scrap and waste. These other applications are not addressed in this test method.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 and health practices and determine the applicability of regulatory limitations prior to use. Specific precautionary statements are given in Section 8.

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This specification applies to pellets of stabilized zirconium oxide used in nuclear reactors. The chemical composition requirements such as the stabilizing additive (calcium oxide or yttrium oxide), analytical chemistry methods, impurity concentration (including hafnium, boron, gadolinium, samarium, europium, dysprosium, cobalt, silicon, iron, calcium, magnesium, aluminum, titanium, thorium, fluorine, chlorine, bromine, iodine, and hydrogen), and moisture concentration are prescribed. The nuclear grade pellets shall conform to the specified physical requirements which includes the following: physical dimensions, density, mechanical properties and test methods such as compressive test and thermal cycling test, and visual appearance such as end chips, circumferential chips, cracks, and fissures. The requirements for cleanliness before and after sampling and packaging are given.1.1 This specification applies to pellets of stabilized zirconium oxide used in nuclear reactors.1.2 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.1.3 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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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 threshold energy required to break the specific water-material bonds is applied to the system. For spent nuclear fuel this threshold energy may come from the combination of thermal input from decay heat, externally applied heat, or from the ionizing radiation itself.4.2.6 The adequacy of a drying procedure may be evaluated by measuring the response of the system after the drying operation is completed. For example, if a vacuum drying technology is used for water removal, a specific vacuum could be applied to the system, the vacuum pumps turned off, and the time dependence of pressure rebound measured. The rebound response could then be associated with the residual water, especially unbound water, in the system.4.2.7 Residual water associated 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 under dry storage conditions.4.2.8 Thermal gradients within the container evolve with time, and as a result water vapor will tend to migrate to the cooler portions of the package. Water may condense in these areas. Condensed water will tend to migrate to the physically lower positions under gravity such as the container bottom.4.2.9 Radiolytic decomposition of hydrated and other water-containing compounds may release moisture, oxygen and hydrogen to the container.4.2.10 Extended time at temperature, coupled with the presence of ionizing radiation, may provide the energy necessary to release bound or trapped water to the container.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 container and may hold water and resist drying.1.1.4 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 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.1.3 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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This specification applies to nuclear-grade aqueous uranyl nitrate solution or crystals not exceeding 5% 235U intended for subsequent manufacture into either UF6 or direct conversion to uranium oxide. This specification is intended to provide the nuclear industry with a general standard for aqueous uranyl nitrate solution or crystals. The purpose of this specification is: to define the impurity and uranium isotope limits for commercial standard uranyl nitrate, and to define additional limits for reprocessed uranyl nitrate (or any mixture of reprocessed and commercial standard uranyl nitrate).1.1 This specification applies to nuclear-grade aqueous uranyl nitrate solution or crystals not exceeding 5 % 235U intended for subsequent manufacture into either UF6 (for feed to an enrichment plant) or direct conversion to uranium oxide (for use in reactors).1.2 This specification is intended to provide the nuclear industry with a general standard for aqueous uranyl nitrate solution or crystals. It recognizes the diversity of manufacturing methods and the processes to which it is subsequently to be subjected. It is therefore anticipated that it may be necessary to include supplementary specification limits by agreement between purchaser and manufacturer. Different limits are appropriate depending on whether or not the uranyl nitrate is to be converted to UF6 for subsequent processing.1.3 The purpose of this specification is: (a) to define the impurity and uranium isotope limits for commercial standard uranyl nitrate, and (b) to define additional limits for reprocessed uranyl nitrate (or any mixture of reprocessed and commercial standard uranyl nitrate). For such uranyl nitrates, special provisions may need to be made to ensure that no extra hazard arises to the employees, the process equipment, or the environment.1.4 The scope of this specification does not comprehensively cover all provisions for preventing criticality accidents, for health and safety, or for shipping. Observance of this standard does not relieve the user of the obligation to conform to all international, federal, state and local regulations for processing, shipping, or any other way of using the uranyl nitrate. An example of a U.S. Government Document is the Code of Federal Regulations, Title 10, Part 50 (latest edition).1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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5.1 There are several methods for managing non-conforming coatings in an operating nuclear power plant. This guide outlines methods that have been determined to be acceptable to the nuclear industry.5.2 Managing the amount of non-conforming coatings is key to ensuring the amount assumed, in the licensing bases is not exceeded.5.3 EPRI Report 1019157 provides additional information on the selection, application, inspection and maintenance of nuclear plant safety-related protective coatings. This reference offers a detailed discussion of important considerations related to protective coatings and can be used to supplement this guide as deemed necessary.1.1 This guide provides the user with guidance on developing a program for managing non-conforming coatings in Coating Service Level I areas of a nuclear power plant.1.2 Non-conforming coatings include degraded qualified or acceptable coatings, unqualified coatings, unknown coatings, and unacceptable coatings.1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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4.1 These test methods are useful as rapid, nondestructive techniques for the in-place determination of the density of unhardened concrete. The backscatter test method is also useful for the same purpose on hardened concrete. The fundamental assumptions inherent in the test methods are that Compton scattering is the dominant interaction and that the material under test is homogeneous.4.2 These test methods are suitable for control and for assisting in acceptance testing during construction, for evaluation of concrete quality subsequent to construction, and for research and development.NOTE 1: Care must be taken when using these test methods in monitoring the degree of consolidation, which is the ratio of the actual density achieved to the maximum density attainable with a particular concrete. The test methods presented here are used to determine the actual density. A density measurement, by any test method, is a function of the components of the concrete and may vary, to some extent, in response to the normal, acceptable variability of those components.4.3 Test results may be affected by reinforcing steel, by the chemical composition of concrete constituents, and by sample heterogeneity. The variations resulting from these influences are minimized by instrument design and by the user's compliance with appropriate sections of the test procedure. Results of tests by the backscatter test method may also be affected by the density of underlying material. The backscatter test method exhibits spatial bias in that the apparatus's sensitivity to the material under it decreases with distance from the surface of the concrete.NOTE 2: Typically, backscatter gauge readings represent the density in the top 75 to 100 mm [3 to 4 in.] of material.1.1 These test methods cover the determination of the in-place density of unhardened and hardened concrete, including roller compacted concrete, by gamma radiation. For notes on the nuclear test see Appendix X1.1.2 Two test methods are described, as follows:  Section    Test Method A—Direct Transmission   Test Method B—Backscatter 891.3 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.

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1.1 This specification covers the classification, processing, and typical properties of as-manufactured nuclear grade graphite billets with dimensions sufficient to meet the designer’s requirements for fuel elements, moderator or reflector blocks, in a high temperature reactor. The graphite classes specified here may be suitable for reactor core applications where dimensional change due to fast neutron irradiation has a significant impact on design, provided they meet the requirements of the ASME code.1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. (See IEEE/ASTM SI 10.)1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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4.1 Boron carbide is used as a control material in nuclear reactors. In order to be suitable for this purpose, the material must meet certain criteria for assay, isotopic composition, and impurity content. These methods are designed to show whether or not a given material meets the specifications for these items as described in Specifications C750 and C751.4.1.1 An assay is performed to determine whether the material has the specified boron and carbon content.4.1.2 Determination of the isotopic content of the boron and the free carbon content is made to establish whether the content is in compliance with the purchaser’s specifications.4.1.3 Impurity content is determined to ensure that the maximum concentration limit of certain impurities (chloride, fluoride, water, metallic impurities, soluble boron) is not exceeded.1.1 These test methods cover procedures for the chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade boron carbide powder and pellets to determine compliance with specifications.1.2 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.1.3 The analytical procedures appear in the following order:  SectionsTotal Carbon by Combustion in an Inductive Furnace and    Infrared Measurement 8 – 17Total Boron by Titrimetry and ICP OES 18 – 28Isotopic Composition by Mass Spectrometry 29 – 33Pyrohydrolysis 34 – 41Chloride by Constant-Current Coulometry 42 – 50Chloride and Fluoride by Ion-Selective Electrode 51 – 59Water by Constant-Voltage Coulometry and Weight Loss on    Drying 60 – 63Metallic Impurities by DCArc OES and wet chemical methods 64 and 65Soluble Boron by Titrimetry and ICP OES 66 – 80Free Carbon by a Coulometric Method 81 – 901.4 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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4.1 This is one of a series of guides designed to provide guidance for implementing activities that meet the requirements of a sound laboratory quality assurance program. The first of these, Guide C1009, is an umbrella guide that provides general criteria for ensuring the quality of analytical laboratory data. Other guides provide expanded criteria in various areas affecting quality, producing a comprehensive set of criteria for controlling data quality. The approach to ensuring the quality of analytical measurements described in these guides is depicted in Fig. 1.FIG. 1 Quality Assurance of Analytical Laboratory Data4.2 The training and qualification of analysts is one of the elements of laboratory quality assurance presented in Guide C1009, which provides some general criteria regarding qualification. This guide expands on those criteria to provide more comprehensive guidance for qualifying analysts. As indicated in Guide C1009, the qualification process can vary in approach; this guide provides one such approach.4.3 This guide describes an approach to analyst qualification that is designed to be used in conjunction with a rigorous program for the qualification and control of the analytical measurement system. This requires an existing data base which defines the characteristics (precision and bias) of the system in routine use. The initial development of this data base is described in Guide C1068. The process described here is intended only to qualify analysts when such a data base exists and the method is in control.4.4 The qualification activities described in this guide assume that the analyst is already proficient in general laboratory operations. The training or other activities that developed this proficiency are not covered in this guide.4.5 This guide describes a basic approach and principles for the qualification of laboratory analysts. Users are cautioned to ensure that the qualification program implemented meets the needs and requirements of their laboratory.1.1 This guide covers the qualification of analysts to perform chemical analysis or physical measurements of nuclear fuel cycle materials. The guidance is general in that it is applicable to all analytical methods, but must be applied method by method. Also, the guidance is general in that it may be applied to initial qualification or requalification.1.2 The guidance is provided in the following sections:    Section  Qualification Considerations 5  Demonstration Process 6  Statistical Tests 71.3 This standard does not apply to maintaining qualification during routine use of a method. Maintaining qualification is included in Guide C1210.1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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5.1 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 hydrogen effects, fracture toughness of the cladding and the creep behavior are important parameters to be evaluated and controlled during the dry storage phase of the life cycle. These degradation mechanisms are discussed in Annex A2 and Annex A4.1.1 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 addresses the determination of the conditions of the spent fuel and storage cask materials at the end of the initial 20-year license period as the result of normal events and conditions. However, the guide also addresses the analysis of potential spent fuel and cask materials degradation as the result of off-normal, and accident-level events and conditions that may occur during any period.1.4 This guide provides information on materials behavior to support continuing compliance with the safety criteria, which are part of the regulatory basis, for licensed storage of SNF at an ISFSI. The safety functions addressed and discussed in this standard guide include thermal performance, radiological protection, confinement, sub-criticality, and retrievability. The regulatory basis includes 10 CFR Part 72 and supporting regulatory guides of the U.S. Nuclear Regulatory Commission. The requirements set forth in these documents indicate that the following items were considered in the original licensing decisions: properties of materials, design considerations for normal and off-normal service, operational and natural events, and the bases for the original calculations. These items may require reconsideration of the safety-related arguments that demonstrate how the systems continue to satisfy the regulatory requirements. Further, to ensure continued safe operation, the performance of materials must be justified in relation to the effects of time, temperature, radiation field, and environmental conditions of normal and off-normal service. Arguments for long-term performance must account for materials alterations (especially degradations) that are expected during the service periods, which include the periods of the initial license and of the license renewal. This guide pertains only to structures, systems, and components important to safety during extended storage period and during retrieval functions, including transport and transfer operations. Materials information that pertains to safety functions, including retrieval functions, is pertinent to current regulations and to license renewal process, and this information is the focus of the guide. This guide is not intended to supplant the existing regulatory process.1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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4.1 Uranyl nitrate solution is used as a feed material for conversion to the hexafluoride as well as for direct conversion to the oxide. In order to be suitable for this purpose, the material must meet certain criteria for uranium content, isotopic composition, acidity, radioactivity, and impurity content. These methods are designed to show whether a given material meets the specifications for these items described in Specification C788.4.1.1 An assay is performed to determine whether the material has the specified uranium content.4.1.2 Determination of the isotopic content of the uranium is made to establish whether the effective fissile content is in accordance with the purchaser’s specifications.4.1.3 Acidity, organic content, and alpha, beta, and gamma activity are measured to establish that they do not exceed their maximum limits.4.1.4 Impurity content is determined to ensure that the maximum concentration limit of certain impurity elements is not exceeded. Impurity concentrations are also required for calculation of the equivalent boron content (EBC), and the total equivalent boron content (TEBC).1.1 These test methods cover procedures for the chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade uranyl nitrate solution to determine compliance with specifications.1.2 The analytical procedures appear in the following order:  SectionsDetermination of Uranium 8Specific Gravity by Pycnometry  16 – 21Free Acid by Oxalate Complexation  22 – 28Determination of Thorium 29Determination of Chromium 30Determination of Molybdenum 31Halogens Separation by Steam Distillation  32 – 36Fluoride by Specific Ion Electrode  37 – 43Halogen Distillate Analysis: Chloride, Bromide, and Iodide by  Amperometric Microtitrimetry 44Determination of Chloride and Bromide 45Determination of Sulfur by X-Ray Fluorescence 46Sulfate Sulfur by (Photometric) Turbidimetry 47Phosphorus by the Molybdenum Blue (Photometric) Method  55 – 62Silicon by the Molybdenum Blue (Photometric) Method 63 – 70Carbon by Persulfate Oxidation-Acid Titrimetry 71Conversion to U3O8 72 – 75Boron by Emission Spectrography AImpurity Elements by Spark Source Mass Spectrography 77Isotopic Composition by Thermal Ionization Mass Spectrometry 78Uranium-232 by Alpha Spectrometry 79 – 85Total Alpha Activity by Direct Alpha Counting 86 – 92Fission Product Activity by Beta Counting 93 – 99Entrained Organic Matter by Infrared Spectrophotometry 100Fission Product Activity by Gamma Counting 101Determination of Arsenic 102Determination of Impurities for the EBC Calculation 103Determination of Technetium 99 104Determination of Plutonium and Neptunium 1051.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use. Specific precautionary statements are given in Section 6.1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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4.1 Because of concerns for safety and the protection of nuclear materials from theft, stringent specifications are placed on chemical processes and the chemical and physical properties of nuclear materials. Strict requirements for the control and accountability of nuclear materials are imposed on the users of those materials. Therefore, when analyses are made by a laboratory to support a project such as the fabrication of nuclear fuel materials, various performance requirements may be imposed on the laboratory. One such requirement is often the use of qualified methods. Their use gives greater assurance that the data produced will be satisfactory for the intended use of those data. A qualified method will help assure that the data produced will be comparable to data produced by the same qualified method in other laboratories.4.2 This guide provides guidance for qualifying measurement methods and for maintaining qualification. Even though all practices would be used for most qualification programs, there may be situations in which only a selected portion would be required. Care should be taken, however, that the effectiveness of qualification is not reduced when applying these practices selectively. The recommended practices in this guide are generic; based on these practices, specific actions should be developed to establish a qualification program.1.1 This guide provides guidance for selecting, validating, and qualifying measurement methods when qualification is required for a specific program. The recommended practices presented in this guide provide a major part of a quality assurance program for the laboratory data (see Fig. 1). Qualification helps to assure that the data produced will meet established requirements.1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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