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4.1 This guide presents concise guidance and approach to developing a test document for qualifying a coating for CSLI service, whether a new or existing coating. Guidance for evaluating existing qualification test data for applicability is presented in Guide D8104.4.2 The requirements for qualification testing can be found in Quality Assurance Criteria III (Design Control), IX (Control of Special Processes), and XI (Test Control) of 10 CFR 50, Appendix B, as implemented, respectively, by Requirements III, IX, and XI of NQA-1. A test document developed per this guide is intended to be compliant with these requirements.4.3 This guide implements the guidance provided in Guide D5144 for qualification of coatings for use in CSLI service. Additional guidance is provided in Regulatory Guide 1.54, Revisions 0 through 3, as may be invoked by the licensee.4.4 For plants with a license basis that predates the requirements of ANSI N5.12 and N101.2, this guide also is applicable. For these plants, the coatings or coating systems may be designated as acceptable, rather than qualified.4.5 All qualification testing shall comply with the licensee’s approved quality assurance program.1.1 This guide provides an approach to identifying the need for and development of a test document to qualify coatings for Coating Service Level I (CSLI) service in nuclear power plants.1.2 It is the intent of this guide to provide a recommended basis for establishing a coatings qualification test document, not to mandate a singular basis for all test documents. Variations or simplifications of the process described in this guide may be appropriate for any given operating or new construction nuclear power plant depending on its licensing commitments.1.3 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard.1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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5.1 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 barrier system (EBS) with respect to these regulatory constraints requires knowledge of the chemical/physical characteristics and degradation behavior of the SNF that could be provided by the testing and data evaluation methods provided by this guide, using the United States as an example, as follows:5.2.1 In the United States, for example, Code of Federal Regulations, Title 10, Part 60 Sections 135 and 113 require that the WF be a material that is solid, non-particulate, non-pyrophoric, and non-chemically reactive, that the waste package contain no liquid, particulates, or combustible materials and that the materials/components of the EBS be designed to provide—assuming anticipated processes and events—substantially complete containment of the HLW for the NRC-designated regulatory period.5.2.2 In the United States, for example, Code of Federal Regulations, Title 10, Part 63 Section 113 requires that the EBS be designed such that, working in combination with the natural barriers, the performance assessment of the EBS demonstrates conformance to the annual reasonably expected individual dose protection standard of Code of Federal Regulations, Title 10, Part 63 Section 311 and the reasonably maximally exposed individual standard of Code of Federal Regulations, Title 10, Part 63 Section 312, and shall not exceed EPA dose limits for protection of groundwater of Code of Federal Regulations, Title 10, Part 63 Section 331 during the NRC-designated regulatory compliance period after permanent closure.5.2.3 In the United States, for example, Code of Federal Regulations, Title 10, Part 63 Section 114 (e), (f), and (g) and Code of Federal Regulations, Title 10, Part 63 Section 115 (c) require that a technical basis be provided for the inclusion or exclusion of degradation/alteration processes pertinent to the barriers of the EBS, and that likewise a technical basis be provided for the degradation/alteration models used in the post-closure performance assessment of the capability of the EBS barriers to isolate waste.5.3 The enhanced chemical reactivity and degraded condition of corroded/damaged uranium metal-based SNF must be accounted for in both the pre-closure safety analyses and the post-closure performance assessment of the geologic repository. An example of this would be the potential for pyrophoric behavior in uranium metal-based SNF (see Guide C1454). Due to the combustibility of the metallic uranium or uranium hydride (or both), and the enhanced aqueous dissolution rates for the exposed uranium metal, the potential for enhanced chemical activity or pyrophoric behavior must be factored into the repository or interim storage facility safety analyses, and assessments of the potential for radionuclide releases from the repository site boundary after repository closure.5.4 Characterization of several key properties of SNF may be required to support the design and performance analyses of both repository above-ground SNF receipt and lag storage facilities, the WP into which the SNF is placed, and the subsurface permanent emplacement drift EBS.5.4.1 Repository waste package design must ensure that the waste to be placed in the repository can be accommodated within the radionuclide and thermal loading ranges of the waste package drift emplacement licensing conditions. To do this the radionuclide content and oxidation rate when exposed to oxygen/water environments should be determined.5.4.2 The condition of the LWR spent fuel cladding (particularly with respect to hydride content and morphology) could potentially influence the performance of the cladding in interim storage, transportation, and geologic repository disposal. The corrosion and consequent failure rate of cladding with high hydride content may be greater than that of low or no hydride content. If the performance assessment is found to be sensitive to the failure rate of the cladding, it may be necessary to perform zirconium hydride content and orientation testing, particularly for high burnup LWR SNF.5.4.3 Metallic uranium-based spent fuel introduces aspects of chemical reactivity, such as combustibility and pyrophoricity (see C1454), that should be addressed in WP design and performance assessment, and in safety analyses associated with interim storage and transportation prior to repository emplacement. Metallic uranium-based nuclear fuel has been widely used in nuclear reactors; sometimes for commercial reactors (for example, Magnox) but more often in plutonium and tritium production reactors. The manner of discharge of metallic uranium SNF from these production reactors, and/or the manner of temporary wet storage of that portion of the spent fuel that was not reprocessed has in many instances resulted in significant corrosion and mechanical damage to the SNF assemblies. This damage has resulted in the direct exposure of the metallic uranium to the basin water. The relatively high chemical reactivity of uranium in contact with water can result in significant physical damage to the assemblies as the result of corrosion product buildup, and the creation in the exposed fuel surface and fuel matrix of uranium hydride inclusions which in turn further increase the chemical activity of the material. The reaction of this spent fuel with air, water vapor, or liquid water can introduce a significant heat source term into design basis events. In order to support the evaluation of these events, the physical condition (that is, the degree of optically/visually observable damage), the chemical oxidation kinetics, the ignition characteristics, and radionuclide release characteristics of the SNF should be investigated.5.4.4 The thermal analysis of the waste package/engineered barrier system requires quantification of the potential chemical heat source. To determine this, the amount of reactive uranium metal in the waste canisters sent to the repository should be provided so the thermal analysis of the waste package/engineered barrier system can be performed.5.4.5 Radionuclide inventories and physical/chemical characteristics are required to enable storage canister, transportation package, and WP loading and emplacement configurations to be developed.5.4.6 Repository WP materials selection and design must account for the potential interactions between the waste and WP. The potential chemical forms of the wastes must be considered, and the effects of residual water or impurities (or both) should be evaluated.5.4.7 The history of the SNF interim storage and transportation conditions prior to delivery to the repository is important whenever the storage conditions may have altered the degradation characteristics of the SNF (for example, with respect to hydride content and morphology in high burnup LWR SNF cladding). Interim dry storage of commercial SNF requires that the fuel cladding should not sustain gross damage during the storage period to the extent that fuel is released from the fuel rods into the canister. Small pinholes or cracks may exist in the cladding during the storage period without violating this interim storage requirement, but may cause the fuel to be classified as failed fuel for repository disposal purposes. The objective of drying commercial SNF fuel is thus to preclude gross damage for interim storage purposes. If the conditions of transport or interim storage are such that there is a significant potential for further degradation of the SNF or change in properties important to the repository pre-closure safety or post-closure performance analyses, the characterization should provide sufficient information to evaluate these changes.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 fuel storage installation (ISFSI)), or both, and that the manner of interim storage may affect the SNF characteristics.1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.6 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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定价： 515元 / 折扣价： 438 元 加购物车

定价： 515元 / 折扣价： 438 元 加购物车

5.1 Use of this guide will ensure that the potential impact on the surrounding environment from planned decommissioning activities has been properly assessed.5.2 Use of this guide will ensure that the adequacy of environmental sampling has been assessed for location, frequency, analytical techniques, and media type to monitor the environment and to detect site-related releases and their impact.1.1 This guide covers the development or assessment of environmental monitoring plans for decommissioning nuclear facilities. This guide addresses: (1) development of an environmental baseline prior to commencement of decommissioning activities; (2) determination of release paths from site activities and their associated exposure pathways in the environment; and (3) selection of appropriate sampling locations and media to ensure that all exposure pathways in the environment are monitored appropriately. This guide also addresses the interfaces between the environmental monitoring plan and other planning documents for site decommissioning, such as radiation protection, site characterization, and waste management plans, and federal, state, and local environmental protection laws and guidance. This guide is applicable up to the point of completing D&D activities and the reuse of the facility or area for other purposes.1.2 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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3.1 Reactor vessels made of ferritic steels are designed with the expectation of progressive changes in material properties resulting from in-service neutron exposure. In the operation of light-water-cooled nuclear power reactors, changes in pressure-temperature (P – T) limits are made periodically during service life to account for the effects of neutron radiation on the ductile-to-brittle transition temperature material properties. If the degree of neutron embrittlement becomes large, the restrictions on operation during normal heat-up and cool down may become severe. Additional consideration should be given to postulated events, such as pressurized thermal shock (PTS). A reduction in the upper shelf toughness also occurs from neutron exposure, and this decrease may reduce the margin of safety against ductile fracture. When it appears that these situations could develop, certain alternatives are available that reduce the problem or postpone the time at which plant restrictions must be considered. One of these alternatives is to thermally anneal the reactor vessel beltline region, that is, to heat the beltline region to a temperature sufficiently above the normal operating temperature to recover a significant portion of the original fracture toughness and other material properties that were degraded as a result of neutron embrittlement.3.2 Preparation and planning for an in-service anneal should begin early so that pertinent information can be obtained to guide the annealing operation. Sufficient time should be allocated to evaluate the expected benefits in operating life to be gained by annealing; to evaluate the annealing method to be employed; to perform the necessary system studies and stress evaluations; to evaluate the expected annealing recovery and reembrittlement behavior; to develop and functionally test such equipment as may be required to do the in-service annealing; and, to train personnel to perform the anneal.3.3 Selection of the annealing temperature requires a balance of opposing conditions. Higher annealing temperatures, and longer annealing times, can produce greater recovery of fracture toughness and other material properties and thereby increase the post-anneal lifetime. The annealing temperature also can have an impact on the reembrittlement trend after the anneal. On the other hand, higher temperatures can create other undesirable property effects such as permanent creep deformation or temper embrittlement. These higher temperatures also can cause engineering difficulties, that is, core and coolant removal and storage, localized heating effects, etc., in preventing the annealing operation from distorting the vessel or damaging vessel supports, primary coolant piping, adjacent concrete, insulation, etc. See ASME Code Case N-557 for further guidance on annealing conditions and thermal-stress evaluations (2).3.3.1 When a reactor vessel approaches a state of embrittlement such that annealing is considered, the major criterion is the number of years of additional service life that annealing of the vessel will provide. Two pieces of information are needed to answer the question: the post-anneal adjusted RTNDT and upper shelf energy level, and their subsequent changes during future irradiation. Furthermore, if a vessel is annealed, the same information is needed as the basis for establishing pressure-temperature limits for the period immediately following the anneal and demonstrating compliance with other design requirements and the PTS screening criteria. The effects on upper shelf toughness similarly must be addressed. This guide primarily addresses RTNDT changes. Handling of the upper shelf is possible using a similar approach as indicated in NRC Regulatory Guide 1.162. Appendix X1 provides a bibliography of existing literature for estimating annealing recovery and reembrittlement trends for these quantities as related to U.S. and other country pressure-vessel steels, with primary emphasis on U.S. steels.3.3.2 A key source of test material for determining the post-anneal RTNDT, upper shelf energy level, and the reembrittlement trend is the original surveillance program, provided it represents the critical materials in the reactor vessel.6 Appendix X2 describes an approach to estimate changes in RTNDT both due to the anneal and reirradiation. The first purpose of Appendix X2 is to suggest ways to use available materials most efficiently to determine the post-anneal RTNDT and to predict the reembrittlement trend, yet leave sufficient material for surveillance of the actual reembrittlement for the remaining service life. The second purpose is to describe alternative analysis approaches to be used to assess test results of archive (or representative) materials to obtain the essential post-anneal and reirradiation RTNDT, upper shelf energy level, or fracture toughness, or a combination thereof.3.3.3 An evaluation must be conducted of the engineering problems posed by annealing at the highest practical temperature. Factors required to be investigated to reduce the risk of distortion and damage caused by mechanical and thermal stresses at elevated temperatures to relevant system components, structures, and control instrumentation are described in 5.1.3 and 5.1.4.3.4 Throughout the annealing operation, accurate measurement of the annealing temperature at key defined locations must be made and recorded for later engineering evaluation.3.5 After the annealing operation has been carried out, several steps should be taken. The predicted improvement in fracture toughness properties must be verified, and it must be demonstrated that there is no damage to key components and structures.3.6 Further action may be required to demonstrate that reactor vessel integrity is maintained within ASME Code requirements such as indicated in the referenced ASME Code Case N-557 (2). Such action is beyond the scope of this guide.AbstractThis guide covers the general procedures to be considered (as well as the demonstration of their effectiveness) for conducting in-service thermal anneals of light-water moderated nuclear reactor vessels. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods, as well as to provide direction for the development of a vessel annealing procedure and a post-annealing vessel radiation surveillance program. Guidelines are provided to determine the post-anneal reference nil-ductility transition temperature (RTNDT), the Charpy V-notch upper shelf energy level, fracture toughness properties, and the predicted reembrittlement trend for these properties for reactor vessel beltline materials.1.1 This guide covers the general procedures for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1).21.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constraints resulting from attached piping, support structures, and the primary system shielding; the mechanical and thermal stresses in the components and the system as a whole; and, material condition changes that may limit the annealing temperature.1.3 This guide provides direction for development of the vessel annealing procedure and a post-annealing vessel radiation surveillance program. The development of a surveillance program to monitor the effects of subsequent irradiation of the annealed-vessel beltline materials should be based on the requirements and guidance described in Practices E185 and E2215. The primary factors to be considered in developing an effective annealing program include the determination of the feasibility of annealing the specific reactor vessel; the availability of the required information on vessel mechanical and fracture properties prior to annealing; evaluation of the particular vessel materials, design, and operation to determine the annealing time and temperature; and, the procedure to be used for verification of the degree of recovery and the trend for reembrittlement. Guidelines are provided to determine the post-anneal reference nil-ductility transition temperature (RTNDT), the Charpy V-notch upper shelf energy level, fracture toughness properties, and the predicted reembrittlement trend for these properties for reactor vessel beltline materials. This guide emphasizes the need to plan well ahead in anticipation of annealing if an optimum amount of post-anneal reembrittlement data is to be available for use in assessing the ability of a nuclear reactor vessel to operate for the duration of its present license, or qualify for a license extension, or both.1.4 The values stated in either SI units or inch-pound units are to be regarded separately as standard. The values stated in each system are not necessarily exact equivalents; therefore, to ensure conformance with the standard, each system shall be used independently of the other, and values from the two systems shall not be combined.1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.6 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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4.1 A laboratory quality assurance program is an essential program for laboratories within the nuclear industry. Guide C1009 provides guidance for establishing a quality assurance program for an analytical laboratory within the nuclear industry. This guide deals with the control of measurements aspect of the laboratory quality assurance program. Fig. 1 shows the relationship of measurement control with other essential aspects of a laboratory quality assurance program.FIG. 1 Quality Assurance of Analytical Laboratory Data4.2 The fundamental purposes of a measurement control program are to provide the with-use assurance (real-time control) that a measurement system is performing satisfactorily and to provide the data necessary to quantify measurement system performance. The with-use assurance is usually provided through the satisfactory analysis of quality control samples (reference value either known or unknown to the analyst). The data necessary to quantify measurement system performance is usually provided through the analysis of quality control samples or the duplicate analysis of process samples, or both. In addition to the analyses of quality control samples, the laboratory quality control program should address (1) the preparation and verification of standards and reagents, (2) data analysis procedures and documentation, (3) calibration and calibration procedures, (4) measurement method qualification, (5) analyst qualification, and (6) other general program considerations. Other elements of laboratory quality assurance also impact the laboratory quality control program. These elements or requirements include (1) chemical analysis procedures and procedure control, (2) records storage and retrieval requirements, (3) internal audit requirements, (4) organizational considerations, and (5) training/qualification requirements. To the extent possible, this standard will deal primarily with quality control requirements rather than overall quality assurance requirements, which are addressed in Guide C1009.4.3 Although this guide uses suggestive rather than prescriptive language (for example, “should” as opposed to “shall”), the elements being addressed should not be interpreted as optional. An effective and comprehensive laboratory quality control program should, at minimum, completely and adequately consider and include all elements listed in Section 1 and in the corresponding referenced sections of this guide.1.1 This guide provides guidance for establishing and maintaining a measurement system quality control program. Guidance is provided for general program considerations, preparation of quality control samples, analysis of quality control samples, quality control data analysis, analyst qualification, measurement system calibration, measurement method qualification, and measurement system maintenance.1.2 This guidance is provided in the following sections:  SectionGeneral Quality Control Program Considerations 5Quality Control Samples 6Analysis of Quality Control Samples 7Quality Control Data Analysis 8Analyst Qualification 9Measurement System Calibration 10Qualification of Measurement Methods and Systems 11Measurement System Maintenance 121.3 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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定价： 646元 / 折扣价： 550 元 加购物车

Plans for sampling and analysis of nuclear material are designed with two purposes in mind: the first is related to material accountability and the second to material specifications. For the accounting of special nuclear material, sampling and analysis plans should be established to determine the quantity of special nuclear material held in inventory, shipped between buyers and sellers, or discarded. Likewise, material specification requires the determination of the quantity of nuclear material present. Inevitably there is uncertainty associated with such measurements. This practice presents a tool for developing sampling plans that control the random error component of this uncertainty. Precision and accuracy statements are highly desirable, if not required, to qualify measurement methods. This practice relates to“ precision” that is generally a statement on the random error component of uncertainty. 1.1 This practice provides an aid in designing a sampling and analysis plan for the purpose of minimizing random error in the measurement of the amount of nuclear material in a lot consisting of several containers. The problem addressed is the selection of the number of containers to be sampled, the number of samples to be taken from each sampled container, and the number of aliquot analyses to be performed on each sample. 1.2 This practice provides examples for application as well as the necessary development for understanding the statistics involved. The uniqueness of most situations does not allow presentation of step-by-step procedures for designing sampling plans. It is recommended that a statistician experienced in materials sampling be consulted when developing such plans. 1.3 The values stated in SI units are to be regarded as the standard. 1.4 This standard does not purport to address all of the safety problems, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

定价： 0元 / 折扣价： 0 元

定价： 590元 / 折扣价： 502 元 加购物车

定价： 590元 / 折扣价： 502 元 加购物车

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定价： 590元 / 折扣价： 502 元 加购物车

4.1 The practice contained herein can be used as a basis for establishing conditions for the safe operation of critical structural components. The practices provide for general plant assessment and verification that materials continue meet design criteria and may in addition be of use for asset protection or life extension. The test specimens and procedures presented in this practice are for guidance when establishing a surveillance program.4.2 This practice for high-temperature materials surveillance programs is used when nuclear reactor component materials are monitored by specimen testing. Periodic testing is performed through the service life of the components to assess changes in selected material properties that are caused by neutron irradiation, thermal effects, chemical reactions, and mechanical stress. The properties of interest are those used as design criteria for the respective nuclear components or well correlated to said criteria (see 5.1.6). The need for surveillance arises from the need to assess predictions of aging material performance to ensure adequate component performance.4.3 This practice describes specimens and procedures required for the surveillance of multiple components. A surveillance program for a particular component will not necessarily require all test types described herein.1.1 This practice covers procedures for surveillance program design and specimen testing to establish changes occurring in the mechanical properties of ferrous and nickel-based materials due to irradiation and thermal effects of nuclear component metallic materials used for high-temperature structural applications above 370 °C (700 °F). This should include consideration of gamma heating. This practice currently only applies to an initial program based on initial estimates of design life of components.1.2 This practice was developed for non-light-water moderated nuclear power reactors.1.3 This practice does not provide specific procedures for extending surveillance programs beyond their original design lifetimes.1.4 This practice does not consider in-situ monitoring techniques but may provide insights into the proper periodicity and design of such.1.5 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.1.6 This standard does not purport to address all of the safety 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.

定价： 590元 / 折扣价： 502 元 加购物车