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4.1 This guide is intended for use by those undertaking the development of fire hazard assessment standards for electrotechnical products. Such standards are expected to be useful to manufacturers, architects, specification writers, and authorities having jurisdiction.4.2 As a guide, this document provides information on an approach to the development of a fire hazard assessment standard; fixed procedures are not established. Any limitations in the availability of data, of appropriate test procedures, of adequate fire models, or in the advancement of scientific knowledge will place significant constraints upon the procedure for the assessment of fire hazard.4.3 The focus of this guide is on fire assessment standards for electrotechnical products. However, insofar as the concepts in this guide are consistent with those of Guide E1546, the general concepts presented also may be applicable to processes, activities, occupancies, and buildings. Guide E2061 contains an example of how to use information on fire-test-response characteristics of electrotechnical products (electric cables) in a fire hazard assessment for a specific occupancy (rail transportation vehicle).4.4 A standard developed following this guide should not attempt to set a safety threshold or other pass/fail criteria. Such a standard should specify all steps required to determine fire hazard measures for which safety thresholds or pass/fail criteria can be meaningfully set by authorities having jurisdiction.1.1 This guide provides guidance on the development of fire hazard assessment standards for electrotechnical products. For the purposes of this guide, products include materials, components, and end-use products.1.2 This guide is directed toward development of standards that will provide procedures for assessing fire hazards harmful to people, animals, or property.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 fire standard cannot be used to provide quantitative measures.1.5 This standard is used to predict or provide a quantitative measure of the fire hazard from a specified set of fire conditions involving specific materials, products, or assemblies. This assessment does not necessarily predict the hazard of actual fires which involve conditions other than those assumed in the analysis.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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This PDF includes Update #2 1 Scope O437.0-1.1 This Standard pertains specifically to a class of mat-formed structural panelboards made predominantly of wood strands or wafers of a minimum and controlled length, a controlled thickness, and a variab

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5.1 This guide provides a list of the standards within Committee D04 that address the use of materials, specifications, and construction practices that could have broader sustainability benefits. This list is current, relative to the approval date of the standard.5.2 The standards discussed are listed in the Referenced Documents section.5.3 This guide is intended to be used as a reference for an owner, engineer, contractor, or combinations thereof, to identify potential sustainability strategies and the respective material and construction standards and specifications. It is important to note that these standards do not ensure sustainability goals are achieved; rather, they may be useful in determining inputs for sustainability metrics.1.1 This guide is intended to be a reference for locating specific test methods relating to materials and construction standards within the jurisdiction of Committee D04 on Road and Paving Materials that could be a strategy used to meet project sustainability goals.1.2 The guide needs to be reviewed and updated by Subcommittee D04.99 on Sustainable Asphalt Pavement Materials and Construction, on an as-needed basis, to remain viable.1.2.1 Additions or deletions to the reference list in Section 2 shall be submitted to Subcommittee D04.99 and balloted.1.3 Units—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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4.1 Master Matrix—This matrix document is written as a reference and guide to the use of existing standards and to help manage the development and application of new standards, as needed for LWR-PV surveillance programs. Paragraphs 4.2 – 4.5 are provided to assist the authors and users involved in the preparation, revision, and application of these standards (see Section 6).4.2 Approach and Primary Objectives: 4.2.1 Standardized procedures and reference data are recommended in regard to (1) neutron and gamma dosimetry, (2) physics (neutronics and gamma effects), and (3) metallurgical damage correlation methods and data, associated with the analysis, interpretation, and use of nuclear reactor test and surveillance results.4.2.2 Existing state-of-the-art practices associated with (1), (2), and (3), if uniformly and consistently applied, can provide reliable (10 to 30 %, 1σ) estimates of changes in LWR-PV steel fracture toughness during a reactor’s service life (36).4.2.3 Reg. Guide 1.99 and Section III of the ASME Boiler and Pressure Vessel Code, Part NF2121 require that the materials used in reactor pressure vessels support “...shall be made of materials that are not injuriously affected by ... irradiation conditions to which the item will be subjected.”4.2.4 By the use of this series of standards and the uniform and consistent documentation and reporting of estimated changes in LWR-PV steel fracture toughness with uncertainties of 10 to 30 % (1σ), the nuclear industry and licensing and regulatory agencies can meet realistic LWR power plant operating conditions and limits, such as those defined in Appendices G and H of 10 CFR Part 50, Reg. Guide 1.99, and the ASME Boiler and Pressure Vessel Code.4.2.5 The uniform and consistent application of this series of standards allows the nuclear industry and licensing and regulatory agencies to properly administer their responsibilities in regard to the toughness of LWR power reactor materials to meet requirements of Appendices G and H of 10 CFR Part 50, Reg. Guide 1.99, and the ASME Boiler and Pressure Vessel Code.4.3 Dosimetry Analysis and Interpretation (1, 3-5, 21, 28, 29, 35, 37, 38)—When properly implemented, validated, and calibrated by vendor/utility groups, state-of-the-art dosimetry practices exist that are adequate for existing and future LWR power plant surveillance programs. The uncertainties and errors associated with the individual and combined effects of the different variables (items 1.4.1 – 1.4.10 of 1.4) are considered in this section and in 4.4 and 4.5. In these sections, the accuracy (uncertainty and error) statements that are made are quantitative and representative of state-of-the-art technology. Their correctness and use for making EOL predictions for any given LWR power plant, however, are dependent on such factors as (1) the existing plant surveillance program, (2) the plant geometrical configuration, and (3) available surveillance results from similar plants. As emphasized in Section III-A of Ref (7), however, these effects are not unique and are dependent on (1) the surveillance capsule design, (2) the configuration of the reactor core and internals, and (3) the location of the surveillance capsule within the reactor geometry. Further, the statement that a result could be in error is dependent on how the neutron and gamma ray fields are estimated for a given reactor power plant (1, 11, 28, 36, 39, 40). For most of the error statements in 4.3 – 4.5, it is assumed that these estimates are based on reactor transport theory calculations that have been normalized to the core power history (see 4.4.1.2) and not to surveillance capsule dosimetry results. The 4.3 – 4.5 accuracy statements, consequently, are intended for use in helping the standards writer and user to determine the relative importance of the different variables in regard to the application of the set of ASTM standards, Fig. 1, for (1) LWR-PV surveillance program, (2) as instruments of licensing and regulation, and (3) for establishing improved metallurgical databases.4.3.1 Required Accuracies and Benchmark Field Referencing: 4.3.1.1 The accuracies (uncertainties and errors) (Note 1) desirable for LWR-PV steel exposure definition are of the order of ±10 to 15 % (1σ) while exposure accuracies in establishing trend curves should preferably not exceed ±10 % (1σ) (1, 11, 21, 36, 40-46). In order to achieve such goals, the response of neutron dosimeters should therefore also be interpretable to accuracies within ±10 to 15 % (1σ) in terms of exposure units and be measurable to within ±5 % (1σ).NOTE 1: Uncertainty in the sense treated here is a scientific characterization of the reliability of a measurement result and its statement is the necessary premise for using these results for applied investigations claiming high or at least stated accuracy. The term error will be reserved to denote a known deviation of the result from the quantity to be measured. Errors are usually taken into account by corrections.4.3.1.2 Dosimetry “inventories” should be established in support of the above for use by vendor/utility groups and research and development organizations.4.3.1.3 Benchmark field referencing of research and utilities’ vendor/service laboratories has been completed that is:(1) Needed for quality control and certification of current and improved dosimetry practices; and(2) Extensively applied in standard and reference neutron fields, PCA, PSF, SDMF, VENUS, NESDIP, PWRs, BWRs (1), and a number of test reactors to quantify and minimize uncertainties and errors.4.3.2 Status of Benchmark Field Referencing Work for Dosimetry Detectors—PCA, VENUS, NESDIP experiments with and without simulated surveillance capsules and power reactor tests have provided data for studying the effect of deficiencies in analysis and interpretations; the PCA/PSF/SDMF perturbation experiments have provided data for more realistic PWR and BWR power plant surveillance capsule configurations and have permitted utilities’ vendor/service laboratories to test, validate, calibrate, and update their practices (1, 4, 5, 47). The PSF surveillance capsule test provided data, but of a more one-dimensional nature. PCA, VENUS, and NESDIP experimentation together with some test reactor work augmented the benchmark field quantification of these effects (1, 3, 4, 28, 36, 48-51).4.3.3 Additional Validation Work for Dosimetry Detectors: 4.3.3.1 Establishment of nuclear data, photo-reaction cross sections, and neutron damage reference files.4.3.3.2 Establishment of proper quality assurance procedures for sensor set designs and individual detectors.4.3.3.3 Interlaboratory comparisons using standard and reference neutron fields and other test reactors that provide adequate validations and calibrations, see Guide E2005.4.4 Reactor Physics Analysis and Interpretation (1, 3, 5, 11, 28, 35, 36, 39, 52)—When properly implemented, validated, and calibrated by vendor/utility groups, state-of-the-art reactor physics practices exist that are adequate for in- and ex-vessel estimates of PV-steel changes in fracture toughness for existing and future power plant surveillance programs.4.4.1 Required Accuracies and Benchmark Field Referencing: 4.4.1.1 The accuracies desirable for LWR-PV steel (surveillance capsule specimens and vessels) exposure definition are of the order of ±10 to 15 % (1σ). Under ideal conditions benchmarking computational techniques are capable of predicting absolute in- and ex-vessel neutron exposures and reaction rates per unit reactor core power to within ±15 % (but generally not to within ±5 %). The accuracy will be worse, however, in applications to actual power plants because of geometrical and other complexities (1, 3, 4, 11, 21, 36-39, 52).4.4.1.2 Calculated in-and ex-vessel neutron and gamma ray fields can be normalized to the core power history or to experimental measurements. The latter may include dosimetry from surveillance capsules, other in-vessel locations, or ex-vessel measurements in the cavity outside the vessel. In each case, the uncertainty arising from the calculation needs to be considered.4.4.2 Power Plant Reactor Physics Analysis and Interpretation: 4.4.2.1 Result of Neglect of Benchmarking—One quarter thickness location (1/4T) vessel wall estimates of damage exposure are not easily compared with experimental results. “Lead factors,” based on the different ways they can be calculated (fluence >0.1 or >1.0 MeV and dpa) may not always be conservative; that is, some surveillance capsules have been positioned in-vessel such that the actual lead factor is very near unity—no lead at all. Also the differences between lead factors based on fluence E > 0.1 or > 1 MeV and dpa can be significant, perhaps 50 % or more (1, 11, 21, 28, 36-38, 52).4.4.3 PCA, VENUS, and NESDIP Experiments and PCA Blind Test: 4.4.3.1 Test of transport theory methods under clean geometry and clean core source conditions shall be made (1, 4, 11, 52).4.4.3.2 This is a necessary but not sufficient benchmark test of the adequacy of a vendor/utility group’s power reactor physics computational tools.4.4.3.3 The standard recommendation should be that the vendor/utility group’s observed differences between their own calculated and the PCA, VENUS, and NESDIP measured integral and differential exposure and reaction rate parameters be used to validate and improve their calculational tools (if the differences fall outside the PCA, VENUS, and NESDIP experimental accuracy limits).4.4.4 PWR and BWR Generic Power Reactor Tests: 4.4.4.1 Test of transport theory methods under actual geometry and variable core source conditions (1, 3, 4, 28, 35, 36, 53).4.4.4.2 This is a necessary and partly sufficient benchmark test of the adequacy of a vendor/utility group’s power reactor physics computational tools.4.4.4.3 The standard recommendation should be that the vendor/utility group’s observed differences between their own calculated and the selected PWR or BWR measured integral and differential exposure and reaction rate parameters be used to validate and improve their calculation tools (if the differences fall outside of the selected PWR or BWR experimental accuracy limits).4.4.4.4 The power reactor “benchmarks” that have been established for this purpose are identified and discussed in Refs (1, 3, 4, 35, 53) and their references and in Guide E2006.4.4.5 Operating Power Reactor Tests: 4.4.5.1 This is a necessary test of transport theory methods under actual geometry and variable core source conditions, but for a particular type or class of vendor/utility group power reactors. Here, actual in-vessel surveillance capsule and any required ex-vessel measured dosimetry information will be utilized as in 4.4.4 (1, 3, 4, 28, 35, 36, 53). Note, however, that operating power reactor tests are not sufficient by themselves (Reg. Guide 1.190, Section 4.4.5.1).4.4.5.2 Accuracies associated with surveillance program reported values of exposures and reaction rates are expected to be in the 10 to 30 % (1σ) range (36).4.5 Metallurgical Damage Correlation Analysis and Interpretation (1-8, 10, 11, 13, 15-29, 36-38)—When properly implemented, validated, and calibrated by vendor/utility groups, state-of-the-art metallurgical damage correlation practices exist that are adequate for in- and ex-vessel estimates of PV-steel changes in fracture toughness for existing and future power plant surveillance programs.4.5.1 Required Accuracies and Benchmark Field Referencing: 4.5.1.1 The accuracies desirable and achievable for LWR-PV steel (test reactor specimens, surveillance capsule specimens, and vessels and support structure) data correlation and data extrapolation (to predict fracture toughness changes both in space and time) are of the order of ±10 to 30 % (1σ). In order to achieve such a goal, however, the metallurgical parameters (ΔNDTT, upper shelf, yield strength, etc.) must be interpretable to well within ±20 to 30 % (1σ). This mandates that in addition to the dosimetry and physics variables already discussed that the individual uncertainties and errors associated with a number of other variables (neutron dose rate, neutron spectrum, irradiation temperature, steel chemical composition, and microstructure) must be minimized and results must be interpretable to within the same ±10 to 30 % (1σ) range.4.5.1.2 Advanced sensor sets (including dosimetry, temperature and damage correlation sensors) and practices have been established in support of the above for use by vendor/utility groups (1, 4, 5, 11, 39, 50, 54, 55).4.5.1.3 Benchmark field referencing of utilities' vendor/service laboratories, as well as advanced practices, is in progress or being planned that is (1, 3-6, 28, 50, 54-56):(1) Needed for validation of data correlation procedures and time and space extrapolations (to PV positions: surface, 1/4 T, etc.) of test reactor and power reactor surveillance capsule metallurgical and neutron exposure data.(2) Being or will be tested in test reactor neutron fields to quantify and minimize uncertainties and errors (included here is the use of damage correlation materials—steel, sapphire, etc.).4.5.2 Benchmark Field Referencing—The PSF (all positions: surveillance, surface, 1/4T, 1/2T, and the void box) together with the Melusine PV-simulator and other tests, such as for thermal neutron effects, provide needed validation data on all variables—dosimetry, physics, and metallurgy (1, 4, 10, 19, 21, 22, 37, 38). Other test reactor data comes from surveillance capsule results that have been benchmarked by vendor/service laboratory/utility groups (1, 3, 4, 6, 11, 18, 27, 28, 36, 40-44, 47).4.5.3 Reg. Guide 1.99, NRC, EPRI Databases—NRC and Electric Power Research Institute (EPRI) databases have been studied on an ongoing basis by ASTM Subcommittees E10.02 and E10.05, vendors, utilities, EPRI, and NRC contractors to establish improved databases for existing test and power reactor measured property change data. ASTM task groups recommend the use of updated and new exposure units (fluence total >0.1, >1.0 MeV, and dpa) for the NRC and EPRI databases (1, 2, 6, 7, 13, 18, 27, 36, 40-44, 47), and incorporate these recommendations in the appropriate standards. ASTM subcommittee E10.02 has updated the embrittlement database and the prediction model in E900–15. The exposure unit used is total fluence for E > 1 MeV. The basis of the prediction model is documented in an adjunct associated with E900, available from ASTM.4 The success of this effort depends on good cooperation between research and individual service laboratories and vendor/utility groups. An ASTM dosimetry cross section file based on the latest evaluations, as detailed in Guide E1018, and incorporating corrections for all known variables (perturbations, photo-reactions, spectrum, burn-in, yields, fluence time history, etc.) will be used as required and justified. Test reactor data will be addressed in a similar manner, as appropriate.1.1 This master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel (PV) and support structure steels throughout a pressure vessel’s service life (Fig. 1). Referenced documents are listed in Section 2. The summary information that is provided in Sections 3 and 4 is essential for establishing proper understanding and communications between the writers and users of this set of matrix standards. It was extracted from the referenced standards (Section 2) and references for use by individual writers and users. More detailed writers’ and users’ information, justification, and specific requirements for the individual practices, guides, and methods are provided in Sections 3 – 5. General requirements of content and consistency are discussed in Section 6.FIG. 1 Organization and Use of ASTM Standards in the E706 Master Matrix1.2 This master matrix is intended as a reference and guide to the preparation, revision, and use of standards in the series.1.3 To account for neutron radiation damage in setting pressure-temperature limits and making fracture analyses ((1-12)2 and Guide E509), neutron-induced changes in reactor pressure vessel steel fracture toughness must be predicted, then checked by extrapolation of surveillance program data during a vessel’s service life. Uncertainties in the predicting methodology can be significant. Techniques, variables, and uncertainties associated with the physical measurements of PV and support structure steel property changes are not considered in this master matrix, but elsewhere ((2, 6, 7, 11-26) and Guide E509).1.4 The techniques, variables, and uncertainties related to (1) neutron and gamma dosimetry, (2) physics (neutronics and gamma effects), and (3) metallurgical damage correlation procedures and data are addressed in separate standards belonging to this master matrix (1, 17). The main variables of concern to (1), (2), and (3) are as follows:1.4.1 Steel chemical composition and microstructure,1.4.2 Steel irradiation temperature,1.4.3 Power plant configurations and dimensions, from the core periphery to surveillance positions and into the vessel and cavity walls,1.4.4 Core power distribution,1.4.5 Reactor operating history,1.4.6 Reactor physics computations,1.4.7 Selection of neutron exposure units,1.4.8 Dosimetry measurements,1.4.9 Neutron special effects, and1.4.10 Neutron dose rate effects.1.5 A number of methods and standards exist for ensuring the adequacy of fracture control of reactor pressure vessel belt lines under normal and accident loads ((1, 7, 8, 11, 12, 14, 16, 17, 23-27), Referenced Documents: ASTM Standards (2.1), Nuclear Regulatory Documents (2.3) and ASME Standards (2.4)). As older LWR pressure vessels become more highly irradiated, the predictive capability for changes in toughness must improve. Since during a vessel's service life an increasing amount of information will be available from test reactor and power reactor surveillance programs, procedures to evaluate and use this information must be used (1, 2, 4-9, 11, 12, 23-26, 28). This master matrix defines the current (1) scope, (2) areas of application, and (3) general grouping for the series of ASTM standards, as shown in Fig. 1.1.6 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.1.7 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.8 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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3.1 The purpose of this standard practice is to provide the minimum requirements for the conduct of compliance audits.3.2 The intended use of standard is to provide a basis for an internal or external entity to develop an audit program. An audit program defines specific requirements for the execution of audits for a particular objective. An example of an audit program would be an external (third party) audit of LSA manufacturer’s quality assurance system.3.3 Compliance to this standard would insure that audit programs and those who develop and execute them are following a consensus set of minimum requirements.3.4 This standard does not mandate either internal or external audits.3.5 An auditing entity cannot request or approve an audit.3.6 Other Audit Criteria—Other audit criteria may be included in the audit scope if specified in the audit plan. Examples include safety, technical, operational, and management requirements. Items that are outside the scope of auditable criteria may be submitted as observations for possible resolution. However these are not binding and are not mandatory.3.7 Additional Services—Additional services are outside the scope of an audit objective. Examples of such services are consultation to resolve negative or open findings or any other service where the auditing entity conducts an activity other than an audit for the audited entity.3.8 Compliance Assurance—An audit is only an indicator of the compliance health of the facility and/or organization during only the period under review and therefore has limited compliance assurance and is not assumed to be exhaustive.3.9 Level of Review is Variable—The audit scope may vary to meet different audit objectives. For example, the audit scope may include only selected audit criteria, selected period under review, or selected portions of a facility or organization.1.1 This standard practice establishes the minimum set of requirements for auditing programs, methods, and systems, the responsibilities for all parties involved, and qualifications for entities conducting audits against ASTM standards on Light Sport Aircraft.1.2 This standard provides requirements to enable consistent and structured examination of objective evidence for compliance that is beneficial for the LSA industry and its consumers. It is the intent of this standard to provide the necessary minimum requirements for organizations to develop audit programs and procedures.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.

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Preface This is the second edition of the CSA N285.6 Series, Material standards for reactor components for CANDU nuclear power plants. It supersedes the first edition published in 1988. This Standard is written in SI (metric) units. N285.6.1-05 Pr

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This guide has been prepared to aid in the writing of material standards using the Classification D 4000 format.1.1 This classification system covers ___ materials suitable for _____________. The inclusion or exclusion of recycled plastics in this classification system must be addressed here.1.2 The properties included in this standard are those required to identify the compositions covered. Other requirements necessary to identify particular characteristics important to specialized applications are to be specified by using suffixes as given in Section 5.1.3 This classification system and subsequent line callout (specification) are intended to provide a means of calling out plastic materials used in the fabrication of end items or parts. It is not intended for the selection of materials. Material selection can be made by those having expertise in the plastic field only after careful consideration of the design and the performance required of the part, the environment to which it will be exposed, the fabrication process to be employed, the costs involved, and the inherent properties of the material other than those covered by this standard.NOTE 1: Insert Note 1 here to show the appropriate ISO equivalency statement.1.4 The following precautionary caveat pertains only to the test method portion, Section 11, of this classification system: 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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