Developed by Subcommittee: E10.05
WITHDRAWN, NO REPLACEMENT
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-Fig. 2).
Formerly under the jurisdiction of Committee E10 on Nuclear Technology and Subcommittee E10.05 on Nuclear Radiation Metrology, this Master Matrix was withdrawn in July 2011 in accordance with section 10.5.3.1 of the Regulations Governing ASTM Technical Committees, which requires that standards shall be updated by the end of the eighth year since the last approval date.
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). Some of these are existing ASTM standards, some are ASTM standards that have been modified, and some are proposed ASTM standards. General requirements of content and consistency are discussed in Section 6 . More detailed writers' and users' information, justification, and specific requirements for the nine practices, ten guides, and three methods are provided in Sections 3-5. Referenced documents are discussed in Section 2. The summary-type 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 documents, Section 2 and references (1-106) for use by individual writers and users.
1.2 This master matrix is intended as a reference and guide to the preparation, revision, and use of standards in the series and for planning and scheduling purposes. This index is to ensure the accomplishment of an objective irrespective of the time required, the number of ASTM committees concerned, or the complexity of the issues involved.
1.3 This master matrix standard provides a guide to ASTM standards related to the energy-critical areas that are to be developed under the category of Fission Reactor Development, Section 10, of Guide E584-77 and as discussed in Practice E583-97.
1.4 To account for neutron radiation damage in setting pressure-temperature limits and making fracture analyses (see Refs 2-7, 9-14, 21-57, 63, 69-71, 77, 78, 83-104 and Recommended 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 (1, 3, 4, 10-13, 17, 21, 22-27, 32-39, 42, 43, 45, 49-57, 71, 77, 78, 83, 91-104, and Recommended Guide E509). 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 this master matrix (2,34 ). 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 edge 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 spectral effects, and
1.4.10 Neutron dose rate effects.
1.5 A number of potential methods and standards exist for ensuring the adequacy of fracture control of reactor pressure vessel belt lines under normal and accident loads (1-4, 6, 7, 13, 14, 21-28, 29-34, 52-57, 71, 77, 78, 91, 93, Recommended Guide E509, and 2.3 ASME Standards). 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, better procedures to evaluate and use this information can and must be developed (1-4, 6, 7, 9-15, 17, 21-34, 52-57, 69, 71-73, 77, 78, 91-104 and Recommended Guide E509). This master matrix, therefore, defines the current (1) scope, (2) areas of application, and (3) general grouping for the series of 22 ASTM standards, as shown in Figs. 1-3.
1.6 The values stated in SI units are to be regarded as the standard.
1.7 This standard may involve hazardous materials, operations, and equipment. 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.
2. Referenced Documents (purchase separately) The documents listed below are referenced within the subject standard but are not provided as part of the standard.
C859 Terminology Relating to Nuclear Materials
E170 Terminology Relating to Radiation Measurements and Dosimetry
E184 Practice for Effects of High-Energy Neutron Radia-tion on the Mechanical Properties of Metallic Materials, E 706 (IB),
E185 Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E 706 (IF),
E482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E 706 (IID),
E509 Guide for In-Service Annealing of Light-Water Cooled Nuclear Reactor Vessels
E560 Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results, E 706 (IC),
E583 Practice for Systematizing the Development of (ASTM) Voluntary Consensus Standards for the Solution of Nuclear and Other Complex Problems
E584 Guide for Developing the (ASTM) Voluntary Consensus Standards Required to Help Implement the National Energy Plan
E636 Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH),
E646 Test Method for Tensile Strain-Hardening Exponents (n-Values) of Metallic Sheet Materials
E693 Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (dpa), E 706 (ID),
E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706 (IIC) ,
E853 Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, E 706 (IA),
E854 Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance, E 706 (IIIB),
E900 Guide for Predicting Neutron Radiation Damage to Reactor Vessel Materials, E 706 (IIF),
E910 Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance, E 706 (IIIC),
E944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA),
E1005 Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, E 706 (IIIA),
E1006 Practice for Analysis and Interpretation of Physics Dosimetry Results for Test Reactors, E 706 (II),,
E1018 Guide for Application of ASTM Evaluated Cross Section Data File E 706 (IIB)
E1035 Practice for Determining Radiation Exposures for Nuclear Reactor Vessel Support Structures
E1214 Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance, E 706 (IIIE)
E1253 Guide for Reconstitution of Irradiated Charpy Specimens,
E2005 Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields,
E2006 Guide for Benchmark Testing of Light Water Reactor Calculations ,
E2059 Practice for Application and Analysis of Nuclear Research Emulsions for Fast Neutron Dosimetry,
SI10 Standard for the Use of the International System of Units (SI): The Modern Metric System
Nuclear Regulatory Documents1.150 Regulatory Guide
American Society of Mechanical Engineers StandardBoilerandPressureVes Sections III and XI