ASTM E706-16

4.2.1 Standardized procedures and reference data are recommended in regard to (1) neutron and gamma dosimetry, 4.2.2 Existing state-of-the-art practices associated with 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.

4.3.1.1 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).

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:

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.

4.4.1.1 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 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.

4.4.2.1 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 4.4.3 PCA, VENUS, and NESDIP Experiments and PCA Blind Test: 

4.4.4 PWR and BWR Generic Power Reactor Tests: 

4.5 Metallurgical Damage Correlation Analysis and Interpretation—4.5.1 Required Accuracies and Benchmark Field Referencing: 

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 4.5.3 Reg. Guide 1.99, NRC, EPRI Data Bases—NRC and Electric Power Research Institute (EPRI) data bases have been studied on an ongoing basis by ASTM Subcommittees E10.02 and E10.05, vendors, utilities, EPRI, and NRC contractors to establish improved data bases 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 data bases. 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.

1.2 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 1.4 The techniques, variables and uncertainties related to (1) neutron and gamma dosimetry, 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, and

1.4.10 Neutron dose rate effects.