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 and 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 data bases.

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:

–needed for quality control and certification of current and improved dosimetry practices;

–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, 37, 38, 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, 37, 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 groups’ power reactor physics computational tools.

4.4.3.3 The standards recommendation should be that the vendor/utility groups’ 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 groups’ power reactor physics computational tools.

4.4.4.3 The standards recommendation should be that the vendor/utility groups’ 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, 37, 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):

–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.

–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 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, 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>1MeV. 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.

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 (1-12)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, and

1.4.10 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 and health practices and determine the applicability of regulatory limitations prior to use.