| ||Format||Pages||Price|| |
|9||$54.00||  ADD TO CART|
|Hardcopy (shipping and handling)||9||$54.00||  ADD TO CART|
Significance and Use
4.1 The ENDF/B library in the United States and similar libraries elsewhere, such as JEFF (, JENDL )(, and BROND )(, provide a compilation of neutron cross section and other nuclear data for use by the nuclear community. The availability of these excellent evaluations makes possible standardized usage, thereby allowing easy referencing and intercomparisons of calculations. However, as the first ENDF/B files were developed it became apparent that they were not adequate for all applications. This need resulted in the development of the specialized ENDF/B Dosimetry File )(, consisting of activation cross sections important for dosimetry applications. This file was made available worldwide. Later, other “Special Purpose” files were introduced , )(. In the ENDF/B-VI compilation )(, dosimetry files no longer appeared as separate evaluation files. The ENDF/B-VII.0 compilation )( removed most of the reaction-specific covariance files used by the dosimetry community. It kept the covariance files for the “standard cross sections” in a special sub-library, but the covariance data in this sub-library are only provided over the energy range in which each reaction is considered to be a “standard”, and does not include the full energy range required for LWR PVS dosimetry applications. Later updates to the ENDF/B releases added covariance files for some reaction channels but these covariance files were often based solely on calculations and were not representative of the methodology used to derive the underlying ENDF/B cross section. In response to the need for a dosimetry-specific library, the International Atomic Energy Agency convened a Coordinated Research Project (CRP) that drew upon the set of international experts to provide a recommended set of dosimetry cross sections and to compile a set of validation evidence that supported the use of this recommended dataset. This file, the International Reactor Dosimetry and Fusion File (IRDFF) )(, draws upon other national nuclear evaluations and supplements these evaluations with a set of reactions evaluated by expert international groups. The IRDFF library was developed to support the LWR dosimetry application as well as other dosimetry applications that go beyond the scope of this standard and, as part of its development process, it incorporates validation data acquired in reference and standard benchmark neutron fields. Some of the IRDFF supplemental reactions represent material evaluations that are currently being examined by the CSEWG for inclusion within updated ENDF/B evaluations. The supplemental IRDFF evaluations only include the specific reactions of interest to the dosimetry community and not a full material evaluation. The ENDF community requires a complete evaluation before including it in the main ENDF/B evaluated library. , )
4.2 The application to LWR surveillance dosimetry introduced new data needs that can best be satisfied by the creation of a dedicated cross section file. This file shall be maintained in a form designed for easy application by users (minimal processing). The file shall continue to incorporate the following types of information or indicate the sources of the following type of data that should be used to supplement the file contents:
4.2.1 Dosimetry cross sections for fission, activation, helium production sensor reactions in LWR environments in support of radiometric, solid state track recorder, helium accumulation dosimetry methods (see Test Methods , , , and ).
4.2.2 Other cross sections or sensor response functions useful for active or passive dosimetry measurements, for example, the use of neutron absorption cross sections to represent attenuation corrections due to covers or self-shielding.
4.2.3 Cross sections for damage evaluation, such as displacements per atom (dpa) in iron.
4.2.4 Related nuclear data needed for dosimetry, such as branching ratios, fission yields, and atomic abundances.
4.3 The ASTM-recommended cross sections and uncertainties are based mostly on the IRDFF (version 1.05) dosimetry files. Damage cross sections for materials such as iron have been added in order to promote standardization of reported dpa measurements within the dosimetry community. Integral measurements from benchmark fields and reactor test regions have been considered in order to ensure self-consistency (. The total dosimetry file is intended to be as self-consistent as possible with respect to both differential and integral measurements as applied in LWR environments. This self-consistency of the data file is mandatory for LWR-pressure vessel surveillance applications, where only very limited dosimetry data are available. Where modifications to an existing evaluated cross section have been made to obtain this self-consistence in LWR environments, the modifications shall be detailed in the associated documentation (see )(). , )
1.1 This guide covers the establishment and use of an ASTM evaluated nuclear data cross section and uncertainty file for analysis of single or multiple sensor measurements in neutron fields related to light water reactor LWR-Pressure Vessel Surveillance (PVS). These fields include in- and ex-vessel surveillance positions in operating power reactors, benchmark fields, and reactor test regions.
1.2 Requirements for establishment of ASTM-recommended cross section files address data format, evaluation requirements, validation in benchmark fields, evaluation of error estimates (covariance file), and documentation. A further requirement for components of the ASTM-recommended cross section file is their internal consistency when combined with sensor measurements and used to determine a neutron spectrum.
1.3 Specifications for use include energy region of applicability, data processing requirements, and application of uncertainties.
1.4 This guide is directly related to and should be used primarily in conjunction with Guides and , and Practices , , and .
1.5 The ASTM cross section and uncertainty file represents a generally available data set for use in sensor set analysis. However, the availability of this data set does not preclude the use of other validated data, either proprietary or nonproprietary. When alternate cross section files that deviate from the requirements laid out in this standard are used, the deviations should be noted to the customer of the dosimetry application.
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.
2. Referenced Documents (purchase separately) The documents listed below are referenced within the subject standard but are not provided as part of the standard.
E170 Terminology Relating to Radiation Measurements and Dosimetry
E185 Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
E482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance
E560 Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results, E 706(IC)
E693 Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA)
E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance
E853 Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results
E854 Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance
E910 Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance
E944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance
E1005 Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance
E2005 Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields
ICS Number Code 35.240.50 (IT applications in industry)
|Link to Active (This link will always route to the current Active version of the standard.)|
ASTM E1018-20e1, Standard Guide for Application of ASTM Evaluated Cross Section Data File, ASTM International, West Conshohocken, PA, 2020, www.astm.orgBack to Top