Active Standard ASTM E2006 | Developed by Subcommittee: E10.05
Book of Standards Volume: 12.02
Historical (view previous versions of standard)
Significance and Use
This guide deals with the difficult problem of benchmarking neutron transport calculations carried out to determine fluences for plant specific reactor geometries. The calculations are necessary for fluence determination in locations important for material radiation damage estimation and which are not accessible to measurement. The most important application of such calculations is the estimation of fluence within the reactor vessel of operating power plants to provide accurate estimates of the irradiation embrittlement of the base and weld metal in the vessel. The benchmark procedure must not only prove that calculations give reasonable results but that their uncertainties are propagated with due regard to the sensitivities of the different input parameters used in the transport calculations. Benchmarking is achieved by building up data bases of benchmark experiments that have different influences on uncertainty propagation. For example, fission spectra are the fundamental data bases which control propagation of cross section uncertainties, while such physics-dosimetry experiments as vessel wall mockups, where measurements are made within a simulated reactor vessel wall, control error propagation associated with geometrical and methods approximations in the transport calculations. This guide describes general procedures for using neutron fields with known characteristics to corroborate the calculational methodology and nuclear data used to derive neutron field information from measurements of neutron sensor response.
The bases for benchmark field referencing are usually irradiations performed in standard neutron fields with well-known energy spectra and intensities. There are, however, less well known neutron fields that have been designed to mockup special environments, such as pressure vessel mockups in which it is possible to make dosimetry measurements inside of the steel volume of the “vessel”. When such mockups are suitably characterized they are also referred to as benchmark fields. A benchmark is that against which other things are referenced, hence the terminology “to benchmark reference” or “benchmark referencing”. A variety of benchmark neutron fields, other than standard neutron fields, have been developed, or pressed into service, to improve the accuracy of neutron dosimetry measurement techniques. Some of these special benchmark experiments are discussed in this standard because they have identified needs for additional benchmarking or because they have been sufficiently documented to serve as benchmarks.
One dedicated effort to provide benchmarks whose radiation environments closely resemble those found outside the core of an operating reactor was the Nuclear Regulatory Commission's Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) (1) . This program promoted better monitoring of the radiation exposure of reactor vessels and, thereby, provided for better assessment of vessel end-of-life conditions. An objective of the LWR-PV-SDIP was to develop improved procedures for reactor surveillance and document them in a series of ASTM standards (see Matrix E706). The primary means chosen for validating LWR-PV-SDIP procedures was by benchmarking a series of experimental and analytical studies in a variety of fields (see Guide E2005).
1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with a higher degree of confidence.
1.2 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.
E261 Practice for Determining Neutron Fluence, Fluence Rate, and Spectra by Radioactivation Techniques
E262 Test Method for Determining Thermal Neutron Reaction Rates and Thermal Neutron Fluence Rates by Radioactivation Techniques
E706 Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E 706(0)
E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706 (IIC)
E944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA)
E1018 Guide for Application of ASTM Evaluated Cross Section Data File, Matrix E706 (IIB)
E2005 Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields
ICS Number Code 27.120.10 (Reactor engineering)
UNSPSC Code 39121015(Reactors); 26142100(Nuclear reactor equipment)