Active Standard ASTM E266 | Developed by Subcommittee: E10.05
Book of Standards Volume: 12.02
Historical (view previous versions of standard)
Significance and Use
Refer to Guide E844 for the selection, irradiation, and quality control of neutron dosimeters.
Refer to Practice E261 for a general discussion of the determination of fast-neutron fluence rate with threshold detectors.
Pure aluminum in the form of foil or wire is readily available and easily handled. 27Al has an abundance of 100 % (1) .
24Na has a half-life of 14.9574 h (2) and emits gamma rays with energies of 1.368626 and 2.754007 MeV (2).
Fig. 1 shows a plot of cross section versus neutron energy for the fast-neutron reaction 27Al(n,α)24Na (3) along with a comparison to the current experimental database (4). This figure is for illustrative purposes only to indicate the range of response of the 27Al(n,α) reaction. Refer to Guide E1018 for descriptions of recommended tabulated dosimetry cross sections.
Two competing activities, 28Al and 27Mg, are formed in the reactions 27Al(n,γ) 28Al and 27Al(n,p) 27Mg, respectively, but these can be eliminated by waiting 2 h before counting.
1.1 This test method covers procedures measuring reaction rates by the activation reaction 27Al(n,α)24Na.
1.2 This activation reaction is useful for measuring neutrons with energies above approximately 6.5 MeV and for irradiation times up to about 2 days (for longer irradiations, see Practice E261).
1.3 With suitable techniques, fission-neutron fluence rates above 106 cm−2·s−1 can be determined.
1.4 Detailed procedures for other fast neutron detectors are referenced in Practice E261.
1.5 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.
E170 Terminology Relating to Radiation Measurements and Dosimetry
E181 Test Methods for Detector Calibration and Analysis of Radionuclides
E261 Practice for Determining Neutron Fluence, Fluence Rate, and Spectra by Radioactivation Techniques
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)
E1005 Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, E 706 (IIIA)
E1018 Guide for Application of ASTM Evaluated Cross Section Data File, Matrix E706 (IIB)
ICS Number Code 17.240 (Radiation measurements); 27.120.30 (Fissile materials and nuclear fuel technology)
UNSPSC Code 11101705(Aluminum)