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Significance and Use
5.1 Refer to Guide for the selection, irradiation, and quality control of neutron dosimeters.
5.2 Refer to Practice for a general discussion of the determination of fast-neutron fluence rate with threshold detectors.
5.3 Pure nickel in the form of foil or wire is readily available, and easily handled.
5.4 58Co has a half-life of 70.85 (3) days (Refs ( and )() ) and emits a gamma ray with an energy of 810.7602 (20) keV (Refs ( and )(). )
5.5 Competing activities 65Ni(2.5172 h) and 57Ni(35.9 (3) h (Ref () are formed by the reactions 64Ni(n,γ) 65Ni, and 58Ni(n,2n)57Ni, respectively. )
5.6 A second 9.04 h isomer, 58mCo, is formed that decays to 70.85-day 58Co. Loss of 58Co and 58mCo by thermal-neutron burnout will occur in environments (Refs ( and )( having thermal fluence rates of 3 × 1012 cm−2·s −1 and above. Burnout correction factors, ) R, are plotted as a function of time for several thermal fluxes in . Tabulated values for a continuous irradiation time are provided in Hogg, et al. (Ref () )
5.7 shows a plot of cross section (Ref () versus energy for the fast-neutron reaction 58Ni(n,p) 58Co. This figure is for illustrative purposes only to indicate the range of response of the 58Ni(n,p) reaction. Refer to Guide ) for descriptions of recommended tabulated dosimetry cross sections.
FIG. 2 58Ni(n,p)58Co Cross Section
Note 1: The data is taken from the Evaluated Nuclear Data File, ENDF/B-VI, rather than the later ENDF/B-VII. This is in accordance with (. ), section 6.1, since the later ENDF/B-VII data files do not include covariance information. For more details see Section H of Ref
1.1 This test method covers procedures for measuring reaction rates by the activation reaction 58Ni(n,p)58Co.
FIG. 1 R Correction Values as a Function of Irradiation Time and Neutron Flux
Note 1: The burnup corrections were computed using effective burn-up cross sections of 1650 b for 58Co(n,γ) and 1.4E5 b for 58mCo(n,γ).
1.2 This activation reaction is useful for measuring neutrons with energies above approximately 2.1 MeV and for irradiation times up to about 200 days in the absence of high thermal neutron fluence rates, provided that the analysis methods described in Practice are followed. If dosimeters are analyzed after irradiation periods longer than 200 days, the information inferred about the fluence during irradiation periods more than 200 days before the end of the irradiation should not be relied upon without supporting data from dosimeters withdrawn earlier.
1.3 With suitable techniques fission-neutron fluence rates densities above 107 cm−2·s −1 can be determined.
1.4 Detailed procedures for other fast-neutron detectors are referenced in Practice .
1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.
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
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
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
E1018 Guide for Application of ASTM Evaluated Cross Section Data File
ICS Number Code 17.240 (Radiation measurements); 27.120.30 (Fissile materials and nuclear fuel technology)
UNSPSC Code 26142004(Neutron irradiators)
|Link to Active (This link will always route to the current Active version of the standard.)|
ASTM E264-19, Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Nickel, ASTM International, West Conshohocken, PA, 2019, www.astm.orgBack to Top