5.1 Refer to Guide E844 for the selection, irradiation, and quality control of neutron dosimeters.

5.2 Refer to Practice E261 for a general discussion of the determination of fast-neutron fluence rate with threshold detectors.

5.3 Titanium has good physical strength, is easily fabricated, has excellent corrosion resistance, has a melting temperature of 1668 °C, and can be obtained with satisfactory purity.

5.4 ⁴⁶Sc has a half-life of 83.787 (16)4 days (2). The ⁴⁶Sc decay emits a 0.889271 (2) MeV gamma 99.98374 (35) % of the time and a second gamma with an energy of 1.120537 (3) MeV 99.97 (2) % of the time.

5.5 The recommended “representative isotopic abundances” for natural titanium (3) are:

8.25 (3) % ⁴⁶Ti

7.44 (2) % ⁴⁷Ti

73.72 (2) % ⁴⁸Ti

5.41 (2) % ⁴⁹Ti

5.18 (2) % ⁵⁰Ti

5.6 The radioactive products of the neutron reactions ⁴⁷Ti(n,p)⁴⁷Sc (τ_1/2 = 3.3485 (9) d) (2) and ⁴⁸Ti(n,p)⁴⁸Sc (τ_1/2 = 43.67 h), (3) might interfere with the analysis of ⁴⁶Sc.

5.7 Contaminant activities (for example, ⁶⁵Zn and ¹⁸²Ta) might interfere with the analysis of ⁴⁶Sc. See 7.1.2 and 7.1.3 for more details on the ¹⁸²Ta and ⁶⁵Zn interference.

5.8 ⁴⁶Ti and ⁴⁶Sc have cross sections for thermal neutrons of 0.59 ± 0.18 and 8.0 ± 1.0 barns, respectively (4); therefore, when an irradiation exceeds a thermal-neutron fluence greater than about 2 × 10²¹ cm^–2, provisions should be made to either use a thermal-neutron shield to prevent burn-up of ⁴⁶Sc or measure the thermal-neutron fluence rate and calculate the burn-up.

5.9 Fig. 1 shows a plot of the International Reactor Dosimetry and Fusion File, IRDFF-II cross section (5) versus neutron energy for the fast-neutron reactions of titanium which produce ⁴⁶Sc (that is, ^natTi(n,X)⁴⁶Sc). Included in the plot is the ⁴⁶Ti(n,p) reaction and the ⁴⁷Ti(n,np:d) contributions to the ⁴⁶Sc production, normalized per ^natTi atom with the individual isotopic contributions weighted using the natural abundances (3). This figure is for illustrative purposes only and should be used to indicate the range of response of the ^natTi(n,X)⁴⁶Sc reaction. Refer to Guide E1018 for descriptions of recommended tabulated dosimetry cross sections. Fig. 2 compares the cross section for the ⁴⁶Ti(n,p)⁴⁶Sc reaction to the current experimental database (6, 7). Fig. 3 compares the cross section for the ⁴⁷Ti(n,np:d) reaction to the current experimental database (6, 7).

FIG. 1 SAND-II 640-Group Histogram Representation of the ^natTi(n,X)⁴⁶Sc Cross Section (Normalized per Elemental Ti Atom Using Natural Abundance Data), Represented By the Sum of the ^natTi(n,p)⁴⁶Sc, ^natTi(n,np)⁴⁶Sc, and ^natTi(n,d)⁴⁶Sc Cross Section Components

FIG. 2 ⁴⁶Ti(n,p)⁴⁶Sc Cross Section (Normalized per Isotopic ⁴⁶Ti Atom), from IRDFF-II, with EXFOR Experimental Data

FIG. 3 ⁴⁷Ti(n,np:d)⁴⁶Sc Cross Section (Normalized per Isotopic ⁴⁷Ti Atom), from IRDFF-II, with EXFOR Experimental Data

1.1 This test method covers procedures for measuring reaction rates by the activation reaction ^natTi(n,X)⁴⁶Sc. The “X” designation represents any combination of light particles associated with the production of the residual ⁴⁶Sc product. Within the applicable neutron energy range for fission reactor applications, this reaction is a properly normalized combination of three different reaction channels: ⁴⁶Ti(n,p)⁴⁶Sc; ⁴⁷Ti(n, np)⁴⁶Sc; and ⁴⁷Ti(n,d)⁴⁶Sc.

Note 1: The ⁴⁷Ti(n,np)⁴⁶Sc reaction, ENDF-6 format file/reaction identifier MF=3, MT=28, is distinguished from the ⁴⁷Ti(n,d)⁴⁶Sc reaction, ENDF-6 format file/reaction identifier MF=3/MT=104, even though it leads to the same residual product (1).2 The combined reaction, in the IRDFF-II library, has the file/reaction identifier MF=10/MT=5.

Note 2: The cross section for the combined ⁴⁷Ti(n,np:d) reaction is relatively small for energies less than 12 MeV and, in fission reactor spectra, the production of the residual ⁴⁶Sc is not easily distinguished from that due to the ⁴⁶Ti(n,p) reaction.

1.2 The reaction is useful for measuring neutrons with energies above approximately 4.4 MeV and for irradiation times, under uniform power, up to about 250 days (for longer irradiations, or for varying power levels, see Practice E261).

1.3 With suitable techniques, fission-neutron fluence rates above 10⁹ cm^–2·s^–1 can be determined. However, in the presence of a high thermal-neutron fluence rate, ⁴⁶Sc depletion should be investigated.

1.4 Detailed procedures for other fast-neutron detectors are referenced in Practice E261.

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.