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

5.2 ²³⁸U is available as metal foil, wire, or oxide powder (see Guide E844). It is usually encapsulated in a suitable container to prevent loss of, and contamination by, the ²³⁸U and its fission products.

5.3 One or more fission products can be assayed. Pertinent data for relevant fission products are given in Table 1 and Table 2.

5.3.1 ¹³⁷Cs-^137mBa is chosen frequently for long irradiations. Radioactive products ¹³⁴Cs and ¹³⁶Cs may be present, which can interfere with the counting of the 0.662 MeV  ¹³⁷Cs-^137mBa gamma rays (see Test Methods E320).

5.3.2 ¹⁴⁰Ba-¹⁴⁰La is chosen frequently for short irradiations (see Test Method E393).

5.3.3 ⁹⁵Zr can be counted directly, following chemical separation, or with its daughter ⁹⁵Nb using a high-resolution gamma detector system.

5.3.4 ¹⁴⁴Ce is a high-yield fission product applicable to 2- to 3-year irradiations.

5.4 It is necessary to surround the ²³⁸U monitor with a thermal neutron absorber to minimize fission product production from a quantity of ²³⁵U in the ²³⁸U target and from  ²³⁹Pu from (n,γ) reactions in the ²³⁸U material. Assay of the ²³⁹Pu concentration when a significant contribution is expected.

5.4.1 Fission product production in a light-water reactor by neutron activation product ²³⁹Pu has been calculated to be insignificant (<2 %), compared to that from ²³⁸U(n,f), for an irradiation period of 12 years at a fast-neutron (E > 1 MeV) fluence rate of 1 × 10¹¹ cm⁻² · s⁻¹ provided the ²³⁸U is shielded from thermal neutrons (see Fig. 2 of Guide E844).

5.4.2 Fission product production from photonuclear reactions, that is, (γ,f) reactions, while negligible near-power and research-reactor cores, can be large for deep-water penetrations (1).⁴

5.5 Good agreement between neutron fluence measured by ²³⁸U fission and the ⁵⁴Fe(n,p)⁵⁴Mn reaction has been demonstrated (2). The reaction  ²³⁸U(n,f) F.P. is useful since it is responsive to a broader range of neutron energies than most threshold detectors.

5.6 The ²³⁸U fission neutron spectrum-averaged cross section in several benchmark neutron fields is given in Table 3 of Practice E261. Sources for the latest recommended cross sections are given in Guide E1018. In the case of the ²³⁸U(n,f)F.P. reaction, the recommended cross section source is the ENDF/B-VI release 8 cross section (MAT = 9237) (3). Fig. 1 shows a plot of the recommended cross section versus neutron energy for the fast-neutron reaction ²³⁸U(n,f)F.P.

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 Guide E1018, Section 6.1, since the later ENDF/B-VII data files do not include covariance information. Some covariance information exists for ²³⁸U in the standard sublibrary, but this is only for energies greater than 1 MeV. For more details, see Section H of Ref 4.