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.

1.1 This test method covers procedures for measuring reaction rates by assaying a fission product (F.P.) from the fission reaction ²³⁸U(n,f)F.P.

1.2 The reaction is useful for measuring neutrons with energies from approximately 1.5 to 7 MeV and for irradiation times up to 30 to 40 years.

1.3 Equivalent fission neutron fluence rates as defined in Practice E261 can be determined.

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 and health practices and determine the applicability of regulatory limitations prior to use.