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

5.2 ²³⁷Np is available as metal foil, wire, or oxide powder. For further information, see Guide E844. It is usually encapsulated in a suitable container to prevent loss of, and contamination by, the   ²³⁷Np and its fission products.4

5.3 One or more fission products can be assayed. Pertinent data for relevant fission products are given in Table 15 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 0.661657 MeV  ¹³⁷Cs-^137mBa gamma ray (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 ²³⁷Np monitor with a thermal neutron absorber to minimize fission product production from trace quantities of fissionable nuclides in the ²³⁷Np target and from  ²³⁸Np and  ²³⁸Pu from (n,γ) reactions in the   ²³⁷Np material. Assay of   ²³⁸Pu and   ²³⁹Pu concentration is recommended when a significant contribution is expected.

5.4.1 Fission product production in a light-water reactor by neutron activation products   ²³⁸Np and   ²³⁸Pu has been calculated to be insignificant (1.2 %), compared to that from  ²³⁷Np(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  ²³⁷Np 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 (3).

5.5 This dosimetry reaction is important in the area of reactor retrospective dosimetry (4, 5). Good agreement between neutron fluence measured by  ²³⁷Np fission and the  ⁵⁴Fe(n,p) ⁵⁴Mn reaction has been demonstrated (6, 7). The reaction  ²³⁷Np(n,f) F.P. is useful since it is responsive to a broader range of neutron energies than most threshold detectors.

5.5.1 Fig. 1 shows the energy-dependent cross section for this dosimetry reaction. The figure shows that, while it is not strictly a threshold detector, because of its sensitivity in the greater than 0.1 MeV neutron energy range it can function as a detector with good sensitivity in the fast neutron region. In the fast fission ²⁵²Cf spontaneous fission benchmark field, ~1 % of the ²³⁷Np fission dosimeter response comes from neutrons with an energy less than 0.1 MeV. In the cavity of a fast burst ²³⁵U reactor, ~5 % of the ²³⁷Np ifssion dosimeter response comes from neutrons with an energy less than 0.1 MeV. In the cavity of a well-moderated pool-type research reactor ~50 % of the fission response from the ²³⁷Np(n,f) reaction comes from energies less than 0.1 MeV. The importance of this low neutron energy sensitivity should be determined based on the aplication.

5.6 The  ²³⁷Np fission neutron spectrum-averaged cross section in several benchmark neutron fields are given in Table 3 of Practice E261. Sources for the latest recommended cross sections are given in Guide E1018. In the case of the  ²³⁷Np(n,f)F.P. reaction, the recommended cross section source is the Russian Reactor Dosimetry File, RRDF (8). This recommended cross section is identical, for energies up to 20 MeV, to what is found in the latest International Atomic Energy (IAEA) International Reactor Dosimetry and Fusion File, IRDFF-1.05 (9) . Fig. 1 shows a plot of the recommended cross section versus neutron energy for the fast-neutron reaction   ²³⁷Np(n,f)F.P.

FIG. 1 RRDF/IRDFF-1.05 Cross Section Versus Energy for the ²³⁷Np(n,f)F.P. Reaction

1.1 This test method covers procedures for measuring reaction rates by assaying a fission product (F.P.) from the fission reaction  ²³⁷Np(n,f)F.P.

1.2 The reaction is useful for measuring neutrons with energies from approximately 0.7 to 6 MeV and for irradiation times up to 90 years, provided that the analysis methods described in Practice E261 are followed. If dosimeters are analyzed after irradiation periods longer than 90 years, the information inferred about the fluence during irradiation periods more than 90 years before the end of the irradiation should not be relied upon without supporting data from dosimeters withdrawn earlier.

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, 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.