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Significance and Use
5.1 Refer to Practice 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 . It is usually encapsulated in a suitable container to prevent loss of, and contamination by, the 237Np and its fission products.
5.3 One or more fission products can be assayed. Pertinent data for relevant fission products are given in and .
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 ).
5.3.2 140Ba-140La is chosen frequently for short irradiations (see Test Method ).
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 ).
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 ().
5.5 This dosimetry reaction is important in the area of reactor retrospective dosimetry (. Good agreement between neutron fluence measured by 237Np fission and the 54Fe(n,p) 54Mn reaction has been demonstrated , )(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 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 . Sources for the latest recommended cross sections are given in Guide . In the case of the 237Np(n,f)F.P. reaction, the recommended cross section source is the Russian Reactor Dosimetry File, RRDF (. 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 )( . ) 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 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 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
E262 Test Method for Determining Thermal Neutron Reaction Rates and Thermal Neutron Fluence Rates by Radioactivation Techniques
E320 Test Method for Cesium-137 in Nuclear Fuel Solutions by Radiochemical Analysis
E393 Test Method for Measuring Reaction Rates by Analysis of Barium-140 From Fission Dosimeters
E704 Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238
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); 26141702(Dosimeters)
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ASTM E705-18, Standard Test Method for Measuring Reaction Rates by Radioactivation of Neptunium-237, ASTM International, West Conshohocken, PA, 2018, www.astm.orgBack to Top