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Significance and Use
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.
5.3.1 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).
5.4 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).
5.5 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.
1.3 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.
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
E705 Test Method for Measuring Reaction Rates by Radioactivation of Neptunium-237
E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706 (IIC)
E944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA)
E1005 Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, E 706 (IIIA)
E1018 Guide for Application of ASTM Evaluated Cross Section Data File, Matrix E706 (IIB)
ICS Number Code 17.240 (Radiation measurements); 27.120.30 (Fissile materials and nuclear fuel technology)
UNSPSC Code 15131500(Nuclear fuel)
ASTM E704-13, Standard Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238, ASTM International, West Conshohocken, PA, 2013, www.astm.orgBack to Top