ASTM E704 - 13

    Standard Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238

    Active Standard ASTM E704 | Developed by Subcommittee: E10.05

    Book of Standards Volume: 12.02

      Format Pages Price  
    PDF Version 4 $37.00   ADD TO CART
    Print Version 4 $37.00   ADD TO CART
    Standard + Redline PDF Bundle 8 $44.40   ADD TO CART

    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 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 1The 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.

    1. Scope

    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.

    2. Referenced Documents (purchase separately) The documents listed below are referenced within the subject standard but are not provided as part of the standard.

    ASTM Standards

    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 Code

    ICS Number Code 17.240 (Radiation measurements); 27.120.30 (Fissile materials and nuclear fuel technology)

    UNSPSC Code

    UNSPSC Code 15131500(Nuclear fuel)

    DOI: 10.1520/E0704

    ASTM International is a member of CrossRef.

    ASTM E704

    Citing ASTM Standards
    Back to Top