ASTM C1769 - 15

    Standard Practice for Analysis of Spent Nuclear Fuel to Determine Selected Isotopes and Estimate Fuel Burnup

    Active Standard ASTM C1769 | Developed by Subcommittee: C26.05

    Book of Standards Volume: 12.01


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    Significance and Use

    5.1 This standard practice defines a measure of heavy element atom percent fission from which the output of heat during irradiation can be estimated.

    5.2 This standard practice is restricted in use to samples where accurate pre-irradiation U and Pu isotopic analysis is available. This data should be available from the fuel manufacture.

    5.3 The contribution of 238U fast fission is not subject to measurement from isotopic analysis. For reactors in which the majority of fissions are caused by thermal neutrons, the contribution may be estimated from the fast fission factors, ε, found in each reactor design document.

    5.4 In post-irradiation isotopic analysis, take extreme care to avoid environmental uranium contamination of the sample. This is simplified by using sample sizes in which the amount of each uranium isotope is more than 1000 times the levels observed in a blank carried through the complete chemistry and mass spectrometry procedure employed.

    5.5 Take care to make sure that both the pre-irradiation and the post-irradiation samples analyzed are representative. In the pre-irradiation fuel, the 235U and 236U atom ratio content may vary from lot to lot. 236U is not found in naturally uranium in measurable quantity (<2 ppm of a u basis) but forms during irradiation and increases with each successive pass through the fuel cycle. In the post-irradiation examination of a large fuel element, the atom percent fission normally varies radially and axially. Radial and axial profiles of atom percent fission can be determined by analyzing samples obtained from along the radius or axis of the fuel element. An average value of atom percent fission can be obtained by totally dissolving the fuel to be averaged, and then mixing and analyzing an aliquot of the resultant solution.

    5.6 The burnup of an irradiated nuclear fuel can be determined from the amount of a fission product formed during irradiation. Among the fission products, 148Nd has the following properties to recommend it as an ideal burnup indicator: (1) It is not volatile. (2) It does not migrate in solid fuels below their recrystallization temperature. (3) It has no volatile precursors. (4) It is nonradioactive and requires no decay corrections. (5) It has a low destruction cross section. (6) Formation of 148Nd from adjacent mass chains can be corrected for. (7) It has adequate emission characteristics for mass analysis. (8) Its fission yield is nearly equivalent for 235U and 239Pu. (9) Its fission yield is essentially independent of neutron energy (11). (10) It has a shielded isotope, 142Nd, which can be used for correcting natural neodymium contamination. (11) It is an atypical constituent of unirradiated fuel.

    1. Scope

    1.1 A sample of spent nuclear fuel is analyzed to determine the quantity and atomic ratios of uranium and plutonium isotopes, neodymium isotopes, and selected gamma-emitting nuclides (137Cs, 134Cs, 154Eu, 106Ru, and 241Am). Fuel burnup is calculated from the 148Nd-to-fuel ratio as described in this method, which uses an effective 148Nd fission yield calculated from the fission yields of 148Nd for each of the fissioning isotopes weighted according to their contribution to fission as obtained from this method. The burnup value determined in this way requires that values be assumed for certain reactor-dependent properties called for in the calculations (1, 2).2

    1.2 Error associated with the calculated burnup values is discussed in the context of contributions from random and potential systematic error sources associated with the measurements and from uncertainty in the assumed reactor-dependent variables. Uncertainties from the needed assumptions are shown to be larger than uncertainties from the isotopic measurements, with the largest effect arising from the value of the fast fission factor. Using this factor will provide the most consistent burnup value between calculated changes in heavy element isotopic composition.

    1.3 This standard practice contains explanatory notes that are not part of the mandatory portion of the standard.

    1.4 The values stated in SI units are to be regarded as the standard. Mathematical equivalents are given in parentheses.

    1.5 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

    C859 Terminology Relating to Nuclear Materials

    C1625 Test Method for Uranium and Plutonium Concentrations and Isotopic Abundances by Thermal Ionization Mass Spectrometry

    D1193 Specification for Reagent Water

    E244 Test Method for Atom Percent Fission in Uranium and Plutonium Fuel (Mass Spectrometric Method)


    ICS Code

    ICS Number Code 27.120.30 (Fissile materials and nuclear fuel technology)

    UNSPSC Code

    UNSPSC Code 15131500(Nuclear fuel)


    Referencing This Standard
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    DOI: 10.1520/C1769-15

    Citation Format

    ASTM C1769-15, Standard Practice for Analysis of Spent Nuclear Fuel to Determine Selected Isotopes and Estimate Fuel Burnup, ASTM International, West Conshohocken, PA, 2015, www.astm.org

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