Значение и использование

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, does not migrate in solid fuels below their recrystallization temperature, and has no volatile precursors. (2) It is nonradioactive and requires no decay corrections. ( 3) It has a low destruction cross section and formation from adjacent mass chains can be corrected for. (4) It has good emission characteristics for mass analysis. (5) Its fission yield is nearly the same for ²³⁵U and ²³⁹Pu and is essentially independent of neutron energy (6). (6) It has a shielded isotope, ¹⁴²Nd, which can be used for correcting natural Nd contamination. (7) It is not a normal constituent of unirradiated fuel.

The analysis of ¹⁴⁸Nd in irradiated fuel does not depend on the availability of preirradiation sample data or irradiation history. Atom percent fission is directly proportional to the ¹⁴⁸Nd-to-fuel ratio in irradiated fuel. However, the production of ¹⁴⁸Nd from ¹⁴⁷Nd by neutron capture will introduce a systematic error whose contribution must be corrected for. In power reactor fuels, this correction is relatively small. In test reactor irradiations where fluxes can be very high, this correction can be substantial (see Table 1).

The test method can be applied directly to U fuel containing less than 0.5 % initial Pu with 1 to 100 GW days/metric ton burnup. For fuel containing 5 to 50 % initial Pu, increase the Pu content by a factor of 10 to 100, respectively in both reagents 5.3 and 5.4.

TABLE 1 K Factors to Correct ¹⁴⁸Nd for ¹⁴⁷Nd Thermal Neutron Capture^A

Total Neutron Flux,
ϕ (neutrons/cm²/s)Total Neutron Exposure, ϕI (neutrons/cm ²) 1 × 10 ²⁰3 × 10 ²⁰1 × 10 ²¹2 × 10 ²¹3 × 10 ²¹ 3 × 10^{1 2}0.99850.99850.99850.9985 0.9985 1 × 10^{1 3}0.99560.99520.99500.9950 0.9950 3 × 10^{1 3}0.99060.98700.98560.9853 0.9852 1 × 10^{1 4}0.98580.97160.95980.9569 0.9559 3 × 10^{1 4}0.98350.95920.91870.9008 0.8941 1 × 10^{1 5}0.98260.95260.88160.8284 0.8006

^A Assuming continuous reactor operation and a 274 ± 91 barn ^{1 47}Nd effective neutron absorption cross section for a thermal neutron power reactor. This cross section was obtained by adjusting the 440 ± 150 barn ¹⁴⁷Nd cross section (7) measured at 20°C to a Maxwellian spectrum at a neutron temperature of 300°C.

Область применения

1.1 This test method covers the determination of stable fission product ¹⁴⁸Nd in irradiated uranium (U) fuel (with initial plutonium (Pu) content from 0 to 50 %) as a measure of fuel burnup (1-3).

1.2 It is possible to obtain additional information about the uranium and plutonium concentrations and isotopic abundances on the same sample taken for burnup analysis. If this additional information is desired, it can be obtained by precisely measuring the spike and sample volumes and following the instructions in Test Method E267.

1.3 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.