Evaluation of the irradiation-averaged fission yield for burnup determination in spent fuel assays

In order to derive the burnup of spent nuclear fuel from the concentration of selected fission products (typically the Nd isotopes and 137Cs), their irradiation-averaged fission yields need to be known with sufficient accuracy, as they evolve with the changes in the actinide vector over the irradiation history. To obtain irradiation-averaged values, radiochemists often resort to robust generic methods – i.e., based on simple mathematical relations – that weight the fission yields according to the actinides contributing to fission, without performing core physics calculations. In order to assess the performance of those generic methods, a database of about 30 000 spent nuclear fuel inventories has been constructed from neutron transport and depletion simulations, covering a representative range of fuel enrichment, burnup, assembly designs and reactor types. When testing several existing methods for effective fission yield calculation, some inaccuracies were identified, originating from improper one-group cross-section parameters that do not accurately reflect resonance and self-shielding effects, and too crude approximations in the estimation of the actinide concentration evolution. Revised effective fission and absorption cross-section parameters are then proposed here, as a first improvement to the earlier burnup determination methods. As a second step, a novel method is proposed that reduces the error on their radiation-averaged fission yield values, and hence on burnup, while retaining a straightforward calculation scheme

Optimization of isolation and mass spectrometric analysis of lanthanides from spent nuclear fuel

status: publishe

Solvent extraction of Am(III), Cm(III) and Ln(III) ions from simulated highly active raffinate solutions by TODGA diluted in Aliquat-336 nitrate ionic liquid

The extraction behaviour of americium(III), curium(III) (minor actinides, MA), fission products (lanthanides (Ln(III)), Mo, Ru, Pd, Rh) and corrosion products (Zr, Fe) was studied in batch solvent extraction experiments using the room-temperature ionic liquid Aliquat-336 nitrate ([A336][NO3]) and a solvent composed of 0.05 M TODGA in [A336][NO3]. From acidic, dilute Ln(III) feed solutions, [A336][NO3] extracts nitric acid (D ≈ 0.5) and partially An(III) as well as Ln(III) (DAm = 0.02 -0.1, DEu and DCm = 0.01 - 0.04). The influence of the acid concentrations and kinetics on extraction and back-extraction of Ln(III) and An(III) by 0.05 M TODGA in [A336][NO3] was investigated using radiotracer-spiked dilute Ln(III) feed solutions. With the solvent composed of 0.05 M TODGA in [A336][NO3], DAn and DLn increased as a function of aqueous feed acidity. In the case of a spiked, simulated highly active raffinate (HAR) feed solution, [A336][NO3] extracted La(III) (DLa = 1.36), Ru (DRu = 1.64) and Pd (DPd = 38), while the distribution ratios of other Ln(III) and An(III) were lower than unity. The solvent composed of 0.05 M TODGA in [A336][NO3] co-extracted from HAR Zr(IV) (DZr > 300), Pd (DPd = 206) and Ln(III) (DLn > 1, except for Nd(III)), but An(III) were retained in the aqueous phase. The interference caused by the co-extraction of several fission (Zr, Pd, Ru, Mo) and corrosion (Zr) products, which are present in the HAR at relatively high concentrations, was suppressed using masking agents (oxalic acid and trans-1,2-diaminocyclohexane-N,N,N’,N’-tetraacetic acid or CDTA). In the case of the actual HAR solution, the kinetics were found to be faster compared to the extraction from dilute Ln(III) feed solutions, possibly due to the different aqueous speciation of the Ln(III) and An(III)