Inference on fission timescale from neutron multiplicity measurement in18O+184W

The pre-scission and post-scission neutron multiplicities are measured for the 18O + 184W reaction in the excitation energy range of 67.23−76.37 MeV. Langevin dynamical calculations are performed to infer the energy dependence of fission decay time in compliance with the measured neutron multiplicities. Different models for nuclear dissipation are employed for this purpose. Fission process is usually expected to be faster at a higher beam energy. However, we found an enhancement in the average fission time as the incident beam energy increases. It happens because a higher excitation energy helps more neutrons to evaporate that eventually stabilizes the system against fission. The competition between fission and neutron evaporation delicately depends on the available excitation energy and it is explained here with the help of the partial fission yields contributed by the different isotopes of the primary compound nucleus.

Determinations of fission cross-sections and fission yields from proton induced fission reactions of 232Th, 233,235,236,238U, 237Np and 239Pu

In this work, the proton induced fission reaction cross-sections and fission yields are calculated for some actinides [Formula: see text], [Formula: see text], [Formula: see text], and [Formula: see text] using the fission barrier models of the TALYS 1.95 code. Cross-sections and fission yield calculations are carried out up to 100 MeV incident proton energies. The calculation results are compared with the available experimental data in the EXFOR library. In addition, a relative variance analysis of fission barrier models was done to determine the fission barrier model whose results best matched with the experimental results. Among the fission barrier models, the best agreement with the experimental data is obtained from the rotating-finite-range fission barrier model calculation for the [Formula: see text] reaction of the studied nuclei having the atomic mass number larger than 230. On the other hand, fission barrier heights for the studied reactions are determined using the same models.

Separation of Lanthanide Isotopes from Mixed Fission Product Samples

The measurement of radioactive fission products from nuclear events has important implications for nuclear data production, environmental monitoring, and nuclear forensics. In a previous paper, the authors reported the optimization of an intra-group lanthanide separation using LN extraction resin from Eichrom Technologies®, Inc. and a nitric acid gradient. In this work, the method was demonstrated for the separation and quantification of multiple short-lived fission product lanthanide isotopes from a fission product sample produced from the thermal irradiation of highly enriched uranium. The separations were performed in parallel in quadruplicate with reproducible results and high decontamination factors for 153Sm, 156Eu, and 161Tb. Based on the results obtained here, the fission yields for 144Ce, 153Sm, 156Eu, and 161Tb are consistent with published fission yields. This work demonstrates the effectiveness of the separations for the intended application of short-lived lanthanide fission product analysis requiring high decontamination factors.

Cross Section and Fission Yields Induced Uncertainty in the Vver-440 Burnup Calculation

The properties of nuclear fuel depend on the actual isotopic composition which develops during a reactor operation. In practice, the prediction accuracy of burnup calculations serves as the basis for the future precise estimation of a core lifetime and other safety-based core characteristics. The present study quantifies nuclear data induced uncertainties of nuclide concentrations and multiplication factors in VVER-440 fuel depletion analysis. The well-known SCALE system and the TRITON sequence are used with the NEWT deterministic solver in the SAMPLER module that implements stochastic techniques to assess the uncertainty in computed results. The propagation of uncertainties in neutron cross section and fission yields is studied through the depletion calculation of 2D heterogeneous VVER-440 fuel assembly with an average enrichment of 4.87 wt % of 235U and six gadolinium rods with 3.35 % of Gd2O3. In the paper, fixed nominal depletion conditions are based on the real operational data of the Slovak NPP Bohunice unit 4 during cycle 30. In total 250 cases with uncertain parameters are computed and the results are evaluated by an auxiliary tool.

APPLICATION OF A CUSTOM DEPLETION FRAMEWORK TO THE PREDICTION OF NEUTRON FLUX DISTRIBUTION THROUGH DEPLETION

Previous works by the authors have introduced the spatial flux variation method (SFV) for predicting the changes in neutron flux due to a change in material compositions. In order to remove a full transport solution at the end-of-step, this work presents a framework responsible for computing macroscopic cross sections after a depletion event. These end of-step cross sections are estimators of changes in neutron loss and production, and enable the prediction of neutron flux using only information obtained from a single beginning of-step transport solution. The framework reads in all relevant data needed to model the depletion system, including one-group cross sections and effective fission yields to reproduce the problem using an external solver. The framework also supports extrapolating microscopic cross sections in order to rebuild the end-of-step macroscopic cross sections needed for the flux prediction. Results indicate that the SFV method is not adversely effected by the external depletion solution, and can be implemented alongside an existing transport-depletion framework.