A. A. Dudnikov

Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle

Abstract A subcritical molten salt reactor is proposed for minor actinides (separated from spent fuel VVER-1000 light water reactor) incineration and for 233U conversion from 232Th. Here the subcritical molten salt reactor with fuel composition of heavy nuclide fluorides in molten LiF–NaF–KF salt and with external neutron source, based on 1 GeV proton accelerator and molten salt cooled tungsten target is considered. The paper presents the results of parametrical analysis of equilibrium nuclide composition of molten salt reactor with minor actinides feed in dependence of core dimensions, average neutron flux and external neutron source intensity. Reactor design is defined; requirements to external neutron source are posed; heavy nuclides equilibrium and fuel cycle main parameters are calculated.

Calculations of spent fuel isotopic composition for fuel rod from VVER-440 fuel assembly benchmark using several evaluated nuclear data libraries

Abstract This article presents a comparison of results of simulation of VVER-440 reactor fuel burnup by means of ISTAR set of tools using ENDF/B-VII.0 and JEFF-3.1 evaluated nuclear data libraries. Code verification is required on the basis of both experimental data and computational uncertainty analysis. A VVER-440 fuel rod burnup simulation using ENDF/B-VII.0 and JEFF-3.1 evaluated nuclear data libraries was performed as preliminary step before comparison of nuclear data uncertainty analysis and calculation/experiment. Fuel irradiation conditions and SNF nuclide composition experimental data were obtained from radiochemical assays of irradiated VVER-440 fuel report. Fuel burnup simulation was carried out by means of ISTAR set of tools using Monte-Carlo code for spectrum averaged reaction rates evaluation required for solution of Batemans equations. As a result, Kinf and nuclide composition across burnup were estimated for the sample considered. On the basis of two nuclear data libraries’ implementation for depletion simulation it was shown that Kinf uncertainties rise across burnup from 0.4% at the beginning of life up to 0.7% at the end of life. Heavy nuclides composition uncertainties run up to 9.5% for Am-241 at the end of the 4th life time; however for main uranium and plutonium isotopes the uncertainties do not exceed 1.2%. Most fission products composition uncertainties stay below 7.3% except Sm-150 nuclide (15.1%). The next step will be a comparison with experimental spent nuclear fuel isotopic composition data considering the above mentioned uncertainties.

Influence of uncertainties of isotopic composition of the reprocessed uranium on effectiveness of its enrichment in gas centrifuge cascades

The effect of the uncertainties of the isotopic composition of the reprocessed uranium on its enrichment process in gas centrifuge cascades while diluting it by adding low-enriched uranium (LEU) and waste uranium. It is shown that changing the content of 232U and 236U isotopes in the initial reprocessed uranium within 15% (rel.) can significantly change natural uranium consumption and separative work (up to 2-3%). However, even in case of increase of these parameters is possible to find the ratio of diluents, where the cascade with three feed flows (depleted uranium, LEU and reprocessed uranium) will be more effective than ordinary separation cascade with one feed point for producing LEU from natural uranium.

The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing 233U from 232Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

Evaluating the effectiveness of dilution of the recovered uranium with depleted uranium and low-enriched uranium to obtain fuel for VVER reactors

The possibility of the recovered uranium enrichment in a cascade of gas centrifuges with three feed flows (depleted uranium, low-enriched uranium, recovered uranium) with simultaneous dilution of U-232,234,236 isotopes was shown. A series of numerical experiments were performed for different content of U-235 in low-enriched uranium. It has been demonstrated that the selected combination of diluents can simultaneously reduce the cost of separative work and the consumption of natural uranium, not only with respect to the previously used multi-flow cascade schemes, but also in comparison to the standard cascade for uranium enrichment.