Vol 124, No 2 (2018)

The possibility of using the LUCKY-A three-dimensional multiprocessor transport code as a neutronics module for the COREMELT integrated non-stationary code for performing safety analysis, which is used for fast sodium reactor safety validation, is discussed. In this application, more stringent requirements are imposed on the computing speed of the LUCKY-A code. To test LUCKY-A possibilities, a typical computational problem is examined in a formulation characteristic for the proposed use of the code. The possibilities of the code for parallelized calculations are shown, and a comparison is made with calculations performed with the MMKKENO and TRIGEX codes. The efficacy of a parallel computational process is investigated as a function of the number of computing kernels used.

The possibilities for reactivity runaway reduction in a fast reactor when using enriched uranium are examined. Specifically, an approach supposing partial compensation of reactivity runaway by the addition of Np or Am, as consumable absorbers, into the starting load is analyzed. It is shown that from the standpoint of the development of large-scale nuclear power generation based on BREST-type reactors the use of a uranium starting load becomes inexpedient in implementing this approach. A starting load consisting of a mixture of enriched uranium and plutonium from spent RBMK fuel is studied. The results indicate that it is possible to develop a core with reactivity runaway not exceeding the effective delayed-neutron fraction during the entire period of operation. The neutronics characteristics were calculated using the Arktika multigroup diffusion code. The RISK code was used to determine the change in the nuclide composition.

The results of a study of the long-term strength of austenitic and ferrite-martensite steel after irradiation in the BR-10, BOR-60, BN-350, and BN-600 reactors are presented. It is shown that irradiation of austenitic-class steel (EI-847, 0Kh18N10T, and ChS-68) greatly reduces the time to failure and the plasticity. This behavior of austenitic steel is associated with the phenomenon of high-temperature radiation embrittlement. Neutron irradiation does not degrade the long-term mechanical characteristics of ferrite-martensite steel, which remain either at the level of the unirradiated material (EP-823, 05Kh12N2M steels) or surpass it (EP-450 steel).

The results of studies on the processing of carbonate ores with extraction of uranium and molybdenum by means of autoclave and atmospheric leaching and sorption extraction of valuable components using carboxyl cationionites and strongly basic anionites are presented. The most stable and highest indices for the extraction of uranium and molybdenum into solution were obtained for autoclave leaching with the following process parameters: pressure 12–14 psig, temperature ~140°C, pulp processing time 4–6 h, and initial sodium carbonate concentration 40–50 g/dm³. The degree of extraction of uranium and molybdenum into solution was ≥96%. Almost complete separation of uranium from molybdenum was obtained on carboxyl cationites without bicarbonate solutions in the pH range 6.5–7.5. An efficient technology is proposed for combined processing of silicate and carbonate ores using a unified sorption scheme making it possible to extract uranium and molybdenum separately while obtaining ammonium uranyl tricarbonate and ammonium paramolybdate.

The possible consequences of flooding of nuclear and radiation-hazardous sites and objects should be included among the environmental risks arising during the rapid development of the infrastructure in the Arctic. A mathematical model, algorithms, and software codes for evaluating the condition and predicting the collapse of the protective barriers, and determining the emission rate of radionuclides and the possible radioactive contamination of the environment, including long-term prediction and situation analysis, have been developed. The software codes have been accepted into the government registry at Rospatent.