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ETVS fuel assemblies containing nitride fuel rods with different structural implementation and cladding made of different materials are tested in the cores of the BN-600 and BOR-60 reactors in order to substantiate the efficiency of fuel rods with mixed uranium-plutonium nitride fuel used in sodium-cooled (BN-1200) and lead-cooled (BREST-OD-300) reactors. We report here the results of post-reactor studies of BN-600, -800, -1200 and BREST type fuel elements with a helium sublayer after irradiation to maximum burnup 7.5, 6, and 4.5% h.a., respectively, in the composition of 10 ETVS BN-600 fuel assemblies as well BREST type fuel rods with helium and lead sublayer after tests to maximum burnup 5 and 4% h.a. as fuel elements of the BREST type with a helium and lead sublayer after testing to maximum burnup 5 and 4% h.a. in the composition of two ETVS BOR-60 fuel assemblies. The obtained results confirmed that the nitride fuel rods with gas and lead sublayers and EP-823 steel cladding remain efficient at the attained test parameters and, possibly, at higher fuel burnup.

Experimental fuel rods with mixed uranium-plutonium nitride fuel are being tested in the BOR-60 reactor as part of collapsible irradiation devices, which are prototypes of fuel rods of the BREST-OD-300 reactor being designed. Currently, reactor tests of fuel rods with a helium sublayer as part of OU-2, as well as lead alloyed with components of EP823-Sh steel as part of OU-4, are continuing. In 2018, one fuel rod was removed from OU-2 and one OU-4 after the first stage of reactor tests with the rods replaced by others from those previously irradiated and post-reactor studies were performed. Information on the features of the radiation swelling of the uranium-plutonium nitride fuel of these fuel rods is presented.

The results of an investigation of fission gas release from mixed nitride fuel after irradiation to maximum burnup 7.5% h. a. in 10 experimental FA in BN-600 are reported. It is shown that gas release commences at burnup >2% h.a. and increases with increasing burnup. The burnup dependence of gas release for mixed and uranium nitride fuel coincide quite well. The fission gas beneath the fuel-rod cladding consists of three basic components: xenon, helium, and krypton. The content of the remaining gases (N2, Ar, O2, CO2) does not exceed tenths of a percent. Xenon has the highest specific release, followed by helium and krypton. Helium has the highest relative release (>40%). Xenon and krypton have almost identical release not exceeding 20% for most of the investigated fuel rods. A higher release of fission gas is associated with either an increase of the fuel temperature (>1500°C) or features of fuel microstructure.

The automated system ASCRO developed and adopted in the nuclear industry for monitoring radiation conditions and on-site monitoring of subsoil status is continually improving, and the Rosatom State Corporation is continually expanding the sphere of activity of these systems. In the present article, the experience gained in the implementation of on-site monitoring of subsoil status in 2017 and the principles of its integration into the complex environmental monitoring system are expounded. Data on the development and integration of information-geoenvironmental packets of nuclear industry enterprises into the complex environmental monitoring system are presented.

Reactor tests in BN-600 and post-reactor studies of experimental FA of fuel rods with mixed nitride fuel and different cladding are being conducted in Project Breakthrough. In the present article, the irradiation parameters of fuel rods with mixed nitride fuel and cladding made from ChS68-ID c.d. austenitic steel are presented: maximum linear power density 38.3 kW/m, maximum fuel burnup 7.5% h.a., and maximum damage dose 73.6 dpa. No depressurization of the fuel elements was recorded. The results of post-reactor examination (PIE) of the irradiated fuel rods are presented – swelling, gas release from fuel, and mechanical and corrosion characteristics of the cladding.

The complex program of computational and experimental validation of dense fuel for fast reactors in Project Breakthrough provides for testing of fuel rods with mixed uranium-plutonium nitride fuel in the BOR-60 and BN-600 reactors. The irradiation parameters and post-reactor studies of the fuel rods of a BREST reactor are presented. Testing of fuel rods in BOR-60 to maximum burnup 5% h.a. and damaging dose 74.8 dpa and in BN-600 to maximum burnup 4.5% h.a. and damaging dose 53 dpa did not result in depressurization of the rods. The results of post-reactor tests of irradiated fuel rods – swelling of fuel, gas release from the fuel, mechanical and corrosion properties of the cladding – are presented.

The environmental and economic efficiency of different energy sources – coal, gas, shale gas, oil, wind, sun, water, nuclei, and others – is examined. The environmental indices of different methods of power generation are presented, the contribution of different types of energy carriers to electricity production and their prospects are indicated, and the environmentally most acceptable energy sources for sustainable development are named. A comprehensive index of the environmental impact is proposed.

Mixed uranium-plutonium nitride fuel is seen as a promising fuel for fast reactors with a closed fuel cycle. A complex program of computational and experimental validation of this fuel up to 2020 has been developed within the scope of Project Breakthrough. Pre-reactor and post-reactor studies of mixed uranium-plutonium nitride fuel are part of this program. The basic research results obtained by the end of 2016 are examined.

The parameters of the neutron flux in the experimental channels of VVR-Ts are determined. It is shown that to evaluate the 99Mo yield it is necessary to take account of, aside from the neutron flux, the dependence of the fission cross section of 235U on the temperature of the neutron gas. The blocking coefficient is calculated for an experimental uranium target. The computed radiochemical yield of 99Mo is compared with the experimentally measured value. It is shown that the average yield equals 0.75. A correlation is found between the neutron flux density, which is directly proportional to the power of the VVR-Ts nuclear reactor, at the target location and the radiochemical yield of 99Mo.

In the Proryv project, mixed uranium-plutonium nitride fuel is to be used in the future nuclear power with fast reactors with sodium coolant (BN-1200) and with lead coolant (BREST). At present, experimental fuel elements and fuel assemblies with mixed nitride fuel are fabricated by means of a technology developed at the Bochvar All-Russia Research Institute for Inorganic Materials for testing in the BN-600 reactor and the MIR and BOR-60 research reactors. The program of reactor tests of the experimental fuel assemblies and some preliminary results of these tests are presented. World experience in research on mixed uranium-plutonium nitride fuel is limited to ~200 fuel elements, irradiated to maximum burnup ~18% h.a. at maximum linear power density 410–1300 W/cm [1, 2]. In our country, we have the results of the irradiation of ~1300 fuel elements with uranium nitride fuel in the BR-10 reactor [3]. Even though many properties of uranium nitride and mixed nitride fuel are close, the results of the research cannot fully serve as validation for the BREST and BN-600 fuel elements, for which the operating parameters determining their serviceability differ considerably from those of fuel elements with uranium nitride fuel. The results of the irradiation of four fuel elements with mixed nitride fuel to maximum burnup 12.1% h.a. in the BOR-60 reactor as part of a Russian–French experiment must also be taken into account [4]. Low swelling and gas-release and the corrosion damage to the cladding in this experiment were obtained for high-purity mixed nitride fuel fabricated from the initial metals with oxygen content

For future nuclear power with fast reactors with sodium (BN-1200) and lead (BREST) as coolants, plans are being made to use mixed uranium-plutonium nitride fuel as part of the Proryv project. A technology for carbothermal synthesis of mixed uranium-plutonium mononitride fuel and fabrication of pellets from it has now been developed. A program of comprehensive computational and experimental research on fuel to study the initial properties and their change under irradiation in reactor experiments is being implemented. Some results of pre-reactor studies of the properties of fuel and a program for future research are presented.

This article is devoted to the development of the production of 99Mo used in nuclear medicine. The stages in the development of the production of 99Mo and molybdenum-technetium generators are shown. Production has increased several-fold over the period 2009–2013 and meets the demand of domestic nuclear medicine. The problem of recycling the uranium from an irradiated target after 99Mo is extracted from the target is important and strongly influences the price of 99Mo. A technology for reprocessing uranium-containing tailings with the uranium obtained being returned into the 99Mo production cycle has been developed at the Scientific-Research Institute for Physical Chemistry. This has eliminated the need to purchase expensive high-enrichment uranium for targets and has made it possible to use the uranium remaining from the reprocessing of irradiated targets.

The present status of research on dense nitride fuel for liquid-metal cooled fast reactors is examined. It is noted that within the framework of project Proryv nitride fuel is to be used in reactors with sodium (BN-1200) as well as lead coolant (BREST). Carbothermal synthesis of nitride from oxide powders has been chosen to fabricate the experimental fuel elements. Several experimental assemblies with mixed uranium-plutonium nitride fuel have been fabricated for tests in BOR-60 and BN-600. It is shown that the existing global experience in studying nitride fuel is insufficient for validating such fuel for use in BN-1200 and BREST. A complex program for computational-experimental validation of fuel elements is being implemented for validating the serviceability of fuel elements.

It is shown that simulation experiments on the heavy-ion irradiation of structural materials used in nuclear power facilities can be used for tomographic atomic-stand studies. Experiments have been performed on irradiation of samples of advanced steel used in nuclear and fusion reactors EK-181 and ODS Eurofer by iron ions to damaging dose 10 dpa. The steel is characterized by the presence of a large number of nanosize (2–3 nm) clusters, dispersion-hardening these materials. Irradiation was performed on the TIPr-1 heavyion linear accelerator. Atomic-probe studies of irradiated samples reveal a change in the composition of nanosize clusters under irradiation.

A γ-ray detector based on pressurized xenon is described, and the characteristics of the processing of the spectrum of the detector are examined. It is shown that the SpectraLine computer program, where these characteristics are taken into account, makes it possible to process successfully the spectra measured with a pressurized-xenon detector, which greatly expands the possibilities of instruments of this class for performing spectrometric measurements. The possibilities of using pressurized-xenon detectors to solve complex problems of gamma-spectrometric analysis using the GammaLab program system are analyzed. The results obtained show that it is indeed possible, in principle, to use a pressurized-xenon detector together with the SpectraLine computer program for quantitative analysis of samples with a complex radionuclide composition, specifically, to determine the isotopic composition of plutonium.

Tests of prefabricated VVER fuel elements burnup 50–60 MW·days/kg in regimes with cyclic power variation have been performed in a circuit setup of the MIR research reactor. The testing procedures are described, and the designs of the irradiation setups are presented. Some fuel elements are equipped with sensors for performing in-reactor measurements, which yielded the experimental data on the variation of the fuel element parameters during the tests (gas pressure, fuel temperature, length). Some results of post-reactor materials-engineering investigations are presented. All fuel elements remained airtight.

Lead–bismuth and lead coolants have special features which make it possible to regard them not only as coolants for nuclear power systems but also as the working body for obtaining valuable technological products. In this article, computational and experimental results concerning the production of hydrogen, synthesis gas, fresh water, nanomaterials, water vapor, and petroleum distillates in the processes and apparatus used for direct-contact heat transfer from liquid-metal coolants to various liquids and gases by direct mixing in the absence of intermediate solid heat-exhange surfaces are presented. The results substantiate the possibilities and prospects for developing nuclear-technological systems for the production of heat, electricity, and valuable technological products.

A pyroelectrochemical process for reprocessing spent fuel and fabricating granular oxides UO2, PuO2 or (U, Pu)O2 from chloride melts has been developed at the Scientific-Research Institute of Nuclear Reactors for a prospective nuclear fuel cycle. The basic equipment has been developed. The basic results of a comprehensive study of fuel elements with vibrationally compacted (U, Pu)O2 fuel for fast reactors are presented. The performance of the reactors remains high up to 30% burnup in standard BOR-60 reactor fuel assemblies and 32% burnup in experimental fuel elements. An assessment is made of the effectiveness of the pyroelectrochemical methods and vibrational compaction technology for plutonium utilization.

The current status and prospects of work on the Moscow meson factory are described. The maximum energy obtained at the present time is limited by the existing powerful klystrons and is 247 MeV. The maximum beam intensity is 150 μA. The accelerator beam is used for fundamental research and for solving applied problems. The average duration of the sessions is 2000 h per year. Three of the five branches of the proton channel which are provided in the design are in operation in the experimental complex. This includes the branch on the multifunctional neutron complex, which is to be used for, among other things, research on nuclear transmutation of long-lived fission products, Np, Am, Cm, and the electronuclear method of generating electricity. The first phase of the proton therapy complex is under construction.

Certain problems associated with the design and construction of and the results of work performed on large-scale safety test stands for water-moderated water-cooled, channel, and boiling-water vessel reactors are presented. It is shown that inadequate adherence to simulation principles in experimental setups can result in incorrect results concerning the effectiveness of emergency core-cooling systems in reactors. The results obtained in this country and abroad show that a successful design, technological effectiveness, and reliability of all components of the model of a reactor core prevent the loss of time and resources in performing experimental programs. For example, such losses resulting from frequent failures of fuel element simulators, which are technologically highly effective and critical components of test stands, can be eliminated.