Simulation of high-temperature swelling of uranium dioxide and deformation behavior of a fuel element

Models and computer codes, developed based on them, for simulating the swelling of uranium dioxide (BARS) and the stress-deformation state of a fuel element (SDS) under high-temperature irradiation are presented. It is shown that when developing a design for high-temperature fuel elements and validating their serviceability the quantitative indicator required for the swelling of uranium dioxide in the range ≥1400°C is the change in the external dimensions of the fuel caused by constant formation and growth of bubbles containing gaseous fission products during irradiation. The results of computational investigations using the models indicated are examined. These results eliminate the inconsistency of the data on the effect of the main operating parameters — the temperature and burnup — on the radiation characteristics and service life behavior of a fuel element. It is shown that the central channel in the fuel kernel and strengthening of the cladding improve the dimensional stability fuel elements. __________ Translated from Atomnaya énergiya, Vol. 103, No. 3, pp. 172–179, September, 2007.     

Radiation resistance of high-density uranium—molybdenum dispersion fuel for nuclear research reactors

The results of radiation tests are discussed and the character of the failure of fuel compositions and the operability of fuel elements is analyzed as a function of the type of fuel and the irradiation conditions. The intense interaction of the fuel with the matrix material is considered as the main factor limiting the operability of fuel elements in power-dense high-flux nuclear reasearch reactors. It is concluded that low-enrichment high-density uranium—molybdenym fuel can provide reliable operation of dispersion fuel elements in low-and medium-power research reactors. Such fuel can be used in power-dense high-flux research reactors if the fuel load is decreased below the maximum admissible amount, the compatibility of the uranium—molybdenum alloy with an aluminum matrix is radically improved, or fuel elements with a different construction, for example, monolithic, are used. __________ Translated from Atomnaya énergiya, Vol. 100, No. 1, pp. 35–44, January, 2005.     

Materials-technology investigations of fuel elements, irradiated in a reactor at the obninsk nuclear power plant, after standing for 38 years in a depository

The results of materials-technology investigations of a spent fuel assembly from a reactor at the Obninsk nuclear power plant, the first nuclear power plant in the world, before the rated burnup and after prolonged dry storage (for about 40 years) were presented. It was established that the fuel elements from the fuel assembly studied are in satisfactory condition. No appreciable damage due to the prolonged storage was found: the outer diameter remains within the technological tolerance limits and the strength and the plasticity of the jackets are high. Only surface corrosion damage to 10 μm depth was found on the fuel-element jackets. The fuel composition remained whole. 6 figures, 1 table, 3 references. State Science Center of the Russian Federation—A. I. Leipunskii Physics and Power-Engineering Institute. Translated from Atomnaya énergiya, Vol. 88, No. 3, pp. 183–188, March, 2000.     

Preparation and removal for reprocessing of spent VVR-2 and or nuclear fuel from the Russian Science Center Kurchatov Institute

The details of the preparation and removal of spent nuclear fuel from the Institute’s VVR-2 and OR research reactors for chemical reprocessing are presented. The spent fuel is represented by fuel assemblies which have different shapes and contain EK-10 fuel elements with similar construction and UO₂–Mg 10% enrichment kernels or S-36 fuel elements with U–Al alloy kernels with 36% enrichment. The storage conditions for the spent fuel are described. The details of the procedures developed to identify fuel assemblies by type of fuel elements are presented. The choice of the TUK-19 shipment container for loading and transporting spent fuel for reprocessing is validated. The details of the loading of spent fuel assemblies into TUK-19 are described; these operations are performed by workers under a protective layer of water in a handling room specially designed for such purposes. Translated from Atomnaya énergiya, Vol. 106, No. 4, pp. 201–209, April, 2009.     

Investigation of fuel structure and composition after irradiation in VVÉR-440 to burnup 70 MW·days/kg

The results of structural investigations performed on fuel and fission products — neodymium, xenon, and cesium — along the radius of a fuel kernel after irradiation in VVéR-440 to burnup 70.2 MW·days/kg are presented. The radial distribution of neodymium is used to calculate the radial distribution of burnup and the accumulation of xenon and cesium. It is shown that a decrease of the xenon content in the fuel matrix as compared with the amount formed over the irradiation time is observed over the entire cross section of the pellet and is due to complete or partial fuel recrystallization occurring predominately along the boundaries of the initial grains and characterized by the formation of a fine-grain structure together with submicron and micron pores. __________ Translated from Atomnaya énergiya, Vol. 101, No. 4, pp. 286–289, October, 2006.     

Safety assessment methodology and analysis of production during plutonium recycling at the State Science Center of the Russian Federation—Scientific-Research Institute of Nuclear Reactors

The methodology and criteria for safety assessment of nuclear fuel cycle technological processes are proposed, substantiated, and checked in large-scale recycling of plutonium (500 kg). The results of comprehensive investigations of the radiation-ecological conditions during the experimental production of mixed uranium-plutonium fuel and fuel assemblies at the State Science Center of the Russian Federation— Scientific-Research Institute of Nuclear Reactors are presented. A methodology and an experimental data bank can be used for safety assessment of commercial recycling of plutonium and Np, Am, and Cm in the nuclear fuel cycle. 4 figures, 3 tables, and 13 references. State Science Center of the Russian Federation—Scientific-Research Institute of Nuclear Reactors. Translated from Atomnaya énergiya, Vol. 87, No. 4, pp. 266–275, October, 1999.     

Thermophysical properties of irradiated uranium-zirconium fuel

The dependence of the thermophysical properties of metallic nuclear fuel — the alloy Zr-40U — in a wide temperature range on the amount of fission products accumulated is presented. Non-irradiated and irradiated samples with different degree of accumulation of fission products — 0.4, 0.6, and 0.9 g/cm³ — are investigated. The specific heat is measured in the range 50–1000°C, the temperature diffusivity is measured in the range 300–1000°C, and the variation of the dimensions and density of the samples on heating is also investigated. The thermal conductivity in the range 50–1000°C is calculated on the basis of the experimental data. __________ Translated from Atomnaya énergiya, Vol. 108, No. 1, pp. 6–9, January, 2008.     

MIR reactor tests of VVER fuel elements in cyclic power variation regimes

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. __________ Translated from Atomnaya énergiya, Vol. 104, No. 2, pp. 80–84, February, 2008.     

Graphite for the GT-MGR high-temperature gas-cooled reactor

The properties of GR-1 graphite based on plentiful and economical raw material — unfired pitch coke for replaceable fuel blocks of the GT-MGR core — are presented. It is established that the optimal variant of this graphite meets the technical requirements, and its high linear thermal expansion coefficient makes it possible to expect adequate radiation dimensional stability. It is shown that with respect to a set of characteristics GR-1 graphite can be regarded as a candidate material. __________ Translated from Atomnaya énergiya, Vol. 103, No. 4, pp. 235–237, October, 2007.     

HTGR nuclear fuel based on plutonium oxide microspheres

The results of investigations performed at the conceptual design stage of the development of rod-shaped fuel elements based on spherical plutonium dioxide particles with protective coatings (fuel microelements) for a modular HTGR (GT-MGR type) with a gas turbine are presented. The basic requirements for fuel elements and their components, ensuring that the elements are used effectively in a reactor for weapons plutonium utilization, are formulated. Technological-material-engineering investigations have been performed on UO₂ and PuO₂−x fuel cores. The basic parameters of the technological processes have been determined, and the required setups have been developed and put into operation. Certain physical and mechanical properties of fuel elements and their components without irradiation have been investigated, 5 figures, 3 references. Scientific and Industrial Association “Luch”. Translated from Atomnaya énergiya, Vol. 88, NO. 1, pp. 35–38, January, 2000.     

Interaction of a model melt of a reactor core with zirconium dioxide ceramic with different porosity

The results of an experimental investigation of the interaction of zirconium dioxide ceramic with different porosity with a model composition of a core melt are presented. The experiments were performed with melt composition (in mass%) UO₂ 46.6, ZrO₂17.6, and Fe₂O₃ under isothermal conditions at 1800°C in an argon atmosphere. Data were obtained on the rate of erosion of the dense ceramic, the character of the permeation of the pores and the pore morphology, and the distribution of the melt elements along the height of the porous layer, 11 figures, 1 table, 3 references. Deceased. Russian Science Center “Kurchatov Institute.” Translated from Atomnaya énergiya, Vol. 88, No. 4, pp. 266–277, April, 2000.     

Some aspects of nuclear and radiological safety of operating a spent nuclear fuel repository

A model of an irregular situation in a spent nuclear fuel repository with the introduction of excess reactivity into the system, consisting of containers with spent fuel assemblies and water, is examined. The neutron kinetics of a critical system is calculated taking account of the thermohydraulics of the system. The character of the flow of a short-time self-sustained chain reaction — “neutron burst” — is described. It is found that an excursion of the system in the range of reactivity introduction rates examined will result in heating of the system and self-quenching of the chain reaction by negative reactivity effects with respect to fuel temperature. Intense fluxes of fission neutrons and prompt gamma rays, accompanying a self-sustained chain reaction, are formed in the excursion process. A mixed neutron and gamma ray field near the system considered is investigated. __________ Translated from Atomnaya énergiya,Vol. 104, No. 3, pp. 141–147, March, 2008.     

Handling spent nuclear fuel from the research reactors of the Russian Science Center Kurchatov Institute

Shipping out the spent fuel of the research reactors at the Institute for reprocessing is examined. The spent fuel is characterized by a great diversity of structural characteristics of the fuel assemblies and fuel elements, fuel compositions, and the enrichment, burnup, and cool-down times of the fuel as well as the state of the components of the assemblies and the structural materials. A classification and quantitative indicators of the accumulated spent fuel from the standpoint of the modern state of its reprocessing technology and the requirements for delivery to the Mayak Industrial Association are presented. The structural features of the TKU-19 and -128 shipment containers are presented, and the loading of spent fuel assemblies into them for shipment to reprocessing is described. The plans and goals of further work on the removal of spent fuel from the Institute’s territory are presented. Translated from Atomnaya énergiya, Vol. 106, No. 2, pp. 99–105, February, 2009.     

Single-purpose electronuclear installation using natural uranium as fuel

The results of investigations of the parameters of an electronuclear setup, operating on solid fuel, for producing useful power are presented. The objective of the investigations is to show the attractiveness and practical possibility of producing a safe (subcritical) setup for producing energy with unlimited fuel resources. The setup contains an accelerator, two targets, and two blankets. The fissioning isotopes accumulate in one blanket, supplied with natural uranium. Actually, his blanket performs the function of enrichment for the uranium-plutonium fuel cycle in modern nuclear power. Power is generated mainly in the other blanket, which is supplied with fuel assemblies that are extracted from the first blanket. In contrast to reactors operating on natural uranium, in an electronuclear setup a high degree of fuel burnup can be achieved by converting part of the generated energy into neutrons. 1 figure, 9 references. State Science Center of the Russian Federation—Institute of Theoretical and Experimental Physics. Translated from Atomnaya énergiya, Vol. 87, No. 3, pp. 199–204, September, 1999.     

Computational-experimental studies of the neutronics of the SM reactor with different variants of the neutron-trap arrangement

The results of experimental studies of the neutronics of the high-flux SM reactor with different arrangements of the neutron trap are presented. The MCU series of high-precision computer programs implementing the Monte Carlo method is used for computations. Experimental data on reactivity effects, the effectiveness of safety and control rods, and the coefficients of nonuniformity of energy release in the core have been obtained in experiments on a critical assembly – a physical model of the SM reactor – and directly in experiments in the reactor. The error is 4.2–10% in determining the reactivity parameters and 5–10% for the relative energy release in the fuel elements. Information on the neutron field formed in the volume of the neutron trap has been obtained for two arrangements of the beryllium and water moderators. The differential and integral energy spectra of the neutrons in the energy interval from 0.5 eV to 20 MeV are obtained for three points inside the trap (external, central series, center). The flux density of thermal, superthemal, and fast neutrons are determined.     

Nuclear central heating — prospects and solutions

Large-scale development of nuclear central heating — a radical expansion of a sphere of application, large increase of cost-effectiveness and self-financing of the construction of nuclear sources of energy, increase of their fraction in the base part of the load schedule, and large-scale displacement of fossil fuel — is validated. Suggestions for a program for developing nuclear heat and power plants are examined. It is shown that the power generating units of nuclear heat and power plants must satisfy specific requirements, which requires developing specialized reactor systems. The main technical and economic characteristics of an innovative simplified boiling water reactor VK-300, specially designed for central heating power generating units, the parameters of a central heating power generating unit with VK-300, and the results of validation of investments in the construction of the VK-300 nuclear heat and power plant in Arkhangel’sk are presented. __________ Translated from Atomnaya énergiya, Vol. 103, No. 1, pp. 36–40, July, 2007.     

Simulation of the physical effects of the interaction of molten fuel with the coolant

Conclusions The method proposed makes it possible to obtain computational estimates of the intensity of a steam explosion inside a reactor vessel and in the space below the reactor inside the melt trap. The computational investigations of the intensity of a steam explosion inside a VVéR vessel in the most likely scenario of a serious accident with efflux of melt into the bottom pressurized chamber show that under certain conditions a high pressure capable of destroying separate structural elements can develop. The mass of the interacting melt, the initial temperature, the fragmentation time, and the final size of the fragments, as well as the type of contact realized, have the greatest effect on the intensity of the steam explosion. Local steam explosions in pipes of the melt trap have a relatively low intensity and cannot have a large effect on the construction in the space below the reactor and on the containment envelope. Deceased. State Science Center of the Russian Federation — Physics and Power Engineering Institute. Translated from Atomnaya énergiya, Vol. 80, No. 1, pp. 3–10, January, 1996.     

Numerical simulation of the thermohydraulics of reactor fuel assemblies with blockages

Computer codes for analyzing the thermohydraulics of fuel assemblies with blocked coolant flow cross sections are examined. The computational results obtained with these codes are analyzed, and a conclusion is drawn on the basis of this analysis about the most important computational codes in this field. The results of the analysis are illustrated. An extensive literature from the worldwide practice of the thermohydraulics of blocked fuel assemblies is presented. Further work required to improve the computer codes is indicated. 7 figures. 2 tables, 65 references. State Science Center of the Russian Federation—A. I. Leipunskii Physics and Power Engineering Institute. Translated from Atomanya énergiya, Vol. 87, No. 5, pp. 342–356, November, 1999.     

Experimental investigation of the physical properties of salt melts containing sodium and lithium fluorides and beryllium difluoride

Experimental measurements of the basic physical properties of the melt NaF-LiF-BeF₂ are presented as validation of the concept of a molten-salt reactor for burning actinides from spent fuel from light-water reactors. Compositions which are characterized by the minimal molar fraction LiF 15–17% and BeF₂ 25–27% and meet the special requirements for a fuel salt for the concept under study are found. The melts of the fluorides of three metals have an acceptable melting temperature (<500°C), permit dissolution of actinide and lanthanide trifluorides to molar fraction 2% and higher at 600°C, possess good neutron-physical (even without enrichment with respect to ⁷Li) and thermophysical properties, are compatible with nickel-molybdenum alloys to temperature 750°C, are inexpensive and are not strongly activated by neutrons so that they do not present a long-term disposal risk. __________ Translated from Atomnaya énergiya, Vol. 101, No. 5, pp. 364–372, November, 2006.     

Computational method for and analysis of the application of granular absorbers in VVÉR reactors

Questions concerning the compensation of excess reactivity in pressurized-water reactors by using consumable granular absorbers are examined. A method of computing the spatial-energy distribution of the neutrons in cells with a granular absorber is presented. The neutron-physical and thermophysical characteristics of fuel assemblies with fuel elements based on homogenized and heterogeneous arrangements of gadolinium in them are compared. It is shown that granular absorbers have certain advantages, specifically, they decrease the gadolinium content in the fuel elements and at the same time increase the total number of gadolinium-containing fuel elements in the fuel assemblies. This decreases the maximum power released in the gadolinium-containing fuel elements and the temperature of the fuel during the entire run. __________ Translated from Atomnaya énergiya, Vol. 100, No. 1, pp. 8–13, January, 2006.