Uranium Silicide Research Articles

Phase-Field Simulation of Grain Growth in Uranium Silicide Nuclear Fuel

Uranium silicide (U3Si2) is regarded as a viable fuel option for improving the safety of nuclear power plants. In the present work, phase-field simulations were employed to investigate grain growth phenomena, encompassing both isotropic and anisotropic grain growth. In simulations of isotropic grain growth, it is commonly assumed that the energy and mobility of the grain boundaries (GBs) remain constant, represented by average values. The calculated grain growth kinetic rate constant, K, exhibits a close correspondence with the experimental measurements, indicating a strong agreement between the two. In simulations of anisotropic grain growth, the values of GB energy and mobility are correlated with the angular disparity between GBs. The simulation results demonstrated that the growth rate of U3Si2 can be influenced by both the energy anisotropy and mobility anisotropy of GBs. Furthermore, the anisotropy in mobility results in a greater prevalence of low-angle GB distribution in comparison to high-angle GBs. However, the energy anisotropy of GBs does not impact the frequency distribution of the angle difference between GBs.

Exploring the effects of Al and Si dopants on the accident tolerant fuels of UO2 pellets for light water reactor

This study investigates the potential of Al and Si dopants to enhance the performance of UO2 pellets used in accident tolerant fuel for light water reactors. These dopants are anticipated to improve the uranium loading level, fission product retention, and thermal conductivity of UO2 pellets, while preserving their neutronic performance. Utilizing first-principles calculations, this study analyzes the kinetics of UO2-Al and UO2-Si interactions, predicting the resulting compound types, compositions, properties, and doping techniques. The calculations consider the reaction products and their thermodynamic properties. The study proposes the use of uranium aluminide, uranium silicide, or uranium aluminum silicide (e.g., UAl3, USi, U3Si2, or U3Al2Si3) to fill the UO2 grain boundaries, as opposed to Al2O3 and SiO2. This could be achieved by synthesizing uranium aluminide and uranium silicide separately, and then mixing them with UO2.

The effect of fission products Xe and Cs on the thermal conductivity of the U3Si2 lattice: a first-principles study

In the past decades, uranium silicide (U3Si2) as a promising accident tolerant fuel (ATF) has drawn considerable attention in the field of nuclear physics. In comparison with traditional nuclear fuel (UO2), the U3Si2 has higher thermal conductivity and uranium density, thereby resulting in lower centerline temperatures and better fuel economy. However, during the nuclear fission reaction, some unexpected fission products, such as Xe and Cs, are released and form the defective states. In this study, we explore the influence of Xe and Cs on the thermal conductivity of the U3Si2 lattice from 200 to 1500 K using density functional theory calculations combined with Boltzmann transport equation. Our results reveal that the lattice and electronic thermal conductivities of defective U3Si2 are reduced at a constant temperature, as compared with that of ideal system, thus resulting in a decrease of the total thermal conductivity. In the case of Cs occupation at U1 site, the total thermal conductivity (4.42 W mK−1) is decreased by ∼56% at 300 K, as compared with the value of 9.99 W mK−1 for ideal system. With U1 and Si sites being occupied by Xe, the total thermal conductivities (4.45 and 6.52 W mK−1) are decreased by ∼55% and 35% at 300 K, respectively. The presented results suggest that the U3Si2 has potential as a promising ATF at high temperatures.

Metamagnetic transition in single-crystalline UIr_{2}Si_{2}

A single crystal of the ternary uranium silicide UIr_{2}Si_{2} was studied by means of of X-ray diffraction, magnetization, heat capacity and electrical transport measurements. The studied compound orders antiferromagnetically at the Néel temperature of 5.5 K and undergoes a metamagnetic transition at 1.8 K in a field of 1.52 T. The metamagnetic transition has a spin-flop character.

HPLC separation of REE, U and Pu from uranium silicide fuel

ABSTRACT Experimental burn-up determination of spent nuclear fuel samples by mass spectrometry requires separation of major actinides and rare earth fission products. An HPLC method is developed for the separation of uranium, plutonium and REEs from the dissolved solution of spent U3Si2 fuel sample. The task was challenging due to the presence of aluminum and uranium in large proportions. The established ion interaction chromatography-based method was found unsuccessful to overcome the interference caused by aluminum matrix. Extraction chromatography using reversed phase stationary phase modified with HDEHP and TODGA was explored for the separation problem. Good separation among aluminum, REEs, U and plutonium was achieved using a column modified with TODGA. The optimized method exhibited good tolerance to large proportions of aluminum, uranium and nitrate salt present in the dissolved fuel solution.

First-principles calculations to investigate mechanical, thermal, oxidation and hydrogen properties of Y, Zr and Nb alloyed U3Si2

Uranium silicide (U3Si2) is considered a potential accident tolerant fuel for next-generation light water reactors, and the reliability of U3Si2 in extreme environments can be further improved by alloying. In this study, the effects of Y, Zr and Nb alloying on the mechanical and thermal properties, oxidation resistance and hydrogen embrittlement resistance of U3Si2 were systematically investigated by first principles. The enthalpy of formation and phonon dispersion curves indicate that the structures are still thermodynamically and dynamically stable after alloying. The addition of Nb improves the mechanical and thermal properties of U3Si2 and its oxidation resistance. However, U2.5Si2Nb0.5(II) exhibited the worst resistance to hydrogen embrittlement. Furthermore, O and H atoms tend to dissolve in the OI(2) and TI sites, respectively. Alloying with elements of smaller atomic radius leads to a reduction in interstitial volume, which facilitates the bonding of O and H atoms with the interstitial atoms and enhances solubility. Moreover, the addition of Nb strengthens the electron bonding and local hybridization between O and H atoms and U atoms, as revealed by the PDOS, charge density and ELF. The findings provide useful information for the improvement and design of novel accident-tolerant fuel materials.

Fuel cycle depletion validation and code-to-code verification studies for High Flux Isotope Reactor highly and low-enriched uranium fuel designs

This paper documents fuel cycle depletion validation and code-to-code verification studies for the High Flux Isotope Reactor (HFIR) highly enriched uranium (HEU) and proposed low-enriched uranium (LEU) fuel designs. In support of HFIR’s world-leading performance, transport and depletion simulations are performed to ensure safe operations, design and qualify irradiation experiments, enhance core components and irradiation facilities, and design and characterize LEU fuel designs. Identifying well-validated, computationally efficient codes is required for the success of these efforts. The HFIR Controller, Shift, and VESTA codes were deployed to simulate HEU uranium–oxide dispersion fuel cycles at 85, 95, and 100 MW operations, as well as LEU fuel cycles operating at 95 MW with uranium–silicide dispersion and uranium–molybdenum monolithic alloy fuel forms. Excellent agreement between the codes and with experimentally obtained 235U enrichment distributions provides increased confidence in the ability of these codes to model and simulate HFIR’s unique core design.

DETERMINING GAMMA SOURCE IN URANIUM MOLYBDENUM OF FUEL IN G.A SIWABESSY MULTI PURPOSE REACTOR

Nuclear fission reactions produce a lot of radionuclides that release energy, one of which is in the form of gamma radiation. Gamma radiation is produced by various types of radionuclides, and nuclear reactor fuel will produce different values of gamma intensity. Uranium Molybdenum (U7Mo-Al) is the type of nuclear fuel for future research reactors that possesses many advantages. For the application of molybdenum-based fuel, it is necessary to determine the resulting gamma radiation. The purpose is to determine the gamma radiation produced from molybdenum-based fuel with various densities. This study begins with the determination of the mass composition of the reactor component, calculations with ORIGEN2.1, and data output analysis. The U7Mo-Al density was varied, namely 2.96 gU/cm3, 3.85 gU/cm3, 4.44 gU/cm3, 5.43 gU/cm3, 6.91 gU/cm3, and 8.29 gU/cm3. The gamma radiation yield of U7Mo-Al is lower than that of uranium silicide (U3Si2) with the same density of 2.96 gU/cm3. The result will add to the justification for the superiority of U7Mo-Al compared to U3Si2/Al. For U7Mo-Al with densities of 3.85 gU/cm3, 4.44 gU/cm3, 5.43 gU/cm3, 6.91 gU/cm3, and 8.29 gU/cm3, the one that produced the lowest gamma radiation intensity is 3.85 gU/cm3 while the highest is 8.29 gU/cm3. This explains that the intensity of the gamma radiation produced is directly proportional to the fuel density. The low intensity of gamma radiation in molybdenum-based fuel can be used as a suggestion in shielding design to ensure the operational safety of reactors.

Structural, electronic, elastic and thermal properties of Cr-doped U3Si2: A DFT study

Alloying has been considered as an effective strategy to make up for the poor oxidation corrosion resistance of U3Si2. However, the theoretical studies on alloyed U3Si2 are rarely reported. In this study, the structural, electronic, elastic and thermal properties of 10at% Cr-doped U3Si2 were studied using first principles calculations and semiclassical Boltzmann transport theory. Our results show that the U3Si1.5Cr0.5 phase is determined to be thermodynamically, dynamically and elastically stable, and U2.5Si2Cr0.5 phase is dynamically and elastically stable. U3Si1.5Cr0.5 is more stable than U2.5Si2Cr0.5 from the thermodynamic calculation results. From the calculated electronic structure results, a mixture of ionic and metallic bonds dominates the bonding of Cr-doped U3Si2. They are ductile materials compared to the brittleness of U3Si2 according to the results of B/G ratio, Cauchy pressure and Poisson's ratio. Moreover, U3Si1.5Cr0.5 is a nearly elastic isotropic material, while U2.5Si2Cr0.5 exhibits strong anisotropy. The thermal conductivity of Cr-alloyed U3Si2 is dominated by the electronic contribution at high temperature, which is lower than U3Si2 but higher than UO2 at service temperature and increases with temperature. Therefore, in addition to significantly improving the resistance to oxidation and corrosion of U3Si2, Cr doping can also convert the brittleness of U3Si2 into toughness and improve its machinability, albeit at the reasonable expense of some mechanical properties and thermal conductivities, while maintaining the high uranium density of U3Si2 fuel form. The results of this work may provide useful clues for the application and improvement of uranium silicide fuels.

Automated reactor physics analysis framework of High Flux Isotope Reactor low-enriched uranium silicide dispersion fuel designs

The High Flux Isotope Reactor (HFIR) is a versatile research reactor that provides one of the highest steady-state neutron fluxes of any reactor in the world. The HFIR reactor physics team investigated the conversion of the current 93 wt% highly enriched uranium U3O8-Al dispersion fuel to a 19.75% low-enriched uranium (LEU) U3Si2-Al dispersion fuel. The team continuously develops a Python module to streamline the analysis steps required for an LEU core design to ensure reproducible and agile design iteration. The Python module automates the data processing between analysis steps and automates the input perturbation for branch calculations and design changes. The automated framework has proven to significantly increase the efficiency and reproducibility of the reactor physics team to design High Flux Isotope Reactor (HFIR) LEU cores and thoroughly analyze performance metrics, safety metrics, and thermal safety margins. Consequently, the team can now respond rapidly to fuel fabrication engineer and thermal-hydraulic-structural analyst requests.Numerous combinations of LEU fuel designs are explored, of which two LEU fuel designs are presented in this paper: a low density silicide design, and a high-density silicide design. Results show that both designs meet or exceed safety and performance metrics with exception for minor differences caused by the hardened spectrum from LEU.

Separation and determination of boron, chlorine, fluorine and molybdenum in uranium silicide using pyrohydrolysis and ion chromatography

Separation and determination of boron, chlorine, fluorine and molybdenum in uranium silicide using pyrohydrolysis and ion chromatography

The design from scratch to an alternative experimental assembly for the LVR-15 research reactor – A case study in Czech Republic

Project LEU-FOREvER put together a consortium of 9 European stakeholders, including research centers, designers and fuel manufacturers, with the specific aim to secure the nuclear fuel supply for European research reactors. One of the shortage risks has been addressed by the project with the definition of a second supply chain, complementary to the current one, with the aim to fuel medium power research reactors (MPRRs) originally manufactured with the Soviet technology. For this reason, one activity was the definition of a core which could accommodate both the original Russian technology (uranium oxide extruded fuel meat) and new European designed fuel elements (uranium silicide dispersed fuel meat) that are being developed, licensed and qualified in the frame of the 4EVERTEST within LEU-FOREvER project. The Czech reactor LVR-15 was selected for performing the developed assembly irradiation test, which will be followed by a basic post irradiation inspection for the assessment of the irradiation performance. For this reason, the project is proceeding through various logical and operative steps. They are described in the present work, starting with the collection of the relevant information about reactor characteristics by fuel designer, the design phase (performed in parallel with reactor operator in order to tune the codes and data coherently), then the detailed safety analysis study which was submitted to national regulator in order to get the irradiation license approval. The neutronic and thermohydraulic analysis performed shows that the fuel assembly can be safely irradiated in the LVR-15 core. The fuel assembly is being irradiated in the reactor at present. Further work for the next future will consist in collecting relevant data from the irradiation in the reactor and the post irradiation inspection activities.

Assessment of neutronics performance of U3Si, U3Si2, and U–9Mo fuels for the support of MNSR conversion using SCALE 6.2.3 computational tools

In spite of the acceptability and production of Uranium Oxides (UO2) with Low Uranium Enrichment (LEU-UO2) fuel in commercial scale, the use of uranium–silicide (U–Si) in place of the UO2 is one of the promising concepts being proposed to increase the accident tolerance of nuclear fuels. In this study, SCALE 6.2.3 computational tools have been used to model the Nigerian Research Reactor-1 (NIRR-1) which is a typical Miniature Neutron Source Reactor (MNSR) and calculated the core neutronic parameters such as Clean Core Cold Excess Reactivity (CCER), Control Rod Worth (CRW), Shutdown Margin (SDM), Safety Reactivity Factor (SRF) and Neutron flux distribution for the Highly Enriched Uranium (HEU) and candidates LEU cores. The results showed a close agreement between the calculated and measured data. The bias between calculated and measured data was attributed to the computational limitation of this study that was caused by a limited number of particles histories used in criticality calculations. The neutronic performance of the U3Si2–Al, U3Si–Al and U9Mo–Al fuels indicated the suitability of these candidates LEU fuels for conversion of MNSRs from HEU to LEU. The observed reduction in the thermal flux of the candidate LEU fuels was attributed to higher inventory of U-238 in LEU fuels when compared with the HEU and further underscore the need to raise their thermal power by approximately 10% for efficient utilization of the MNSR in Neutron Activation Analysis.

Critical Velocity Experimental Assessment in Flat Plate Fuel Element for Nuclear Research Reactor

Aluminum-coated plates, containing a uranium silicide (U3Si2) meat dispersed in an aluminum matrix, are commonly used in the fuel elements of Material Testing Reactors (MTRs). These fuel elements are typically comprised of narrow channels formed by parallel flat plates, which allow coolant flow to remove the heat of fission reactions. It is important to mention that the thickness of the plates is much smaller than their width and height. The high flow rates needed to ensure efficient fuel-element cooling may cause fuel-plate mechanical failures due to instability induced by the flow in the channels. In the case of critical velocity, excessive permanent deflections of these plates can cause blockage of the flow channels and lead to overheating. An experimental facility that simulates a plate-like fuel element with three coolant channels was developed for this work. The test-section dimensions were based on the Fuel Element design of the Brazilian Multipurpose Reactor (RMB), project being coordinated by the National Commission of Nuclear Energy (CNEN). Experiments were performed to reach Miller's critical velocity condition. This critical condition was reached at 14.5 m/s leading to consequent plastic deformation of the fuel plates

I2S-LWR Activation Analysis of Heat Exchangers Using Hybrid Shielding Methodology With SCALE6.1

The Integral Inherently Safe Light Water Reactor (I2S-LWR) concept developed by a team lead by Georgia Tech is a novel PWR reactor delivering electric power of 1000 MWe while implementing inherent safety features typically reserved for Generation III+ small modular reactors. The main safety feature is based on an integral primary circuit configuration, bringing together a compact core design with 121 fuel assemblies (FA), control rod drive mechanism (CRDM), 8 primary heat exchangers (PHE), 4 passive decay heat removal systems (DHRS), 8 pumps, and other integral components. A high power density core based on uranium silicide fuel (U3Si2) is selected to achieve high thermal power which is extracted with PHEs placed in the annular region between the barrel and the vessel. The compact and integrated design of I2S-LWR leads to activation of integral components, mainly made from stainless steel, so accurate and precise Monte Carlo (MC) simulations are needed to quantify potential dose rates to personnel during routine maintenance operation. This shielding problem is therefore very challenging, posing a non-trivial neutron flux solution in a phase space. This paper presents the performance of the hybrid shielding methodologies CADIS/FW-CADIS implemented in the MAVRIC sequence of the SCALE6.1 code package. The main objective was to develop a detailed MC shielding model of the I2S-LWR reactor along with effective variance reduction (VR) parameters and to calculate neutron fluence rates inside PHEs. Such results are then utilized to find the neutron activation rate distribution via 60Co generation inside of a stack of microchannel heat exchangers (MCHX), which will be periodically withdrawn for the maintenance. 59Co impurities are the main cause of (n,γ) radiative gamma dose to personnel via neutron activation since 60Co has half-life of 5.27 years and is emitting high energy gamma rays (1.17 MeV and 1.33 MeV). The developed MC model was successfully used to find converged fluxes inside all 8 stacks of PHEs with respect to MC statistics using the FW-CADIS methodology. For that purpose the SN module Denovo, based on forward-adjoint transport theory, was used to find VR parameters (importance map and biased source) to effectively bias MC simulation. Further research is required to account for other activation pathways, i.e. isotopes of iron which may generate 59Co through neutron activation and beta decay.

Oxidation of U3Si2: The role of exothermic energy

Uranium silicide (U3Si2) is one of the candidate accident tolerant fuel materials under consideration for adoption in power producing nuclear reactors. The development of this fuel material including the bulk scale manufacturing route, and factors which may affect this route, are still under investigation. One such factor is the influence of oxygen on the material and the consequent release of exothermic energy. This work uses combined thermogravimetric and differential scanning calorimetry (DSC) methods to study oxidation of arc melted U3Si2 material, including ramp tests, isothermal experiments and varying particle sizes. It is found that there is a difference in mass gain profile between fragments and small powder samples, with fragments delivering the same heat load (∼4.6 ± 0.5 kJ.g−1) over a significantly shorter period of time. The measured exothermic heat output is in agreement with thermodynamic modelling once corrections have been made for the presence of elemental silicon in the reaction product. The onset temperature of rapid oxidation also reduces with particle size. Initial reactivities of freshly prepared powder samples during isothermal experiments give appropriate reaction constants, which are used to construct an Arrhenius plot. Supporting DSC data indicate that one of the higher temperature data points (275 °C in this instance) had exhibited exothermic energy-induced self-heating and an accelerated rate of reaction. The resultant calculated activation energy from oxidation of freshly prepared U3Si2 powder is approximately 36 kJ.mol−1.

Structural, electronic, mechanical and thermodynamic properties of U–Si intermetallic compounds: A comprehensive first principles calculations

Uranium silicides are proposed as the prominent accident tolerant fuels for the light water reactors (LWR) due to their high metal density and high thermal conductivity. Among the U–Si alloys, the alloy with high U:Si ratio is more favorable for nuclear fuel application due to the higher uranium density. Thus, the crystal structure, mechanical property, electronic structure, phonon band structure and thermodynamic property of U3Si, U3Si2 and USi compounds, along with the thermodynamic reaction between U–Si intermetallic compounds are systematically studied in our work. The optimized structural parameters of these U–Si alloys are comparable with previous results. Besides, all of them are metallic in nature. Since the calculated elastic constants satisfy the Born stability criteria, one can know U–Si alloys are mechanically stable. The phonon dispersion curves are obtained based on the density functional perturbation theory (DFPT). Accordingly, various thermophysical properties, such as Helmholtz free energy, heat capacity, internal energy and entropy are calculated. Furthermore, the reaction energies related to the formation of U3Si, U3Si2 and USi as well as transformation between them are calculated. It is revealed at the same chemical environment the reaction to form USi occurs more easily, whereas the high temperature and sufficient uranium environment are more in favor for fully silicification of uranium metal into U3Si. Theoretical investigation of this work is expected to provide some new insights for the application of uranium silicides as nuclear fuels and future exploration on the design and synthesis of new-type uranium silicides.

The effect of adding impurities to the high density fuel matrix on the reactivity of the RSG-GAS core

The safety calculation of the RSG-GAS reactor core needs to be done if new high-density fuel is used in the RSG-GAS core. The use of new high-density fuels is very much needed in the RSG-GAS core so that it is necessary to calculate its safety. When using high density fuel there will be a “bottle neck” effect where there is a problem with the fuel and cladding interaction. According to previous researchers this can be overcome by adding metal elements Si, Ni or Ti to fuel. However, with the addition of metal elements Si, Ni or Ti to the fuel, it needs to be analyzed in the RSG-GAS core. The analysis was carried out calculations using a computer program by simulating the Si metal used 1%, 2%, 3% 4% and 5% in the fuel. The calculation of the additional metal element in the high-density fuel effect on the reactivity of the RSG-GAS core was carried out with the WIMSD-5B and Batan-FUEL programs. The WIMSD-5B program is a computer program consisting of several modules including the transport module which is used for the calculation of the generation of macroscopic constants for the 19.75% enriched uranium silicide fuel and the density of 4.8 gU/cc as a function of degree of burn. Core calculations were performed using a neutron diffusion program, namely Batan-FUEL. By calculating the core criticality at operating temperature, the core neutronic parameter values are obtained. The results of the analysis show that the amount of Si 5% in the high density fuel matrix can solve the bottle neck problem and the neutronic parameters are still within the safety limit. This value will determine how much the value of the fuel burn up and control rods are in accordance with the acceptance criteria so that the safety analysis can be carried out on the RSG-GAS core using high density. From the analysis, it was found that the addition of the element Si in the high-density fuel matrix did not find any significant changes in core reactivity and neutronic parameter.

Post-transient examination of performance of uranium silicide fuel and silicon-carbide composite cladding under reactivity-initiated accident conditions

• First transient testing of accident-tolerant fuel designs under transient nuclear heating reactivity-initiated accident conditions. • Claddings experienced dimensional changes but maintained primary rod-like geometries. • Elemental species diffusion from the U 3 Si 2 fuel to the zircaloy cladding was observed. • No elemental species diffusion from the U 3 Si 2 fuel to the silicon-carbide cladding was observed. • Through wall thickness cracking observed in the silicon-carbide cladding resulting from pellet cladding interaction from thermal expansion of the pellet. New fuels are being developed for light water nuclear reactors with the initial goal of improving accident tolerance. More recently, these new fuel systems are also being recognized for their potential to support higher fuel burnups and thus improved economics for reactor plants through improved fuel utilization. Fuel safety testing of these new fuels systems is being conducted under transient irradiation conditions in the Transient Reactor Test facility at Idaho National Laboratory. A test campaign including two rodlets of U 3 Si 2 fuel in Zircaloy-4 cladding and two rodlets of U 3 Si 2 fuel in Silicon-Carbide Composite cladding has been performed to begin the development of fuel-safety criteria in design-basis reactivity-initiated accident conditions. Post-Transient-Irradiation Examinations have been performed and found that the accident tolerant materials maintained their primary rod-like geometry. Elemental species diffusion was found to occur in the Zircaloy rodlets indicating rapid diffusion kinetics at the temperatures achieved. No elemental species diffusion was found in the Silicon-Carbide rodlets. However, through-wall thickness cracks were observed due to displacement loading of the cladding from thermal expansion of the fuel pellets. This work presents the first experimental evidence of the performance of these fuel systems under transient nuclear heating and representative reactivity-initiated accident energy depositions.

On the Development of Tolerant Fuel --- a New Generation Fuel

The concept of tolerant fuel is considered as applied to water-cooled power reactors. The concept is based on eliminating the steam-zirconium reaction. For this, two work areas, i.e., using the physical and the thermodynamic barriers, were considered. Physical barrier presupposes exclusion of contact between water and zirconium, and the thermodynamic barrier (the most radical method) envisages replacement of the alloy containing zirconium with other materials inert to water when exposed to high temperature in the reactor core (∼ 1200 °C). Consequences of the most devastating accidents at the nuclear power plants in the world were discussed: Three Mile Island, Chernobyl and Fukushima. The latest accident in Japan brought to the fore the concept of tolerant nuclear fuel, i.e., being resistant to accidents. Work orientation in creating the tolerant fuel is indicated. Main attention is paid to materials and technologies applied to tolerant fuel. General requirements to safety analysis of the reactor facility fuel system currently developed in the Russian Federation and abroad, as well as current safety criteria for fuel elements, under design-based accidents are presented. Procedure for calculating justification of the safety criteria fulfillment for fuel elements under design-basis accidents is briefly considered. Main characteristics of the new generation materials under development for reactor cores as applied to tolerant fuel are presented. Based on comparing the proposed materials as the tolerant fuel for the fuel element claddings, composite materials based on the heat-resistant SiC/SiC ceramic system could be recommended, and as far as fuel materials are concerned --- materials with increased density, uranium capacity and thermal conductivity values, i.e., nitride fuel and fuel made of uranium silicide