Fuel Element Cladding Research Articles

Оценка возможности использования пакета COMSOL для моделирования теплогидравлических процессов в реакторах ВВЭР

Thermophysical modeling software packages are widely used for modeling various thermohydraulic processes including modeling nuclear reactor fuel element. Verification of such calculations is possible only if we compare the results obtained in different packages (cross-verification) due to the extreme complexity or impossibility to carry out analytical calculations or experiments. It seems relevant to carry out cross-verification of calculations of WWER reactor fuel elements in COMSOL Multiphysics software package using another popular ANSYS software package. Models of fuel elements for the analysis of thermohydraulic processes are developed using thermophysical modeling packages ANSYS Fluent and COMSOL Multiphysics for cross-verification of calculation results in the latter. Modeling of temperature fields is carried out based on the heat conduction equation with an internal heat source. The authors have considered several formulations of problems: with fuel element cladding and without it and the central hole in the fuel. Models of fuel elements of the WWER-1000 reactor have been designed for three formulations of problems with different designs in ANSYS Fluent and COMSOL Multiphysics software packages. Temperature fields have been determined using the finite element method. The results of the calculations of the fuel core maximum temperature are presented and compared for two software packages. The maximum relative deviation of the temperatures obtained is 0,97 %. A comparison of the temperature calculations results obtained using models developed in two packages shows that the relative deviation of the values corresponds to the error of the measuring instruments. It confirms that COMSOL Multiphysics can be successfully used to simulate processes in WWER reactors along with ANSYS Fluent.

Construction of a Strain-Hardening Diagram to Analyze the State of Stress in the Fuel-Element Cladding Material

Construction of a Strain-Hardening Diagram to Analyze the State of Stress in the Fuel-Element Cladding Material

The modeling of interaction between fuel element cladding and steam

This work simulates a close to emergency case in the reactor core - the material of fuel element cladding is subjected to steam flow. The presented model takes into account several processes occurring together - the admittance of oxygen and hydrogen, the propagation of heat, diffusion and mechanical disturbances, chemical reaction. The transfer mechanism of oxygen and hydrogen under the action of the stress gradient is taken into account. The model is suitable to describe the processes at very small times, corresponding to short thermal impulses and initial stages of surface interrelation with steam flow. The all short-time accompanying processes (10-9 −10-12 s) affect the changes in the material exposed to steam contact, but they are not easily accessible for direct observation. The problem was solved numerically. The temperature in our numerical experiment is not too high, but it is already sufficient for introduction of hydrogen and oxygen into the Zr-1%Nb surface, as well as for oxidation of zirconium. Examples of calculations are presented. It shows how different-scale processes interact with each other and to what distortions in the distributions of all interacting quantities this leads. Examples of calculations for different durations of vapor interaction with cladding surface are presented. It is shown that the concentration of oxygen and hydrogen in Zr alloy increases when the vapor action is too long-term.

Comparison of thermal expansion path of the Fe2Mo Laves phase calculated using the DFT-based phonon and Debye-Grüneisen approaches

Precipitation of the intermetallic Fe2Mo Laves phase at the interface between the nuclear fuel and the fuel element cladding can significantly weaken the strength characteristics of the cladding and fuel. Despite the importance of designing materials for the cladding of fuel rods, the thermodynamic properties and the trajectory of the thermal expansion of the Fe2Mo remain poorly understood. The thermodynamic properties of the Fe2Mo have been studied using the finite-temperature quantum mechanical calculations within the frame of the density functional theory under the quasiharmonic approximation. The vibrational contribution to the free energy was obtained using phonon calculations. The thermal expansion of Fe2Mo was predicted by comparing between free energies calculated in different directions. A path with the least energy was chosen as the trajectory of thermal expansion. The obtained result was compared with the direction calculated in previous theoretical work used the Debye–Grüneisen approach and accounted magnetic subsystem to calculate the vibrational and magnetic contributions to the free energy. This comparison reveals that these two approaches are in good agreement with each other. The work shows that the Fe2Mo possesses a non-isotropic thermal expansion. The heat capacity and volumetric expansion at constant pressure are modeled. The calculated results analyzed and are in satisfactory agreement with the experimental data. Obtained results can be useful for further design of fuel element cladding materials intended for generation IV reactors, the operating temperature of which should be above 873 K.

Local Disturbance of the Water-Chemical Regime as a Cause of Increased Oxidation of Cladding of VVER-1000 Fuel Elements

Local Disturbance of the Water-Chemical Regime as a Cause of Increased Oxidation of Cladding of VVER-1000 Fuel Elements

High-temperature oxidation of zirconium alloys with magnetron sputtered chromium coatings

Loss-of-coolant accidents (LOCAs) often lead to dire consequences. That’s why the global community are actively engaged in research aimed at creating accident tolerant fuels. One of the areas of such research is preserving zirconium alloys as the material for fuel-element claddings by creating corrosion-resistant coatings that would be in contact with the coolant. Chromium could be a promising material to be used in such protective coatings as chromium oxide (Cr2O3) serves as an effective barrier for oxygen both in normal operation and in case of LOCA. Chromium coating hinders oxygen diffusion into the metal substrate and thus prevents embrittlement and failure of the fuel-element cladding. This paper describes the results of research studies that have been carried out in recent years and that look at the resistance of magnetron sputtered chromium coatings to high-temperature oxidation in water vapour up to 1,500 oC. The paper demonstrates advantages and drawbacks of chromium coatings in hightemperature oxidation conditions. The focus is on understanding how regimes of magnetron sputtering influence the resistance of chromium coatings to high-temperature oxidation and how the structure and phase state of chromium coatings are related to their properties. The authors describe the optimal regimes of magnetron sputtering for obtaining dense coatings and examine the effect of the substrate temperature and the bias voltage on the structure and density of resulting coatings. High Power Impulse Magnetron Sputtering (HiPIMS) serves as an effective technique that helps enhance the density of coatings. The conclusion drawn is that in order to broaden the temperature range in which chromium coatings can effectively protect zirconium alloys from failure up to 1500 oC, there should be a diffusion barrier between the surface of the fuel-element cladding and the chromium coating. The authors would like to thank B. A. Kalin, supervisor of this project, who had passed away before this publication was made. The staff of the Laboratory of Ion-Plasma and Ion-Beam Machining of Materials, a part of Department No. 9 at the National Research Nuclear University MEPhI, devote this paper to his memory.

Decomposition of hydrogen solution in the neighborhood of the coolant saturation temperature as a possible cause of increased oxidation of VVER fuel element

The analysis carried out in this work has shown that a sharp local decrease in the hydrogen concentration and a corresponding increase in the radiolytic oxygen content is caused by two phase transitions occurring in the reactor coolant at successive temperatures. Decomposition of hydrogen solution in the coolant in the neighborhood of the saturation temperature is followed by the liquid–vapor phase transformation. In this case, the coolant may in some zones deviate from the strict water-chemistry regime aimed to suppress radiolysis. It has been shown that the hydrogen concentration in the coolant may decrease by orders of magnitude in a short time. In this case the efficiency of radiolytic oxygen suppression reduces drastically, and concentration of the radiolytic oxygen in the coolant may increase considerably, thus promoting local oxidation of VVER fuel element claddings in E110.

Separation of ultrafine uranium oxide particles from zirconium alloys in remelting processes with fluoride and oxide slags: thermodynamic analysis

This paper examines how ultrafine particles of uranium oxide interact with zirconium alloys in the presence of oxide and fluoride slags. The authors analyzed the probability of the components of Zr – 1% Nb alloy contaminated with ultrafine uranium oxide particles transitioning into slag phase during the process of remelting fuel element claddings using different fluxing agents. By means of thermodynamic modelling, the authors looked at the interaction of the cladding material of fuel elements used in thermal reactors (zirconium-niobium alloy) contaminated with traces of nanosized uranium oxide with molten oxide and fluoride fluxes in the temperature range of 1,000 to 2,000 oC. It was found that, for the interaction of Zr – 1% Nb alloy with fluoride and oxide melts, the metal should be slightly dissolved in the slag phase. Analysis of the interactions of ultrafine uranium compounds with slag mixtures for the temperature range of 1,000 to 2,000 oC showed a predominant transition of uranium, when reacting with fluoride fluxes, into the metal phase in its elementary form – U and into slag phase (about 4%) in the form of UF3 and UOF2 compounds. During reactions with oxide fluxes, uranium mainly distributes in the metal in its elementary form — U. At the same time, 18 to 19% of uranium can transition into slag phase in the forms of UO, UO2, and UOF2. It should be noted that that fact that uranium oxide particles have an ultradispersed structure should contribute to their transition into slag and thus help speed up the treatment process.This research was carried out in the framework of the Single Industry-Specific Plan by Rosatom; Project No. EOTP-TTsPM-263. Equipment owned by the shared knowledge centre Ural-M was utilized for this research.The authors would like to thank T.V. Osinkina and A.S. Russkikh, junior researchers at the Laboratory of Rare Refractory Metals of the Institute of Metallurgy at the Ural Branch of the Russian Academy of Sciences, for their contribution.

Development of Radiation Porosity in the Material of Fuel-Element Cladding Produced by Different Technologies for Fast Neutron Reactors and Its Influence on the Fuel Assembly Service Life in the BN-600 Reactor

Development of Radiation Porosity in the Material of Fuel-Element Cladding Produced by Different Technologies for Fast Neutron Reactors and Its Influence on the Fuel Assembly Service Life in the BN-600 Reactor

РАСЧЁТНАЯ ОЦЕНКА ТРЕБУЕМОГО ТЕПЛОВЫДЕЛЕНИЯ В ОПЫТНЫХ ТВЭЛАХ ДЛЯ ДОСТИЖЕНИЯ СВЕРХКРИТИЧЕСКИХ ПАРАМЕТРОВ ТЕПЛОНОСИТЕЛЯ ПРИ ИХ ОБЛУЧЕНИИ В ИССЛЕДОВАТЕЛЬСКОМ ЯДЕРНОМ РЕАКТОРЕ

The paper considers the possibility of achieving the required neutron-physical parameters during irradiation of experimental fuel elements circumfluent with a SCW parameters coolant in the RR IR-8 in order to study the behavior of fuel element cladding under the influence of reactor radiation and coolant with SCW parameters under conditions closest to real operation in nuclear power plant. The possibility of carrying out such experiments in the RR IR-8 will make it possible to contribute to the development of new generation IV nuclear power plants, one of the areas of which is a reactor with super-critical parameters coolant. This direction is promising, as the use of the SCW coolant technology will increase the efficiency of the reactors by 10-12 %. Irradiation experiments in RR IR-8 are carried out using experimental ampoule rigs. During developing the design of an ampoule rig with experimental fuel elements, it is required to provide the necessary heat transfer from the fuel element to the water of the reactor pool, and to maintain the pressure of the SCW coolant by the structure of the ampoule rig vessel. The calculations of heat release in the experimental fuel elements and the accumulation of fluence in the cladding were carried out using the MCU-PTR code with the MDBPT-50 database, which implements the Monte Carlo method. The assessment of heat release in fuel elements during irradiation of an ampoule rig in the cells of the IR-8 core and reflector was carried out. The calculation results of heat release in experimental fuel elements and fast neutron fluxes in the cladding during irradiation in the central channels of fuel assemblies and special blocks made of aluminum, stainless steel, as well as water are presented. An assessment of the contribution of heat release from gamma heating of the ampoule rig structure elements was carried out, which showed that the amount of heat release can be comparable with the energy release due to the fuel fission reaction in an experimental fuel element, which must be taken into account when calculating the temperature irradiation regimes of experimental fuel elements. Obtained results showed the principle possibility of achievement the parameters of the SCW coolant in the ampoule rig on the outer surface of the experimental fuel elements cladding in the RR IR 8.

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

Modeling of elastic-plastic deformation of the ring parts as applied to the analysis of the results of testing the material of the fuel-element cladding

An analytical formula for a smooth description of the tension diagram of EK-181 steel and a method for rearranging the diagram when changing the direction of deformation are proposed for the first time. The process of straightening a quarter of an annular sample and further stretching is numerically modeled. It is shown that the conditional yield strength of the material of the straightened sample is 7.5% less than the actual conditional yield strength of steel. It is shown that the test for pure bending of a cantilever sample in the form of a semicircle with the processing of the bending diagram (by analogy with GOST 3565–80 for torsion) provides an estimate of the conditional yield strength which is 32% higher than the actual yield strength. The possibility of numerical reconstruction of the tension diagram from the diagram of pure bending of a cantilevered semi-ring sample is proved. It is shown that this procedure really gives the value of the conditional yield strength of steel EK-181 with a tolerance for the residual deformation of 0.2%. The analysis of the test procedure for the rings of fuel element cladding and the proposed algorithm for determination of the conditional yield stress of the ring material is carried out. Attention is drawn to the arbitrariness of the choice of the designed load on the two-stage diagram of the diametrical tension of the ring and to the lack of scientific substantiation of the possibility of determining the yield stress on the second part of the diagram. It is shown that this method in the current form contradicts GOST for tensile testing due to the absence of a base with a uniform stress state on the ring. Therefore, the considered method is not recommended for determining the values of the conditional yield strength suitable for strength calculations.

ATWS Analysis for Innovative Lead-Cooled High-Power Fast Reactor with Accident Tolerant Materials of Core

The main purpose of the article is to substantiate the safety of a lead-cooled fast reactor of high power. Thermal power of the reactor is about 5600 MW, electric power is about 2400 MW. The design is similar to the BREST-OD-300 project developed at NIKIET JSC. The safety of a high-power reactor is ensured only with the use of innovative core materials (previously the author proposed to use fuel based on micrograins of UN-PuN mononitride and uranium metal nanopowder, a coolant based on lead of thorium ores, fuel element cladding based on EP823 steel with nanopowder of titanium oxide and yttrium, tungsten-coated cladding are used). Research methods are mathematical modeling of the reactor and reactor emergency modes. The most dangerous emergency modes (ATWS – anticipated transients without scram and their combinations, taking into account the non-simultaneous start) are considered. The studies used the codes developed by the author (Dragon-M, FRISS-2D) and the well-known code for precision neutron-physical calculations (MCU). A safe reactor corresponds to the fulfillment of restrictions for a number of functionals characterizing emergency modes. Studies have shown that when using innovative core materials, any combination of emergency modes does not lead to unacceptable releases of radioactive substances outside the Nuclear Power Plant. The results of the study show the possibility of developing reliable, safe, environmentally acceptable nuclear power based on reactors of this type.

Neutronography of Irradiated Reactor Austenitic Steels

The neutron diffracttion studies of the samples of fuel element claddings made of EK-164 and ChS-600 austenitic steels in the initial state and after operation in a BN-600 fast neutron reactor are reviewed. The samples were investigated in a wide range of fast neutron doses and irradiation temperatures, up to ∼80 displacements per atoms (dpa) and 628°C, respectively. Various microstructural characteristics, such as microstresses, dislocation density, and crystallographic texture are determined using full-profile analysis of neutron diffraction data. At high doses studied in this work, the irradiation temperature is the dominant factor responsible for the dislocation density. The procedures of analyzing and revealing defect types were preliminary developed on specially prepared nickel samples and austenitic alloys upon irradiation in the IVV-2M reactor core at a temperature near 80°С, which provided a low thermal mobility of lattice defects. The results obtained in some first diffraction studies on samples after their operation in the reactor core show that neutron data can be used to characterize the microstructures formed due to irradiation and, at least, estimate the mechanical properties of irradiated materials without their destruction and with no radiation risk for personnel.

Neutron diffraction study of samples of fuel element claddings made of austenitic steel

The paper presents the results of investigation of the performance of industrial fuel elements operating in the active zone of a fast–neutron reactor BN-600. As a continuation of previous works, we have studied fuel elements made of austenitic steel EK-164 after their irradiation under a maximal dose of 72.3 dpa at temperatures ranging from 370 to 628 °C. As in the earlier work on studying spent fuel elements fabricated from standard steel ChS-68, neutron-diffraction method was used that demonstrates a number of advantages over traditional methods, in particular, it is the least-exposing workers to radiation hazards and most suitable for studying bulk irradiated materials. Based on the experience gained, some quite typical microstructure distortions arisen under irradiation were revealed, changes in the density and type of dislocations depending on the dose and temperature were determined, and the dominant role of temperature at high doses of fast neutrons was confirmed. The behavior of EK-164 steel samples selected from different series is shown to depend on their initial structural state.

Low activation materials of transport packaging sets

One of the tools for ensuring nuclear safety during the transportation of spent nuclear fuel is the creation of a high-strength transport container that effectively absorbs radioactive radiation from fission products. This raises the problem of creating appropriate structural materials. In this regard, it seems very promising to use a metal absorbing coating on high-strength corrosion-resistant steel, which is well mastered by the industry. However, ensuring the effectiveness of such a coating is possible only at a high concentration of neutron-absorbing elements in the cladding layer. One way to reduce the induced radioactivity of packaging sets is to use low-activated ferritic-martensitic steels such as EP450 (12Kh13M2BFR), EP823 (16Kh12SMVFBR) and 16Kh12V2FTaR. These steels have shown a high level of serviceability as structural materials for fuel element cladding and RBN fuel assemblies (BN-600, BN-350, BOR-60, etc.). Currently, a promising direction is the laser coating on the surface of transport packaging complexes. This paper explores the method of protection transport packaging sets against neutron radiation by laser surfacing of metallic powder.

Low activation materials of transport packaging sets

One of the tools for ensuring nuclear safety during the transportation of spent nuclear fuel is the creation of a high-strength transport container that effectively absorbs radioactive radiation from fission products. This raises the problem of creating appropriate structural materials. In this regard, it seems very promising to use a metal absorbing coating on high-strength corrosion-resistant steel, which is well mastered by the industry. However, ensuring the effectiveness of such a coating is possible only at a high concentration of neutron-absorbing elements in the cladding layer. One way to reduce the induced radioactivity of packaging sets is to use low-activated ferritic-martensitic steels such as EP450 (12Kh13M2BFR), EP823 (16Kh12SMVFBR) and 16Kh12V2FTaR. These steels have shown a high level of serviceability as structural materials for fuel element cladding and RBN fuel assemblies (BN-600, BN-350, BOR-60, etc.). Currently, a promising direction is the laser coating on the surface of transport packaging complexes. This paper explores the method of protection transport packaging sets against neutron radiation by laser surfacing of metallic powder.

IMPROVEMENT OF FUEL ELEMENT SHELL CONTROL METHODS TO INCREASE NUCLEAR REACTOR SAFETY

The analysis of the existing methods of control of the surface of the fuel element cladding material was carried out, which showed that their use for detecting surface and internal defects, such as local inhomogeneities, micro- and macropores, various cracks, axial looseness, etc. is characterized by low efficiency, is a laborious process that requires additional surface treatment, material of the fuel elements cladding. In addition, the investigated methods of controlling the surface of the fuel element cladding material make it possible to visually identify only rough external cracks and large slag inclusions, small cracks and non-metallic inclusions invisible under the slag layer. It is proposed to assess the quality of the surface of the shell material in case of its damage and destruction, the use of a computational apparatus based on the method of the theory of fractals. It is proposed to use the fractal properties of the shell material structure and a quantitative fractal value – the fractal dimension, which makes it possible to determine the degree of filling of the volume of the shell material structure during fuel element depressurization. A mathematical model of damage to the structure of the fuel element cladding material is developed depending on the simultaneous effect of high temperature and internal pressure caused by the accumulation of nuclear fuel fission products between the nuclear fuel pellet and the inner surface of the fuel element cladding, taking into account the fractal increases in the geometric parameters of the fuel element cladding. It is shown that damaged structures of the fuel rod cladding material depend on the pressure and temperature inside the fuel rod cladding, as well as the fractal increase in geometric parameters, such as: volume and surface area, outer and inner diameters, height and cross-sectional area, cladding length and height of nuclear pellets, gap between the inner surface of the cladding and nuclear fuel. A criterion for assessing the integrity of the fuel rod cladding is determined, which depends on the change in geometric values in the event of damage and destruction of the structure of the fuel rod cladding material. Practical recommendations are given on the use of the proposed method for monitoring the tightness of the fuel element cladding for processing information obtained from the computational module of the system for monitoring the tightness of the cladding for the automated process control system of the nuclear power plant power unit, which makes it possible to detect the depressurization of fuel elements at an earlier stage in comparison with the standard procedure.

ТЕПЛОГИДРАВЛИЧЕСКИЕ ИССЛЕДОВАНИЯ ВЫСОКОТЕМПЕРАТУРНОЙ ЯЭУ ДЛЯ ПРОИЗВОДСТВА ВОДОРОДА И ДРУГИХ ТЕХНОЛОГИЧЕСКИХ ПРИЛОЖЕНИЙ

An important problem determining the development of clean energy is the involvement of hydrogen in the fuel cycle. At present, the main method of hydrogen production is steam methane conversion. In the long term, large-scale hydrogen production, this method is not viable due to the consumption of non-renewable resources and the emission of greenhouse gases. Alternative methods of hydrogen production by water splitting methods using thermochemical or electrolysis processes require a high-temperature heat source. Nuclear reactors can serve as the most widely used high-temperature heat sources. The performed neutron-physical and thermophysical studies have shown that there is a fundamental possibility to provide the required parameters of a high-temperature (900-950 °C) with a 600 MW (thermal) fast neutron reactor with a sodium coolant for hydrogen production. It’s possible on the basis of one of the thermochemical cycles or high-temperature electrolysis with a high coefficient of thermal utilization of energy. It is shown that the temperature regime of core fuel elements is determined by a large number of parameters that have a regular and statistical nature. The developed methodology and numerical program allows to take into account, in the fuel assemblies shaped during the campaign, the effect on the temperature distribution of the fuel element cladding and temperature irregularities along the fuel element perimeter in the interchannel exchange fuel assembly, the random distribution of channel cross-sections and the heat generation of fuel elements using the Monte Carlo method, also other factors. For various reactor operating regimes, zones with stable temperature stratification with large gradients and temperature fluctuations have been identified. The results obtained make it possible to judge the amplitude and frequency characteristics of temperature pulsations in these potentially dangerous areas. The relative small size, the type of coolant, the choice of fissile material and structural materials make it possible to create a reactor with inherent properties that ensure increased nuclear and radiation safety.

Study of high temperature oxidation of zirconium fuel element claddings with chromiferous coatings

This paper presents the results of the study of resistance to corrosion of zirconium fuel element cladding fragments protected with nichrome and chrome coatings. Experiments were conducted on coatings deposition on fuel element claddings by cathodic arc deposition and electrochemical deposition methods. Argon ion beam pretreatment of the cladding surface is shown to have a beneficial effect on coating adhesion. Studies of resilience of the coatings were conducted in the atmosphere of super-heated steam at 900 °C and with an exposure of up to 10000 seconds. It has been shown that the coatings deposited impede formation of ZrO2 oxide layer at the outer surface and reduce the depth of oxygen penetration into the metallic sublayer as compared with the samples in their original condition.