BN-600 Reactor Research Articles

Applying the Hollomon-Jaffe parameter to predict changes in mechanical properties of irradiated austenitic chromium-nickel steels during isothermal exposure

<abstract> <p>The Hollomon-Jaffe parameter is widely used in metallurgy and materials science to characterize the behavior and predict the various metals' physical-mechanical properties under different temperature and time modes. The possibility of predicting changes in the mechanical properties of structural steels due to thermal influences has been studied. The paper presents the results of a study of the mechanical properties of the materials of the core components of the BN-350 reactor facility (RF) made of austenite chromium-nickel steel 12Cr18Ni10Ti (a spent fuel assembly's jacket) and 09Cr16Ni15M3Nb (an intro-channel displacer). The samples were studied both before and after radiation annealing. Annealing of steel samples at 550 ℃ reduced the yield strength and significantly restored the plasticity and ability of the material to strain hardening. The efficiency of post-radiation annealing of the materials increases with annealing temperature and leads to a transition to the reduction process. It was established that medium of high temperature annealing during heat treatment does not lead to significant changes in the mechanical properties of irradiated materials. The microstructure studied using a scanning electron microscope reasonably correlates with the results of mechanical tests. The possibility of using the Hollomon-Jaffe parameter to predict the properties of austenite chromium-nickel steel, which received damaging doses in the range from 12 to 59 dpa, was shown for the first time. Thus, for the first time, the unique coefficient (<italic>C</italic>) of the Holloman-Jaffe parameter for irradiated materials of chromium-nickel steel was experimentally determined, and dependencies characterizing the change in hardness of chromium-nickel steel on temperature and duration of post-radiation thermal exposure were established.</p> </abstract>

INVESTIGATION OF RESISTANCE TO PITTING CORROSION OF STRUCTURAL MATERIALS OF THE BN-350 REACTOR

The article is devoted to the study of the resistance of structural materials of the BN-350 reactor to pitting corrosion in an aqueous medium containing chlorine ions in the presence of organic corrosion inhibitors. The results of accelerated testing of samples of austenitic steels 12Cr18Ni10Ti, 08Cr16Ni11Mo3Ti and Cr13Mo2NbVB (EP-450) ferrite-martensitic steel for pitting corrosion in a 10% aqueous solution of iron three-chloride hexahydrate without an inhibitor and in the presence of various concentrations of corrosion inhibitor are presented. The effect of heat treatment on the pitting resistance of structural steels is studied and the role of carbide precipitates of MC and M23C6 types in pitting defects formation is discussed.

Computational simulation of minor actinide burning in a BN-600 reactor with fuel without uranium and plutonium

The paper presents the results of studies on the burning of minor actinides (MA) extracted from SNF of thermal reactors in a BN-600 reactor, which uses the complete set of MAs instead of traditional nuclear fuel types: uranium and/or plutonium. The advantages of such approach to MA burning are that long-lived waste is recycled and energy is produced that can be used, e.g., to generate electricity. Besides, where, e.g., a reactor with uranium or MOX fuel is used for transmutation, apart from burning “foreign” minor actinides, it will additionally generate “its own” MAs. The studies have shown that such reactor can be efficient only if based on fast neutrons, which is due to the specific properties of the minor actinide neutron capture and fission cross-sections as compared with traditional fuel nuclides. The calculation results have shown rather a high rate of MA transmutation and burning in a reactor fueled with minor actinides.

Heavy metals and arsenic in foodstuffs in the vicinity of industrial enterprises and nuclear power plant

Introduction. There is evaluated the content of seventeen elements in local agricultural and natural foodstuffs of the 30-km zone of influence of the Beloyarsk NPP, located in the industrially developed area of the Sverdlovsk region. Materials and methods. The study was conducted in 2013 and 2019 in the private sector of 27 settlements, 20 collective farms and forests of the 30-km zone around the Beloyarsk NPP, and a cooling reservoir. In crop products (grain, potatoes, roots, vegetables), animal husbandry (milk, meat, poultry) and natural foods (mushrooms, berries, fish) the content of Pb, Zn, Cd, As, Hg, Mo, Cu, Ni, Co, Cr, Sb, Sr, Mn, V, W, Ba and Fe was estimated. Elements in the samples were detected by atomic absorption and plasma emission methods of analysis. The results obtained for the most toxic Pb, Cd, As and Hg were compared with sanitary-hygienic standards (SanPiN 2.3.2.1078-01). Results. The trend in the content of heavy metals and arsenic in local foodstuffs before and after the commissioning of the BN-800 reactor is presented. The types of foods with the maximum accumulation of pollutants are identified. In the majority of local foodstuffs in the vicinity of the Beloyarsk NPP there was shown no risk of exceeding the sanitary-hygienic standards for the content of the most toxic elements. An excess of SanPiN standards was noted for arsenic content in 10-30% of grain samples, for cadmium in 15-20% of wild berry samples. Limitations. The content of toxicants was assessed in the main groups of local agricultural and natural foods available during the sampling period, and expanding the list of analyzed products and the number of elements may be the subject of further research. Conclusion. The assessment of the safety of foodstuffs in the vicinity of the Beloyarsk NPP showed that the operation of the BN-600 and BN-800 reactors doesn’t significantly affect the increase in the content of toxic elements in locally produced foods, and the accumulation of pollutants in them is determined by long-term emissions of industrial enterprises of Sverdlovsk region.

Stability of the Y–Ti–O oxides in reactor materials under neutron irradiation at high temperatures

The paper presents the results of electron microscopic studies of ferrite-martensitic steel samples hardened with Y-O oxides, EP-450 DUO in the initial state and after neutron irradiation in the BN-600 reactor at 1000°C to a damaging dose of 77.5 dpa. These studies showed that the main types of oxide phases were Y2(Si, Ti)2O7 and Y2(Si, Ti)O5. These precipitates at sizes less than 10-20 nm were semi-coherent with a ferritic matrix of steel EP-450 DUO with the ratio (110)malrix//(221)palticle. Some of the Y-Ti-O oxides in the initial state were Y2Ti2O7-type with some deviations from the stoichiometric composition.However, after neutron irradiation under BN-600 conditions at temperature ~ 1000°C, oxide particles could not be described by the indicated stoichiometry. Besides, after irradiation, silicon and aluminum were found in the oxide's composition. In the case of taking these elements into account during the construction of a triple composition diagram, it was shown that the oxide phases had Y2(Ti, Si, Al)2O7 and Y2(Ti, Si, Al)O5 types. It was established that in samples of EP-450 DUO steel in the initial state with oxide particles up to 20 nm in size, the yttrium content is generally lower than the titanium concentration. The titanium and yttrium concentrations corresponded to the stoichiometric composition Y2Ti2O7 (1:1) with a further increase in the average diameter of these phases. After irradiation, the situation changed somewhat: the yttrium content in most oxide phases exceeds the total concentration of titanium, silicon, and aluminum.The paper also presents the analysis of porosity and evolution of grain structure in EP-450 DUO steel after neutron irradiation.

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

Determination of the deformation deceleration coefficient for calculating the initial effective length of the working part of an annular specimen made of fuel cladding

When studying the plastic properties of metallic materials under tension, the initial calculated length of the sample which affects the final results is of significant importance. In the case of determining the short-term mechanical properties of the materials of fuel cladding pipes using annular samples, calculation of the initial effective length of their working part is rather difficult since the plastic deformation of the annular sample is unevenly distributed along the perimeter. This feature is taken into account using the deformation deceleration coefficient. We present the results of determining the deformation deceleration coefficient for calculation of the initial effective length of the working part of an annular sample made of fuel cladding. An experimental method for determining the coefficient is proposed, taking into account the measurements of the distribution of plastic deformation in certain areas along the perimeter of the sample. The method consists in stretching annular and ovalized samples with marks on the end surface. It is shown that for annular and ovalized specimens from the fuel cladding of a BN-600 reactor, the values of the deformation coefficient and effective length are k = 0.50 ± 0.04 and l 0 = 6.1 ± 0.4 mm, respectively. Analysis of the actual relative deformation of the working parts and comparison with the relative elongation showed that when the ovalized specimens are stretched, the value of the relative plastic deformation is higher than that for annular specimens. The proposed method and the results obtained can be used in determination of the parameters of the materials used in the cladding pipes of the of fast neutron reactors.

Experience Gained with Chemically Cleaning the Evaporators of the Beloyarsk NPP BN-600 Reactor Plant Sodium–Water Steam Generators from Deposits

Experience Gained with Chemically Cleaning the Evaporators of the Beloyarsk NPP BN-600 Reactor Plant Sodium–Water Steam Generators from Deposits

Radiochemical and Elemental Analysis of Mixed Uranium-Plutonium Fuel Irradiated in the BN-600 Reactor

Radiochemical and Elemental Analysis of Mixed Uranium-Plutonium Fuel Irradiated in the BN-600 Reactor

Experience of mixed nitride uranium–plutonium fuel fabrication at the siberian chemical plant jsc site

The “Proryv” project is significant for the Russian Federation and relates to the activity that has been assigned the status of “especially important” for the Rosatom State Corporation. It is planned to develop a new generation of nuclear energy technologies based on the closed nuclear fuel cycle of fast neutron reactors with a lead coolant and mixed nitride uranium-plutonium nuclear fuel as a part of this direction (hereinafter – MNUP fuel).The first method for MNUP fuel producing was the direct synthesis of nitrides from starting metals and alloys of uranium and plutonium (hydrogenation-nitration), implemented in laboratory conditions. The disadvantage of this method is its explosion and fire hazard.A method for carbothermal reduction in the atmosphere of nitrogen from the initial oxides of uranium and plutonium is proposed for the production of MNUP fuel on an industrial scale.According to the results of the literature data analysis of considered above methods, carbothermic reduction in a nitrogen atmosphere from the starting oxides of uranium and plutonium is the most optimal for the industrial production of MNUP fuel.Based on the foregoing as part of the «Proryv» project the concept of MNUP fuel production by the carbothermic synthesis method was adopted as the main method at the site of Siberian Chemical Plant (Joint stock company). For this purpose in 2012 a set of experimental facilities (a unique platform for the pilot industrial production of fuel components (from pellets to fuel assemblies) of fast reactors) was created.During this period for the first time in the world at the set of experimental facilities full-scale nuclear fuel assemblies based on MNUP fuel composition were made, data for verifying the technological model of the processes for manufacturing MNUP fuel were prepared, using experimental models of the equipment a complex technological process with experimental scaling of the manufacturing technology of MNUP fuel was developed, an automatic algorithm sorting of tablets and systems for providing and controlling the parameters of the box atmosphere was developed, as well as construction materials of furnaces for the synthesis and sintering of nitride fuel were tested. In total, 24 nuclear fuel assemblies — more than 1000 fuel elements of various designs with different shells — were transferred for loading into the BN-600 reactor. Successfully completed irradiation of 21 nuclear fuel assemblies.

Status on performance study of mixed nitride fuel pins of BREST reactor type

Within the framework of the “PRORYV” project the “The comprehensive program for the calculation and experimental justification of fuel pins with mixed uranium-plutonium nitride fuel of BN-1200 and BREST reactors for the period up to 2020” has been developed Tpoянoв et al., 2015 [1]. The purpose of the program is to judge fuel pins performance, to ensure stability and required quality of the experimental technology for manufacturing of mixed nitride, fuel pins and experimental fuel assemblies (EFA) for testing in BOR-60 and BN-600 reactors, to optimize the design and technology of fuel pins with mixed nitride for the developed BN-1200 and BREST-OD-300 reactors based on the results of reactor testing. Currently, the program aims at justifying the starting operation stages of BREST-OD-300 and BN-1200 reactors. The following results were obtained to judge the fuel performance at the starting operation stage of BREST-OD-300 reactor (the maximum mixed nitride burn-up of 6 at%, the maximum damage dose of 89 dpa): 1. A set of out-of-pile investigations has been performed in order to obtain the necessary data on the properties of fuel and cladding materials, which are required for the development of calculation models of mixed nitride behavior under irradiation and design justification of their performance, including in design accidents. 2. To March 2021 the irradiation of eight BN-600 EFAs with EP823-Sh cladding was finished. In EFA-11 the maximum burn-up 9.0 at% and the maximum damage dose 107.6 dpa have been reached. PIE of fuel pins of two OU irradiated in BOR-60 and four EFAs irradiated in BN-600 with maximum burn-up of 5 at% and damage dose of 74.7 dpa are completed. Experimental data required for verification of calculation results of temperature and stress-strain state of mixed nitride pins with EP823-Sh steel cladding, as well as for evaluation of fuel pin performance and establishment of allowable lifetime of the BREST-type fuel pins with mixed nitride have been obtained. 3. The KORAT, DRAKON and BERKUT-U codes were modified and verified. Certification of the DRAKON code has been completed and the BERKUT-U code has been submitted for certification. Obtaining the permission to extend the irradiation of EFA beyond the parameters planned in the Program was possible due to the fact that independent concurrent calculations with the use of KORAT, DRAKON and BERKUTU codes were performed during the development of technical designs of fuel pins with mixed nitride in justification of the possibility to extend the irradiation of EFA-11, EFA 14, EFA-15 in BN-600 reactor. All codes confirmed this possibility. Positive results of irradiation of BREST-type fuel pins in BOR-60 reactor were also the most important argument for obtaining permission. In order to improve fuel pins with (U, Pu)N and further burn-up increase, the technologies of their production were developed and optimized, the design of the pellets was changed (chamfer) and also fuel pins with liquid-metal sub-layer have been considered.

Results of post-irradiation examinations of mixed nitride pins with gas and liquid metal sub-layers

Today the investigations have been completed on helium-bonded fuel pins with mixed uranium-plutonium nitride of BN-600/BN-800, BN-1200 and BREST reactors types after irradiation as a part of ten EFAs of BN-600 reactor up to maximum burn-up of 7.5; 6.0 and 4.5at%, respectively. Also the PIE of mixed nitride pins of BREST type with helium and lead sub-layers after intermediate tests to maximum burn-up of 4,8 and 3,9at% as part of dismountable EFAs of BOR-60 reactor are over. As a result of postirradiation examinations, the main intra-fuel pin processes that affect on materials properties change and fuel pins state, their relationship with the initial nitride state were identified. The regularities and quantitative characteristics of nitride swelling, the behavior of fission products, cladding corrosion and mechanical properties change are revealed. The average for the irradiation time rate of fuel swelling in cross sections near the core midplane in gas-bonded pins of different EFAs irradiated in BN-600 reactor is equal to (1.6 - 2.0) %/at%. According to the results of irradiation in BOR-60 reactor, the fuel swelling rate in lead-bonded pins is lower than in helium-bonded pins: 1.4 ± 0.2 and 1.7 ± 0.2 %/at%, respectively. The presence of carbon and oxygen impurities in the fuel can lead to local areas of claddings carburization and oxidation, randomly distributed along the height and perimeter of its inner surface. The key characteristics of the fuel that determine the cladding carburization and oxidation, and ways to prevent them, are determined. Cladding corrosion in lead-bonded pin is caused by selective dissolution of cladding steel components. The maximum depth of the corrosion zone is located in the upper part of the fuel column and is equal to 40 microns. The results of cladding mechanical tests obtained using longitudinal segment samples and tubular samples loaded by internal gas pressure showed the significant margin of strength and ductility of cladding materials along the entire height of fuel pins both at nominal operating temperatures and at room temperature. High values of strength characteristics indicate a weak influence of corrosion on the strength of studied claddings materials. The mechanical cladding properties of a lead – bonded pins have the same pattern of change along the pin height depending on irradiation and testing temperatures, as of fuel pins with helium sublayer, keeping enough ductility in the area of low-temperature irradiation embrittlement. So, for example, at irradiation and test temperatures of 350-380 °C, the values of the strength limit from 1070 to 1200 MPa were obtained with the values of the total elongation of not less than 2 %. Thus, the results of post-irradiation examinations confirmed the performance of mixed nitride pins at the achieved test parameters and the possibility of their irradiation prolongation to higher fuel burn-up.

Unification of the BN-1200 reactor core designs with MOX and MNUP fuel

Two types of mixed uranium-plutonium fuel are considered for BN-1200 reactor (Matveev and Khomyakov, 2012) within the framework of Proryv Project: conventional, well-mastered MOX fuel and advanced nitride fuel (MNUP fuel) having higher density. Decision about the expediency of use of this or that type of fuel in the first BN-1200 reactors at the initial stage of their operation will be made taking into account the results of MNUP fuel pins tests carried out in BOR-60 and BN-600 reactors. In this report the information on basic engineering decisions taken in the design of BN-1200 reactor core (Vasiliev et al., 2010; Shepelev et al., 2013; Vasiliev et al., 2017) is given, features of cores with MOX and MNUP fuels are analyzed, possible variants of cores with unified design of fuel pins and FSA are presented, estimates of core characteristic change with increased fuel burnup are given, problems of transition from MOX fuel to MNUP fuel are considered.

РАСЧЕТНЫЙ АНАЛИЗ ОДНОВРЕМЕННО ВОЗНИКАЮЩИХ ЗАПРОЕКТНЫХ АВАРИЙ НА ЭНЕРГОБЛОКАХ № 3 И 4 БАЭС

The paper presents the results of a computational analysis of simultaneously occurring beyond design basis accidents at power generating units No. 3 (BN-600) and No. 4 (BN-800) of the BNPP. Safety analysis for operating power generating units at the same time occurring beyond design basis accidents is carried out in accordance with the requirements of NP-001-15. Computational studies were performed with the SOCRAT-BN, ANSYS, ORIGEN2, and VIBROS 2.2 codes. There are considered beyond design basis accidents caused by seismic impact with intensity of 7 points on MSK-64 scale (exceeding the maximum allowed earthquake (MAE) for the site of power unit No. 3 and the corresponding MAE for the site of power unit No. 4). As a result of seismic impact at all the power generating units of the BNPP, a loss of system power supply occurs, which, together with additional failures of the main equipment, leads to beyond design basis accidents. The paper presents the results of computational studies of beyond design basis accidents in terms of reactor plant, assembly cooling pond and holding pond and radiation effects assessment. The results of a computation analysis of the beyond design basis accident showed that the release of radioactivity occurs only for the BN-600 reactor facility (where are breaking of the fuel elements cladding). At the same time, the predicted radiation doses for the first years after the accident are lower than the NRB-99/2009 criteria for making a decision on the evacuation and resettlement of inhabitants.

The evolution of helium-vacancy bubbles in austenitic steels under neutron irradiation

The paper describes an evolution model for helium-vacancy bubbles in austenitic steels under neutron irradiation. A condition for the growth of helium-vacancy void nucleus was found using the model of point defect migration. An equation to determine the critical diameter, when a nucleus transforms into a vacancy void was obtained. The critical diameter value under irradiation in the fast reactor spectrum was calculated. The found characteristics were compared to the experimental data obtained from samples of the austenitic reactor steel (Russian Grade EK-164) cladding operated in the fast BN-600 reactor. It has been shown that radiation porosity growth leads to increase in critical diameter, thus significantly affecting the swelling process.

Determining the Specific Activity of Tritium in Samples of Spent Nuclear Fuel

Procedure for determining the specific activity of tritium in samples of spent nuclear fuel (SNF) was developed and methodologically certified. The procedure was used to determine the specific activity of tritium in samples of mixed nitride uranium-plutonium nuclear fuel irradiated in BOR-60 and BN-600 reactors to be 8.5×106 ± 1.1×105 and 2.6×106 ± 2.0×105 Bq/g, respectively. The content of tritium in the shell of a nuclear fuel element irradiated in a BN-600 reactor was estimated, and the efficiency of the voloxidation procedure of the mixed nitride uranium-plutonium spent nuclear fuel for removing tritium was evaluated.

Разработка и исследование плазмохимической технологии получения смешанных оксидов урана и плутония из растворов

This paper presents the results of work on the application of plasma-chemical technology for producing uranium and plutonium oxides and their homogeneous mixtures from nitrate solutions of a given composition for the purpose of manufacturing fuel element cores. The technology is based on the process of high-temperature decomposition of aqueous solutions of salts by spraying them into a flow of the heat-transfer gas (air) heated to 5000 – 6000 K in a plasma torch. It is a highly productive short-stage process. No reagents are required for salt precipitation; there are no filtration and high-temperature treatment precipitation operations; the resulting powders are chemically active. It was found that the transformation time of salt solution drops to oxides is 0.05 – 0.10 s for the length of 0.5 – 0.7 m. The temperature in the reactor varies in the range of 1200 – 1800 K; no sintering and fusion of particles occurs. The apparatus-technological scheme of the plant has been developed. The technological equipment is made in a nuclear-safe design and installed in the exhaust boxes. The scheme includes a high-frequency induction plasmatron (plasma torch) with coaxial gas inlet into the discharge chamber. The high-frequency induction discharge is excited by inserting a grounded ignition electrode into the plasmatron discharge chamber. The plasmatron design provides the possibility of long continuous operation at high plasma temperatures with the creation of intensive plasma-jet processes for processing radioactive solutions. The reactor provides ways to prevent product accumulation on the walls (gas curtain method, electric pulse cleaning of cylindrical surfaces, vortex tangential gas supply at the level of the outlet holes of the injectors). Equipment for the separation of nanodispersed oxide powders from high-temperature steam- and powder-gas flows (vortex precipitators, filters with a metal-woven partition) has been developed and applied. The degree of oxide recovery in vortex precipitators is 88 – 92 % per stage, in filters with a metal-cloth partition the degree of purification reaches 105 – 108. Plasma-chemical oxides are nanocrystalline and nanoscale materials represented by crystalline and amorphous structures. Mixed powders are a homogeneous mixture of oxides in the form of solid solutions. The particle size does not exceed 3 μm; the crystallite size is about 10 nm. Fuel elements were manufactured from mixed powders at FGUP “Mayak”, which passed successful tests in the “Bor-60” reactor. Seven industrial fuel elements were manufactured for the BN-600 reactor, six of which were loaded into the reactor.

ВЛИЯНИЕ ЗАМЕНЫ КОНСТРУКЦИОННОЙ СТАЛИ ТВС НА ЗАПАС РЕАКТИВНОСТИ В РЕАКТОРЕ БН-600

The current program in Russia to increase the fuel consumption of fast reactors and increase its burn-out causes the transition to new structural materials, which, in turn, leads to changes in the neutron-physical characteristics of reactors. In particular, the drop in the reactivity reserve noted in the BN-600 reactor of the Beloyarsk NPP at the end of 76 operational cycles, as will be shown below, is due to the transition to a new type of shell steel with an increased content of nickel, which strongly affects the reactivity. Design support for the operation of the BN-600 and BN-800 fast reactors, as well as the experiments carried out on them, is performed by IPPE. This article presents the results of a calculated analysis of the expected changes in the reactivity reserve at the end of 76 operational cycles when replacing the shell steel in BN-600. In addition, the influence of experimental assemblies located in the core on the reactivity reserve of the BN-600 is analyzed. Analysis of calculations of the actual loading of the BN-600 reactor at 76 operational cycle using the methods of the 1st-order perturbation theory, strict perturbation theory, and the Monte Carlo method showed that a partial transition at 76 operational cycle to EK-164 shell steel leads to a decrease in the reactivity margin by 0.030±0.004 %Δk/k. Replacement of steel for the entire core will reduce the reactivity margin by ~0.12 %Δk/k, which is confirmed by Monte Carlo calculations. The calculated reactivity margin obtained at the end of 76 operational cycles for the hot state of the BN-600 reactor is in good agreement with the measured reactivity margin.

Evaluation of the burnout efficiency of Np-237 and Am-241 in a BN-600 reactor with a modified core

From the results of comparing the fission and capture cross sections, it follows that a reactor with fuel in the form of Am-241 or Np-237 can only be on fast neutrons, because in the thermal and intermediate spectra, the capture cross section significantly exceeds the fission cross section. Therefore, this paper analyzes the possibility of achieving criticality when loading the BN-600 reactor core with fuel from Am-241 or Np-237 alone, as well as the efficiency of their burning out in various modifications and variants of core loading. The calculation results show that when the reactor is fully loaded with fuel consisting of minor actinides of interest, we have a huge reactivity reserve, which does not decrease, and in some cases increases, during 450 days of reactor operation at rated power. Also, when calculating the modified core with fuel from minor actinides and uranium dioxide, it was possible to achieve high burnout rates of Am-241 and Np-237.

Modeling of BN-600 Hybrid Core Benchmark (HEX-Z) with MCU-PTR and SERPENT 2

For the computational support of innovative projects it is necessary to use high-fidelity codes, which must be verified and validated. In Russia, MCU and SERPENT codes are widely used. The MCU code was previously chosen as the ‘Proryv’ project code. Both of these codes are used for calculation of fast reactors, although they are certified for thermal reactors. Therefore, there is a need to show how accurately they allow calculating the characteristics of fast reactors. For cross-verification of these codes, the main characteristics of the benchmark of BN-600 reactor with a hybrid core were calculated: neutron multiplication factors, reactivity coefficients, distribution of energy release over zones, reaction rates. The values of the neutron multiplication factors for various states of the model obtained with the MCU-PTR are systematically higher than the values obtained with the SERPENT 2 (for majority of the states by 0.4%). The values of the reactivity coefficients have the same sign, however, difference between both codes can reach 160% and if compared with the benchmark – 181%. For the majority of reactivity coefficients this compares with the distribution of the results presented in the benchmark. The MCU results are closer to the benchmark than the Serpent ones.