Mixed Nitride Uranium-plutonium Research Articles

POSSIBILITIES OF NEUTRON ACTIVATION ANALYSIS FOR STUDYING THE CORROSION BEHAVIOR OF METALLIC MATERIALS IN MOLTEN SALTS

For the BREST-OD-300 reactor facility [1, 2], the technology for evaluating mixed nitride uranium-plutonium spent economical fuel is being determined [3–9]. To separate MNUP SNF from fuel claddings made of materials with high radiation resistance – ferritic-martensitic steel EP-823 [10–16], it is planned to use pyrometallurgical grades of “soft chlorination” [17]. When alloying and impurity elements of steel EP-823 are dissolved in molten salts of eutectic composition based on lithium and potassium chlorides, the melt will be contaminated. For the same reason, the formation of volatile compounds will occur, with their further mass transfer from hot to cold sections of process equipment. When studying the corrosion behavior of metals and alloys in liquid media, the problem often arises of determining small amounts of dissolution products in solution. This problem arises, for example, the rate of dissolution of microimpurities. The sensitivity of the usual, traditional methods used in corrosion testing such as mass loss or colorimetric determination of corrosion products in solution is often insufficient to make appropriate measurements. In these cases, the most effective is the use of the radiochemical method of neutron activation analysis based on. qualitative and quantitative determination of chemical elements, based on the measurement of the radiation characteristics of radionuclides formed during the irradiation of materials with neutrons. This paper presents the results of a study of the corrosion behavior and mass transfer of corrosion products of EP-823 steel pre-irradiated in the IVV-2M research nuclear reactor in molten salts 2KCl–3LiCl and 2KCl–3LiCl–PbCl2 at temperatures of 500 and 650°C for 24 h. It is shown that the method of neutron activation analysis can be used to study the corrosion behavior of EP-823 steel in molten salts of various compositions.

POSSIBILIITY OF URANIUM EXTRACTION FROM SPENT NUCLEAR FUEL IN FUSED ELECTROLITES CONTAINING RARE ELEMENTS

Link for citation: Nikitin D.I., Polovov I.B., Rebrin O.I. Possibility of uranium extraction from spent nuclear fuel in fused electrolytes containing rare elements . Bulletin of the Tomsk Polytechnic University. Geo Аssets Engineering, 2023, vol. 334, no. 10, рр. 210-218. In Rus. The relevance of the research is caused by the plans of using electrolytic separation of metalized spent nuclear fuel as a stage in pyrochemical reprocessing of mixed nitride uranium-plutonium fuel. The main aim is to determine the parameters of uranium electrolytic separation from an alloy with simulators of fission products (precious metals and rare earth elements) simulating spent nuclear fuel in salt mixtures based on 3LiCl–2KCl with additives of rare earth element chlorides. Objects: model spent nuclear fuel – uranium alloy with simulators of fission product (precious metals and rare earth elements) with a mass compound of Pd:Ru:Ag:Rh=25:1:3:3, Nd:Ce:La:Pr:Sm:Y=15:10:5:5:5:1). Methods: electrorefining, X-ray fluorescence and X-ray diffraction method of analysis, inductively coupled plasma mass spectrometry, assessment of distribution of components in the system. Results. The data obtained showed that uranium deposits have dendrite formations of alpha-uranium at 550 °C in orthorhombic crystal system toward needle cathode current density. The resulting cathode deposits are free from impurities of ruthenium, rhodium, molybdenum, praseodymium and yttrium. The purification coefficient for palladium reaches 3000, and for silver 1700. Noble metals accumulate in the anode sludge, and even with the complete depletion of the anode material. The concentration of noble metals in the cathode deposit does not exceed 0,0015 wt %. Despite the high concentration of rare-earth chlorides in the electrolyte, which simulates the accumulation of rare-earth ions during repeated reprocessing of spent nuclear fuel, the concentration of rare-earth metals in the cathode product did not exceed 0,007 wt %. During uranium electrowinning from a model spent nuclear fuel in the 3LiCl–2KCl–UCl3 electrolyte (10,1 wt %) with REE chlorides, which imitate their accumulation in the electrolyte during repeated processing, at 550 °C, as well as the initial cathode current density of 0,2 A/cm2, the specific amount of electricity is 1,0 A∙h/cm2. A cathode uranium deposit is released with a current efficiency exceeding 90 % with all anode mass depletion, and purification coefficient 1800 for the sum of precious metals and for the sum of rare earth metals.

ELECTROREFINING OF URANIUM ALLOYS CONTAINING PALLADIUM AND NEODYMIUM IN 3LiCl–2KCl–UCl<sub>3</sub> MELTS

The technology of pyrochemical processing of mixed nitride uranium-plutonium spent fuel, realizable at the experimental and demonstration energy complex of the site of the Siberian Chemical Plant, includes several operations with the ultimate goal of isolating the target fission products. It’s planned to use the electrofining of the products of the previous stage, metallized spent nuclear fuel, аs the penultimate stage of processing. It’s necessary to determine the processes and technological modes of electrolytic refining of alloys modeling the product of this stage of the processing module to implement electrolytic refining. This paper presents the results of electrofining of model alloys (simulating the raw materials of the stage of electrofining processing) on an enlarged laboratory electrolyzer. The initial parameters of uranium refining processes in melts based on 3LiCl–2KCl–UCl3 were determined earlier. The basic parameters of refining were the use of electrolyte 3LiCl–2KCl–UCl3 (10.1 wt % UCl3) and conducting experiments at 550°C. Uranium alloys containing palladium and neodymium were prepared by direct fusion of uranium metal, PdAP-1 grade palladium metal powders and neodymium metal (99.99%) in a medium of high-purity argon (99.998%). The data obtained showed that at a temperature of 550°C, cathode precipitates are typical dendritic forms of alpha-uranium in rhombic syngony with a tendency to needle formation with an increase in cathode current density. An increase in the company time and cathode current density leads to a decrease in the current output due to short-circuiting of the electrodes with cathode sediment needles or metal shedding from the cathode. The modes of the cathode process have been experimentally refined as a result of electrofining. When electrofining alloys U–Pd(1.59 wt %), U–Pd(1.62 wt %), U–Pd(1.54 wt %), U–Pd(1.58 wt %)–Nd(5.64 wt %), U–Pd(1.84 wt %)–Nd(6.49 wt %), U–Pd(1.79 wt %)–Nd(6.54 wt %), uranium cathode precipitates were obtained, which were subjected to chemical analysis, which showed the high purity of the resulting metallic uranium, as well as the absence of metallic palladium and molybdenum in it. The palladium purification coefficient exceeds 5000, the neodymium purification coefficient exceeds 1000, which meets the requirements for purification from fission products at this stage of pyrochemical processing of spent fuel. Palladium accumulates in anode slime, while the bulk of neodymium passes into the molten electrolyte.

Determining the swelling of mixed nitride uranium-plutonium fuel based on the post-reactor studies of experimental BN-600 and BOR-60 fuel rods

Determining the swelling of mixed nitride uranium-plutonium fuel based on the post-reactor studies of experimental BN-600 and BOR-60 fuel rods

Voloxidation of Mixed Nitride Uranium–Plutonium Spent Nuclear Fuel

The efficiency of voloxidation of mixed nitride uranium–plutonium spent nuclear fuel (MNUP SNF) for the separation of the fuel composition from the fuel rod cladding and removal of 3H and 14C was estimated. It was shown that the completeness of SNF separation from fuel rod cladding under optimal conditions is at the level of 98–99%. The residual content of tritium in the voloxidized fuel does not exceed 0.2% of its content in the original SNF sample; radiocarbon is removed by 98%.

ANODIC PROCESSES OF URANIUM ALLOYS CONTAINING PALLADIUM AND NEODYMIUM IN 3LiCl–2KCl-UCl<sub>3</sub> MELTS

At the reprocessing module of the pilot demonstration power complex site of the Siberian Chemical Combine, a combined technological scheme for the reprocessing of mixed nitride uranium-plutonium spent fuel consisting of pyrochemical operations, hydrometallurgical refining of uranium, plutonium and neptunium is implemented step by step. According to this scheme, the target pyrochemical reprocessing products, purified from the main mass of fission products with actinoid content not less than 99.9%, are sent for hydrometallurgical reprocessing. For pyrochemical reprocessing it is necessary to develop a technology of electrorefining of metallised spent nuclear fuel. To carry out electrolytic refining it is necessary to define processes and regimes of anodic dissolution of alloys simulating product of this head operation “metallization”. In the present work the results of investigations of processes of anodic dissolution of U–Pd and U–Pd–Nd model alloys with different concentrations of palladium and neodymium in melts based on 3LiCl–2KCl–UCl3 (10.1 wt % UCl3) at 550°C using different methods are presented. Uranium alloys containing palladium and neodymium were prepared by direct alloying of uranium metal and palladium metal powders of PdAP-1 grade, and neodymium metal (99.99%) in high-purity argon medium (99.998%). Electrochemical measurements were performed using an Autolab 302N potentiostat/halvanostat equipped with a Booster 20A high-current module. The anodic polarisation curves consist of only one oxidation wave which was attributed to the dissolution of uranium metal. Increasing the palladium content in the alloy from 1.5 to 10.0 wt %, does not affect the shape of the polarisation curves. The increase of neodymium content in the alloy from 1.0 to 10.0 wt % also does not affect the shape of polarization curves. Electrorefining parameters of uranium alloys containing palladium and neodymium were determined. The limiting current density of uranium evolution from uranium alloys containing palladium and neodymium in the electrolyte 3LiCl–2KCl–UCl3 (10.1 wt % UCl3) at 550°C was 0.4 A/cm2. It was shown that palladium does not diffuse into the melt as a result of anodic dissolution and neodymium accumulates in the electrolyte only when the alloy is refined with 10.0 wt % neodymium, which is much higher than the future real concentrations of electrotreated uranium alloy components in the technological chain of spent nuclear fuel processing.

Incorporation and migration of xenon in uranium-plutonium mixed nitride; A density functional theory study

Actinide nitride materials are promising candidates for advanced nuclear fuels. In this work, we investigate the bulk properties of the mixed nitrides U0.75Pu0.25N and U0.5Pu0.5N, and study the incorporation and migration behaviour of the fission gas Xe. The disordered U0.75Pu0.25N and U0.5Pu0.5N structures are constructed using the special quasi-random structure method. Their lattice parameters are closer to the experimentally determined values than the corresponding ordered structures. The density of states show that Pu f states are located at lower energy than U f, consistent with the trend of increasing f orbital stability across the actinide series. The actinide vacancy formation energy (Ef) in disordered and ordered U0.5Pu0.5N is highly dependent on the chemical environment around the vacancy: it increases as the number of U atoms in the first nearest-neighbour shell (NU(1NN)) increases, but decreases as the number of U atoms in the second nearest-neighbour shell (NU(2NN)) increases. The Xe incorporation energy (Ei) is found to be independent of vacancy species, depending only on the chemical environment of the vacancy. As does Ef, Ei increases with increasing NU(1NN), while decreases with increasing NU(2NN), because the smaller the NU(1NN) and the larger the NU(2NN), the larger the vacancy steric space. The Ei of Xe and Kr are calculated to be within the ranges 4.47–6.01 eV and 3.30–4.64 eV, respectively. The Xe migration energy barrier in ordered U0.5Pu0.5N allows us to set the energy range for Xe diffusion in disordered U0.5Pu0.5N as 0.50–1.75 eV. A lower range of 0.30–1.25 eV is found for Kr diffusion.

Nitriding and Carburization of a Mixed Nitride Uranium-Plutonium Fuel Pin Cladding

Nitriding and Carburization of a Mixed Nitride Uranium-Plutonium Fuel Pin Cladding

Criticality analysis of pyrochemical reprocessing apparatuses for mixed uranium-plutonium nitride spent nuclear fuel using the MCU-FR and MCNP program codes

A preliminary criticality analysis for novel pyrochemical apparatuses for the reprocessing of mixed uranium-plutonium nitride spent nuclear fuel from the BREST-OD-300 reactor was performed. High-temperature processing apparatuses, “metallization” electrolyzer, refinery remelting apparatus, refining electrolyzer, and “soft” chlorination apparatus are considered in this work. Computational models of apparatuses for two neutron radiation transport codes (MCU-FR and MCNP) were developed and calculations for criticality were completed using the Monte Carlo method.The criticality analysis was performed for different loads of fissile material into the apparatuses including overloading conditions. Various emergency situations were considered, in particular, those associated with water ingress into the chamber of the refinery remelting apparatus. It was revealed that for all the considered computational models nuclear safety rules are satisfied.

Synthesis of Mixed Actinide Oxides Using Microwave Radiation

A method has been developed for producing mixed actinide oxides suitable for fabricating mixed nitride uranium plutonium fuel for fast neutron reactors. The method is based on the use of microwave radiation for the direct denitration of actinide nitrate solutions. The possibility of producing uranium, plutonium, and neptunium-mixed oxides was shown. A pilot installation for preparing actinide oxides by microwave denitration was designed and tested. Mixed oxides of uranium and cerium (for plutonium imitation) were successfully used to synthesize uranium cerium nitrides and produce fuel pellets. Compared with the precipitation (ammonia) method of producing mixed oxides, microwave denitration reduces the generation of secondary liquid radioactive waste by more than six times.

Thermogravimetric study of mixed uranium-plutonium fuel for prospective generation IV reactors

The analysis of the accidental regime of the MNIT fuel reactor facility has indicated a possible scenario of decomposition of the fuel composition with the release of metallic plutonium, which is unacceptable during operation of a nuclear reactor. The present paper studies the behaviour of mixed uranium-plutonium nitride nuclear fuel (MNIT fuel) at high temperatures (up to 2400 K) in a helium flow by thermogravimetry using an STA 449 F1 thermal analyser. The nitride fuel is shown to sublimate at a temperature of about 2000 K. X-ray microanalysis of the samples prior to and after high-temperature testing has established that mass loss in the mixed nitride is primarily due to plutonium evaporation.

Certain Aspects of Using 15N-Enriched Mixed Uranium-Plutonium Nitride Fuel in a Fast-Reactor Fuel Cycle

Certain Aspects of Using 15N-Enriched Mixed Uranium-Plutonium Nitride Fuel in a Fast-Reactor Fuel Cycle

SNF processing electrochemical operations: liquid-metal and salt medium purification

The paper investigates the process of regeneration of a liquid metal medium used in the pyroelectrochemical reprocessing of spent mixed uranium-plutonium nitride fuel produced by a fast neutron reactor. The investigation concerns the interaction of liquid cadmium with sludge formed during the anodic dissolution of ceramic nitride pellets in a 3LiCl-2KCl melt medium as well as the possibility of its purification by filtration from individual metal fission products. Anode sludge is represented by fission products of the platinum group, zirconium, molybdenum and technetium. It was determined by scanning electron microscopy that the metal product is composed of several intergrowth phases. It was found that upon contact of a polymetallic alloy simulating anode sludge with a melt, the liquid metal phase is saturated to 0.025 wt% of Pd, 0.01 wt% of Rh for 50 hours at 500 °C, while zirconium forms an insoluble dispersed intermetallic compound ZrCd3. Powders of molybdenum and technetium, which are not wetted with cadmium, can be completely removed using a filter mesh of plain weaving of the P-200 type. It is also possible to remove zirconium from anodic cadmium by filtration. The filtration efficiency of ruthenium and palladium powders did not exceed 54.3 and 13.1 wt%, respectively, due to partial dissolution and thinning of particles, which will lead to saturation of the liquid metal phase and the need to purify it by alternative methods.

Key Outcomes of Comprehensive Computational and Experimental Validation of Fuel Rods with Mixed Uranium-Plutonium Nitride Fuel for the BN-1200 and Brest Reactors

Key Outcomes of Comprehensive Computational and Experimental Validation of Fuel Rods with Mixed Uranium-Plutonium Nitride Fuel for the BN-1200 and Brest Reactors

РАДИАЦИОННО-ГИГИЕНИЧЕСКИЕ ИССЛЕДОВАНИЯ ЭКСПЕРИМЕНТАЛЬНОГО ПРОИЗВОДСТВА СМЕШАННОГО НИТРИДНОГО УРАН-ПЛУТОНИЕВОГО ТОПЛИВА НА АО «СХК». Часть 2: Дозы и риски

Purpose: Assessment of the compliance of the radiation protection of workers at the complex experimental installations of JSC SChC with the requirements of the Russian radiation safety standards NRB-99/2009 to limit the generalized risk of potential exposure and the IAEA recommendations for not exceeding the control level of the minimum significant radiation risk. Materials and methods: The results of radiation-hygienic investigations of radiation exposure factors affecting workers involved in the manufacture of mixed uranium-plutonium nitride (MUPN) fuel at the complex experimental installations of JSC SChC are used as the input data for the preliminary assessment of radiation doses to workers. The models for assessment of radiation risk of potential exposure have been developed in accordance with the recommendations of the ICRP and the IAEA. Results: Preliminary estimates of the doses of external gamma-neutron (2.5 ± 0.5 mSv / year) and internal exposure of workers (~1 mSv/year1) are related to the current levels of exposure of workers of complex experimental installations. These levels are the result of exposure to ionizing radiation sources associated both with the development of new technologies and with residual radioactive contamination resulting from previous activities not related to the manufacture of MUPN fuel. The presented dose estimates are related to the use of raw materials that have undergone deep preliminary purification from radiogenic impurities. When irradiated nuclear materials are used as raw materials, the levels of gamma-neutron exposure to workers will be significantly higher. The maximum increase in the generalized risk of potential exposure due to annual exposure is estimated for women aged 18 years at the beginning of exposure, for which the increase in this risk is 1.45 × 10-4 year-1, which is 1.37 times lower than the limit established by the Russian radiation safety standards NRB-99/2009: 2 × 10-4 year-1. All predicted values of the lifetime attributable fraction of radiation (LARF) in mortality from malignant neoplasms are significantly less than the control level of the minimum significant risk recommended by the IAEA (LARF = 5 %), and the maximum value of LARF = 2.8 % is achieved for women aged 18 years at the beginning of exposure. Conclusion: Restrictions on the radiation risks of potential exposure, established by NRB-99/2009, as well as those recommended by the IAEA for not exceeding the reference level of the minimum significant risk, are met with a large reserve. The results obtained and the developed methods will be used to ensure the radiation safety of workers during the transition from experimental installations to pilot industrial implementation of the technology for the production of MUPN fuel.

Perspective Compounds for Immobilization of Spent Electrolyte from Pyrochemical Processing of Spent Nuclear Fuel

Immobilization of spent electrolyte–radioactive waste (RW) generated during the pyrochemical processing of mixed nitride uranium–plutonium spent nuclear fuel is an acute task for further development of the closed nuclear fuel cycle with fast neutron reactors. The electrolyte is a mixture of chloride salts that cannot be immobilized directly in conventional cement or glass matrix. In this work, a low-temperature magnesium potassium phosphate (MPP) matrix and two types of high-temperature matrices (sodium aluminoironphosphate (NAFP) glass and ceramics based on bentonite clay) were synthesized. Two systems (Li0.4K0.28La0.08Cs0.016Sr0.016Ba0.016Cl and Li0.56K0.40Cs0.02Sr0.02Cl) were used as spent electrolyte imitators. The phase composition and structure of obtained materials were studied by XRD and SEM-EDS methods. The differential leaching rate of Cs from MPP compound and ceramic based on bentonite clay was about 10−5 g/(cm2·day), and the rate of Na from NAFP glass was about 10−6 g/(cm2·day). The rate of 239Pu from MPP compound (leaching at 25 °C) and NAFP glass (leaching at 90 °C) was about 10−6 and 10−7 g/(cm2·day), respectively. All the synthesized materials demonstrated high hydrolytic, mechanical compression strength (40–50 MPa) even after thermal (up to 450 °C) and irradiation (up to 109 Gy) tests. The characteristics of the studied matrices correspond to the current requirements to immobilized high-level RW, that allow us to suggest these materials for industrial processing of the spent electrolyte.

РАДИАЦИОННО-ГИГИЕНИЧЕСКИЕ ИССЛЕДОВАНИЯ ЭКСПЕРИМЕНТАЛЬНОГО ПРОИЗВОДСТВА СМЕШАННОГО НИТРИДНОГО УРАН-ПЛУТОНИЕВОГО ТОПЛИВА НА АО «СХК». Часть 1: Методы и результаты

Purpose: To present the methods and results of studies of the factors of radiation exposure to workers involved in the manufacture of mixed uranium-plutonium nitride (MUPN) fuel at the complex experimental installations CEI-1 and CEI-2 of JSC SChC.
 Material and Methods: Regularities of the formation of external exposure doses have been revealed based on the study of the dynamics of the ambient dose equivalent rate (ADER) of photon and neutron radiation at the CEI-1 and CEI-2 workplaces, as well as instrumental individual dosimetric control of the equivalent doses to workers. In order to assess the inhalation intake and possible doses from internal irradiation, studies of the physicochemical properties of radioactive aerosols were carried out.
 Results: It has been found that the main sources of penetrating radiation in the premises of CEI-1 are boxes where tablets are pressed, chips and rejected tablets are crushed, as well as temporary storage of products is occurred. The highest ADER values have been measured in those boxes, where the radiation exposure was due to radioactive contamination caused by past activity, and is not associated with fabrication of MUPN fuel. A significant contribution of neutron exposure to individual doses of workers was measured, which exceeded the contribution of gamma exposure at some workplaces of the CEI-1. At CEI-2, a non-functioning exhaust ventilation pipe passing over the premises was found to be a powerful source of external radiation. This pipe contained a significant amount of radioactive material. Assessment of the contribution of gamma exposure from the ventilation pipe to the external exposure of workers reached 85% at some workplaces. Studies of the physicochemical properties of radioactive aerosols have revealed a high reactivity of MUPN compounds, leading to instant oxidation of the thoracic fraction of MUPN fuel aerosols under contact with air. The complex morphological and dispersed composition of aerosol particles in combination with a complex chemical composition caused by the aging processes of aerosols, can lead to a fundamental difference in the biokinetics of MUPN aerosols, the process of dose formation and, consequently, the degree of radiological hazard compared to those adopted in the ICRP models for U and Pu.
 The results of the current radiation-hygienic research are of a preliminary nature, since the object of this research is an experimental installation, which was used to develop a new technology for the production of MUPN fuel. The instrumental and methodological approaches to assess the factors of radiation exposure to workers tested at these experimental installations, will be used in the future to conduct similar studies during the pilot industrial operation of new modules for fabricating and refurbishing of MUPN fuel.

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.