Mixed Uranium-plutonium Research Articles

The influence of precipitates upon the thermal conductivity of nitride fuel

In the article presented are the results modeling the behavior of mixed uranium-plutonium nitride fuel with an admixture of oxygen and carbon during irradiation in fast reactor. It was shown that accumulation of fission products and formation of secondary phases lead to formation of heterogeneous multicomponent solid solution including, in particular, metal, telluride, oxide, intermetal and sesquialteral nitride secondary phases. Influence of the secondary phases formed during irradiation on the effective thermal conductivity of the fuel is assessed in the framework of a generalized theory by Odelevskii developed for heterogeneous systems with isolated inclusions. It was found that in the irradiated nitride fuel, the main correction of the thermal conductivity is related to formation of the U2N3 phase. As it is shown, that with increase of fuel burnup and growth of secondary condensed phases, the effective thermal conductivity decreases up to about 3% for burnup of 5.5%.

Дисперсность, морфология и элементный состав аэрозольных частиц на производстве смешанного нитридного уран‑плутониевого топлива

Purpose: Study of physical and chemical properties of radioactive aerosols formed on the fabrication of mixed nitride uranium-plutonium (MNUP) fuel.
 Material and methods: The following impactors were utilized for activity particle-size distribution analysis: AIP-2, PHRT, IRAM-2-4I (SRC FMBC), Andersen cascade impactor (Copley Scientific, UK), SKC Sioutas (SKC inc., USA). Scanning electron microscope (SEM) Tescan LYRA-3 equipped the X-ray microanalyzer X-max 80 (Oxford Instruments) was used for study of morphological characteristics of aerosol particles. Secondary ion-mass spectrometer Cameca IMS-1280 and track analysis were used for search of particles containing the uranium and plutonium radionuclides. 
 Results: Values of AMAD (Σa) vary from 12 µm till 30 µm, for 239Pu – from 14 till 27 µm. Lowest AMAD values (0.4–2.5 µm for 239Pu) were found in the repair area near the glove box of synthesis and sintering. The samples contain a much larger number of uranium-containing particles, which is likely due to a larger mass fraction of uranium oxide compared with plutonium dioxide in the initial mixture of reagents. Elemental composition of aerosol particles includes U (63–86 %), Pu (5–10 %) and О (9–47 %), Fe – in some samples till 32 %. Other elements like Nа, S, N, Р are contained in rather lowest amounts not exceeding 1–8 %. Only one nitrogen-containing particle was found (3 % of N).
 Conclusion: Coarse fraction of radioactive aerosols makes a major contribution in activity concentration in air of the operator zone which is driven by such operations as pressing and crushing. Aerosol particles are presented as individual particles or dense aggregates by size of 0,2– 2 µm and also as conglomerates by size 0,5–4 µm consisting of oxides (or carbides) of silicon, iron, calcium, etc. containing inclusions (200–400 nm) or individual particles (20–200 nm) of mixed U-Pu oxide or uranium oxide on the surface of conglomerate as well aslocated in the near-surface layers inside.

ОЦЕНКА РАДИАЦИОННОЙ БЕЗОПАСНОСТИ ТРАНСПОРТНОГО КОНТЕЙНЕРА ПРИ ПЕРЕВОЗКЕ ОТРАБОТАВШЕГО НИТРИДНОГО УРАН-ПЛУТОНИЕВОГО ТОПЛИВА БН-600 С ИСПОЛЬЗОВАНИЕМ ПРОГРАММЫ MCU-FR

At present, the BN-600 and BN-800 reactors are used for tests of experimental fuel assemblies containing MOX and MNUP fuels. For transportation of spent fuel assemblies of BN-600 with mixed nitride uranium-plutonium fuel is planned to use a regular transport container containing 35 fuel assemblies. The purpose of the work is to develop a calculation model of the transport container containing 35 fuel assemblies and radiation safety evaluation of the transport container during carting spent MNUP fuel of BN-600 using the MCU-FR Monte Carlo code. The developed model of the experimental fuel assembly is integrated into the developed calculation model of the transport container. The resulting model was used to calculate the equivalent dose rate from sources of ionizing radiation in the fuel. The azimuthal and radial distributions of the equivalent dose rate in the plane of the FA core center were considered. In the course of radiation safety evaluation, it was shown that under normal transportation conditions, the maximum total calculated radiation levels on the container surface are determined by primary photon radiation and do not exceed the established limit of 10 mSv/h under exclusive use. The developed model of the transport container and the calculation results obtained by means of the MCU-FR code will be used to prepare the validation base of the ODETTA calculation code.

ВОЗМОЖНОСТЬ ОБЛАГОРАЖИВАНИЯ ПЛУТОНИЯ В БН-800

The concept of the development of two-component nuclear energy based on fast and thermal reactors, adopted in Russia, suggests that in case of intensive development of nuclear reactors, fast reactors will provide fuel for thermal neutron reactors, while plutonium from VVER SNF will be used to produce mixed uranium-plutonium (MOX) fuel fast sodium reactors. Currently, in the global nuclear energy sector, some PWR units are licensed to use a combination of conventional and MOX fuel in them or are in the process of licensing. But plutonium obtained during the reprocessing of spent fuel from the MOX fuel cannot be reused in thermal reactors due to an exceedingly high content of heavy (even) isotopes that do not fission in the thermal-neutron spectrum. As a result, plutonium is accumulated, which is a definite problem for European energy. This problem can be partially solved with the help of fast reactors. Currently, fast energy sodium reactors operate only in Russia - BN-600 and BN-800. BN-800 can operate on MOX fuel, industrial remote technologies to fabricate such fuel is now being developed at a plant belonging to the Mining and Chemical Complex. Having a fast reactor and remote MOX fuel technologies, it becomes possible to organize the export of services - ″improve″ of foreign-grade plutonium unsuitable for reuse in thermal reactors. The joint use of Russian fast reactor technologies and French MOX fuel technologies in thermal reactors for Russia opens up the world export market for plutonium enrichment services, and for the whole world - a new stage in the development of global two-component nuclear energy.

Berkut Code Validation on Post-Reactor Studies of Irradiated Bn-600 Fuel Rods with Mixed Uranium-Plutonium Nitride Fuel

An advanced version of the BERKUT-U/V2.1 code intended for modeling the behavior of fuel rods with oxide or nitride fuel in standard and emergency operating regimes of fast reactors with liquid-metal coolant is being developed at the Nuclear Safety Institute, Russian Academy of Sciences (IBRAE RAS), as part of the project Next-Generation Codes of Project Proryv (Breakthrough). The present work reports the results of the validation of an advanced version of the code based on post-reactor studies of BN-600 fuel rods (KETVS-1, -6) and BREST-OD-300 fuel rods (ETVS-5) with mixed uranium-plutonium nitride fuel, irradiated in the BN-600 reactor.

Development and verification of CFD model of carbothermal synthesis furnace for production of mixed nitride uranium-plutonium fuel

The retort batch furnace for the carbothermal synthesis of uranium and plutonium nitrides is a component of the mixed nitride uranium-plutonium (MNUP) fuel Fabrication/Refabrication Module (FRM), a part of the Pilot Demonstration Energy Complex (PDEC). A CFD model of the furnace was built in the SolidWorks Flow Simulation code to check the feasibility of its thermophysical operating modes (heating and cooling rates, temperature of the product loaded into the furnace). The experimental data obtained from the performance tests was used to verify the developed CFD model and to confirm its adequacy. The relative deviation of the calculated temperature of the loaded product from the experimental data in the process of isothermal annealing does not exceed 0.7%. The temperatures of the loaded product predicted by the CFD model were used to justify engineering solutions for the uranium and plutonium nitrides carbothermal synthesis furnace. The CFD model can be used to define the furnace operation mode by selecting the gas flow rates inside and outside the retort, the heater temperature, and heating and cooling rates for the product loaded in the furnace.

Integrating project and the PDEC information model

PRORYV Project is one of the major advanced global projects in the nuclear power sector that provides for creation of the new generation nuclear power technologies based on closed nuclear fuel cycle using fast neutron reactors. On JSC “SGChE” site a Pilot Demonstration Power Complex, which consist of BREST-OD-300 (reactor with a lead coolant) and on-site plant based on closed nuclear fuel cycle is being built. On-site Plant will include the Unit for processing irradiated mixed uranium-plutonium (nitride) fuel and the Unit for fabrication the reprocessed irradiated nuclear fuel. The need to apply modern and progressive tools for process management within PRORYV Project’facilities life cycle is required by following key features: due to the large number of participants in the project and the breadth of the range of tasks to be solved, there are different information, calculation and modeling tools and approaches to their application; the high degree results’ innovation introduces considerable uncertainty in the methods of their achievement and requires a large number of R & D, the results of which constantly cause changes in the facilities’ design. While developing PRORYV Project the Information Model is created. Information Model is a continuously updated structured set of electronic data and documents on each facility and technology, which is necessary and sufficient for each stage of R & D and facilities’ life cycle (location, design, construction, commissioning, operation and decommissioning). The Information Model has several modes that are architectural, construction, functional-technological, and design. The Information Model is linked to the Project Schedule and the requirements management system. The Information Model is the basis for the creation of other systems by linking them with each other. The Integrating Project is the visualized part of the Information Model, which represents a set of electronic data and documents on PRORYV Project and contains the actual design, engineering and technological documentation on the facilities, as well as advanced data monitoring capabilities (in particular: 3D model, 4D simulation model, construction 4D-model and mathematical model). Integrating Project is required for modeling all stages of the life cycle of facilities with the purpose of timely analysis of facilities’ characteristics and technological processes. The Integrating Project is created on the basis of the Information model and is part of it. At the stage of operation and subsequent decommissioning, the Integrating project is used and supplemented by the operating organization. Due to the application of Integrating Project tools, the following works have been executed: activity on optimization and justification of the number of personnel at the Pilot Demonstration Energy Complex is organized; discrepancies in the development of electronic management procedures in the course of verification of the manuals for the operation of Fabrication Unit equipment were revealed; at an early development, conflicts in project documentation were identified. Recommendations for adjusting the technological and design documentation on BREST-OD-300 reactor unit as well as Processing and Fabrication Units were developed and issued.

Optimization of RW volumes from reprocessing of SNF from fast reactors. Fractionation options

In this work, are analyzed the volume of vitrified radioactive waste generated during the processing of spent nuclear fuel from fast reactors (using the example of processing mixed uranium-plutonium nitride spent nuclear fuel from the BREST-OD-300 reactor).The analysis showed- there is a minimum acceptable volume of HLW from spent nuclear fuel reprocessing, determined by the parameters of permissible thermal and radiation effects on the matrices used for RW utilization, as well as by the restriction on the chemical composition of the matrices, fixed in regulatory documents;- the allocation of the fraction of minor actinides (MA) during the reprocessing of the fast reactor spent nuclear fuel significantly reduces the required volume of the matrix for disposal of HLW;- the currently technically achievable level of MA (99.9%) and plutonium (99.975%) recovery during the reprocessing SNF allows reducing the time to achieve radiation equivalence of the generated RW and uranium feed used to produce mixed nitride uranium-plutonium nuclear fuel to 300 years;- a complete rejection of the deep underground disposal of RW from SNF reprocessing (rejection of the formation of radioactive waste of the 1st and 2nd classes) is not economically justified, because leads to the need for the formation of a large volume of RW of the 3rd class;- minimization of the RW volume to be deep disposed is achieved by a combination of fractionation (separation of the MA fraction) and the use of the principle of multi-stage processing of HLW, which assumes long-term intermediate controlled storage of HLW (50-60 years).

Hydrometallurgical and combined technologies fast reactor MNUP and MOX UNF reprocessing

Russian Federation is developing the key technologies for closed nuclear fuel cycle with FR used mixed uranium-plutonium nitride or oxide (MOX) fuel within the framework of the project “Proryv”. This is a new integrated comprehensive product for nuclear power, which provides in the future- no accumulation of spent nuclear fuel;- radioactive waste management based on the principles of radiation-equivalent disposal;- technological support for the non-proliferation treaty;- competitiveness with other large-scale power technologies.The experimental and demonstration energy complex is under construction now on the site of Siberian Chemical Combine for the industrial development of the closed nuclear fuel cycle. ODEC will include a fast reactor BREST-OD-300 with lead coolant, facility for fabricating/refabrication of nitride fuel and the spent nuclear fuel processing facility.Up today, the key technology elements were demonstrated using macro quantities U, Pu, Np, irradiated mixed nitride fuel and real high-level waste. Within the framework of the project “Proryv” a prototype of the fuel cycle of fast reactors is developing.

Application of Dynamic Cross-Flow Filtration for Solving the Technological Problems of Project Breakthrough

The results of experiments on the clarification of solution simulators obtainable by means of dynamic crossflow filtration during the reprocessing of mixed uranium-plutonium nitride spent fuel are presented. The experiments were performed using full-scale, dynamic, cross-fl ow filtration stands at the RT-1 radiochemical plant at the Mayak Production Association. It is shown that this method yields high-quality clarification of solutions without the use of flocculants – the obtained permeates were completely free of the solid phase and silicic acid in the γ-form. In addition, the concentration of the solid phase can reach 100 g/cm3 without loss-of-clarification. It is shown that the method of dynamic cross-fl ow fi ltration is effective for clearing transuranium elements from liquid radwaste. After their co-precipitation with the carrier (trivalent iron) and clarification, the obtained permeates are low-level wastes according to the Basic Sanitary Rules for Radiation Safety (OSPORB-99/2010).

Radiation and Chemical Stability of a Magnesium-Phosphate Matrix for 14C Immobilization

The results of manufacturing, studying the radiation and hydrolytic stability of a magnesium-potassium phosphate (MPP) matrix, are presented, which is being developed for the purpose of immobilizing 14C, a product of processing mixed uranium-plutonium nitride fuel. Samples of MPP matrix containing 20 and 40 wt % calcium carbonate as a waste form fixing 14C were irradiated with accelerated electrons with energies up to 2.5 MeV to doses of 106, 107 and 108 Gy. It was found that the destruction of the target phases begins at a dose of 107 Gy up to their complete decomposition after the accumulation of a dose of 108 Gy. A significant increase in the leaching rates for all studied cations from radiation-damaged samples after a dose accumulation of 108 Gy has not been established; the duration of the geological storage of the matrix until complete decomposition of the main 14C fixing phases has been estimated.

Determination of 232U and 236Pu Weight Contents of Spent Nuclear Fuel

A procedure has been developed for the determination of 232U and 236Pu content by alpha spectrometry by ion-exchange isolation and purification of uranium and plutonium fractions from irradiated fuel solution. The proposed procedure has been utilized to determine weight concentrations of 232U and 236Pu in samples of uranium oxide fuel irradiated in VVER-1000 and RBMK-1000 reactors, as well as in mixed uranium-plutonium nitride fuel irradiated in a BN-600 reactor.

Разработка и верификация CFD-модели печи карботермического синтеза нитридов урана и плутония

Разработка и верификация CFD-модели печи карботермического синтеза нитридов урана и плутония

Effect of Nitride Nuclear Fuel Structure and Phase Composition on Fuel-Rod Life

The behavior of mixed uranium-plutonium nitride fuel and uranium nitride was studied by means of thermogravimetry at high temperatures (to 2173 K) in a helium flow. The mass loss of the nitride samples was found in the low temperature range (<1773 K), which is not associated with decomposition of uranium or plutonium mononitrides. The mass loss occurs in two stages and is accompanied by the release of nitrogen. It is shown that nitride fuel can contain up to several percent of uranium sesquinitride U2N3, which decomposes in the indicated temperature range and can strongly affect pellet integrity and fuel-rod life during operation.

Mathematical Model of Cadmium Distillation at the Stage of Pyrochemical Reprocessing of Spent Mixed Uranium-Plutonium Nitride Nuclear Fuel

A mathematical model of the technological process of cadmium distillation at the stage of pyrochemical reprocessing of spent mixed uranium-plutonium nitride fuel for the purpose of studying the serviceability of and developing monitoring and process control systems is presented. The basic thermodynamic and kinetic characteristics and the principles of mathematical modeling of vacuum distillation of liquid mixtures, specifically, cadmium, are described. The requirements of the model and the assumptions are formulated. The information structure and mathematical description of the model were developed. Computational experiments were performed on the model and showed its serviceability and possibility of use for developing monitoring and process control systems.

Results of Studies of BN-600 Fuel Rods with Mixed Uranium-Plutonium Nitride Fuel and ChS68-ID c.d. Steel Cladding

Reactor tests in BN-600 and post-reactor studies of experimental FA of fuel rods with mixed nitride fuel and different cladding are being conducted in Project Breakthrough. In the present article, the irradiation parameters of fuel rods with mixed nitride fuel and cladding made from ChS68-ID c.d. austenitic steel are presented: maximum linear power density 38.3 kW/m, maximum fuel burnup 7.5% h.a., and maximum damage dose 73.6 dpa. No depressurization of the fuel elements was recorded. The results of post-reactor examination (PIE) of the irradiated fuel rods are presented – swelling, gas release from fuel, and mechanical and corrosion characteristics of the cladding.

Carbothermal Facility for the Synthesis of Nitrides for Mixed Uranium-Plutonium Fuel Production

A facility for carbothermal synthesis of uranium and plutonium nitrides for the production of dense mixed uranium-plutonium nitride fuel for fast reactors is described. The synthesis reaction occurs at 1650 ± 50°C in nitrogen and nitrogen-hydrogen atmospheres in six high-temperature batch furnaces of the horizontal type. Tests performed on the pilot model of a furnace showed good agreement between the experimental and computed temperature distributions along the longitudinal axis of the furnace. The facility operates in an automatic mode; control and monitoring of the technological process are performed remotely.

Results of Investigations of BREST-Type Reactor Fuel Rods with Mixed Uranium-Plutonium Nitride Fuel, Irradiated in BOR-60 and BN-600

The complex program of computational and experimental validation of dense fuel for fast reactors in Project Breakthrough provides for testing of fuel rods with mixed uranium-plutonium nitride fuel in the BOR-60 and BN-600 reactors. The irradiation parameters and post-reactor studies of the fuel rods of a BREST reactor are presented. Testing of fuel rods in BOR-60 to maximum burnup 5% h.a. and damaging dose 74.8 dpa and in BN-600 to maximum burnup 4.5% h.a. and damaging dose 53 dpa did not result in depressurization of the rods. The results of post-reactor tests of irradiated fuel rods – swelling of fuel, gas release from the fuel, mechanical and corrosion properties of the cladding – are presented.

Development of Sodium-Cooled Fast Reactors Under Modern Conditions: Challenges and Stimuli

The factors that can promote stable development of sodium-cooled fast reactors with mixed uranium and plutonium fuel under conditions of variable demand for atomic energy, types and magnitude of NFC services in the domestic and foreign markets are examined. The possibility of effective control of the balance of plutonium in a two-component nuclear power system with possible change in the demand for electricity and nuclear fuel of reactors built at home and abroad is shown for the example of an analysis of variants for managing fuel in an industrial power-fuel complex with BN-1200 (PEK-BN). It is noted that high-quality plutonium, which can be produced in PEK-BN, for repeated recycling in the VVER system with uranium and mixed uranium plutonium fuel, plays a fundamental role, and promising directions for expanding the fuel business of the State Corporation Rosatom by providing new types of services on the domestic and foreign markets are noted.

Capabilities of gamma-spectrometry methods for on-line monitoring of nitride SNF pyrochemical reprocessing

The paper reports the first test results for detectors of various types and equipment of gamma-spectrometry channels under external radiation originating from pyrochemical reprocessing of spent mixed nitride uranium-plutonium (MNUP) fuel. Testing was carried out on a solid-state detector with a CdZnTe crystal, a scintillation detector with a LaBr3crystal, and an ionization chamber based on compressed xenon. Simulated external gamma-radiation was created by means of a Co-based scattered gamma-radiation source. The paper also describes an experimental facility and a measurement technique, and presents the facility testing results for the above three detectors. The solid-state detector was proved to have the best performance. However, achieving the design characteristics of the gamma-spectrometry channel requires new solutions for protection and collimation of gamma-radiation produced by a real MNUP SNF reprocessing facility. What is meant here is the influence of the detectors’ geometry on the configuration of the protective collimator which is proposed to be used in real conditions. Thus, if a Xe-based detector is used, the calculated mass of the protective collimator is 900 kg, while it is possible to use less massive protection for the other detectors. In addition, when manufacturing neutron shielding for detectors based on CdZnTe and LaBr3, it is necessary to consider the neutron radiation factor in MNUP SNF processing. It is possible to surround the collimator with a moderating layer (for example, polyethylene) and create inside it a skin from a thermal neutron absorber (for example, based on cadmium).