Mixed Uranium-plutonium Research Articles

Local Gas Purification System in Spent Nitride Fuel Oxidation

A local gas-purification system of the voloxidation operation is examined within the framework of the technology of regeneration of spent mixed uranium-plutonium nitride fuel from fast reactors. The system includes a metal-fabric filter, HEPA filter, ruthenium capture unit, an adsorber for capturing iodine, a unit for additional oxidation and trapping of hydrogen (tritium), and a unit for capturing 14CO2. A description is given of each unit, the overall scheme, and the possibilities of the gas-purification system.

Complex Method of Non-Destructive Measurements of Plutonium Composition and Mass

A complex method of performing non-destructive measurements of the isotopic composition and mass of plutonium that does not require external data on 242Pu content is described. The method includes a set of measurements performed by means of γ-spectrometry, calorimetry, and neutron multiplicity. Determinations of the metrological characteristics showed that in terms of accuracy it greatly surpasses the existing analogues, approaching the coulometric and isotopic dilution methods. The method is recommended for use in the industrial production of mixed uranium-plutonium fuel.

Radiochemical analysis of a sample of mixed uranium–plutonium nitride fuel

Radiochemical analysis of a mixed uranium–plutonium nitride (MUPN) fuel sample irradiated in a BOR-60 reactor was performed. The study was made using the set of procedures developed at the Research Institute of Nuclear Reactors for determining the nuclide composition and gravimetric content of U, Pu, Am, Cm, Nd and other fission products, platinum group metals, and transition metals, including nuclear physical methods, atomic emission spectrum analysis, mass spectrometry for determining the nuclide composition, and isotope dilution mass spectrometry for determining the gravimetric content of nuclides with preliminary radio-chemical separation of fractions of elements by ion-exchange, extraction-chromatographic, precipitation, and distillation methods. The MUPN fuel burn-up was determined from the ratio of the number of atoms of the fission product selected as a burn-up monitor to the number of heavy atoms in the dissolved fuel sample (method of fission product accumulation, MFP). The 145Nd + 146Nd sum and 148Nd were used as burn-up monitors.

Investigations of Mixed Uranium-Plutonium Nitride Fuel in Project Breakthrough

Mixed uranium-plutonium nitride fuel is seen as a promising fuel for fast reactors with a closed fuel cycle. A complex program of computational and experimental validation of this fuel up to 2020 has been developed within the scope of Project Breakthrough. Pre-reactor and post-reactor studies of mixed uranium-plutonium nitride fuel are part of this program. The basic research results obtained by the end of 2016 are examined.

Effect of carbon impurity in irradiated mixed uranium–plutonium nitride on the phase composition upon β−-decay of fission products

As shown by thermodynamic modeling, accumulation of fission products in the course of irradiation of mixed uranium–plutonium nitride containing 0.25 wt % C impurity leads to the formation of a multicomponent carbonitride solid solution containing U, Pu, Am, Np, Zr, Y, lanthanides, and also separate carbide (UMoC2, U2RuC2, URu3C0,7, BaC2, SrC2) and nitride (U2N3) phases and intermetallic compounds U(Ru, Rh, Pd)3. The β−-decay of metallic radionuclides in separate carbide phases of spent nuclear fuel leads to changes in their chemical compositions. The transformation kinetics of separate dicarbide (89SrC2 → 89YC2, 90SrC2 → 90YC2 → 90ZrC + C, 140BaC2 → 140LaC2 → 140CeC2) and complex carbide (U99MoC2 → UC2 + 99Tc) phases was calculated.

Evaluation of the Effect of Linear Scattering Anisotropy and Boundary Shape on the Calculation of K ∞ for VVER-1000 Cells with Mixed Uranium-Plutonium Power-Grade Fuel

The method of surface pseudo sources for solving the multigroup neutron transport equation with scattering anisotropy and different boundary conditions by the operator splitting method was developed previously. The method was implemented in the RATsIYa option of the WIMS-SH-2.0 complex. The impact of scattering anisotropy and the form of the boundary conditions on K ∞ of VVER-1000 reactor cells with uranium-plutonium power-grade fuel was investigated. The calculations showed that the impact of each of these two effects is about 0.5%; they have different signs, as a result of which they largely compensate one another.

Nuclear fuel cycle and its impact on the environment

In this paper, we consider the present-day situation and outlooks of the development of nuclear power generation in Russia and other countries. It was noted that the implementation of the concept of a closed nuclear-fuel cycle accepted in Russia relies on the solution of the problem of the disposal of spent nuclear fuel (SNF) and radioactive waste (RAW). This paper presents the main results of investigations focused on the development of radiation-safe methods of manufacturing nuclear fuel elements, including mixed uranium–plutonium oxide fuel for fast-neutron reactors; creation of low waste-production technologies of SNF processing and RAW disposal; and the analysis of fundamental features of the behavior and speciation of radionuclides in environmental objects for the development of efficient methods of radioecological monitoring and remediation of radionuclide-contaminated areas.

Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium–plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel assemblies. Based on analysis of modern engineering solutions on SNF regeneration, it has been concluded that new reprocessing technologies should meet the ecological safety requirements, provide a more extensive use of the resource base of nuclear engineering, allow the production of valuable and trace elements on an industrial scale, and decrease radioactive waste release.

РH process as a technology for reprocessing mixed uranium–plutonium fuel from BREST-OD-300 reactor

A technology for reprocessing mixed uranium–plutonium nitride fiel (MUPN) from BREST reactor is considered. The technology should ensure reprocessing of spent nuclear fuel with the storage time after irradiation of no more than 1 year, 10–15% content of fissile materials (FM), and burn-up of 10% of heavy atoms. The target product of the technology is a mixture of actinide oxides separated from fission products with the separation factor of ~106. A PH (Pyro–Hydro) process was suggested for MUPN SNF reprocessing. It involves pyroelectrochemical fuel reprocessing with separation of U, Np, and Pu from the major fraction of fission products responsible for the heat release from the fuel and for the radiation load on process media, a series of hydrometallurgical operations for final purification of the target products (U–Pu–Np–Am), and radioactive waste (RW) management. The PH process is being developed since 2011 by the teams from the Bochvar High-Tech Research Institute of Inorganic Materials, Khlopin Radium Institute, and Research Institute of Atomic Reactors with active participation of the Leading Research Institute of Chemical Technology, Siberian Chemical Combine, and institutes of the Russian Academy of Sciences, primarily Frumkin Institute of Physical Chemistry and Electrochemistry and Vernadsky Institute of Geochemistry and Analytical Chemistry.

Melting behavior of mixed U–Pu oxides under oxidizing conditions

In order to use mixed U–Pu oxide ceramics in present and future nuclear reactors, their physical and chemical properties need to be well determined. The behavior of stoichiometric (U,Pu)O2 compounds is relatively well understood, but the effects of oxygen stoichiometry on the fuel performance and stability are often still obscure. In the present work, a series of laser melting experiments were carried out to determine the impact of an oxidizing atmosphere, and in consequence the departure from a stoichiometric composition on the melting behavior of six mixed uranium plutonium oxides with Pu content ranging from 14 to 62wt%. The starting materials were disks cut from sintered stoichiometric pellets. For each composition we have performed two laser melting experiments in pressurized air, each consisting of four shots of different duration and intensity. During the experiments we recorded the temperature at the surface of the sample with a pyrometer. Phase transitions were qualitatively identified with the help of a reflected blue laser. The observed phase transitions occur at a systematically lower temperature, the lower the Pu content of the studied sample. It is consistent with the fact that uranium dioxide is easily oxidized at elevated temperatures, forming chemical species rich in oxygen, which melt at a lower temperature and are more volatile. To our knowledge this campaign is a first attempt to quantitatively determine the effect of O/M on the melting temperature of MOX.

Fabrication of mixed uranium–plutonium carbide fuel pellets with a low oxygen content and an open-pore microstructure

Mixed uranium–plutonium carbides are considered as potential fuels for Generation IV fast reactors. Compared to that of oxide fuels, their fabrication with regard to required specifications is more difficult. Carbides are obtained by a two-step procedure from the mixture of UO2, PuO2 and graphite powders. Due to their extreme reactivity with oxygen and moisture, fabrication and handling of carbide fuels are performed in glove boxes maintained under a dynamic flow of nitrogen providing operational safety. Nevertheless, oxygen pickup during processing of the material seems unavoidable and thus must be limited by suitable procedures. The oxygen content in the carbides can vary over a wide range of values, which can affect the green density and hence the sintered density. The addition of zinc stearate in a suitable amount to the fuel powder leads to an open-pore microstructure. A high open porosity contributes to minimizing the residual oxygen content in the sintered pellets.

Program and Results of Reactor Tests of Mixed Nitride Fuel for Fast Reactors

In the Proryv project, mixed uranium-plutonium nitride fuel is to be used in the future nuclear power with fast reactors with sodium coolant (BN-1200) and with lead coolant (BREST). At present, experimental fuel elements and fuel assemblies with mixed nitride fuel are fabricated by means of a technology developed at the Bochvar All-Russia Research Institute for Inorganic Materials for testing in the BN-600 reactor and the MIR and BOR-60 research reactors. The program of reactor tests of the experimental fuel assemblies and some preliminary results of these tests are presented. World experience in research on mixed uranium-plutonium nitride fuel is limited to ~200 fuel elements, irradiated to maximum burnup ~18% h.a. at maximum linear power density 410–1300 W/cm [1, 2]. In our country, we have the results of the irradiation of ~1300 fuel elements with uranium nitride fuel in the BR-10 reactor [3]. Even though many properties of uranium nitride and mixed nitride fuel are close, the results of the research cannot fully serve as validation for the BREST and BN-600 fuel elements, for which the operating parameters determining their serviceability differ considerably from those of fuel elements with uranium nitride fuel. The results of the irradiation of four fuel elements with mixed nitride fuel to maximum burnup 12.1% h.a. in the BOR-60 reactor as part of a Russian–French experiment must also be taken into account [4]. Low swelling and gas-release and the corrosion damage to the cladding in this experiment were obtained for high-purity mixed nitride fuel fabricated from the initial metals with oxygen content <0.15 wt.% and carbon content <0.1 wt.% with pellet density 85%. The mixed nitride used in the experiment of [4] was characterized by elevated plutonium content 45 and 60 %, which could have promoted high creep rates and, correspondingly, lower stresses in the fuel-element cladding. For this reason, it is necessary to have information on its in-reactor behavior with the lower plutonium content provided in the fuel elements of the BREST and BN-1200 reactors. Analysis showed that it is necessary to obtain additional data on the pre-reactor and reactor properties of mixed uranium-plutonium nitride fuel. This requires reactor tests of BN-1200 and BREST type fuel elements in the BOR-60 and other research reactors in order to achieve statistically representative validation of the serviceability of mixed nitride fuel as part of full-scale experimental BN-600 fuel assemblies. A complex program of computational and experimental validation of using mixed uranium-plutonium nitride fuel in the BN-1200 and BREST-OD-300 reactors was developed as part of the Proryv project [5]. This program provides for pre-reactor investigations of the properties of fuel fabricated by the carbothermal synthesis technology and reactor tests of fuel elements in the BN-600 reactor and the MIR and BOR-60 research reactors [6]. The serviceability of fuel elements with

Thermodynamic Modeling of the Phase Composition of Mixed Uranium-Plutonium Mononitride with Impurity Oxygen Under Irradiation to Burnup 140 GW·days/ton and Temperature 900–1400 K

Thermodynamic modeling of the phase composition of mixed uranium-plutonium mononitride with impurity oxygen as a function of burnup and temperature under irradiation by fast neutrons has shown that the accumulation of fission products results in the formation of heterogeneous fuel: a multicomponent solid solution based on uranium-plutonium oxynitride that contains yttrium, zirconium, and lanthanides and separate nitride, oxide, metal, and intermetallides phases as well as alkaline-metal iodides and, apparently, bromides. The change in the content of these phases as a function of burnup to 140 GW·days/ton and temperature 900–1400 K is calculated. The effect of the individual phases formed in the course of irradiation on the effective thermal conductivity of the fuel is evaluated together with the diffusion coefficient of xenon atoms in the fuel kernel.

SCALE and SERPENT solutions of the OECD VVER-1000 LEU and MOX burnup computational benchmark

The loading of hybrid cores with Mixed Uranium Plutonium Oxide (MOX) and Low Enriched Uranium (LEU) fuels in commercial nuclear reactors requires well validated computational methods and codes capable of providing reliable predictions of the neutronics characteristics of such fuels in terms of reactivity conditions (kinf), nuclide inventory and pin power generation over the entire fuel cycle length. Within the framework of Joint United States/Russian Fissile Materials Disposition Program an important task is to verify and validate neutronics codes for the use of MOX fuel in VVER-1000 reactors. Benchmark analyses are being performed for both computational benchmarks and experimental benchmarks. In this paper new solutions for the (UO2+Gd) and (UO2+PuO2+Gd) fuel assemblies proposed within the “OECD VVER-1000 Burnup Computational Benchmark” are presented, these being representative of the designs which are expected to be used in the plutonium disposition mission. The objective is to test the SERPENT and SCALE codes against previously obtained solutions and to provide new reference solutions to the benchmark with modern nuclear data libraries.

Program and Some Results of Pre-Reactor Studies of Mixed Uranium-Plutonium Nitride Fuel for Fast Reactors

For future nuclear power with fast reactors with sodium (BN-1200) and lead (BREST) as coolants, plans are being made to use mixed uranium-plutonium nitride fuel as part of the Proryv project. A technology for carbothermal synthesis of mixed uranium-plutonium mononitride fuel and fabrication of pellets from it has now been developed. A program of comprehensive computational and experimental research on fuel to study the initial properties and their change under irradiation in reactor experiments is being implemented. Some results of pre-reactor studies of the properties of fuel and a program for future research are presented.

Analysis, Colligation, and Investigation of the Thermal Conductivity of Fuel Compositions Based on Uranium Mononitride

The aims of this work were to compare the results of studies performed by the present authors using the existing data on the thermal conductivity of uranium mononitride and colligate and analyze the thermal conductivity of mixed uranium-plutonium fuel. The dependence of the thermal conductivity on the content of impurities and the porosity of a material was investigated. It was shown that there is a discrepancy in the experimental data. Since it determines the presence of impurities and porosity, the technology used to fabricate the fuel has a considerable effect on the change in the thermal conductivity.

Prospects for Using Nitride Fuel in Fast Reactors with a Closed Nuclear Fuel Cycle

The present status of research on dense nitride fuel for liquid-metal cooled fast reactors is examined. It is noted that within the framework of project Proryv nitride fuel is to be used in reactors with sodium (BN-1200) as well as lead coolant (BREST). Carbothermal synthesis of nitride from oxide powders has been chosen to fabricate the experimental fuel elements. Several experimental assemblies with mixed uranium-plutonium nitride fuel have been fabricated for tests in BOR-60 and BN-600. It is shown that the existing global experience in studying nitride fuel is insufficient for validating such fuel for use in BN-1200 and BREST. A complex program for computational-experimental validation of fuel elements is being implemented for validating the serviceability of fuel elements.

High temperature X-ray diffraction study of the oxidation products and kinetics of uranium-plutonium mixed oxides.

The oxidation products and kinetics of two sets of mixed uranium-plutonium dioxides containing 14%, 24%, 35%, 46%, 54%, and 62% plutonium treated in air were studied by means of in situ X-ray diffraction (XRD) from 300 to 1773 K every 100 K. The first set consisted of samples annealed 2 weeks before performing the experiments. The second one consisted of powdered samples that sustained self-irradiation damage. Results were compared with chosen literature data and kinetic models established for UO2. The obtained diffraction patterns were used to determine the temperature of the hexagonal M3O8 (M for metal) phase formation, which was found to increase with Pu content. The maximum observed amount of the hexagonal phase in wt % was found to decrease with Pu addition. We conclude that plutonium stabilizes the cubic phases during oxidation, but the hexagonal phase was observed even for the compositions with 62 mol % Pu. The results indicate that self-irradiation defects have a slight impact on the kinetics of oxidation and the lattice parameter even after the phase transformation. It was concluded that the lattice constant of the high oxygen phase was unaffected by the changes in the overall O/M when it was in equilibrium with small quantities of M3O8. We propose that the observed changes in the high oxygen cubic phase lattice parameter are a result of either cation migration or an increase in the miscibility of oxygen in this phase. The solubility of Pu in the hexagonal phase was estimated to be below 14 mol % even at elevated temperatures.

Cross-correlation measurements with the EJ-299-33 plastic scintillator

New organic–plastic scintillation compositions have demonstrated pulse-shape discrimination (PSD) of neutrons and gamma rays. We present cross-correlation measurements of 252Cf and mixed uranium–plutonium oxide (MOX) with the EJ-299-33 plastic scintillator. For comparison, equivalent measurements were performed with an EJ-309 liquid scintillator. Offline, digital PSD was applied to each detector. These measurements show that EJ-299-33 sacrifices a factor of 5 in neutron–neutron efficiency relative to EJ-309, but could still utilize the difference in neutron–neutron efficiency and neutron single-to-double ratio to distinguish 252Cf from MOX.These measurements were modeled with MCNPX-PoliMi, and MPPost was used to convert the detailed collision history into simulated cross-correlation distributions. MCNPX-PoliMi predicted the measured 252Cf cross-correlation distribution for EJ-309 to within 10%. Greater photon uncertainty in the MOX sample led to larger discrepancy in the simulated MOX cross-correlation distribution. The modeled EJ-299-33 plastic also gives reasonable agreement with measured cross-correlation distributions, although the MCNPX-PoliMi model appears to under-predict the neutron detection efficiency.

Thermal conductivity of perspective fuel based on uranium nitride

The comparative analysis of the results obtained by the authors of this paper and the available literature data on the thermal conductivity of uranium mononitride was prepared, analytical review of the thermal properties of mixed uranium–plutonium nitride fuel also was done. The analysis has shown that characteristics such as porosity and amount of impurities (which are determined by fuel fabrication technology) have a significant influence on thermal conductivity of the nitride fuel. Thermal conductivity of mixed uranium–plutonium fuel, in addition, is very sensitive to the content of PuN. The data obtained agree well with those taken from literature sources, and can be reliably described using known correlations. Correlations of thermal conductivity on temperature, porosity and impurity content, presented in this study, can be used to predict fuel behavior.