Mixed Nitride Fuels Research Articles

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

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

Mathematical Model of the Technological Process of Carbothermal Synthesis of Mixed Uranium-Plutonium Nitride Fuel

Mathematical Model of the Technological Process of Carbothermal Synthesis of Mixed Uranium-Plutonium Nitride Fuel

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.

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.

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.

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.

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

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.

Une curieuse carcinose péritonéale

Nitrides have been proposed to be a suitable material for fast neutronic systems from beginning of the development of nuclear fuel. Starting with the production of uranium nitride and sesquinitrides up to mixed plutonium uranium nitrides, today's developments are inert nitride matrix materials to burn plutonium or to transmute long-lived actinides in accelerator-driven sub-critical systems (ADS) or fast reactors (FR). Several authors proposed zirconium nitride as possible inert matrix material for this reason. Mixed zirconium nitrides can be fabricated by carbothermal nitridation of the oxides in a narrow temperature window. Obtaining high quality material with low carbon and oxygen content is still the major challenge. Producing mixed nitride fuels by special shaping methods, as for example direct coagulation casting or freeze drying, in comparison to conventional powder compaction enables to use this material in special shapes to optimize burn up.The history of nitride fuels at PSI will be shown up to today's CONFIRM project, dealing with plutonium zirconium nitride fuels.

On Void Reactivity Limitation for the Core Loaded with Mixed Nitride Fuels

Self-controllability and self-terminability are strongly required as passive safety features in future liquid-metal FBR (LMFBR) commercialization. In this study, the self-terminability of the MN-fueled core was focused on. The FBR cores including the MN-fueled core should be designed to eliminate severe recriticality leading to energetics during core disruptive accidents (CDAs) because the cores are not designed to operate in an optimum configuration from a reactivity point of view. In addition, in the case of the MN-fueled core, the nitrogen gas N2 generated by the thermal dissociation, one of the unique phenomena of the nitride fuel, would give another cause of reactor vessel failure. Such a mechanical effect would strongly depend on the magnitude of the void reactivity. In this study, the allowable maximum void reactivity of the MN-fueled core to maintain the mechanical integrity of the reactor vessel was evaluated against an Unprotected Loss-of-Flow (ULOF) accident by using ARGO taking into account the equation-of-state of the MN fuel. By considering the partial-pressure-dependent thermal dissociation process, the design limit of the core void reactivity (positive sum) was estimated to be 6.2$.

On Void Reactivity Limitation for the Core Loaded with Mixed Nitride Fuels

Self-controllability and self-terminability are strongly required as passive safety features in future liquid-metal FBR (LMFBR) commercialization. In this study, the self-terminability of the MN-fueled core was focused on. The FBR cores including the MN-fueled core should be designed to eliminate severe recriticality leading to energetics during core disruptive accidents (CDAs) because the cores are not designed to operate in an optimum configuration from a reactivity point of view. In addition, in the case of the MN-fueled core, the nitrogen gas N2 generated by the thermal dissociation, one of the unique phenomena of the nitride fuel, would give another cause of reactor vessel failure. Such a mechanical effect would strongly depend on the magnitude of the void reactivity. In this study, the allowable maximum void reactivity of the MN-fueled core to maintain the mechanical integrity of the reactor vessel was evaluated against an Unprotected Loss-of-Flow (ULOF) accident by using ARGO taking into account the equation-of-state of the MN fuel. By considering the partial-pressure-dependent thermal dissociation process, the design limit of the core void reactivity (positive sum) was estimated to be 6.2$.

Oxidative ammonolysis of uranium(IV) fluorides to uranium(VI) nitride

Actinide nitrides, in particular UN, are being considered as fuel types for advanced reactor systems. Here, we demonstrate a low-temperature synthesis route on uranium that could be developed into a commercial fabrication process for UN and mixed actinide nitride fuels. UN was successfully synthesized from UO 2 by first reacting with NH 4HF 2 in a ball mill at 20 °C to form tetravalent ammonium uranium fluorides. Then, reaction with an ammonia atmosphere at 800 °C oxidized tetravalent uranium fluorides to hexavalent UN 2. The final product, UN, was obtained by decomposing UN 2 at 1100 °C under argon to produce UN through an intermediate phase of U 2N 3.

05/01012 Fission gas release and swelling in uranium-plutonium mixed nitride fuels

05/01012 Fission gas release and swelling in uranium-plutonium mixed nitride fuels

Nitrides as a nuclear fuel option

Nitrides have been proposed to be a suitable material for fast neutronic systems from beginning of the development of nuclear fuel. Starting with the production of uranium nitride and sesquinitrides up to mixed plutonium uranium nitrides, today's developments are inert nitride matrix materials to burn plutonium or to transmute long-lived actinides in accelerator-driven sub-critical systems (ADS) or fast reactors (FR). Several authors proposed zirconium nitride as possible inert matrix material for this reason. Mixed zirconium nitrides can be fabricated by carbothermal nitridation of the oxides in a narrow temperature window. Obtaining high quality material with low carbon and oxygen content is still the major challenge. Producing mixed nitride fuels by special shaping methods, as for example direct coagulation casting or freeze drying, in comparison to conventional powder compaction enables to use this material in special shapes to optimize burn up.The history of nitride fuels at PSI will be shown up to today's CONFIRM project, dealing with plutonium zirconium nitride fuels.

Calculation of thermodynamic parameters of U–Pu–N system with carbon and oxygen impurities

Mixed nitride fuels are being considered for advanced FBR, but very little is known about the thermodynamic properties of these fuels. For an overall composition of the nitride fuel with small amounts of oxygen and carbon impurities, thermodynamic properties, e.g. carbon activity and partial pressures of nitrogen, carbon-monoxide, plutonium and uranium, were calculated in present work. These calculations were based on standard Gibbs free energies of the binary compounds, present in this multi-component system (U,Pu)–C–N–O. For an over all composition of the fuel, stable phase-field was determined by minimization of the Gibbs free energy of the system. The fabrication experiences of various workers, reported in literature, have shown that depending on the impurity content, nitride fuel can exist in two phase fields, mono-nitride phase in equilibrium with sesquinitride phase or mono-nitride phase in equilibrium with dioxide phase. Therefore, in present calculations special attention was given to the thermodynamic behavior of these two phase-fields. A comparison of calculated thermodynamic properties indicated that nitride fuel with dioxide as second phase will be superior to the one with sesquinitride.

Phase diagram calculations of the UPuN system with carbon and oxygen impurities

The most common method for the preparation of mixed nitride fuels is the carbothermic reduction of a UO 2 + PuO 2 + C mixture in a nitrogen atmosphere. A mixed nitride fuel thus formed has carbon and oxygen impurities which are kept well below 5000 ppm. For a given overall composition of the nitride fuel and the temperature, the present work calculates the stable phases in equilibrium and the amounts of those phases. These calculations are based on the principle of mass balance and minimization of the Gibbs free energy for the system. The Gibbs free energy of formation of the binary compounds UN, PuN, UO, PUO, UC, PUC, UO 2, PuO 2 and UN 1.5 are used for the calculations. In general, the binary compounds are assumed to form ideal solid solutions, but in certain cases, available or estimated interaction parameters were used to see the effect of deviation from the ideal solution assumption on the phase diagram. Partial phase diagrams of (U, Pu)CN were drawn for different oxygen impurity concentrations at various temperatures. Comparison with the experimental data available in the literature is carried out. Based on these calculations it is suggested that a separate oxide phase has to be left deliberately in the mixed nitride fuel matrix during preparation to ensure fuel clad chemical compatibility.

Fabrication of Uranium-Plutonium Mixed Nitride Fuel Pins for Irradiation Tests in JMTR

Fabrication of Uranium-Plutonium Mixed Nitride Fuel Pins for Irradiation Tests in JMTR

The effect of oxygen impurity on the characteristics of uranium and uranium-plutonium mixed nitride fuels

The effect of oxygen impurity on the typical characteristics of UN and (U, Pu)N fuels was investigated. The oxygen content of the samples was adjusted to ∼0.3, ∼0.6 and ∼1.0 wt%. It was confirmed from chemical, X-ray and ceramographic analyses that oxygen added to the nitrides existed in the oxide precipitates after sintering. While sinterability was observed. Furthermore, the addition of ∼1.0wt% of oxygen impurity decreased the thermal conductivities of UN and (U, Pu)N by 9–10 and 12–13% at 1000 and 1500 K, respectively.

Fabrication of Uranium-Plutonium Mixed Nitride Fuel Pins for Irradiation Tests in JMTR.

Abstract Pour He-bonded fuel pins containing uranium-plutonium mixed nitride pellets were fabricated for the irradiation tests in JMTR, aiming at understanding the irradiation behavior and demonstrating the feasibility of mixed nitride as an advanced fuel for FBRs. Mixed nitride pellets were prepared by carbothermic reduction of the oxides. One of the fuel pins was fitted with a set of W-Re thermocouples in order to trace the temperature change of pellet center during irradiation. Moreover, ferritic stainless steel was adopted as a cladding material in one fuel pin besides type 316 austenitic stainless steel. In this report, the design and characteristics of mixed nitride fuel pins are described.