O2 Fuels Research Articles

Understanding and Predicting the Thermodynamic Behavior of Fission Products Encountered Between the (U,Pu)O2 Fuel Pellet and the Cladding: Characterization and Modeling Approaches

This study provides a comprehensive investigation of the formation and behavior of a layer enriched in fission products encountered between the (U,Pu)O2 fuel pellet and the cladding, designated as JOG (“Joint Oxyde Gaine” in French). Employing a multifaceted approach that combined thermodynamic calculations, experimental synthesis, and advanced characterization techniques, simulated JOG has been synthetized (without radioactive fission products). Using thermodynamic calculations with the TAF-ID database, the phase compositions in the JOG was assessed for various temperatures, pressures, and oxygen potential conditions, revealing insights into the environmental factors influencing JOG formation. Experimental simulation of the JOG composition, exposed to controlled conditions, confirmed the presence of key compounds such as Cs2MoO4, CsI, and PdTe, as evidenced by SEM, EDS, and XRD analyses. The results of the calculations highlighted notable differences in the nature of the phases constituting the JOG under varying pressure and oxygen potential conditions. At 873 K and oxygen partial pressure of 10–4 bar, Cs2MoO4, Pd–Te, and a gas phase rich in tellurium and CsI were predominant, contrasting with the emergence of liquid phases at 70 bar. This study offered a comprehensive understanding of JOG microstructure, and highlighting the importance of accurate characterization for reactor safety. This information lays the foundations for future studies on the chemical interaction between the JOG and the steel cladding.

Stable and High-performance Flow H2 -O2 Fuel Cells with Coupled Acidic Oxygen Reduction and Alkaline Hydrogen Oxidation Reactions.

Conventional H2 -O2 fuel cells suffer from the low output voltage, insufficient durability, and high-cost catalysts (e.g., noble metals). Herein, this work reports a conceptually new coupled flow fuel cell (CF-FC) by coupling asymmetric electrolytes for acidic oxygen reduction reaction and alkaline hydrogen oxidation reaction. By introducing an electrochemical neutralization energy, the newly-developed CF-FCs possess a significantly increased theoretical open-circuit voltage. Specifically, a CF-FC based on a typical transition metal single-atom Fe-N-C cathode catalyst demonstrates a high electricity output up to 1.81V and durability with an ultrahigh retention of 91% over 110h, far superior to the conventional fuel cells (usually, < 1.0V, < 50% retention over 20h). The output performance can even be significantly enhanced easily by connecting multiple CF-FCs into the parallel, series, or combined parallel-series connections at a fractional cost of that for the conventional H2 -O2 fuel cells, showing great potential for large-scale practical applications. Thus, this study provides a platform to transform conventional fuel cell technology through the rational design and development of advanced energy conversion and storage devices by coupling different electrocatalytic reactions.

Comparison of the neutronic properties of the (Th-233U)O2, (Th-233U)C, and (Th-233U)N fuels in small long-life PWR cores with 300, 400, and 500 MWth of power

Abstract The neutronic characteristics of (Th-233U)O2, (Th-233U)C, and (Th-233U)N have been compared in small long-life pressurized water reactors (PWRs). Neutronic calculations were carried out at 300 MWth, 400 MWth, and 500 MWth with two cladding types: zircaloy-4 and ZIRLO (Zr low oxygen). They were performed using the Standard Reactor Analysis Code (SRAC) and JENDL-4.0 nuclide data, dividing the reactor core into three fuel zones with varying 233U enrichment levels, ranging from 3% to 9% and fluctuating by 1%, employing the PIJ module at the fuel cell level and the CITATION module at the reactor core level. In addition, 231Pa was added as burnable poison (BP). The (Th-233U)N fuel demonstrated superior criticality compared to the other fuel types, as it consistently achieves critical conditions throughout the reactor’s operating cycle with excess reactivity &lt;1.00% dk/k for several fuel configurations at the 300 MWth and 400 MWth power levels. Moreover, the (Th-233U)N and (Th-233U)C fuels exhibited similar and flatter power density distribution patterns compared to the (Th-233U)O2 fuel. The power peaking factor (PPF) value was relatively higher for (Th-233U)O2 fuel than the other two fuels. The (Th-233U)N fuel exhibited the most negative Doppler coefficient, followed by (Th-233U)C and (Th-233U)O2 fuels. Analysis of burnup levels revealed that the (Th-233U)O2 fuel achieved significantly higher burnup than the other two fuels.

Decoupled Artificial Photosynthesis via a Catalysis-Redox Coupled COF||BiVO4 Photoelectrochemical Device.

Artificial photosynthesis is an attractive approach to direct fuel production from sunlight. However, the simultaneous O2 evolution reaction (OER) and CO2 reduction reaction (CDRR) present challenges for product separation and safety. Herein, we propose a strategy to temporally decouple artificial photosynthesis through photoelectrochemical energy storage. We utilized a covalent organic framework (DTCo-COF) with redox-active electron donors (-C-OH moieties) and catalytically active electron acceptors (cobalt-porphyrin) to enable reversible -C-OH/-C═O redox reaction and redox-promoted CO2-to-CO photoreduction. Integrating the COF photocathode with an OER photoanode in a photoelectrochemical device allows the effective storage of OER-generated electrons and protons by -C═O groups. These stored charges can be later employed for CDRR while regenerating -C═O to complete the loop, thus enabling on-demand and separate production of O2 or solar fuels. Our work sets the stage for advancements in decoupled artificial photosynthesis and the development of more efficient solar fuel production technologies.

Development of an interatomic potential for mixed uranium-americium oxides and application to the determination of the structural and thermodynamic properties of (U,Am)O2 with americium contents below 50%

Uranium-americium mixed oxides are envisaged as targets in fast neutron reactors to allow the transmutation of americium in view of the reduction of its quantity in nuclear waste. The knowledge of structural and thermodynamic properties of (U,Am)O2 fuels is of crucial importance for fuel manufacturing, as well as for safe reactor operation. Atomic scale modelling has been proved an important complement to experimental characterizations to obtain the necessary data and bring insight into the behaviour of fuel materials over a wide range of compositions and conditions. An important feature of (U,Am)O2 compounds observed experimentally by several authors and confirmed by electronic structure calculations for all compositions, is the presence, not only of cations in the (+IV) oxidation state, but also of the Am3+ and U5+ species, inducing a complex behaviour as a function of composition. To allow the study of these mixed oxides using classical molecular dynamics, we optimized interaction parameters for the Am3+ and U5+ cations for the Cooper-Rushton-Grimes interatomic potential, which has yielded excellent results on several pure and mixed actinide oxides. These potential terms were parametrized on a small set of the experimental data available, namely the room temperature lattice parameters, the thermal expansion and the enthalpy increment for several americium contents. The potential obtained was validated against experimental data not included in the parametrization process and then applied to determine various structural and thermodynamic properties of U1-yAmyO2 compounds with 0≤y≤0.5 as a function of temperature and composition. The results yielded by this parametrization are in excellent agreement with the available experimental data and provide significant insight into the properties of uranium-americium mixed oxides, especially at high temperature.

Electron inelastic mean free path in UO2 and (U, Pu)O2 fuels

Inelastic mean free path (IMFP) of UO2 and (U, Pu)O2 fuels at 0, 20, 50 and 95 at.% of Pu were determined for the first time using a combination of convergent beam electron diffraction (CBED) and electron energy loss spectroscopy (EELS) on TEM FIB lamellae. Non-linear least squares fit was performed on the CBED intensity profiles to only obtain the crystalline thickness of the sample with a precision of up to 0.2% [1] while EELS spectra were used to calculate the relative thickness of the sample: crystalline and amorphous contribution. The IMFP were determined to be 119 – 129 nm for UO2, 107 – 117 nm for (U, Pu)O2 at 20 at.% of Pu, 116 – 125 nm for (U, Pu)O2 at 50 at.% of Pu and 123 – 137 nm for (U, Pu)O2 at 95 at.% of Pu using collection semi-angles β = 23.4 – 11.6 mrad, convergence semi-angle α = 8 mrad and accelerated voltage of 200 keV on a Thermofischer TALOS F200X TEM in the LECA STAR hot laboratory (CEA Cadarache). A comparison with the most relevant models in the literature shows that for compounds with a comparable quantity of light and heavy elements it is better to use a mean atomic number instead of an effective one as proposed by different authors. The importance of accurate density determination is also emphasized if the IMFP values are to be predicted by the existed models. As a second part, a strategy is developed to calculate the effective refractive index at the nanometer scale for the fuels allowing to easily find the IMFP under different measurement parameters such as those in energy filtered transmission electron microscopy (EFTEM) maps.

New recommendation for the thermal conductivity of irradiated (U, Pu)O2 fuels under fast reactor conditions. Comparison with recent experimental data

New experimental results obtained at the Joint Research Centre in Karlsruhe (JRC-Karlsruhe) within the European Sustainable Nuclear Industrial Initiative Plus project (ESNII+) were used to develop an updated recommendation for the thermal conductivity of (U, Pu)O2 fuels irradiated in fast reactor (FR) conditions. Fresh and irradiated FR fuels were characterised, and their thermal diffusivity and heat capacity were measured by the laser flash technique. The thermal conductivity of those fuels was calculated, using hydrostatic density measurements when available or using calculated density.The thermal conductivity model, developed in this work, describes different effects induced by irradiation on the degradation of thermal conductivity: soluble fission products (FPs), precipitated FPs, radiation damage, and porosity. Those effects are considered in separate corrective factors, which depend on the chemical composition of the irradiated fuel, its microstructure, and the temperature. To estimate these parameters, thermomechanical calculations (PLEIADES-GERMINAL V2) and thermochemical calculations (Thermo-Calc V4.1 with the TAF-ID V11 database) were carried out in conjunction. The data provided for each fuel sample consisted of fuel temperature, burn-up, end-of-life (EOL) isotopic composition, molar fraction of the dissolved FPs in the matrix, molar fraction of the oxide, and the metallic precipitated phases, density, produced and released fission gases.In this work, we showed a clear discrepancy between the experimental data and the most widely used thermal conductivity model for irradiated fuels (Lucuta et al., 1996). Lucuta's model was then modified and a new recommendation for FR (U, Pu)O2 fuels was provided. This work demonstrates that by introducing the neodymium content as the parameter driving the thermal conductivity of the matrix, a better agreement between the model and the experimental data can be achieved. This paper discusses the uncertainties related to the irradiation conditions and the data provided by the GERMINAL-Thermo-Calc calculation.

Neutronic Analysis of SMART Reactor Core for (U-Th)O2 and MOX Fuel Hybrid Configurations

Neutronic Analysis of SMART Reactor Core for (U-Th)O2 and MOX Fuel Hybrid Configurations

Electron Inelastic Mean Free Path in UO2 and (U, Pu)O2 Fuels

Electron Inelastic Mean Free Path in UO2 and (U, Pu)O2 Fuels

Preliminary Evaluation of the LASSO Method for Prediction of the Relative Power Density Distribution in Mixed Oxide (Pu, DU)O2 Fuel Pellets

Abstract Predicting the power distribution within nuclear fuel is essential for predicting reactor fuel performance, since power distributions can impact pellet temperature distributions and fission product transport and migration. Analytical expressions for radial power distribution in fuel pellets were sought using lattice physics calculations to generate data and a machine learning technique to find representative expressions. Analytical approximations can be useful in nuclear fuel performance codes, such as element simulation and stresses (ELESTRES)/ element simulation code in a loss of coolant accident (ELOCA) for providing very rapid predictions of power distributions with reduced computational effort and memory requirements, relative to using an embedded or coupled neutron transport/burnup reactor physics code. Radial power distributions were calculated a priori using lattice physics codes to model mixed oxide (MOX) 37-element fuel bundles in pressure tube heavy water reactors. Such advanced fuels are of interest for future fuel cycles. Several datasets were generated with different amounts of PuO2 and variable neutron energy spectrum. Results of preliminary studies with the least absolute shrinkage and selection operator (LASSO) regression machine learning method have obtained analytical fitting functions with a mean maximum relative error (MRE) of 0.056 and a maximum MRE of 0.152 on the test set. However, using LASSO to estimate the coefficients of a physically motivated modified Bessel plus an exponential function, results in a lower MRE (mean MRE 0.041 and maximum MRE 0.11) on the same test set. Further potential improvements in both the curve fit and the machine learning methods are discussed.

Core design and fuel behaviour of a small modular pressurised water reactor using (Th,U)O[formula omitted] fuel for commercial marine propulsion

There is an increasing interest in employing technologies to decarbonise the shipping sector; one option is the use of nuclear energy to power civilian ships. Previous work has investigated the use of UO2 fuel and its behaviour in a novel, soluble-boron-free, low power density PWR that offered a 15 year core life to power a container ship with a demand of 110 MWe. Here, building on that previously-presented UO2 core design, we investigate the analogous design of a (Th,U)O2 core to examine the neutronic and fuel performance benefits of using (Th,U)O2 fuel. The motivations for studying (Th,U)O2 fuel include understanding how (Th,U)O2 would behave in a novel low power density system and its behaviour as a fertile poison in a long-life core. Core design was developed using CASMO-SIMULATE and fuel performance was analysed using ENIGMA. In this new design, the introduction of (Th,U)O2 fuel exhibited similar neutronic characteristics (shutdown margins, xenon transient behaviour and reactivity coefficients) relative to the UO2 core. However, the fuel behaviour predicted by ENIGMA for (Th,U)O2 in this low power density core was found to be significantly impaired compared with using UO2 fuel. In particular, the introduction of uranium to ThO2 was found to degrade thermal conductivity and thorium-based fuels exhibited worse creep behaviour than UO2 fuels; both conclusions are supported by published experimental evidence. A factor behind the degraded fuel performance characteristics predicted by ENIGMA is the assumption that athermal phonon scattering by fission products has the same effect on thermal conductivity as for UO2. This assumption is necessary to address the limited burnup data for (Th,U)O2 fuels.

Effect of PANI–AC Composite on Electrochemical Synthesis of Hydrogen Peroxide by Alkaline H2–O2 Fuel Cell Reactor

Effect of PANI–AC Composite on Electrochemical Synthesis of Hydrogen Peroxide by Alkaline H2–O2 Fuel Cell Reactor

Effect of compaction pressure on the thermal conductivity of UO2-BeO-Gd2O3 pellets

The (U,Gd)O2 fuels are used in pressurized water reactors (PWR) to control the neutron population in the reactor during the early life with the purpose of extended fuel cycles and higher target burnups. Nevertheless, the incorporation of Gd2O3 in the UO2 fuel decreases the thermal conductivity, leading to premature fuel degradation. This is the reason for the addition of beryllium oxide (BeO), which has a high thermal conductivity and is chemically compatible with UO2. Pellets were obtained from powder mixtures of the UO2, Gd2O3 and BeO, being the oxide contents of the beryllium equal to 2 and 3wt%, and the gadolinium fixed at 6wt%. The pellets were compacted at 400, 500, 600, and 700 MPa and sintering under hydrogen reducing atmosphere. The purpose of this study was to investigate the effect of BeO, Gd2O3 and compaction pressure on the thermal conductivity of the UO2 pellets. The thermal diffusivity and conductivity of the pellets were determined from 298 K to 773 K and the results obtained were compared to UO2 fuel pellets. The thermal diffusivity was determined by Laser Flash and Thermal Quadrupole methods and the thermal conductivity was calculated from the product of thermal diffusivity, the specific heat capacity and density. The sintered density of the pellets was determined by the xylol penetration and immersion method. The results showed an increase in the thermal conductivity of the pellets with additions of BeO and with the compaction pressure compared to the values obtained with UO2 pellets.

Multiphysics Modeling of Thorium-Based Fuel Performance With Cr-Coated SiC/SiC Composite Under Normal and Accident Conditions

Using the finite element multiphysics modeling method, the performance of the thorium-based fuel with Cr-coated SiC/SiC composite cladding under both normal operating and accident conditions was investigated in this work. First, the material properties of SiC/SiC composite and chromium were reviewed. Then, the implemented model was simulated, and the results were compared with those of the FRAPTRAN code to verify the correctness of the model used in this work. Finally, the fuel performance of the Th0.923U0.077O2 fuel, Th0.923Pu0.077O2 fuel, and UO2 fuel combined with the Cr-coated SiC/SiC composite cladding and Zircaloy cladding, respectively, was investigated and compared under both normal operating and accident conditions. Compared with the UO2 fuel, the Th0.923U0.077O2 and Th0.923Pu0.077O2 fuels were found to increase the fuel centerline temperature under both normal operating and reactivity-initiated accident (RIA) conditions, but decrease the fuel centerline temperature under loss-of-coolant accident (LOCA) condition. Moreover, compared to the UO2 fuel with the Zircaloy cladding, thorium-based fuels with Cr-coated SiC/SiC composite cladding were found to show better mechanical performance such as delaying the failure time by about 3 s of the Cr-coated SiC/SiC composite cladding under LOCA condition, and reducing the plenum pressure by about 0.4 MPa at the peak value in the fuel rod and the hoop strain of the cladding by about 16% under RIA condition.

Three-dimensional (X-Y-Z) Core Design of Long-Life Pressurized Water Reactor Using (Th-U)O2 Fuels with The Addition of Gd2O3 and Pa-231 as Burnable Poisons

Pressurized water reactors (PWRs) are one of the most dominant types of nuclear power plants that have been operated commercially to produce electricity in the world. The purpose of this study was to perceive a three-dimensional (X-Y-Z) core design of long-life PWR using Thorium-Uranium dioxide ((Th-U)O2) fuels with the addition of Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. A combination of Thorium and enriched Uranium fuels have a higher conversion ratio than other fuels, therefore can guarantee the reactor to operate longer. The burnable poison isotopes could be used to reduce excess reactivity due to the very high thermal neutron absorption cross-section. For core geometry analysis, a three-dimensional (X-Y-Z) geometry and a fuel volume fraction of 40% were applied. The computer code of SRAC 2006 from the Japan Atomic Energy Agency (JAEA) and the JENDL 4.0 as a nuclear data library were used for calculation. In this study, different fractions of Uranium dioxide, Uranium-235, Gadolinium, and Protactinium-231 in fuel were carried out. The result of this study was a three-dimensional core design of 800 MWt PWR using 60% Uranium dioxide fuel with enriched Uranium-235 of 12%-11% and the addition of 0,025% Gd2O3 and 1,0% Pa-231 which could operate for ten years without refueling. This research is expected to be a reference for long-life PWR design using the Thorium and Uranium fuel cycles.

The impact of gadolinium on the reactor production of 153Sm

Radioisotopes represent major sources of ionizing radiation, not least for use in medical applications, brachytherapy and nuclear medicine included. In this, the nuclear reactor is the main source of β- - γ emitting isotopes, an example product being 153Sm used in the treatment of pain arising from bone metastases. Present analysis relates to the potential of gadolinium neutron capture reaction, its impact on reactor production of radioisotopes and the proliferation resistant potential of thorium fuel cycle. A comparative analysis has been made of the impact of gadolinium on the production of 153Sm by UO2 and (Th, U) O2 fuels in a Westinghouse small modular reactor. Five fuel assemblies were investigated: one containing no gadolinium, the other four containing 16, 24, 34 or 44 gadolinium fuel rods. The code Monte Carlo N-Particle eXtended (MCNPX) integrated with the CINDER90 burn-up code was used for calculations. In the production of 153Sm the same trend is followed for the fuels containing gadolinium, increasing significantly with the number of gadolinium fuel rods. Zero production results from fuel assemblies without gadolinium. The concentration of 153Sm increases significantly with burn-up, indicating that gadolinium has a positive impact on the production of 153Sm.

Design Conceptual of 800MWt Long Life Pressurized Water Reactor Using (Th-U)O2 Fuels with Gd2O3 and Pa-231 as Burnable Poisons

A long-life pressurized water reactor (PWR) has been reviewed as an innovative reactor design that can fulfill electricity demand. This study aimed to find out the optimum design of 800MWt long life PWR using Thorium-Uranium dioxide (Th-U)O2 fuels with Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. An established computer code of SRAC 2006 with JENDL 4.0 as data nuclear library had been used for the analysis. A two-dimensional R-Z geometry and fuel volume fraction of 40% were used for core geometry analysis. The different fraction of Uranium dioxide, Uranium-235, Gadolinium, and Protactinium-231 had been carried out. The result of this study was a design of PWR 800MWt using Uranium dioxide fuel of 60% with enrichment 11%-12%-13% Uranium-235 and the addition of 0,025% Gd2O3 and 1,0% Pa-231 that could operate for ten years without refueling. The reactor could produce a power density of 45,4 watts/cc with excess reactivity about 3,6% dk/k. This study is expected to be a reference for a long-life pressurized water reactor using the Thorium-Uranium fuel cycles.

Neutronic Analysis of Small Long-Life Pressurized Water Reactor Using (Th-U)O2 Fuels with Gd2O3 and Pa-231 as Burnable Poisons

The requirement for electricity increases with the growth of the human population. The existing power plants have not been able to fulfill all electricity requirements, especially in remote areas. The small long-life pressurized water reactor (PWR) is one of the solutions and innovations in nuclear technology that can produce electrical energy for a long time without refueling. This study aimed to analyze the neutronic of small long-life PWR that using Thorium-Uranium dioxide ((Th-U)O2) fuels with enriched Uranium-235 (U-235) and the addition of Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. The SRAC Code with the JENDL-4.0 nuclear data library had been used for the calculation method. In this study, the geometry of the two-dimensional (R-Z) reactor core with different fuel volume fraction was analyzed. Moreover, variations of the Uranium-235, Gadolinium, and Protactinium-231 fractions in the fuels were carried out. The result in this study was a PWR 420 MWt design using 60% Uranium dioxide fuel with enriched Uranium-235 of 10%-11%-12% and the addition of 0,0125% Gadolinium and 1,0% Protactinium-231 as the burnable poisons that could operate for thirteen years without refueling. The small long-life PWR design could produce a power density of 85,1 watts/cc with the reactivity for less than 4,6% dk/k.

A comparative study on the impact of Gd2O3 burnable neutron absorber in UO2 and (U, Th)O2 fuels

The performance of gadolinium burnable absorber (GdBA) for reactivity control in UO2 and (U, Th)O2 fuels and its impact on spent fuel characteristics was performed. Five fuel assemblies: one without GdBA fuel rod and four each containing 16, 24, 34 and 44 GdBA fuel rods in both fuels were investigated. Reactivity swing in all the FAs with GdBA rods in UO2 fuel was higher than their counterparts with similar GdBA fuel rods in (U, Th)O2 fuel. The excess reactivity in all FAs with (U, Th)O2 fuel was higher than UO2 fuel. At the end of single discharge burn-up (~ 49.64 GWd/tHM), the excess reactivity of (U, Th)O2 fuel remained positive (16,000 pcm) while UO2 fuel shows a negative value (−6,000 pcm), which suggest a longer discharge burn-up in (U, Th)O2 fuel. The concentration of plutonium isotopes and minor actinides were significantly higher in UO2 fuel than in (U, Th)O2 fuel except for 236Np. However, the concentration of non-actinides (gadolinium and iodine isotopes) except for 135Xe were respectively smaller in (U, Th)O2 fuel than in UO2 fuel but may be two times higher in (U, Th)O2 fuel due to its potential longer discharge burn-up.

Criticality safety analysis of spent fuel pool for a PWR using UO2, MOX, (Th-U)O2 and (TRU-Th)O2 fuels

A spent fuel pool of a typical Pressurized Water Reactor (PWR) was evaluated for criticality studies when it uses reprocessed fuels. PWR nuclear fuel assemblies with four types of fuels were considered: standard PWR fuel, MOX fuel, thorium-uranium fuel and reprocessed transuranic fuel spiked with thorium. The MOX and UO2 benchmark model was evaluated using SCALE 6.0 code with KENO-V transport code and then, adopted as a reference for other fuels compositions. The four fuel assemblies were submitted to irradiation at normal operation conditions. The burnup calculations were obtained using the TRITON sequence in the SCALE 6.0 code package. The fuel assemblies modeled use a benchmark 17x17 PWR fuel assembly dimensions. After irradiation, the fuels were inserted in the pool. The criticality safety limits were performed using the KENO-V transport code in the CSAS5 sequence. It was shown that mixing a quarter of reprocessed fuel withUO2 fuel in the pool, it would not need to be resized