Uranium Oxide Research Articles

A direct-AMS assessment of Re-Os in a Cigar Lake uranium ore specimen

The decay constant of 187Re is known to change when its inner shell electrons are ionized; the possibility of a small change occurring in a terrestrial radioactive environment has not been assessed. A recently completed multi-isotope survey of uranium ore concentrates (UOC) has prompted the question. As an extension of that work, using direct accelerator mass spectrometry (direct-AMS), this paper reports a closer investigation of such possibility in an unaltered natural rock specimen of the Cigar Lake uranium ore, containing wt.30–60 % uranium oxide. Stimulated by this research goal, the direct-AMS quantification methods were first investigated, and later tested in connection with the Re-Os assessment. Complete sets of measurement of the anion beams of 185ReOn−, (187Re+187Os)On− and 189OsOn− (n = 0–5) were carried out to find suitable sputter target matrices that would yield self-consistent solutions of (187Os/189Os)Today and (187Re/185Re)Today. These two quantities are best determined by these multi-beam measurements to be 4400 ± 500 (2σ) and 1.635 ± 0.060 (2σ), respectively. The sample is then reasoned to contain [Re/Os]Initial ∼42,000. These results do not show apparent 187Re decay constant changes, but cannot rule out the possibility, either. As a quick follow-up inspection, a simplified approach was used to surveying 187Re/185Re in a series of uranium-concentrated materials, in which this ratio was all measured somewhat smaller than the natural known value. This work was also intended to be a pilot feasibility study of using the direct-AMS method as a form of sensitive secondary ionization mass spectrometry (SIMS) for rapid Re-Os assessment in mg-sized raw minerals and environmental samples.

Ambient melting behavior of stoichiometric uranium oxides

As UO2 is easily oxidized during the nuclear fuel cycle it is important to have a detailed understanding of the structures and properties of the oxidation products. Experimental work over the years has revealed many stable uranium oxides including UO2, U4O9 (UO2.25), U3O7 (UO2.33), U2O5 (UO2.5), U3O8 (UO2.67), and UO3, all with a number of different polymorphs. These oxides are broadly split into two categories, fluorite-based structures with stoichiometries in the range of UO2 to UO2.5 and less dense layered-type structures with stoichiometries in the range of UO2.5 to UO3. While UO2 is well characterized, both experimentally and computationally, there is a paucity of data concerning higher stoichiometry oxides in the literature. In this work we determine the ambient melting points of all the six stoichiometric uranium oxides listed above and compare them to the available experimental and/or theoretical data. We demonstrate that a family of the six ambient melting points map out a solid-liquid transition boundary consistent with the high-temperature portion of the phase diagram of uranium-oxygen system suggested by Babelot et al.

Directional growth and reconstruction of ultrafine uranium oxide nanorods within single-walled carbon nanotubes.

Understanding the atomic structures and dynamic evolution of uranium oxides is crucial for the reliable operation of fission reactors. Among them, U4O9-as an important intermediate in the oxidation of UO2 to UO2+x -plays an important role in the nucleation and conversion of uranium oxides. Herein, we realize the confined assembly of uranyl within SWCNTs in liquid phase and reveal the directional growth and reconstruction of U4O9 nanorods in nanochannels, enabled by in situ scanning transmission electron microscopy (STEM) e-beam stimulation. The nucleation and crystallization of U4O9 nanorods in nanochannels obey the "non-classical nucleation" mechanism and exhibit remarkably higher growth rate compared to those grown outside. The rapid growth process is found to be accompanied by the formation and elimination of U atom vacancies and strain, aiming to achieve the minimum interfacial energy. Eventually, the segments of U4O9 nanorods in SWCNTs merge into single-crystal U4O9 nanorods via structural reconstruction at the interfaces, and 79% of them exhibit anisotropic growth along the specific 〈11̄0〉 direction. These findings pave the way for tailoring the atomic structures and interfaces of uranium oxides during the synthesis process to help improve the mechanical properties and stability of fission reactors.

Controlled and sequential single-electron reduction of the uranyl dication.

A flexible tripodal pyrrole-imine ligand (H3L) has been used to facilitate the controlled and sequential single-electron reductions of the uranyl dication from the U(VI) oxidation state to U(V) and further to U(IV), processes that are important to understanding the reduction of uranyl and its environmental remediation. The uranyl(VI) complexes UO2(HL)(sol) (sol = THF, py) were straightforwardly accessed by the transamination reaction of H3L with UO2{N(SiMe3)2}2(THF)2 and adopt 'hangman' structures in which one of the pyrrole-imine arms is pendant. While deprotonation of this arm by LiN(SiMe3)2 causes no change in uranyl oxidation state, single-electron reduction of uranyl(VI) to uranyl(V) occurred on addition of two equivalents of KN(SiMe3)2 to UO2(HL)(sol). The potassium cations of this new [UVO2(K2L)]2 dimer were substituted by transmetalation with the appropriate metal chloride salt, forming the new uranyl(V) tetra-heterometallic complexes, [UVO2Zn(L)(py)2]2 and [UVO2Ln(Cl)(L)(py)2]2 (Ln = Y, Sm, Dy). The dimeric uranyl(V)-yttrium complex underwent further reduction and chloride abstraction to form the tetrametallic U(IV) complex [UIVO2YIII(py)]2, so highlighting the adaptability of this ligand to stabilise a variety of different uranium oxidation states.

РАСПРЕДЕЛЕНИЕ РЗЭ И U В СИСТЕМЕ ВОДА – ПОРОДА НА ПРИМЕРЕ СКАЛИНСКОГО ПРОЯВЛЕНИЯ РАДОНОВЫХ ВОД

The paper presents data on geochemical features of the water-rock system of the Skalinskoye radon water manifestation. The waters of the Skalinskoye manifestation by chemical composition belong to the SO4-HCO3 Na-Mg-Ca type with a total mineralization value of 279–787 mg/dm3. The geochemical parameters of the medium correspond to the oxidizing environment with Eh +84,6…+187,0 mV, pH 6.9–8.8 and O2dissolved concentrations of 3.75–9.93 mg/dm3. Radon activity is up to 572 Bq/dm3. The water-bearing rocks are represented by biotite granites of normal and moderately alkaline series, which are permeated by quartz and sericite-quartz veins. With an increase in the total mineralization of radon waters, the prevailing forms of migration are calcium, magnesium, nickel and strontium bicarbonates, nickel carbonate and magnesium sulfate. Uranium migrates mainly in the form of dissolved hydroxide U(OH)4(aq). According to saturation indices, all waters are saturated in relation to calcite, uraninite and amorphous uranium oxide.

Capturing ammonium nitrate in a synthetic uranium oxide hydrate phase: revealing the role of ammonium ions and anion inclusions.

Although uranium oxide hydrate (UOH) minerals and synthetic phases have been extensively studied, the role of ammonium ions in the formation of UOH materials is not well understood. In this work, the stabilization of a synthetic UOH phase with ammonium ions and the inclusion of ammonium nitrate were investigated using a range of structural and spectroscopic techniques. Compound (NH4)2(NO3)[(UO2)3O2(OH)3] (U-N1) crystallises in the orthorhombic Pmn21 space group, having a layered structure with typical α-U3O8 type layers and interlayer (NH4)+ cations as well as (NO3)- anions. The presence of uranyl, (NH4)+ cations and (NO3)- anions were further confirmed with a combination of FTIR and Raman spectroscopies through characteristic vibrational modes. The roles of the (NH4)+ cations for charge compensation and facilitating the inclusion of (NO3)- anions via hydrogen bonding were revealed and discussed. The findings have implications for uranium geochemistry, reprocessing of spent nuclear fuel and possible spent nuclear fuel alteration pathways under geological disposal.

Improving uranium oxide pathway discernment and generalizability using contrastive self-supervised learning

In the field of Nuclear Forensics, there exists a plethora of different tools to aid investigators when performing analysis of unknown nuclear materials. Many of these tools offer visual representations of the uranium ore concentrate (UOC) materials that include complimentary and contrasting information. In this paper, we present a novel technique drawing from state-of-the-art machine learning methods that allows information from scanning electron microscopy images (SEM) to be combined to create digital encodings of the material that can be used to determine the material’s processing route. Our technique can classify UOC processing routes with greater than 96% accuracy in a fraction of a second and can be adapted to unseen samples at similarly high accuracy. The technique’s high accuracy and speed allow forensic investigators to quickly get preliminary results, while generalization allows the model to be adapted to new materials or processing routes quickly without the need for complete retraining of the model.

Unearthing the Chicago Piles: A Monte Carlo perspective

The divergence of the Chicago Pile 1 (CP-1) on December 2nd, 1942, marks the birth of the nuclear age. Built under the West Stands of the Stagg Field in Chicago in slightly over two weeks, CP-1 was the first man-made nuclear reactor, conceived as a proof-of-principle physics experiment to demonstrate the feasibility of a self-sustained neutron chain reaction. The fuel consisted in a mixture of uranium oxide and metal uranium, with lumps of several sizes and shapes stacked within bricks of graphite used as moderator to form a cubic lattice. After only three months of operation, CP-1 was dismantled due to safety reasons and rebuilt at the Argonne Site in twenty-one days, under the name of Chicago Pile 2 (CP-2): the new pile had a longer life, contributing vastly to research on material science and nuclear reactor theory, and was finally decommissioned in 1954. Prompted by the eightieth anniversary of the Chicago Pile(s), in this talk we celebrate these astonishing scientific artifacts by building fully detailed three-dimensional digital twins of CP-1 and CP-2 using TRIPOLI-4®, the Monte Carlo code developed at CEA. Technological data for the models have been retrieved from published and unpublished documents of the Manhattan Project. We compare the simulation results to the available experimental measurements, encompassing the reactivity of the piles, the control rod worth, the kinetics parameters and several reactivity coefficients. For each, we examine the associated uncertainties due to missing or incomplete technological data as well as the bias due to nuclear data libraries.

Determination of critical material damage parameter for Indian clad tube burst behavior under severe accident scenario

During loss of coolant accident scenario in PHWRs, heat removal from clad tubes containing the uranium oxides fuel may get severely impaired. This results in increase in temperature of the fuel clad and subsequent increase in the clad tube internal pressure. This causes the clad tube to deform and balloon leading to the increase in diameter of the fuel clad. After a critical value of deformation is reached, the clad tube gets burst. The burst process is due to initiation of crack in the clad due to creep deformation process in the presence of internal pressure and high temperature environment. In this work, experimental data of fuel clad bursts from literature have been used to evaluate a critical damage parameter that can be used to predict the onset of burst phenomenon of clad tubes. Finite element modelling of the fuel clad has been carried out to simulate the ballooning deformation of the fuel clad tubes due to creep. Norton’s creep model has been used in the simulation. The critical damage parameter evaluated, using analysis and experimental data, has been used subsequently to predict burst time of Zircaloy-4 fuel clad when subjected to different types of loading environments. An exponential decay type correlation has been obtained from the finite element analysis results which can be used by safety analysts to predict burst pressure of Zircaloy-4 fuel clad of Indian PHWR as a function of burst temperature or vice-versa.

The triple oxygen isotope signature of uranium oxides in the nuclear fuel cycle

Uranium oxides, in the form of U3O8 and UO2, are principal compounds in the uranium nuclear fuel cycle and attract significant interest in the general field of nuclear forensics, for the development of new and credible signatures to support attribution. This study presents, for the first time, the analytical method and high-precision measurements of the triple oxygen isotope composition of U3O8 and UO2. We show that Δ’17O can be measured at the accuracy and reproducibility needed for nuclear forensics. Δ’17O exhibits a wide range, from -450 to -120 ppm, for both commercial and synthetic uranium oxides. The presentation of the data in Δ’17O-δ’18O space provides a two-dimension identification of uranium oxides for nuclear forensic characterization of samples. The relation between Δ’17O and the process from which the oxide was formed or manipulated, has the potential to trace the chemical history of the material. We determined the change in Δ’17O of the system U3O8 – atmospheric O2 resulting from isotope exchange at the temperature range of 298 K to 973 K. The dependency of the triple isotope exponent, θ, on temperature was calculated for this exchange reaction, concluding that the phase transition from pseudo-hexagonal to hexagonal structure plays a major role in determining Δ17O and θ. A set of reduction reactions of U3O8 to UO2 indicate that Δ’17O undergoes minimal change during this process.

Morphology of uranium oxides reduced from magnesium and sodium diuranate

Abstract Morphological analysis of uranium materials has proven to be a key signature for nuclear forensic purposes. This study examines the morphological changes to magnesium diuranate (MDU) and sodium diuranate (SDU) during reduction in a 10 % hydrogen atmosphere with and without steam present. Impurity concentrations of the materials were also examined pre and post reduction using energy dispersive X-ray spectroscopy combined with scanning electron microscopy (SEM-EDX). The structures of the MDU, SDU, and UO x samples were analyzed using powder X-ray diffraction (p-XRD). Using this method, UO x from MDU was found to be a mixture of UO2, U4O9, and MgU2O6 while UO x from SDU were combinations of UO2, U4O9, U3O8, and UO3. By SEM, the MDU and UO x from MDU had identical morphologies comprised of large agglomerates of rounded particles in an irregular pattern. SEM-EDX revealed pockets of high U and high Mg content distributed throughout the materials. The SDU and UO x from SDU had slightly different morphologies. The SDU consisted of massive agglomerates of platy sheets with rough surfaces. The UO x from SDU was comprised of massive agglomerates of acicular and sub-rounded particles that appeared slightly sintered. Backscatter images of SDU and related UO x materials showed sub-rounded dark spots indicating areas of high Na content, especially in UO x materials created in the presence of steam. SEM-EDX confirmed the presence of high sodium concentration spots in the SDU and UO x from SDU. Elemental compositions were found to not change between pre and post reduction of MDU and SDU indicating that reduction with or without steam does not affect Mg or Na concentrations. The identification of Mg and Na impurities using SEM analysis presents a readily accessible tool in nuclear material analysis with high Mg and Na impurities likely indicating processing via MDU or SDU, respectively. Machine learning using convolutional neural networks (CNNs) found that the MDU and SDU had unique morphologies compared to previous publications and that there are distinguishing features between materials created with and without steam.

Electrolytic Reduction of UO2 Microspheres Synthesized Via Internal Gelation Method

Pyroprocessing is the combination of process steps to extract and recycle actinides from fission products in spent nuclear fuels in high-temperature molten salt media. Pyroprocessing has been studied extensively over the past decades. One of the key sub-processes is the electrolytic reduction of uranium oxides. The electrolytic reduction process originated from the FFC Cambridge process, an approach for electrochemically reducing titanium dioxide into titanium metal in molten CaCl2. In the electrolytic reduction of uranium oxides, the uranium oxide feed is applied as the cathode in LiCl-Li2O molten salt electrolyte at 650°C. LiCl appears more promising than conventional CaCl2 owing to the lower melting point (878 K), and higher decomposition voltage (3.46 V at 923 K). Additionally, Li2O is expected to speed up the reduction via an additional chemical reduction pathway with the participation of Li metal, and prevents the direct dissolution of the Pt anode.This work aims to study the morphology effect of the oxide feed on the reduction kinetics and the incorporation of impurities during the electrolytic reduction process. The electrolytic reduction of UO2 is carried out on sintered UO2 pellets, as well as various UO2 microspheres synthesized via internal gelation. The introduction of uranium oxide microspheres holds several advantages. The spheres provide a higher surface area, which is expected to improve the overall reduction efficiency of uranium oxide during electrolysis. In addition, it is easier to handle the feed during loading and unloading of the cathode basket compared to powder feeds.The potentiostatic and galvanostatic electrolysis of UO2 is performed, and the comparison between the electrolytic reduction of UO2 pellets and microspheres is investigated. The SEM micrographs in figure 1show the morphology change of a UO2 pellet to uranium metals particles. A significant morphology change between the flat surface of the sintered UO2 feed and the reduced U particles is observed at the surface. Further characterization is performed using XRD and TGA to identify the products of the electrolytic reduction process qualitatively and quantitatively. Results show that galvanostatic electrolysis shows a better reduction efficiency compared to potentiostatic electrolysis, applied both below and above the Li formation onset potential. Overall, there is barely un-reacted UO2 visible in the XRD patterns of the reduced feed after galvanostatic electrolysis. U metal and UO are identified in the XRD patterns depending on electrolysis parameters. The presence of UO can be as an intermediate of UO2 reduction during electrolysis, which is a rare compound to be reported in literature. Overall, the reduction of microspheres does improve the overall reduction efficiency compared to the conventional pellet feeds. Figure 1

Encapsulation of Uranium Oxide in Multiwall WS2 Nanotubes.

Uranium is a high-value energy element, yet also poses an appreciable environmental burden. The demand for a straightforward, low energy, and environmentally friendly method for encapsulating uranium species can be beneficial for long-term storage of spent uranium fuel and a host of other applications. Leveraging on the low melting point (60°C) of uranyl nitrate hexahydrate and nanocapillary effect, a uranium compound is entrapped in the hollow core of WS2 nanotubes. Followingly, the product is reduced at elevated temperatures in a hydrogen atmosphere. Nanocrystalline UO2 nanoparticles anchor within the WS2 nanotube lumen are obtained through this procedure. Such methodology can find utilization in the processing of spent nuclear fuel or other highly active radionuclides as well as a fuel for deep space missions. Moreover, the low melting temperatures of different heavy metal-nitrate hydrates, pave the way for their encapsulation within the hollow core of the WS2 nanotubes, as demonstrated herein.

A science‐based mixed oxide property model for developing advanced oxide nuclear fuels

AbstractHerein, a science‐based uranium and plutonium mixed oxide (MOX) property model is derived for determining the properties of MOX fuel and analyzing their performance as functions of Pu content, minor‐actinide content, oxygen‐to‐metal ratio, and temperature. The property model is constructed by evaluating the effect of phonons and electronic defects on the heat capacity and thermal conductivity of MOX fuels. The effect of phonons was evaluated based on experimental datasets related to lattice parameters, thermal expansion, and sound speeds. Moreover, the effect of electronic defects was determined by analyzing oxygen‐potential data based on defect chemistry. Furthermore, the model evaluated the effect of the Bredig transition on the thermal properties of MOX fuel by analyzing the irradiation test results. The derived property model is applied to the performance code to analyze fast reactor fuel pins.

Monitoring the reduction of UO3 thin film by hydrogen atoms using valence-level spectroscopy: correlating the U5f1 signal to surface hydroxyls

The reaction of a UO3 thin film with atomic hydrogen was studied by He(II) ultraviolet photoelectron spectroscopy (UPS) in the temperature range 190–300 K. UO3 reduction was instantaneously observed once it contacted H atoms at 10–7 torr. The reduction was manifested by the presence of U5f1 electrons in He(II) UPS at approximately 1.5 eV below the Fermi level. Based on the peak characteristics, the valence band shape (composed largely of O2p orbitals in addition to some contribution from U6d and U5f orbitals), and X-ray photoelectron spectroscopy (XPS) U4f lines, the reduction of U6+ in UO3 only results in the formation of U5+ cations and was largely limited to those on the surface. Associated with the reduction was the formation of surface hydroxyls (-OH species) due to the transfer of a proton of the H atom (H.) to surface oxygen ions, while the electron of H. is transferred to a U5f orbital. The pseudo-first-order rate constant of the initial rate of reduction at 10–7 torr and 190 K was found to be approximately 0.01 s–1. Qualitative analysis of the valence band before and after reduction indicates that O2p hybridization with U6d and U5f orbitals leads to well-distinguished features that are characteristic of UO3, U2O5, and UO2. These features, which were quantitatively reversed during the redox process, furthers the assessment of the stoichiometry of a given binary uranium oxide.

Influence of Temperature History and Flow Mixing on the Vapor-Phase Speciation of Uranium Oxide Nanoparticles

Influence of Temperature History and Flow Mixing on the Vapor-Phase Speciation of Uranium Oxide Nanoparticles

Dissolution behavior and aging of iron–uranium oxide

Understanding the dissolution behavior of the fuel debris is necessary for the safe decommissioning of Fukushima Daiichi Nuclear Power Plants. The dissolution behavior of FeUO4 compounds formed by a high-temperature reaction of UO2 with iron, a stainless-steel component of reactor structural materials, was investigated under atmospheric conditions. The compounds were prepared in an electric furnace using U3O8 and Fe3O4 as starting materials, and their solid states were analyzed using X-ray diffraction, scanning electron microscopy–energy-dispersive X-ray spectroscopy, and X-ray absorption fine-structure spectroscopy. The fission products were produced via thermal-neutron irradiation. The concentration of nuclides dissolved in water was examined by performing static leaching tests of FeUO4 compounds for up to three months. A redox reaction was proposed to occur between trivalent Fe and pentavalent U ions in the early stage of FeUO4 dissolution. It was thermodynamically deduced that the reduced divalent Fe ion was finally oxidized into a trivalent ion in the presence of dissolved oxygen, and iron hydroxide limited the solubility of Fe. Meanwhile, the concentration of hexavalent U (i.e., uranyl ion) was limited owing to the presence of secondary minerals such as metaschoepite and sodium uranate and subsequently decreased, possibly owing to sorption on Fe oxides, for example. The concentrations of multivalent ions of fission products, such as Ru and Ce, also decreased, likely for the reason above. By contrast, the concentration of soluble Cs ions did not decrease. The validity of this interpretation was supported by comparing the results with the dissolution behavior of a reference sample (Fe-free U3O8).

Uranium in Fluorite, a Case Study: The La Azul Fluorspar Deposit, Taxco, Guerrero, Mexico

Uranium can be found in the Earth’s crust in different reservoirs, with igneous rocks being the primary source of this element from which many types of secondary deposits are formed. Fluorspar deposits generally do not contain uranium, but in some cases, fluorite can carry both uranium in solid solutions and inclusions of uranium minerals. We studied the concentration (ICP-MS), composition (electronic microprobe), and spatial distribution (microscopy and auto-radiography) of elemental uranium and uranium minerals at different scales (microscopy and auto-radiography in fluorite from the La Azul fluorspar deposit (Taxco, Mexico) to assess the origin of uranium and its significance in this ore deposit. Auto-radiography images with the CR-39 detector were found to be impressive in their ability to elucidate uranium distribution at the millimeter scale. The limit between the solid solution of elemental uranium in natural fluorite and the appearance of uranium oxides as inclusions appeared to be between 20 μg g−1 and 40 μg g−1 bulk uranium concentration in this fluorspar ore. The maximum concentration of U in fluorite from the La Azul deposit was about 100 μg g−1. Using Raman spectroscopy and microprobe analysis, we identified the micro-inclusions of uranium minerals as uraninite (of the pitchblende variety); its composition suggested a hydrothermal origin for this fluorspar deposit. We also calculated a chemical age that can be compared with the previously published regional geology and isotopic (U-Th-Sm)/He ages in fluorite. Micro-thermometric studies of fluid inclusions were carried out in different samples of uranium-rich fluorite to identify the nature and origin of the mineralizing fluid and the precipitation mechanisms of uranium minerals. We concluded that the uranium-rich fluorite precipitated in the initial phases of mineralization from a reducing fluid, with low salinity (<8% NaCl eq.) and an intermediate temperature (110–230 °C), and that the presence of organic compounds and sulfides (mainly pyrite) favored the simultaneous precipitation of uraninite (pitchblende variety) and fluorite.

High-gradient magnetic separation of colloidal uranium oxide particles from soil components in aqueous suspensions

The separation of uranium oxide (UO2) particles from soil-surrogate particles in aqueous suspensions was achieved using filtration enhanced by a magnetic field. Enhanced attraction of paramagnetic UO2 colloids to a ferromagnetic stainless-steel filter placed in a strong magnetic field arises because of the positive magnetic susceptibility of the particles and the high-gradient field generated near ferromagnetic fibers. Enhanced uptake of smaller particles over larger ones occurs through Brownian motion that promotes the collision of particles with the ferromagnetic fibers of the filter. Hence, this work focused on UO2 particles in the colloidal size range. Experiments used a water-cooled electromagnet and an array of permanent magnets. Chemical analysis showed that the magnetic field increased the capture efficiency of uranium particles from a range of 27–53% with the magnet off up to 98% with the magnet on after a single pass of the suspension through the filter. The recovery of the UO2 particles from the filter, however, was more difficult to achieve. Small amounts of UO2, together with significant amounts of background SiO2 particles, were removed from the filter during a first flush with the magnetic field on. A much larger recovery of UO2 was not observed until a second out-of-field flush was performed, which also released some SiO2. The degree to which particle separation was enhanced through the use of multi-stage filtration compared to single pass-through filtration was also examined. A design was suggested that could be used to optimize the separation efficiency for a continuous process.

The Benefits of Incorporating Nanomaterials in Nuclear Power

Nuclear power has been a prominently considered power source for decades. Potentially being the most efficient fuel source surpassing hydro-electric, coal, and wind. The splitting of atoms and the chain reaction that results can release tremendous amounts of energy from very little fuel. However, it is also very controversial which accidents such as chornobyl and extremely long-lasting waste products. But with the incorporation of nanotechnologies such as nano-crystallized uranium oxide as fuel, and oxide dispersion-strengthened (ODS) alloys as a way to strengthen reactor hardware even under high temperatures. Nuclear energy could be made much safer, and even introduce commercialized fusion power with efficient deuterium extraction from seawater, and copper carbon nano alloys to act as an efficient heat conductor. Nuclear energy in the production of power has a huge future potential and could make its way into many more applications such as space exploration, shipping, and more. Nanomaterials used could help speed up the process, making it efficient, smaller, and safer.