Nuclear Fuel Research Articles

Evaluation of Desiccation Behavior in Basalt Microfiber–Reinforced Bentonite Clay for Geological Repositories of Nuclear Spent Fuel Using Digital Image Correlation

Abstract Secure storage of nuclear spent fuel (NSF) is of great concern for protecting public health and safety. The preferred long-term solution is underground containment in geological repositories, where one or more engineered barrier materials (EBM) encapsulate the NSF and separate it from the natural rock. Bentonite clay is commonly used as an EBM due to its many advantageous properties including low hydraulic conductivity, which ensures limitation of water infiltration to the system and the subsequent risk of corrosion in NSF canisters. However, bentonite clay subjected to heating from nuclear decay may form desiccation cracking. This study conducted disk-shaped free shrinkage tests and ring-shaped restrained shrinkage tests of bentonite clay samples reinforced with basalt microfibers. Digital image correlation was used as a noncontact full-field displacement measurement to track the time-evolving shrinkage and desiccation cracking phenomena and make quantified comparisons between plain bentonite and bentonite with varying contents of basalt microfibers (i.e., 0.0, 0.5, 1.0, and 1.5 % wt.). Results indicate that plain bentonite and basalt microfiber-reinforced samples showed similar free shrinkage behavior, while desiccation cracking behavior was significantly altered by adding basalt microfibers. Microfiber reinforcement effectively reduced major cracks through a “crack-bridging” effect while causing minor cracks to initiate earlier and at higher moisture contents than plain bentonite. Results infer that reinforcing plain bentonite with inorganic microfibers can potentially control desiccation cracking, leading to safer and improved nuclear waste management.

Reductive hydrothermal conversion of uranyl oxalates into UO2+x monitored by in situ XANES analyses.

Hydrothermal conversion of actinide oxalates has recently gained attention as an innovative fabrication route for nuclear fuels but has remained mainly limited to tetra- or tri-valent cations. We report herein the reductive conversion of mixtures of uranyl and oxalate ions into UO2+x oxides under mild hydrothermal conditions (T = 250 °C). A multi-parametric study first led to specifying the optimal conditions in terms of pH, oxalate/U ratio and duration to provide a quantitative precipitation of uranium in the hyper-stoichiometric dioxide form with pH = 0.8; R = noxalate/nU = 3, and t = 72 hours. Particularly, pH was evidenced as a key parameter, with 3 different compounds obtained over a range of only 0.4 units. The mechanism leading to the formation of UO2+x was then investigated thanks to an in situ XANES study. Analysis of the supernatant showed that U(VI) was quickly reduced into U(IV) thanks to the presence of oxalates and/or their decomposition products in solution, following first-order kinetics. Tetravalent uranium was then hydrolysed, leading to the precipitation of UO2+x as the only crystalline phase. This study thus demonstrates that the hydrothermal conversion of actinide oxalates into oxides is an extremely versatile tool that can be implemented in a large variety of chemical systems, particularly in terms of the oxidation state of the cations initially present in solution.

Study on the effect of graphite-based materials on the reactivity of activated aluminum composites in different forms of water

For further energy enhancement of water-reactive metal fuels, it is crucial to find aluminum composites that can react synchronously and quickly with multiple types of water (ice, steam, and water). In this paper, two graphite-based aluminum composites, aluminum/bismuth/graphite (Al/Bi/GI) and aluminum/bismuth/graphene nanosheets (Al/Bi/GE), were prepared. The morphology and crystal structure of the composites were characterized by scanning electron microscopy (SEM) and X-ray diffraction (XRD). The reactivity of Al/Bi/GI and Al/Bi/GE with water from 20 °C to 80 °C and ice at 0 °C and −15 °C was characterized using hydrogen generation properties. Their ignition and combustion properties in high-temperature water steam from 300 °C to 900 °C were also investigated. The results show that Al/Bi/GI and Al/Bi/GE composites prepared by ball milling have blocky and multilayered lamellar structures, respectively. That ball milling does not result in the formation of new alloys. The thickness of graphite material affects the reactivity of activated aluminum composites. The hydrogen generation rates of Al/Bi/GE in water and ice are much higher than that of Al/Bi/GI. The maximum hydrogen generation rates of Al/Bi/GE were 28.8 and 46.6 times higher than those of Al/Bi/GI in water at 20 °C and ice at −15 °C, respectively. In high-temperature water steam, the ignition temperature and ignition delay time of Al/Bi/GE are lower than those of Al/Bi/GI. Al/Bi/GE achieves a rapid reaction that can be synchronized with three forms of water and is expected to be applied to water-reactive metal fuel propellants.

A grain-scale modeling on the surface burning and agglomeration of the metalized solid propellants

This study delves into the dynamic behavior of reactive grains within metalized propellants during surface deflagration. Our grain-scale simulations account for both fast and slow deflagration processes involving two distinct metal fuels, namely zirconium and magnesium, with potassium perchlorate (KClO4) as the common oxidizer. Our primary goal is to develop a model and numerical technique capable of replicating the surface reactions observed in experiments. To achieve this, we utilize the compressible Navier–Stokes equations, incorporating the Arrhenius rate law and Johnson-Cook stress model. We pay particular attention to the melt layer, where surface burning occurs in an open atmosphere, and we develop a mechanism to describe the interaction between reactive grains and exhaust gases. We differentiate between two limiting scenarios: convective burning, where unreacted particles experience high-speed compression waves leading to reduced thermal diffusion within the cold reactant, and diffusive burning, characterized by insufficient pressure from particles undergoing both reaction and deformation to erode the burning surface. This results in pronounced agglomeration and the formation of gaseous metal oxides above the melt layer. The results shed light on critical factors influencing surface combustion, including the thermal softening of reactive particles, the time required for particle agglomerations, and tangential forces at multi-element interfaces within the grain-scale simulation domain.

Iron particles ignition in different hot coflow temperatures

Understanding the ignition of iron particle combustion in hot environments is critical for harnessing the use of metal fuels in clean energy production. In this research work, a new single particle burner is implemented to disperse and burn iron particles using hot air coflow. The burner disperses iron particles through Coulomb forces between two air-spaced capacitor electrodes. The oxidizer coflow is heated up as it passes through a newly designed induction gas heater. Electromagnetic induction is used to heat up multiple metallic and porous heat discs by Joule heating. When the coflow gas passes through the pores of the discs, it carries the heat generated within the porous discs and delivers it to the travelling iron particles which provides ultimate heat transfer effectiveness. Coflow temperature homogeneity is examined using Schlieren imaging and confirmed by thermocouple measurements at the outlet of the coflow tube. Particle ignition and combustion occur in open space, providing excellent conditions for optical diagnostics. Two synchronized high-speed cameras were used to simultaneously determine the particle size using green laser shadowgraphy and detect iron particles ignition. Iron particles ignition experiments were conducted for a sieved patch (45–53 μm) at different hot coflow temperatures starting at 700℃ up to 900℃. The burner showed extensive capabilities to maintains a stable and homogeneous hot coflow for a wide range of temperatures up to 900℃. Ignition experiments showed 5.67 % of ejected particles are burnt at 700 ℃, while full ignition (99.33 %) can be achieved at 900 ℃.

A combination strategy targeting the combustion enhancement of electrically controlled rocket fuels exerts synergistic elimination against zirconium's electrostatic hazards

The efforts provide a prospective fate of the zirconium powder with high ESD sensitivity and water-containing electrically controlled rocket fuels possessing poor compatibility with typical high-activity micro-nano metal fuels.

Transmutation of Plutonium and Minor Actinide in PWR Thorium-Transuranic Fuel Assembly

Long-lived radioactive waste is often considered a concerning issue on utilizing nuclear power. This waste is in the form of plutonium and minor actinides (MA), formed due to successive neutron capture of 238U. Both are not particularly hazardous radiologically, but their long half-life caused an issue in the public acceptance of radioactive waste disposal. Thereby, this issue must be resolved, either politically or technically. One of the technical solutions to address the issue of long-lived radioactive waste is the incineration of transuranic (TRU) elements in a pressurized water reactor (PWR) fuel assembly. Mixing TRU with thorium in a PWR fuel assembly can theoretically reduce TRU stockpile more effectively than uranium. This paper discusses plutonium and MA transmutation in a Thorium-Transuranic (Th-TRU) PWR fuel assembly using MCNP6 code and ENDF/B-VII library. A Westinghouse 17×17 PWR fuel assembly was chosen to determine the feasibility of TRU incineration. Three assemblies of 50:50, 55:45, and 60:40 mixture of UO2 and Th-TRU fuel rods are compared to one reference UO2 assembly from the criticality and burn-up point of view. From the calculations, the plutonium incineration rate was observed to be 9.47 %, 10.91 %, and 12.2 %, while MA incineration rates were found to be 11.5 %, 11.93 %, and 12.3 % for Th50-TRU50, Th55-TRU45, and Th60-TRU40, respectively. This observation indicates that a higher thorium fraction in the fuel assembly can increase the TRU transmutation rate. Therefore, a high thorium fraction is recommended to increase the transmutation rate of the TRU.

Transmutation of Plutonium and Minor Actinide in PWR Thorium-Transuranic Fuel Assembly

Long-lived radioactive waste is often considered a concerning issue on utilizing nuclear power. This waste is in the form of plutonium and minor actinides (MA), formed due to successive neutron capture of 238U. Both are not particularly hazardous radiologically, but their long half-life caused an issue in the public acceptance of radioactive waste disposal. Thereby, this issue must be resolved, either politically or technically. One of the technical solutions to address the issue of long-lived radioactive waste is the incineration of transuranic (TRU) elements in a pressurized water reactor (PWR) fuel assembly. Mixing TRU with thorium in a PWR fuel assembly can theoretically reduce TRU stockpile more effectively than uranium. This paper discusses plutonium and MA transmutation in a Thorium-Transuranic (Th-TRU) PWR fuel assembly using MCNP6 code and ENDF/B-VII library. A Westinghouse 17×17 PWR fuel assembly was chosen to determine the feasibility of TRU incineration. Three assemblies of 50:50, 55:45, and 60:40 mixture of UO2 and Th-TRU fuel rods are compared to one reference UO2 assembly from the criticality and burn-up point of view. From the calculations, the plutonium incineration rate was observed to be 9.47 %, 10.91 %, and 12.2 %, while MA incineration rates were found to be 11.5 %, 11.93 %, and 12.3 % for Th50-TRU50, Th55-TRU45, and Th60-TRU40, respectively. This observation indicates that a higher thorium fraction in the fuel assembly can increase the TRU transmutation rate. Therefore, a high thorium fraction is recommended to increase the transmutation rate of the TRU.

수반응 금속연료 추진제의 성능 및 물성 연구

수반응 금속연료 추진제의 성능 및 물성 연구

Preliminary Selection of Safety-Relevant Radionuclides for Long-Term Safety Assessment of Deep Geological Disposal of Spent Nuclear Fuel in South Korea

Preliminary Selection of Safety-Relevant Radionuclides for Long-Term Safety Assessment of Deep Geological Disposal of Spent Nuclear Fuel in South Korea

Design and Structural Safety Evaluation of Transfer Cask for Dry Storage System of PWR Spent Nuclear Fuel

Design and Structural Safety Evaluation of Transfer Cask for Dry Storage System of PWR Spent Nuclear Fuel

Design and Structural Safety Evaluation of Canister for Dry Storage System of PWR Spent Nuclear Fuels

Design and Structural Safety Evaluation of Canister for Dry Storage System of PWR Spent Nuclear Fuels

Assessing the Potential of Small Modular Reactors (SMRs) in Spent Nuclear Fuel Management: A Review of the Generation IV Reactor Progress

Assessing the Potential of Small Modular Reactors (SMRs) in Spent Nuclear Fuel Management: A Review of the Generation IV Reactor Progress

Rolling Test Simulation of Sea Transport of Spent Nuclear Fuel Under Normal Transport Conditions

Rolling Test Simulation of Sea Transport of Spent Nuclear Fuel Under Normal Transport Conditions

Pertechnetate/perrhenate-capped Zr/Hf-Dihydroxide Dimers: Elucidating Zr-TcO4 Co-Mobility in the Nuclear Fuel Cycle.

Spent nuclear fuel contains heavy element fission products that must be separated for effective reprocessing for a safe and sustainable nuclear fuel cycle. 93 Zr and 99 Tc are high-yield fission products that co-transport in liquid-liquid extraction processes. Here we seek atomic-level information of this co-extraction process, as well as fundamental knowledge about ZrIV (and HfIV ) aqueous speciation in the presence of topology-directing ligands such as pertechnetate (TcO4 - ) and non-radioactive surrogate perrhenate (ReO4 - ). In this context, we show that the flat tetrameric oxyhydroxyl-cluster [MIV 4 (OH)8 (H2 O)16 ]8+ (and related polymers) is dissociated by perrhenate/pertechnetate to yield isostructural dimers, M2 (OH)2 (XO4 - )6 (H2 O)6 ⋅ 3H2 O (M=Zr/HfIV ; X=Re/TcVII ), elucidated by single-crystal X-ray diffraction. We used these model compounds to understand the pervasive 93 Zr-99 Tc coextraction with further speciation studies in water, nitric acid, and tetrabutylphosphate (TBP) -kerosene; where the latter two media are relevant to nuclear fuel reprocessing. SAXS (small angle X-ray scattering), compositional evaluation, and where experimentally feasible, ESI-MS (electrospray ionization mass spectrometry) showed that perrhenate/pertechnetate influence Zr/HfIV -speciation in water. In Zr-XO4 solvent extraction studies to simulate fuel reprocessing, we provide evidence that TcO4 - enhances extraction of ZrIV , and compositional analysis of the extracted metal-complexes (Zr-ReO4 study) is consistent with the crystallized ZrIV 2 (OH)2 (ReVII O4 - )6 (H2 O)6 ⋅dimer.

Anodic dissolution of gallium in alkali metal chloride melts

Gallium and gallium based alloys can be potentially used in pyrochemical and pyroelectrochemical reprocessing of spent nuclear fuels, particularly for separating actinides and rare earth fission products. However, the electrochemical behavior of gallium in high temperature molten salt electrolytes is very little studied. The present work was aimed at investigating the processes taking place during anodic dissolution of gallium in fused alkali chlorides (the ternary 6NaCl–9KCl–5CsCl eutectic mixture) and determining the ratio of gallium chloro-species in different oxidation states formed and remained in the melt. The experiments were performed between 550 and 750 °C, and the anodic current density varied from 0.024 to 0.094 A/cm2. Anodic polarization and coulometry were used, and the results obtained demonstrated that two gallium species, Ga(I) and Ga(III), were formed as a result of gallium anodic dissolution. The ratio between these two oxidation states depended on temperature and anodic current density. The method of processing the experimental data and the calculated values of the ratio of gallium ionic forms in the system under various conditions are presented.

An assessment of PWR fuel assembly spacer grid strength by FEM models

Since Fukushima accident, integrity of fuel assembly for seismic load became very important. This study assesses the structural integrity of mid spacer grid (SG) of PLUS7 fuel assembly (FA) which is the weakest part of fuel assembly structural components. Spacer grid FEM models were developed by 3D solid element and 3D shell element. From the FEM analysis, it was found that the structural strength for normal direction was much larger than for shear direction. At core outlet, the normal stiffness was estimated to 36kN/mm for 3D solid model and 33kN/mm for 3D shell model, and the shear stiffness was estimated to 49 N/mm for 3D solid model and 75 N/mm for 3D shell model. From this result normal stiffness is a lot bigger than shear stiffness, and spacer grid structure is more vulnerable against shear loads. It is recommended to do shear test or analysis of fuel assembly in the development stage.

Investigating the Influence of Pre-Grown Copper Oxide Film on Sulfide-Induced Corrosion of Copper

Used nuclear fuel should be confined safely and permanently due to its detrimental effect on human health and the environment. The universally proposed plan for the permanent disposal of used nuclear fuels is to bury the containers containing the nuclear used fuels in a multi-barrier system known as a deep geological repository (DGR) at a depth 500 m underground. The container is the key engineered barrier that can withstand long-term mechanical loads and the corrosive environment. The proposed container in some countries will be designed with either a cylindrical Cu shell containing a nodular cast iron insert (Sweden, Finland) or a Cu‐coated carbon steel vessel (Canada). After emplacement, the container will experience a range of conditions underground, evolving from warm, humid, and oxidizing to cool, dry, and anoxic over time.During the initial oxidizing period, the entrained oxygen, trapped in the DGR upon sealing, and the water-radiolysis will lead to the formation of an oxide/hydroxide film on the Cu container surface. Once anoxic conditions are eventually established, sulfide ions produced by the action of sulfate-reducing bacteria remote from the container will become the main potential oxidant.So far, considerable efforts have been devoted to investigating exclusively either the oxic period or the anoxic period. However, the significant gap that the current research aims to address is how the early period of oxide growth affects the later stages of the repository evolution, at which time the primary threat to the integrity of the Cu container would be sulfide ions within the groundwaters. In this series of experiments, various methods were applied to create different types of copper oxide/hydroxide layers with known compositions and unique structures, to investigate their role in the subsequent sulfide-induced corrosion of the underlying Cu under de-aerated conditions. The morphology of the surface oxide film and the nature and concentration of sulfide species will influence the possible manners in which the sulfide species could interact with the oxide film; for example: Chemical conversion of copper oxide to copper sulfideGalvanically-coupled processes between the copper oxide reduction and Cu oxidation by sulfideDirect corrosion of Cu by sulfide species arriving at the metal surface Our results have shown that regardless of the composition and nature of the oxide film, either a single-phase oxide film or a dual layer of combined copper oxides, the pre-grown oxide film on the Cu surface is partially converted to a copper sulfide film via chemical conversion and/or galvanically-coupled processes. Moreover, an unreacted remnant of the oxide layer was detected on the surface, surviving for longer than the duration of our experiments. This remnant oxide is non-protective, and permits direct corrosion of Cu by sulfide species arriving at the metal surface.Oxide films were grown either electrochemically, radiolytically, or chemically. The electrochemically-grown oxides, upon exposure to sulfide solutions, underwent quick conversion to copper sulfide (Cu2S) as shown by the corrosion potential (Ecorr) measurements and cathodic stripping voltammetry (CSV) at different immersion times. The same results were observed for the radiolytically-grown oxide with a multi-layer structure composed of Cu(I) and Cu(II) oxides. The thickness of the two mentioned oxides was on the nanometre scale.Chemically grown oxides were thicker than the electrochemically- and radiolytically-grown oxides. The results of CSVs showed that some parts of the oxide film and the sulfide film formed on top of the oxide film could not be cathodically stripped from the surface and remained in the form of a defective film, and that direct corrosion of Cu occurred, driven by sulfide species reaching the metal surface. This suggests that conversion of oxide to sulfide films is fastest when an electrochemical pathway is available, and much slower if it must depend on the chemical pathway. The regions of the oxide which were not connected electrically to the surface could only undergo chemical conversion. The results obtained from immersion of copper oxide powder in sulfide solution support the idea that the chemical conversion is not as quick as the electrochemical reaction in the presence of copper metal.

Evaluating the Corrosivity of Relevant Environments That Can Form on the Surface of Spent Nuclear Fuel Canisters

Austenitic stainless steels (SS), utilized extensively in coastal environments due to their enhanced corrosion resistance, are susceptible to localized corrosion and stress corrosion cracking (SCC) when exposed to chloride environments and sufficient stresses. One potential scenario under which SCC may pose a concern is the storage and eventual transport of spent nuclear fuel (SNF) in SS canisters. The SS canisters are stored in many locations across the United States and may be subjected to corrosion from brines formed by deliquescence of salt aerosols of varying compositions. Recently, the compositions of salts collected from the surface of SNF interim storage canisters at four different storage sites have been evaluated; these represent both coastal and inland sites, and widely differing compositions. The observed salt chemistries vary from laboratory salts commonly used in corrosion testing (i.e., NaCl, MgCl2, and artificial seawater). Herein, we evaluate the relative corrosivity of the brines through polarization scans, exposure tests, and modeling efforts. The results of this study are discussed in terms of potential canister degradation and improved methods to determine the corrosivity of field-relevant environments. SNL is managed and operated by NTESS under DOE NNSA contract DE-NA0003525. This document is SAND2023-02443C.

Stress Corrosion Cracking of Austenitic Stainless Steels in Environments Relevant to Spent Nuclear Fuel Storage

Localized corrosion and stress corrosion cracking (SCC) are potential degradation mechanisms for stainless steels (SS) during exposure to corrosive environments. When localized corrosion features are subject to stresses (either external or residual), a crack can nucleate and grow. One potential scenario under which chloride-induced SCC may pose a risk is the interim storage of spent nuclear fuels (SNF) in SS canisters. In the U.S., these SS canisters are located at interim spent fuel storage installations across the country. As canisters cool over time, dust deposition can lead to the formation of corrosive brines, which form by various salt compositions. Initial studies have shown significant influences of environment on corrosion severity and morphology, which may in turn influence crack nucleation and growth. Recent, compositional analysis of salts on the surface of interim SNF SS storage canisters has been carried out with representative brine mixing formulations developed for laboratory exposures. This presentation evaluates the corrosivity, specifically in-situ crack growth rates influenced by the electrochemical environment of the prepared brines as compared to saturated salts and other seawater surrogate brines. The crack growth rates, in-situ electrochemical measurements, as well as fractography will be discussed in terms of testing methodologies and potential risks for canister degradation. SNL is managed and operated by NTESS under DOE NNSA contract DE-NA0003525. This document is SAND2023-02249A.