Inert Matrix Fuels Research Articles

Order-disorder structural transition and thermal expansion properties in Ce-substituted Y2Zr2O7 system

In the pursuit of advancing inert matrix fuel (IMF) applications, zirconate pyrochlores have emerged as promising candidates for the incorporation of minor actinides and plutonium. This study employs CeO2 as a nonradioactive surrogate for PuO2 and focuses on the synthesis of cerium-substituted Y2Zr2O7 samples (Y2-xCexZr2O7, 0.0 ≤ x ≤ 2.0) through a solid-state route under both reducing and oxidizing conditions to mimic plutonium incorporation. The impact of cerium content and its valence state, which in turn is influenced by synthesis conditions, on crystal structure and phase stability has been investigated through X-ray diffraction and Raman spectroscopic studies. Samples synthesized under reduced conditions undergo a defect fluorite to pyrochlore phase transition through a biphasic mixture consisting of these two phases upon increasing Ce-substitution. Conversely, the high temperature oxidizing synthesis conditions yield a defect fluorite phase field over a wide composition range (0.0 ≤ x ≤ 1.6) and a tetragonal phase at x = 2.0. Intriguingly, pyrochlore-type phases obtained under reducing conditions transform into the metastable kappa-phases upon mild oxidation at a relatively low temperature (1073 K), as confirmed by Raman spectroscopic studies. A combination of thermo-gravimetric and Raman spectroscopic investigations confirms the complete reduction of Ce4+ to Ce3+ under reducing conditions. The lattice thermal expansion behavior has also been investigated for the defect fluorite phases, synthesized under oxidizing conditions, by means of high temperature XRD spanning the temperature range 298–1273 K. Notably, the thermal expansion of compositions exhibits an increasing trend with increasing cerium content.

Structural and thermo-physical properties of sol-gel derived yttria stabilized zirconia microspheres: A viable inert matrix for minor actinide incineration

To embark upon the challenges in the field of nuclear industry including proliferation and waste management, the utilization of U-free fuel matrix (Inert Matrix Fuel) are proposed to be utilized to burn and transmute various minor actinides including Pu, Am etc. Thus in this direction, yttria stabilized zirconia (YSZ) microsphere matrix co-doped with different mol% of Ce and Nd were synthesized using internal gelation method. Different metal concentrations [M] and hexamthylnetetramine+urea (HMTA-urea) concentrations ([HMTA-urea]/[M]= [R]) were used in order to prepare crack free YSZ microspheres. The conventional sol gel method used led to the formation of anisotropic tetragonal phase due to the leaching of constituent elements. Thus a modified scheme was developed as elaborated in detail in the manuscript which resulted into an isotropic cubic phase for all compositions. The thermo-physical properties like thermal expansion, heat capacity and thermal conductivity of the samples were evaluated in the temperature range 298–1473 K and are at par with the values reported for other matrices. The thermal conductivity data showed that the conduction in these materials mainly occurs through phonon conduction and the presence of enhanced oxygen vacancies in them due to Ce and Nd doping, diminishes the thermal conductivity significantly. Further, the local structure was probed using EXAFS which confirmed them to be homogenous and coordination number 8 for all the samples. Thus with the development of such kind of fuels, the risk of proliferation can be avoided with additional advantage of energy generation.

Study on damage of Gd2O3-CeO2 under electronic energy loss: comparison between bulk-like and nanostructure.

To understand the physical phenomena responsible for radiation damage of the materials used in nuclear reactors, and thus study their operation life and/or efficiency, it is required to simulate the conditions by exposing the materials to energetic ions. Ceria (CeO2) has been proposed as one of the inert matrices for the transmutation of minor actinides in the futuristic inert matrix fuel (IMF) concept. The inert matrix should also contain burnable poison to compensate for the initial reactivity of fuel. In this context, gadolinium (Gd) is an excellent burnable poison with a high neutron absorption cross-section. In view of this, Gd2O3-CeO2 nano-powders were synthesized and sintered at 800 °C and 1300 °C to obtain different grain sizes and morphologies. FESEM and TEM were carried out to study the grain size of pristine pellets. The sintered pellets were irradiated with 80-MeV Ag ions (electronic energy loss (Se) regime) at room temperature to emulate the effect of fission fragments. For analysis of the effect of grain size on the irradiation-induced structural degradation at different fluences, GIXRD and Raman spectroscopy were performed. Significantly large damage has been observed for the smaller grain-sized samples (sintered at 800 °C) as compared to the large grain-sized sample (sintered at 1300 °C). Neither of the samples amorphized under the present experimental conditions as indicated by the presence of the Raman-active T2g mode (centred at 462 cm-1) and all the XRD peaks of fluorite cubic structure up to the highest fluence employed (1 × 1014 ions cm-2). X-ray photoelectron spectroscopy results demonstrate that Ce4+ to Ce3+ and vacancy-related isolated clusters are the main defects produced in the systems. The radiation tolerance behaviour of the samples is understood with the help of thermal spike simulation, which indicates higher transient lattice temperatures with longer duration in the smaller grain-sized sample upon irradiation. Gd-doped ceria thus possesses good radiation stability in the Se regime, indicating its potential for application in IMFs.

Effects of the mesostructures and irradiation conditions on the thermo-mechanical coupling behaviors in LWR-used fully ceramic-microencapsulated fuel pellets

Fully ceramic microencapsulated (FCMTM) fuels, proposed as an accident tolerant fuel in light water reactors (LWRs), consists of tri-structural isotropic (TRISO) fuel particles and silicon-carbide ceramic matrix. To ensure the in-pile service safety with improving the fuel economy simultaneously, FCM fuels should be optimally designed. The effects of TRISO particle volume fraction, particle distribution pattern and irradiation conditions on the thermal-mechanical behavior in FCM fuel pellets are investigated via the self-developed code TMAIMF(Thermo-mechanical Analysis of Inert Matrix Fuels). The evolution results of temperature, deformation and stress fields with the increase of burnup indicates that (1) when the particle volume fraction increases from ∼15% to ∼43%, the central temperature gives a rise of ∼10% at 19% FIMA (fissions per initial heavy metal atoms); the radial size increment will be shortened; while, the maximum tensile stresses of each coating layer of the TRISO particles are slightly varied; (2) the results of temperature, deformation and maximum tensile stress have weak correlations with the fuel particle distribution pattern; (3) the integrity of FCM pellets can be maintained at high fission rate conditions and extended burnup levels; (4) the finite element models with the axial displacement constrained are representative, which can obtain more conservative analysis results than the free-end models. The numerical simulation results demonstrate that the TRISO-particle-based fuel pellets could be flexibly optimized to satisfy the demands of economy and safety.

Thermal conductivity evaluation of ion irradiated Si3N4 and ZrN ceramics using spatial domain thermoreflectance

Nitride ceramics have been investigated for different applications in the nuclear industry, such as space nuclear power, fusion reactor diagnostics and plasma heating, inert matrix fuels, and accident tolerant fuels. Although thermal conductivity remains one of the most important properties to track following irradiation, traditional techniques such as laser flash and xenon flash are limited to bulk sample characterization, which requires lengthy and cost-consuming neutron irradiation. This work used spatial domain thermoreflectance (SDTR) for the micrometer-scale measurement of thermal conductivity in 15 MeV Ni ion-irradiated silicon nitride and zirconium nitride from 1 to 50 dpa and 300 to 700 °C. The SDTR-measured unirradiated thermal conductivity was found to be consistent with the published data on bulk samples. Electrically conductive ZrN exhibits modest reduction after irradiation which is minimal at the highest irradiation temperatures. In electrically insulating Si3N4, the reduction is more significant and unlike ZrN, the reduction remains significant even at a higher irradiation temperature. The thermal resistance evolution following irradiation was compared with lattice swelling, which was determined using grazing incidence x-ray diffraction, and radiation-induced defects were observed using transmission electron microscopy. A saturation value was observed between 15 and 50 dpa for thermal conductivity degradation in both nitride ceramics and a direct correlation with high-temperature defect recombination was observed, as well as the potential presence of additional carrier scattering mechanisms.

Properties and microstructure evolution of silicon nitride and zirconium nitride following Ni ion irradiation

Silicon nitride and zirconium nitride have been proposed as potential materials for multiple nuclear applications (inert matrix fuels, accident tolerant fuels, space nuclear power, fusion reactor design), yet knowledge on their behavior under irradiation remains limited. Ion irradiations were performed using 15 MeV Ni5+ ions on Si3N4 and ZrN samples, with midrange doses (around 3 µm) from 1 to 50 dpa and temperatures from 300 to 700℃. Volumetric lattice swelling was determined by grazing incidence X-ray diffraction, defect production and evolution were tracked using Transmission Electron Microscopy, and nanoindentation was performed to quantify the ceramics’ mechanical properties evolution. The results from these irradiation studies on nitride ceramics help fill the current gap present in the literature. Behavior consistent with past work on irradiated Si3N4 was observed with respect to mechanical properties and defect formation up to 15 dpa and 500°C. Failure of the grain boundary sintering aid in Si3N4 was observed above these conditions. Different behavior was observed in both nitrides at 50 dpa and 700°C, where lattice swelling increased past potential saturation values. Unreported cavity formation was witnessed in both materials under all irradiation conditions, with stable number density and slight size increase above 15 dpa. The mechanism for the cavity formation remains to be determined.

Inert-matrix fuel for transmutation: Selected mid- and long-term effects on reprocessing, fuel fabrication and inventory sent to final disposal

Partitioning and transmutation (P&T) fuel cycles provide a technical approach to ease the problem of radioactive waste disposal. Some of the partitioned components of the waste stream are irradiated while others can be used for energy production or are sent to final storage. Minor actinides are planned to be irradiated in a fast spectrum nuclear facility to transmute them into stable or short-lived isotopes. As minor actinides have negative effects on reactor dynamics, subcritical, accelerator-driven systems are proposed to increase their fraction in the fuel. An example is the MYRRHA research reactor to be built in Mol, Belgium.This reactor is modeled for depletion calculations. The behavior of special fuel elements that mirror fuel composition as envisioned for large scale transmutation facilities, namely inert-matrix fuels with an increased minor actinide content, are investigated within this reactor environment. It turns out that gamma dose rates, activity and residual heat from the spent fuel elements present significant challenges for implementing a P&T program. Spent inert-matrix fuel element show significantly higher levels than spent fuel elements from fast reactors. This requires long cooling periods and poses unprecedented challenges to reprocessing technology. The problem is amplified by the fact that it is generally agreed upon that due to low transmutation efficiencies several transmutation steps would be necessary. Looking at the radiotoxicity index, the efforts suggested to reduce the minor actinide content in a final repository are justified. The long-term safety case of deep geological repositories, however, implies that certain long-lived fission products are more relevant. The build-up of some of these radionuclides is investigated for two hypothetical German P&T scenarios. Naturally, the amount of fission products increases with continued irradiation. But namely the fraction of Cs-135 increases over-proportionally when inert-matrix fuel rich on minor actinides is used.

Effects of the key parameters of TRISO particle buffer layer on in-pile thermo-mechanical behavior in FCM fuel pellets

TRISO particle buffer layer is a critical functional part in FCM fuels. A skeleton stress model for buffer layer is proposed, dependent on the current porosity, the maximum macroscale tensile stress and the pore pressure, which is incorporated into the self-developed code of Thermo-mechanical Analysis of Inert Matrix Fuels (TMAIMF). Deeper insight is gained from the simulated thermo-mechanical variables in FCM fuel pellets, and the effects of buffer layer porosity and creep coefficient are investigated. This research indicates that (1) the macroscale maximum tensile stresses in the particle coating layers of buffer, IPyC and OPyC as well as the SiC matrix exhibit a significant decrease except those in SiC coating layers, when the initial porosity and creep coefficient of buffer layer increase; (2) with a rise of PyC creep coefficient, the maximum tensile stress of buffer layer skeleton and the buffer/IPyC interfacial tensile stresses will decrease heavily under irradiation conditions of LWRs, which gives an explanation for the no-cracking phenomenon in irradiated FCM surrogate pellets. Optimization of the microstructure of buffer layer and the creep performance of PyC materials might make the integrity and high heat transfer capability of FCM fuels maintained during a long irradiation cycle in LWRs.

Modelling of effective irradiation swelling for inert matrix fuels

The results of effective irradiation swelling in a wide range of burnup levels are numerically obtained for an inert matrix fuel, which are verified with DART model. The fission gas swelling of fuel particles is calculated with a mechanistic model, which depends on the external hydrostatic pressure. Additionally, irradiation and thermal creep effects are included in the inert matrix. The effects of matrix creep strains, external hydrostatic pressure and temperature on the effective irradiation swelling are investigated. The research results indicate that (1) the above effects are coupled with each other; (2) the matrix creep effects at high temperatures should be involved; and (3) ranged from 0 to 300 MPa, a remarkable dependence of external hydrostatic pressure can be found. Furthermore, an explicit multi-variable mathematic model is established for the effective irradiation swelling, as a function of particle volume fraction, temperature, external hydrostatic pressure and fuel particle fission density, which can well reproduce the finite element results. The mathematic model for the current volume fraction of fuel particles can help establish other effective performance models.

Modeling of irradiation-induced thermo-mechanical coupling and multi-scale behavior in a fully ceramic-microencapsulated fuel pellet

Fully ceramic microencapsulated (FCM) fuels with enhanced accident tolerance can provide an increased retention of fission products. In this study, a new method to model their irradiation-induced thermo-mechanical behavior is established. A fuel performance code called Thermo-mechanical Analysis of Inert Matrix Fuels (TMAIMF) is developed based on ABAQUS platform to enable the researchers to allow for all the irradiation effects in the three-dimensional large-deformation constitutive models of the involved materials. Especially, the important gas-release-induced thermal-mechanical effects are considered for the porous carbon layer. Finite element analysis is implemented for the thermo-mechanical coupling and multi-scale behavior in a FCM pellet, with the non-spherically symmetric thermal-mechanical interactions among all the coating layers in a TRISO particle, the TRISO particles and the matrix taken into account. Numerical simulation results indicate that (1) the irradiation shrinkage and gas-induced effective expansion strain in the porous carbon layer are responsible for the accommodation of kernel swelling, so the deformations of FCM pellets are dominated by the thermal expansion and irradiation swelling of SiC matrix; (2) the gas-induced effective expansion strains in the porous carbon layer together with the irradiation creep strains occurred in various materials protect the pellet from cracking during the whole irradiation process, with highly efficient heat transfer maintained. This study supplies a fundamental basis for advanced fabrication and optimization of FCM fuel pellets.

Proliferation resistance evaluation of an HTGR transuranic fuel cycle using PRAETOR code

The proliferation resistance (PR) of an inert matrix fuel (IMF) in the transuranic nuclear fuel cycle (NFC) of a high temperature gas cooled reactor is evaluated relative to the uranium and plutonium mixed-oxide (MOX) NFC of a light water reactor using PRAETOR code and sixty-eight input attributes. The objective is to determine the impacts of chemical stability of IMF and fuel irradiation on the PR. Specific material properties of the IMF, such as lower plutonium content, carbide ceramics coating, and absence of 235U, contribute to enhance its relative PR compared to MOX fuel. The overall PR value of the fresh IMF (an unirradiated direct use material with a one-month diversion detection timeliness goal) is nearly equal to that of the spent MOX fuel (an irradiated direct use nuclear material with a three-month diversion detection timeliness goal). Final results suggest a reduced safeguards inspection frequency to manage the IMF.

One-pot synthesis of Ln2Sn2O7 pyrochlore and MgAl2O4 spinel by soft chemistry route as potential inert matrix fuel system, and the microstructural analysis

MgAl2O4–Ln2Sn2O7 (Ln = Y, La, Nd, Gd and Tb) composite materials with 50–70 vol% of MgAl2O4 as potential inert matrix fuels have been synthesized using a soft chemistry route. X-ray diffraction (XRD), Raman, scanning and transmission electron microscopies (SEM/TEM) are employed to investigate the bulk and micro structures of both ceramic phases. Heat treatment at 1400 °C in air for 24 h leads to the formation of Ln2Sn2O7 pyrochlore and MgAl2O4 spinel which are confirmed by XRD, Raman and selected area electron diffraction (SAED). More importantly, the component in one phase does not affect the crystalline formation of the other. Thermal analysis results show MgAl2O4 crystallizes prior to that of Y2Sn2O7 and the crystallization of Y2Sn2O7 delays due to the presence of MgAl2O4. SEM shows uniformly distributed pyrochlore particles in spinel matrix, especially for the sample in equal volume ratio, with grain sizes being 1–5 μm for both phases. HRTEM and SAED observations confirm that both phases possess a high level of crystallographic perfection at the atomic scale.

Different Radiation Tolerances of Ultrafine-Grained Zirconia-Magnesia Composite Ceramics with Different Grain Sizes.

Developing high-radiation-tolerant inert matrix fuel (IMF) with a long lifetime is important for advanced fission nuclear systems. In this work, we combined zirconia (ZrO2) with magnesia (MgO) to form ultrafine-grained ZrO2–MgO composite ceramics. On the one hand, the formation of phase interfaces can stabilize the structure of ZrO2 as well as inhibiting excessive coarsening of grains. On the other hand, the grain refinement of the composite ceramics can increase the defect sinks. Two kinds of composite ceramics with different grain sizes were prepared by spark plasma sintering (SPS), and their radiation damage behaviors were evaluated by helium (He) and xenon (Xe) ion irradiation. It was found that these dual-phase composite ceramics had better radiation tolerance than the pure yttria-stabilized ZrO2 (YSZ) and MgO. Regarding He+ ion irradiation with low displacement damage, the ZrO2–MgO composite ceramic with smaller grain size had a better ability to manage He bubbles than the composite ceramic with larger grain size. However, the ZrO2–MgO composite ceramic with a larger grain size could withstand higher displacement damage in the phase transformation under heavy ion irradiation. Therefore, the balance in managing He bubbles and phase stability should be considered in choosing suitable grain sizes.

Probing the temperature effects in the radiation stability of Nd2Zr2O7 pyrochlore under swift ion irradiation

The pyrochlore structure is of particular interest to the nuclear industry due to its potential for nuclear waste immobilization and as a host material for inert matrix fuels. Herein, the experimental results of the temperature and ion fluence dependence on the disordered ion tracks produced by the electronic excitation in the pyrochlore compound Nd2Zr2O7 are presented. The samples were exposed to extreme conditions of ion irradiation with 100 MeV 107Ag9+ ions at ion fluences ranging from 3 × 1011 ions/cm2 to 2 × 1013 ions/cm2 at temperatures of 300 K and 1000 K. The in-situ X-ray diffraction along with micro-Raman spectroscopic analysis on Nd2Zr2O7 samples irradiated at 300 K displays order-disorder phase transformation which is consistent with earlier reports. The direct-impact model is used to calculate the amorphous fraction which exists due to massive deposition of electronic-energy in a localized region, and the estimated value of track radius is 2.98 ± 0.56 nm. Furthermore, theoretical thermal spike simulations are performed to estimate the track radius (∼4.2 nm) which provides strong support for experimental results. To simulate the response of Nd2Zr2O7 under nuclear reactor environment, irradiation carried out at an elevated temperature that exhibits significant radiation tolerance. Its substantiality is revealed from comparative temperature dependent irradiation studies where a slow rate of amorphization is found at the elevated temperature because of the competition between defect production and thermal healing mechanism. Thus, obtained results strongly endorse the appositeness of Nd2Zr2O7 as inert matrix material for the fabrication of futuristic nuclear fuel.

Material attractiveness evaluation of inert matrix fuel for nuclear security and non-proliferation

The impacts of chemical stability of inert-matrix-fuel (IMF) on the material attractiveness for states and non-state actors were evaluated in an open transuranic fuel cycle employing high temperature gas cooled reactors. The methodology for material attractiveness evaluation was developed to assess material attractiveness for states and treat IMF which is chemically inert for nitric acid solution. The material attractiveness was relatively assessed with physical properties of material in each of three discrete phases in the development of a nuclear explosive device. The material attractiveness assessment for non-state actors revealed that the non-irradiated TRISO fuel particle and IMF kernel for high temperature gas cooled reactors and the non-irradiated mixed oxide fuel (MOX) powder for MOX light water reactors were the most vulnerable targets in each fuel cycles. The TRISO fuel particle and IMF kernel would have less material attractiveness than the MOX powder because of their greater processing time and complexity. The material attractiveness assessment for states aiming concealed diversion revealed that the TRISO fuel particle and IMF kernel have less material attractiveness than the MOX powder, and are regarded as irradiated uranium fuel grade in the material attractiveness.

Preparation of Fine-Grained Y2.5Nd0.5Al5O12 + MgO composite ceramics for Inert Matrix Fuels by Spark Plasma Sintering

We have studied the feasibility of preparing high-density (98.6–99.5%) Y2.5Nd0.5Al5O12 (YAG)–xMgO (x = 5, 10, 20 vol %) composite ceramics by spark plasma sintering. YAG–MgO powder materials have been prepared via MgO precipitation from an aqueous solution of magnesium nitrate, Mg(NO3)2, on the surface of garnet particles. The sintering rate of the YAG–MgO composites has been shown to be controlled by volume diffusion at low temperatures and by grain-boundary diffusion at elevated temperatures.

A study of fine-grained ceramics based on complex oxides ZrO2-Ln2O3 (Ln = Sm, Yb) obtained by Spark Plasma Sintering for inert matrix fuel

Fine-grained ceramics based on solid solutions of ZrO2-x-Ln2O3 (where Ln = Sm, Yb), which are used to simulate inert matrix fuel (IMF), were obtained by colloidal chemical synthesis of powders and their Spark Plasma Sintering (SPS). It has been established that the SPS activation energy (Qs) for fine-grain ceramics based on ZrO2 corresponds to the activation energy of oxygen diffusion on grain boundaries of zirconium oxide (zirconia). We have demonstrated that introducing oxides of samarium (Sm2O3) and ytterbium (Yb2O3) into zirconia decreases SPS activation energy and stabilizes the phase composition of zirconia. Ceramics with densities up to 100% were obtained by the following modes of SPS: sintering temperature 1000–1250 °C, sintering time 6–10 min. Hydrolysis testing in the static mode (Т = 20 °C, P = 1 atm) in distilled water revealed no leaching of lanthanide ions at detection thresholds of 5 ppm for Sm ions and 0.1 ppm for Yb ions.

Thermal conductivity of Mo-reinforced ZrO2 composites fabricated by spark plasma sintering for inert matrix fuels

ZrO2 is one of the candidate materials for the matrix of inert matrix fuels because of its excellent chemical stability and irradiation resistance. In this paper, the low thermal conductivity of ZrO2 was increased by the addition of Mo reinforcements in order to improve its performance and therefore increase the safety of nuclear reactors. The effect of interfacial thermal resistance and reinforcement structure in terms of its size, shape, and orientation on the thermal conductivity of Mo powder- and Mo wire mesh-reinforced ZrO2 composites fabricated by spark plasma sintering at 1700°C in vacuum was investigated. It was found that structures with a higher degree of interconnectivity of the high thermal conductivity reinforcements were more effective in increasing the thermal conductivity of ZrO2 composites. The Mo wire mesh-ZrO2 composites had a higher thermal conductivity than did the Mo powder-ZrO2 composites, owing to the interconnected structure of the former. The thermal conductivities of the Mo-ZrO2 composites were also analyzed by the finite element analysis and using analytical models of the composites with the dispersed reinforcing particles and the interconnected reinforcement structures. A good agreement was found between the theoretically predicted and experimentally measured thermal conductivity of Mo-reinforced ZrO2 composites.

Oxidation and reduction behaviors of a prototypic MgO-PuO2-x inert matrix fuel

Oxidation and reduction behaviors of prototypic MgO-based inert matrix fuels (IMFs) containing PuO2-x were experimentally investigated by means of thermogravimetry. The oxidation and reduction kinetics of the MgO-PuO2-x specimen were determined. The oxidation and reduction rates of the MgO-PuO2-x were found to be low compared with those of PuO2-x. It is note that the changes in O/Pu ratios of MgO-PuO2-x from stoichiometry were smaller than those of PuO2-x at high oxygen partial pressure.

Preparation of mesoporous zirconia microspheres as inert matrix

Mesoporous zirconia microspheres, with a diameter of 900 μm, were prepared as an inert accelerator driven system (ADS) transmutation element matrix by the sol-gel method. The purpose of mesopores is to improve the adsorption capacity of inert matrix fuel (IMF) for minor actinides. The study indicated that the mesoporous zirconia performance was improved after the microspheres were hydrothermally treated at 150 °C, the specific surface area increased from 28.29 m2/g to 61.28 m2/g, and hydrothermal treatment avoided the cracking of the microspheres. Pre-decomposition of the organics during the hydrothermal process stabilized the mesoporous structure. The average pore diameter of mesoporous microsphere was 14.3 nm.