Inert Matrix Fuels Research Articles

High Temperature, High Pressure Water Corrosion Behavior of MgO-pyrochlore Composites for Inert Matrix Fuel Applications

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MgO–pyrochlore composite as an inert matrix fuel: Neutronic and thermal characteristics

Inert matrix fuels are an important component of advanced nuclear fuel cycles, as they provide a means of utilizing plutonium and reducing the inventory of ‘minor’ actinides. We examine the neutronic and thermal characteristics of MgO–pyrochlore (A 2B 2O 7: La 2Zr 2O 7, Nd 2Zr 2O 7 and Y 2Sn 2O 7) composites as inert matrix fuels in boiling water reactors. By incorporating plutonium with resonance nuclides, such as Am, Np and Er, in the A-site of pyrochlore, the k inf vs. burn-up curves are shown to be similar to those of UO 2, although the Doppler coefficients are less negative than UO 2. The Pu depletion rates are 88–90% ( 239Pu) and 54–58% (total Pu) when the inert matrix fuels experience a burn-up equivalent of 45 GWd/tHM UO 2. Because of the high thermal conductivity of MgO, the center-line temperatures of the MgO–pyrochlore composites at 44.0 kW/m are lower than those of UO 2 pellets. After burn-up, the A-site cation composition is 15–35 at.% lower than that of the B-site cations in pyrochlore (e.g., A 1.84B 2.17O 7.00) due to the fission of Pu in the A-site and the presence of fission product elements in the A- and B-sites of the pyrochlore structure.

Reprocessing Silicon Carbide Inert Matrix Fuel by Molten Salt Corrosion

AbstractSilicon carbide is one of the prime matrix material candidates for inert matrix fuels (IMF) which are being designed to reduce plutonium and long half-life actinide inventories through transmutation. Since complete transmutation is impractical in a single in-core run, reprocessing the inert matrix fuels becomes necessary. The current reprocessing techniques of many inert matrix materials involve dissolution of spent fuels in acidic aqueous solutions. However, SiC cannot be dissolved by that process. Thus, new reprocessing techniques are required.This paper discusses a possible way for separating transuranic (actinide) species from a bulk silicon carbide (SiC) matrix utilizing molten carbonates. Bulk reaction-bonded SiC and SiC powder (1 μm) were corroded at high temperatures (above 850 °C) in molten carbonates (K2CO3 and Na2CO3) in an air atmosphere to form water soluble silicates. Separation of Ceria (used as a surrogate for the plutonium fissile fuel) was achieved by dissolving the silicates in boiling water and leaving behind the solid ceria (CeO2).

Mechanisms of Radiation Damage and Properties of Nuclear Materials

AbstractRadiation damage effects in ceramics, e.g., nuclear waste forms, transmutation targets, and inert matrix fuels, may have important implications for the physical and chemical stability of these materials as the cumulative radiation dose increases over time. A key aspect of scientific research in this area is the ability to understand the fundamental damage mechanisms through the combination of experimental and atomistic modelling techniques. In this paper, we review some of the lessons learned from the significant body of data now available for pyrochlore-defect fluorite based materials, followed by an illustration of the advantages of working on simple compounds with well established interatomic potentials. We conclude the paper with a description of radiation damage processes in the LaxSr1-1.5xTiO3 defect perovskites, a system that includes phase transformations, short-range order effects, and complex defect behavior.

Low Temperature Synthesis of Silicon Carbide Inert Matrix Fuel (IMF)

AbstractA process for the synthesis of silicon carbide (SiC) inert matrix fuels at a low temperature (1050 °C) is reported which utilized a liquid polymer precursor. As the polymer content increased, the theoretical density of the pellet at first increased and then reached a plateau. From the onset of the plateau, the packing of the one micron SiC particles in the green body was determined to be 64-68% at 600 MPa pressing pressure. As expected, mixing coarse and fine SiC particles gave a higher pellet density. The maximum density achieved was 80% of the theoretical density. Mercury porosimetry showed that the largest pore size was around 10% of the largest particle sizes present in the green body. SEM images showed that ceria, which was selected as a surrogate for PuO2 in the present study, was well distributed.

Densification of Inert Matrix Fuels Using Naturally-occurring Material as a Sintering Additive

AbstractWe proposed a new concept for densification of inert matrix fuels containing minor actinides. In this concept, magnesium silicates which are both a naturally-occurring material and asbestos waste were used as a sintering additive which protects public health by safely disposing of the asbestos waste. In this study, the effects of magnesium silicate additives on the densification behaviors of MgO, Mo and CeO2 were experimentally investigated. The densities of MgO and CeO2 pellets increased with only 1 wt.% additives of MgSiO3 and Mg2SiO4. The densities of Mo pellets showed little change with additives.

Analytical investigations of irradiated inert matrix fuel

Yttria stabilised zirconia (YSZ) inert matrix fuel (IMF) fabricated at PSI and irradiated 3 years in the Halden Material Test Reactor (HBWR) since 2000, has been examined by Electron Probe Microanalysis (EPMA) and Secondary Ion Mass Spectroscopy (SIMS) after irradiation and compared with data gained for the unirradiated material. The examined pellet cross-section was estimated to have an equivalent burn-up of 22MWdkg−1. EPMA measurements demonstrate that the burn-up was rather flat over more than the half pellet radius. A Pu consumption of about 2.5 wt% has been measured with a higher rate in the fuel border zone. The high fuel temperature is responsible for a certain homogenisation of the mineral phases in the fuel centre region whereas the border zone has remained rather with an as-fabricated phase distribution. The central part was also characterised by a dense porosity distribution as well as a temperature and relocation driven depletion of the volatile fission products Xe and Cs. In addition, SIMS has been realised on the same specimen in order to determine the semi-quantitative distribution of different isotopes in the pellet.

Inert Matrix Fuel Analysis by Means of the TRANSURANUS Code: The Halden IFA-652 In-Pile Test

Inert matrix fuels (IMFs) are a possible option to reduce separated plutonium stockpiles by burning it in light water reactor (LWR) fleet. A high burning efficiency targeted by preventing new plutonium buildup under irradiation (U-free fuel), a proved high radiation damage, and leaching resistance are fundamental requirements when a once-through fuel cycle strategy is planned. Among other options, both calcia-stabilized zirconia (CSZ) and thoria fulfill these criteria standing as the most promising matrices to host plutonium. While several in-pile tests concerning thoria fuels are found, calcia-stabilized zirconia under-irradiation performance is still to be fully assessed; with this regard the thermal conductivity, markedly lower than the uranium oxide (UOX) and mixed oxide (MOX) cases, plays a fundamental role. For this reason, ENEA has conceived a comparative in-pile testing of three different U-free inert matrix fuel concepts, which have been performed in the OECD Halden HBWR (IFA-652 experiment). The discharge burnup accomplished about 90–97% of the 45MWd∕kgUeq target under typical LWR irradiation conditions. The test rig is a six-rod bundle loaded with IM, IMT, and T innovative fuels. IM and T fuels have, respectively, CSZ and thoria as matrices, the fissile phase being the high enriched uranium (HEU) oxide (UO2 93% U235 enriched). IMT is a ternary fuel composed by CSZ+thoria matrix and HEU oxide as a fissile phase. Thoria is added in IMT fuel to improve the low IM reactivity feedback coefficients. Pins are instrumented providing fuel centerline temperature, pin inner pressure, and fuel stack elongation measurements. Our purpose is to investigate the key processes of IMF under-irradiation behavior by means of the TRANSURANUS fuel performance code. Thermal conductivity and its degradation with burnup, densification-swelling response, and fission gas release (FGR) are tentatively modeled in the burnup range of IFA-652. In particular, the effects of pellet geometry and fuel microstructures in the IM and IMT cases are pointed out. The consistency of our results is discussed aiming at understanding the in-pile response, as a fundamental step, in the perspective of future deployment of the nuclear fuels we are dealing with. Notwithstanding this ambitious objective, it is clear, however, that these results rely on a limited data set and that, as TRANSURANUS is a semi-empirical code mostly tailored for commercial fuels, the modeling of the IMF is still a work in progress.

Thoria and inert matrix fuels for a sustainable nuclear power

Nuclear power to be sustainable requires the fulfilling of peculiar constraints, which include addressing the proliferation risk. One possible route for sustainability is that to adopt a fuel cycle based on thorium. However, comparison with uranium cycle indicates that thorium cycle utilization is premature. Instead, a promising short-term option is the use of inert matrix fuels, possibly containing thoria, in a once-through cycle. Irradiation tests performed in the Halden reactor show encouraging behaviour under irradiation. Furthermore, these fuels are very well suited for a direct disposal in a geological repository.

Effect of swift heavy ion irradiation in pyrochlores

A summary of the results on the radiation stability of pyrochlores (Gd2Zr2O7, Nd2Zr2O7, Gd2Ti2O7) in the electronic stopping regime is given. X-ray diffraction and Raman spectroscopy show a transformation from the ordered pyrochlores (Fd3m) to the anion deficient fluorite (Fm3m) structure in Gd2Zr2O7 or amorphization in Nd2Zr2O7. The track diameter, crystallite size and progression in structural modifications have been monitored. It is found that radiation stability of these pyrochlores depends on the energy to form cation antisites and delocalized anion vacancies. This in turn is correlated to the radius ratio of A–B cations. This criterion for evaluating radiation resistance of pyrochlores has important implications in the field of inert matrix fuels.

Decay heat production in a TRU burner

The design of new reactors such as ADS has been investigated in many countries during the last years for burning transuranic nuclides (TRUs) contained in spent reactor fuel. To increase the TRU incineration rate, fertile-free dedicated fuels, which may contain a large fraction of minor actinides (MAs), are currently considered. Based on past experience, R&D activities for dedicated fuels in Europe concentrate on fuel forms, in which the oxide actinide phase is mixed with oxide or metal inert matrices. Decay heat in a system with inert matrix fuel (IMF) containing MAs may differ from that in a conventional fast reactor. In this paper, several fast reactor designs with different TRU content are considered and related decay heat values, calculated on the basis JEFF 3.0 and JEFF 3.1 nuclear data libraries, are compared. It is shown that some decay heat components for fuels with MAs may be lower than those for MA-free fuels, but the total decay heat may be significantly higher for cooling times exceeding about 1 min.

Zirconia Inert Matrix Fuel for Plutonium and Minor Actinides Management in Reactors and as an Ultimate Waste Form

AbstractA plutonia stabilised zirconia doped with yttria and erbia has been selected as inert matrix fuel (IMF) at PSI. The results of experimental irradiation tests on yttria-stabilised zirconia doped with plutonia and erbia pellets in the Halden research reactor as well as a study of zirconia solubility are presented. Zirconia must be stabilised by yttria to form a solid solution such as MAz(Y,Er)yPuxZr1-yO2-ζ where minor actinides (MA) oxides are also soluble. (Er,Y,Pu,Zr)O2-ζ (with Pu containing 5% Am) was successfully prepared at PSI and irradiated in the Halden reactor. Emphasis is given on the zirconia-IMF properties under in-pile irradiation, on the fuel material centre temperatures and on the fission gas release. The retention of fission products in zirconia may be stronger at similar temperature, compared to UO2. The outstanding behaviour of plutonia-zirconia inert matrix fuel is compared to the classical (U,Pu)O2 fuels. The properties of the spent fuel pellets are presented focusing on the once through strategy. For this strategy, low solubility of the inert matrix is required for geological disposal. This parameter was studied in detail for a range of solutions corresponding to groundwater under near field conditions. Under these conditions the IMF solubility is about 109 times smaller than glass, several orders of magnitude lower than UO2 in oxidising conditions (Yucca Mountain) and comparable in reducing conditions, which makes the zirconia material very attractive for deep geological disposal. The behaviour of plutonia-zirconia inert matrix fuel is discussed within a burn and bury strategy.

Stability of Cubic Zirconia in a Granitic System Under High Pressure & Temperature

AbstractCubic zirconia is a well known, highly durable material with potential uses as an actinide host phase in ceramic waste forms and inert matrix fuels and in containers for very deep borehole disposal of some highly radioactive wastes. To investigate the behaviour of this material under the conditions of possible use, a cube of ∼ 2.5 mm edge was made from a single crystal of yttriastabilized cubic zirconia doped with 0.3 wt.% CeO2. The cube was enclosed in powdered granite within a gold capsule and a small amount of H2O added before sealing. The sealed capsule was held for 4 months in a cold-seal pressure vessel at a temperature of 780°C and a pressure 150 MPa, simulating both the conditions of a deep borehole disposal involving partial melting of the host rock and the conditions under which the actinide waste form might be encapsulated in granite prior to disposal. At the end of the experiment the quenched, largely glassy, sample was cut into thin slices and studied by optical microscopy, EMPA, SEM and cathodoluminescence methods. The results show that no corrosion of the zirconia crystal or reaction with the granite melt occurred and that no detectable diffusion of elements, including Ce, in or out of the zirconia took place on the timescale of the experiment. Consequently, it appears that cubic zirconia could perform most satisfactorily as both an actinide host waste form for encapsulation in solid granite for very deep disposal and as a container material for deep borehole disposal of highly radioactive wastes (HLW), including spent fuel.

Towards a Single Host Phase Ceramic Formulation for UK Plutonium Disposition

AbstractThe UK has a considerable stockpile of separated plutonium; a legacy of over 50 years of civilian nuclear programmes. This material has been considered both as an asset for future energy generation and a liability due to the proliferation threat. A proportion of the PuO2stocks may be consumed by nuclear fission, in mixed oxide (MOx) or inert matrix (IMF) fuels but a quantity of waste PuO2will remain which is unsuitable for fuel manufacture and will require immobilisation. A research program is currently underway to investigate the potential of various single phase ceramic formulations for the immobilisation of this waste PuO2 fraction. In this work a number of synthetic mineral systems have been considered including titanate, zirconate, phosphate and silicate based matrices. Although a wealth of information on plutonium disposition in some of the systems exists in the literature, the data is not always directly comparable which hinders comparison between different ceramic hosts. The crux of this research has been to compile a database of information on the proposed hosts to allow impartial comparison of the relative merits and shortcomings in each system.

Plutonium management with thorium-based fuels and inert matrix fuels in thermal reactor systems

The plutonium that is produced by light water reactors worldwide is currently re-used to a limited extent. In the last century, the expected introduction of fast reactors and the associated need for large amounts of plutonium did not take place. The result is that worldwide a stockpile of excess plutonium has formed, which is the dominant contributor to the radiotoxicity of spent nuclear fuel for storage times from ∼10 2 to 10 5 years. One option to reduce or stabilize the plutonium stockpile is to utilize this plutonium in advanced fuel types, such as thorium-based and inert matrix fuels. Because these fuels do not contain uranium, the plutonium consumption rate is very high. In this paper, the status of the fuel research and some recent developments are given.

Laboratory production of zirconium carbide compacts for use in inert matrix fuels

Zirconium carbide is being actively considered for use as an inert matrix material in composite nuclear fuel for gas-cooled fast reactors. ZrC can be produced either by the endothermic carbothermal reduction of zirconium dioxide or by the direct exothermic reaction of pure zirconium and graphite powder mixtures. The exothermic reaction is classified as combustion synthesis or self heating synthesis. Experiments were conducted to demonstrate the combustion synthesis reaction of zirconium and graphite powders and measure the ignition and adiabatic temperatures. The heat released during this short reaction time was sufficient only to partially sinter the compacts to less than 40% theoretical density. Subsequently, compacts of ZrC were similarly produced by combustion synthesis followed by a short, high temperature hold at 2440 °C to relieve residual stresses in the compacts following the rapid reaction sintering. External pressures of up to 5.2 MPa were used as an additional driving force for sintering. The effects of reactant particle size and degree of uniaxial pressing on the product density and porosity were also studied. Higher densities in the fabricated compacts were noted for higher uniaxial pressures irrespective of powder size. Also, smaller powder sizes produced compacts up to 92% TD, while larger particle sizes produced compacts up to 84% TD for the same pressure. The compacts were characterized based on composition, microstructure, and density/porosity. Results of the different experiments are presented.

Core Burnup Calculation of Plutonium Burning Inert-Matrix Fueled High Temperature Gas Cooled Reactor

For the efficient reduction of excess plutonium amount, Japan Atomic Energy Research Institute (JAERI, now Japan Atomic Energy Agency) has studied a concept of rock-like oxide (ROX) fuel as a kind of inert matrix fuel (IMF). In the JAERI study, ROX fuel is burnt in existing light water reactors (LWRs), while in this study, pebble bed type high temperature gas cooled reactor (HTGR) is studied, mainly because of its high neutron economy and softer neutron spectrum than LWRs. Here, PuO2-yttria stabilized zirconia (YSZ: (Zr,Y)O2-x) particles are dispersed in graphite matrix. In the ROX fueled LWR study, it was necessary to improve fuel temperature reactivity coefficients by adding such additives as 238U and Er. Here in HTGR, although the negative temperature coefficient is much larger than that in LWR without any improvements, temperature coefficient was improved as large as possible to the level of UO2 HTGR case by adding Er in the fuel. Burnup calculations on PuO2-YSZ fueled HTGR, by simulating the continuous refueling of fuel pebbles with the batch fuel loading, showed almost complete transmutation for 239Pu and more than 80% for the total plutonium. As the maximum power density of the fuel pebble obtained by the core burnup calculation was very large in comparison with the UO2 HTGR, the maximum temperature in YSZ fuel particle was also evaluated. Despite the low thermal conductivity of YSZ, the evaluated YSZ temperature was well below the melting point, thanks to the high thermal conductivity of graphite and small YSZ particle size. Here, the high power density in the Pu-YSZ fueled core might become a problem, but is possible to be reduced by adjusting the initial plutonium enrichment.

Effect of sintering conditions on the microstructure and mechanical properties of ZrN as a surrogate for actinide nitride fuels

Pellets of sintered ZrN were studied to optimize the mechanical properties and microstructures needed in nitride fuel pellets, using ZrN as a surrogate for actinide nitrides and as potential component in low fertile and inert matrix fuels. Samples were prepared via sintering in either Ar or N2 (with and without 6% H2) and at 1300°C or 1600°C. A significant difference in the hardness was measured ranging from 1000 (Kg/mm2) in samples sintered at 1600°C in argon to 100 (Kg/mm2) in samples sintered at 1300°C in nitrogen. Samples with 6% hydrogen added to the sintering environment experienced a decrease in hardness, as well as an increase in intergranular cracking as compared to samples sintered without hydrogen, suggesting hydrogen embrittlement. Grain size was more uniform in samples sintered in pure Ar as compared to Ar–H2, while the latter had a larger fraction of high angle grain boundaries than the former. Cracking around indents had a clear tendency to follow high angle boundaries, which were found to be intrinsically weak in ZrN.

Studies of an Accelerator-Driven Transuranium Burner with Hafnium-Based Inert Matrix Fuel

Neutronic and burnup characteristics of an accelerator-driven transuranium burner in a startup mode were studied. Different inert and absorbing matrices as well as lattice configurations were assessed in order to identify suitable fuel and core design configurations. Monte Carlo transport and burnup codes were used in the analyses. The lattice pin pitch was varied to optimize the source efficiency and coolant void worth while respecting key thermal and material-related design constraints posed by fuel and cladding. A HfN matrix appeared to provide a good combination of neutronic, burnup, and safety characteristics: maintaining a hard neutron spectrum, yielding acceptable coolant void reactivity and source efficiency, and alleviating the burnup reactivity swing. A conceptual design of a (TRU,Hf)N fueled, lead-bismuth eutectic-cooled accelerator-driven system was developed. Twice higher neutron fission-to-absorption probabilities in Am isotopes were achieved compared to reactor designs relying on ZrN or YN inert matrix fuel. The production of higher actinides in the fuel cycle is hence limited, with a Cm fraction in the equilibrium fuel being ~40% lower than for cores with ZrN matrix-based fuel. The burnup reactivity swing and associated power peaking in the core are managed by an appropriate choice of cycle length (100 days) and by core enrichment zoning. A safety analysis shows that the system is protected from instant damage during unprotected beam overpower transient.

Zirconia inert matrix for plutonium utilisation and minor actinides disposition in reactors

The radiotoxicity of the UO 2 spent fuel is dominated by plutonium and minor actinides (MA): Np, Am and Cm, after decay of the short life fission products. Zirconia ceramics containing Pu and MA in the form of an inert matrix fuel (IMF) could be used to burn these actinides in light water reactors or in high temperature reactors. Optimisation of the fuel designs dictated by properties such as thermal, mechanical, chemical and physical must be performed with attention for their behaviour under irradiation. Zirconia must be stabilised by yttria to form a solid solution such as MA z Y y Pu x Zr 1− y O 2− ζ where minor actinide oxides are also soluble. MA may act as a burnable poison reducing the reactivity at the beginning of life and yielding fertile nuclides improving the reactivity at the end of life. These zirconia cubic solid solutions are stable under heavy ion irradiation. The retention of fission products in zirconia, under similar thermodynamic conditions, is a priori stronger, compared to UO 2, the lattice parameter being larger for UO 2 than for (Y,Zr)O 2− x . (Er,Y,Pu,Zr)O 2− x in which Pu contains 5% Am was successfully irradiated in the Proteus reactor at PSI, in the HFR facility, Petten as well as in the Halden reactor. These tests support potential irradiations of such IMF in a commercial reactor. This would allow later a commercial deployment of such a zirconia fuel for Pu and MA utilisation in a last cycle. The fuel forms namely pellet-fuel, cercer, cermet or coated particle fuel are discussed considering the once through strategy. For this strategy, low solubility of the inert matrix is required for geological disposal. As spent fuels these IMFs must be excellent materials from the solubility point of view, this parameter was studied in detail for a range of solutions corresponding to groundwater under near field conditions. Under these conditions the IMF solubility is about 10 6 times smaller than glass, which makes the zirconia material very attractive for deep geological disposal. The desired objective would be to use IMF to produce energy in reactors, opting for an economical and ecological solution.