Thorium-uranium Dioxide Research Articles

Assessment of alternative nuclear fuel cycles for the Brazilian nuclear energy system

Alternative nuclear fuel cycles could contribute to the future of nuclear fission technology reducing the use of natural resources, waste management and lessening greenhouse gas emissions. To increase understanding of future technologies, analysis of power generation systems are essential and supports long-term decision-making. On this basis, this research presents an evaluation and comparison considering the use of alternative nuclear fuels in Pressurized Water Reactors (PWR) based on the Brazilian nuclear energy system. Assessment results show consumption of resources, inventories of spent fuel, nuclear waste, greenhouse gas emissions and costs of nuclear fuels. Among the main results, Mixed Oxide (MOX) fuel appears as the most promising by multi-criteria analysis considering two of the three evaluated areas. On the other hand, if costs have an importance greater than 55%, the conventional fuel UOX appears as the most advantageous. Thorium-uranium dioxide (TRU-U) and Transuranic-thorium dioxide (TRU-Th) fuels are in intermediate positions, along with Uranium oxide (UOX), and may still be options depending on the intentions of decision-makers.

Nanofibers of solid-solution thorium(IV)-uranium(IV) oxides by electrospinning

Thorium-uranium dioxide nanofibers were prepared in a three-step process. Green composite microfibers were electrospun from solutions composed of Th(IV) and U(VI) acetylacetonate complexes in different molar ratios, polyvinylpyrrolidone, and acetone/ethanol mixed solvents. The second step converted the green composite fibers by calcination in the air to mixed Th/U oxide fibers. In the final step, the heat treatment under forming gas provided the nanofibers with diameters of 28−46 nm composed of solid solutions ThxU1−xO2 (x = 1, 0.75, 0.5, 0.25, 0). The XRD and STEM-EDS analyses of the prepared Th/U mixed oxide nanofibers attested to their atomic homogeneity.

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

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

Comparisons of Reduced Moderation Small Modular Reactors With Heavy Water Coolant

This paper presents the comparison of two reduced moderation small modular reactor concepts with heavy water coolant. Two reduced moderation small modular reactors, RMSMR-Th and RMSMR-MOX, are proposed for the sustainable utilization of nuclear resources. The design concepts are established on modifications of the well-experienced pressurized water reactor technology. Tightly-packed lattice and heavy water coolant are employed to yield a hard neutron spectrum, which proved advantageous for increasing the conversion ratio as well as lowering the burnup reactivity swing between beginning of cycle and end of cycle. Thorium-uranium dioxide fuel and MOX fuel are compared using the same core arrangement, and the small modular reactor concept is adopted to reduce void coefficients. Radial blanket region and axial blanket region are adopted to enhance the fissile breeding and a three-zone fuel configuration is adopted to flatten the power distribution. Core burnup calculations were carried out to investigate the available cycle length, the conversion ratio, the power peaking factors, reactivity coefficients, etc. Light-water based thermal-hydraulic models were employed to examine the safety features of the concepts. Numerical simulations indicate that both RMSMR-Th and RMSMR-MOX can sustain the power generation of 100MWe by 7 years without refueling. Compared with thorium-uranium dioxide fuel, MOX fuel is helpful in reducing the burnup reactivity swing, increasing the conversion ratio and increasing the mDNBR value. However, the positive void coefficient becomes a serious problem making RMSMR-MOX less attractive than the RMSMR-Th concept.

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

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

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

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

A Thoria and Thorium Uranium Dioxide Nuclear Fuel Performance Model Prototype and Knowledge Gap Assessment

Thorium-based fuel cycles can improve fuel sustainability within the nuclear power industry. The Canadian supercritical water-cooled reactor (SCWR) concept uses this path to achieve the sustainability requirement of the Gen-IV Forum. The study of thorium dioxide/thoria ThO2-based fuel irradiation behavior is significantly less advanced than that of uranium dioxide (UO2) fuel, although ThO2 possesses superior thermal conductivity, thermal expansion, higher melting temperature, and oxidation resistance that may improve both fuel performance and safety. The fuel and sheath modeling tool (FAST), a fuel performance model for UO2 fuel, was developed at the Royal Military College of Canada (RMCC). FAST capability has been extended to include thoria (ThO2), thorium uranium dioxide (Th,U)O2, and thorium plutonium dioxide (Th,Pu)O2 as fuel pellet materials, to aid in designing and performance assessment of Th-based fuels, including SCWR (Th,Pu)O2 fuel. The development and integration of ThO2 and (Th,U)O2 models into the existing FAST model led to the multipellet material FAST (MPM-FAST). Model development was performed in collaboration between RMCC and Canadian Nuclear Laboratories (CNL). This paper presents an outline of the ThO2 and (Th,U)O2 MPM-FAST model, a comparison between modeling results with postirradiation examination (PIE) data from a test conducted at CNL, and an account of the knowledge gap between our ability to model ThO2 and (Th,U)O2 fuel compared to UO2. Results are encouraging when compared to PIE data.

Fabrication and Thermal Conductivity of (Th,U)O2 Pellets

Techniques to fabricate thorium-uranium dioxide fuel [(Th,U)O2] have been developed, and the thermal conductivity of (Th,U)O2 pellets has been measured. Mixtures of thorium dioxide (ThO2) and uranium dioxide (UO2) powders were successfully wet-milled, compacted, and sintered at 1700°C to fabricate (Th,U)O2 pellets. The wet-milling process results in a fuel density of 96 to 98% of theoretical density and a uniform distribution of the uranium and thorium in the (Th,U)O2 pellet. The laser flash method was used to measure the thermal diffusivity of the ThO2 and (Th,U)O2 pellets, and the thermal conductivities of (Th0.655U0.345)O2 and (Th0.355U0.645)O2 fuel were found to be lower than that of ThO2 or UO2 fuel. The degradation of the thermal conductivity by the UO2 substitution is great at low temperatures but decreases as the temperature increases.

Costs for Manufacturing Thorium-Uranium Dioxide Fuels for Light Water Reactors

The results of a 2-yr effort to determine the capability of U.S. fuel manufacturers to economically manufacture thorium-uranium dioxide (ThO2-UO2) fuel in plants that have previously only manufactured UO2 fuel with <5% 235U enrichment are presented. It was determined that there were no fundamental obstacles to converting the current plants that manufacture a uranium-oxide-only fuel to a mixed thorium-uranium dioxide fuel. However, the differential costs for manufacturing a 75% ThO2–25% UO2 fuel, with the uranium enriched with 20% 235U, as compared to a 100% UO2 fuel, was between $269 and $291/kg of metal oxide fuel, depending on the manufacturing method used to convert the uranium and thorium feeds to the dioxide powders. More than 90% of this cost was associated with the increased cost of the uranium feed and the addition of the thorium feed. If a 70% ThO2-30% UO2 fuel were used, the differential costs would increase to between $519 and $542/kg of metal oxide fuel, of which >95% is associated with the uranium and thorium feed materials.

Use of Thoria-Urania Fuels in PWRs: A General Review of a NERI Project to Assess Feasible Core Designs, Economics, Fabrication Methods, In-Pile Thermal/Mechanical Behavior, and Waste Form Characteristics

This paper provides an introduction to and a summary of the remaining papers in this issue of Nuclear Technology. The papers in this issue present the important results from a U.S. Department of Energy–sponsored Nuclear Engineering Research Initiative (NERI) project to study the efficacy of the thorium-uranium dioxide (ThO2-UO2) once-through fuel cycle in current light water reactors. The project addressed fuel cycle neutronics and economics; ThO2-UO2 fuel manufacturing; the in-pile thermal/mechanical behavior of ThO2-UO2 fuel during normal, off-normal, and accident conditions; and the long-term stability of ThO2-UO2 waste. Results from this work show that a small-scale separation of the uranium and thorium will enhance the fuel reactivity and achievable burnup from uranium-thorium dioxide fuels. Under conditions that meet the thermal requirements in present pressurized water reactors (PWRs), a properly designed microheterogeneous fuel will have more reactivity than all-uranium fuel, and the overall production of plutonium is significantly reduced. The use of thorium as a host for actinide fuels when PWRs are used for actinide transmutation was also explored. It was also determined that there were no fundamental obstacles to converting the current plants that manufacture uranium oxide–only fuel to a mixed ThO2-UO2 fuel. Also, the in-service and transient thermal and mechanical performance of homogeneous ThO2-UO2–based fuels with respect to safety is generally equal to or better than that of all-uranium fuel. Furthermore, a mixed thorium-uranium dioxide spent fuel appears to be a much more stable waste form than uranium oxide spent fuel.

Microheterogeneous Thoria-Urania Fuels for Pressurized Water Reactors

A thorium-based fuel cycle for light water reactors will reduce the plutonium generation rate and enhance the proliferation resistance of the spent fuel. However, priming the thorium cycle with 235U is necessary, and the 235U fraction in the uranium must be limited to below 20% to minimize proliferation concerns. Thus, a once-through thorium-uranium dioxide (ThO2-UO2) fuel cycle of no less than 25% uranium becomes necessary for normal pressurized water reactor (PWR) operating cycle lengths. Spatial separation of the uranium and thorium parts of the fuel can improve the achievable burnup of the thorium-uranium fuel designs through more effective breeding of 233U from the 232Th. Focus is on microheterogeneous fuel designs for PWRs, where the spatial separation of the uranium and thorium is on the order of a few millimetres to a few centimetres, including duplex pellet, axially microheterogeneous fuel, and a checkerboard of uranium and thorium pins. A special effort was made to understand the underlying reactor physics mechanisms responsible for enhancing the achievable burnup at spatial separation of the two fuels. The neutron spectral shift was identified as the primary reason for the enhancement of burnup capabilities. Mutual resonance shielding of uranium and thorium is also a factor; however, it is small in magnitude. It is shown that the microheterogeneous fuel can achieve higher burnups, by up to 15%, than the reference all-uranium fuel. However, denaturing of the 233U in the thorium portion of the fuel with small amounts of uranium significantly impairs this enhancement. The denaturing is also necessary to meet conventional PWR thermal limits by improving the power share of the thorium region at the beginning of fuel irradiation. Meeting thermal-hydraulic design requirements by some of the microheterogeneous fuels while still meeting or exceeding the burnup of the all-uranium case is shown to be potentially feasible. However, the large power imbalance between the uranium and thorium regions creates several design challenges, such as higher fission gas release and cladding temperature gradients. A reduction of plutonium generation by a factor of 3 in comparison with all-uranium PWR fuel using the same initial 235U content was estimated. In contrast to homogeneously mixed U-Th fuel, microheterogeneous fuel has a potential for economic performance comparable to the all-UO2 fuel provided that the microheterogeneous fuel incremental manufacturing costs are negligibly small.

Xenon Diffusivity in Thoria-Urania Fuel

Postirradiation annealing tests were performed to obtain the 133Xe diffusion coefficients in uranium dioxide (UO2) and mixed thorium-uranium dioxide [(Th-U)O2] fuels. Specimens were a single-grained UO2, a polycrystalline UO2, and a polycrystalline (Th-U)O2. The (Th-U)O2 specimen was a mixture of 35% ThO2 and 65% UO2. Each 300-mg specimen was irradiated to a burnup of 0.1 MWd/t U. Postirradiation annealing tests were performed at 1400, 1500, and 1600°C, continuously. The xenon diffusion coefficients for the nearly stoichiometric single-grained UO2 agree well with the data of others. The xenon diffusion coefficients in the polycrystalline (Th-U)O2 are approximately one order lower than those in the polycrystalline UO2. The xenon diffusion coefficient in the (Th-U)O2 increases with the increasing oxygen potential of the ambient gas.

ThO 2–UO 2 annular pins for high burnup fuels

The main purpose of this work is to investigate the use of annular fuel pins (particularly pins containing thorium dioxide) for high burnup fuel. The following parameters were evaluated and compared between postulated mixed thorium–uranium dioxide standard and annular (9% void fraction) type fuel assemblies, as a function of burnup: the infinite multiplication factor, the uranium and plutonium isotopic compositions, the fuel temperature coefficient of reactivity and the conversion ratio. We used the SCALE-4.3 code system. The calculation method consisted in obtaining actinide and fission product number densities as functions of assembly burnup, by means of a 1-D transport calculation combined with a 0-D burnup calculation. These number densities were then used in a 3-D Monte Carlo code for obtaining k ∞ from two-dimensional-symmetry “snapshots”.

Dynamic ion exchange chromatography for determination of number of fissions in thorium-uranium dioxide fuels

High-performance liquid chromatography and the use of /sup 138/La as a fission monitor have been examined for the determination of the number of fissions and burnup in (Th, U)O/sub 2/ fuels. The fission product /sup 139/La in a solution of irradiated fuel was separated on a reversed phase dynamically modified with 1-octanesulfonate, and the eluted metal ions were monitored by an ''on-line'' postcolumn reaction system. The relative standard deviation of the peak heights for lanthanum in 13 fuel solutions over a 1 month period was 0.96%, and the agreement between the HPLC burnup results and those obtained by standard mass spectrometric techniques was within 0.1%; the largest source of uncertainty for the determination of the number of fissions was that associated with fission yield values. Other uses of this chromatographic system were also examined briefly. 24 references, 3 figures, 1 table.

Separation and mass spectrometry of nanogram quantities of uranium and thorium from thorium-uranium dioxide fuels

A microchemical procedure was developed for the separation and isotopic analysis of U and Th from irradiated (Th,U)O/sub 2/ fuel. The separation procedure consisted of two stages; in the first a tributyl phosphate impregnated resin bead was equilibrated with the dissolved fuel in 0.08 M HF/6 M HNO/sub 3/ solution. The bead sorbed approximately 1.7 ..mu..g of U and 4.8..mu..g of Th and provided good separation of these from the fission products. In the second stage, the U and Th were back-extracted into 0.025 M HF/8 M HNO/sub 3/ solution, which contained a small anion-exchange membrane disk. The disk adsorbed approximately 14 ng of U and 45 ng of Th, and subsequently was transferred to the ionizing filament of a thermal-ionization mass spectrometer and covered with a starch deposit. Sensitivities were sufficiently high for sequential analysis of these quantities of Th and U from a single disk. Isotopic data obtained for a combined U and Th standard showed excellent agreement with certified values: overall bias and precision were < 0.03% and 0.2% relative standard deviation, respectively, for both elements. The applicability of the procedure to uranium fuels was also demonstrated. 6 figures, 2 tables.