Transmutation Fuels Research Articles

Characterization of minor actinide mixed oxide fuel

The global nuclear energy partnership (GNEP) was created in order for ‘fuel-cycle supplier’ nations to provide assured supplies of nuclear fuel to ‘fuel-cycle customer’ nations. The customer nations would utilize the fuel for electricity generation and subsequently return it to the supplier nation after it is spent. This spent fuel would then be reprocessed by the supplier nation in order to recycle the actinide constituents, mainly uranium and plutonium, in advanced nuclear power reactors, and thus reduce waste volumes [1,2]. The International Atomic Energy Agency would control the nuclear materials. One of the thrust areas for the GNEP program is the development of these actinide bearing fuels for transmutation in a fast reactor.

The development of metallic nuclear fuels for transmutation applications: Materials challenges

Metallic alloys show great potential to serve as transmutation fuels that could be used to burn long-lived and high-heat-producing minor actinides and fission products in nuclear reactors as part of the Global Nuclear Energy Partnership program. To implement these fuels, work is ongoing to develop fuel fabrication processes, characterize alloy microstructures, measure fuel properties, determine the compatibility of fuel and cladding alloys, and understand the performance of the fuel alloys during irradiation.

Local atomic structure in (Zr 1 − xU x)N

(Zr 1 − x U x )N solid solutions were prepared for EXAFS measurements by a sol–gel route combined with infiltration and carbothermic reduction. The lattice parameter and the more distant coordination shells (Me 2 and Me 3) around the Zr and U atoms follow the Vegard law. In the first coordination shell, the U–N distance also follows the Vegard law. Though the Zr–N bond distance increases with the lattice expansion caused by increasing U content, it remains constant at 232–235 pm in U-rich (Zr 1 − x U x )N ( x > 0.6). The measurements indicate that U accommodates the lattice contraction with increasing Zr content, whereas Zr is able to expand its Zr–N bond only at lower U content. In the composition range of transmutation fuels, (Zr 1 − x U x )N is homogeneous at the local atomic scale.

ICONE15-10617 COMMERICALIZATION OF THE GLOBAL NUCLEAR ENERGY PARTNERSHIP (GNEP)

In February 2006 President Bush announced the Advanced Energy Initiative, which included the Department of Energy's (DOE) Global Nuclear Energy Partnership (GNEP). GNEP has seven broad goals, one of the major elements being to develop and deploy advanced nuclear fuel recycling technology. DOE is contemplating accelerating the deployment of these technologies to achieve the construction of a commercial scale application of these technologies. DOE now defines this approach as "…two simultaneous tracks: (1) deployment of commercial scale facilities for which advanced technologies are available now or in the near future, and (2) further research and development of transmutation fuels technologies." GE believes an integrated technical solution, using existing reactor and fuel reprocessing technologies, is achievable in the near term to accelerate the commercial demonstration of GNEP infrastructure. The concept involves a single, integrated, commercial scale, recycling facility consisting of the Consolidated Fuel Treatment Center (CFTC), capable of processing LWR and fast reactor Spent Nuclear Fuel (SNF) and fabricating Advanced Recycling Reactor (ARR) actinide fuel. The integrated facility would include a fast reactor that uses actinide-bearing fuel to produce electricity. For optimal performance, GE believes this integrated facility should be co-located to eliminate transportation between the CFTC and ARR, and enhance proliferation resistance. This Advanced Recycling Center takes advantage of previous investments by government and industry in fast reactor technology research and development. To allow for commercial acceptance, a prototypical demonstration reactor and associated fuel cycle facility will be constructed, tested, and licensed. Taking advantage of GE's NRC-reviewed modular sodium-cooled PRISM reactor, only a single reactor will be needed and the cost and risk minimized in the initial phase of the program. This paper outlines a process and a schedule to deploy the necessary domestic infrastructure to achieve the goals of GNEP. It is recognized that there are many types of public private partnerships that could demonstrate the technologies to close the fuel cycle. It is important to acknowledge the realities in the marketplace when developing an approach to advance nuclear technologies that can gain widespread commercial acceptance.

Studies on the Feasibility of the LWRs Waste-Thorium In-Core Fuel Cycle in the Gas Turbine-Modular Helium Reactor

The capability to operate on LWRs waste constitutes one of the major benefits of the Gas Turbine-Modular Helium Reactor; in this paper, it has been evaluated the possibility to incinerate the LWRs waste and to simultaneously breed fissile 233U by fertile thorium. Since a mixture of pure 239Pu-thorium has shown a quite poor neutron economy, the LWRs waste-thorium fuel performance has been also tested when plutonium and thorium are allocated in different TRISO particles. More precisely, when fissile and fertile actinides share the same TRISO kernel, the resonance at 0.29 eV of the fission and capture microscopic cross sections of 239Pu diminishes also the absorption rate of fertile 232Th and thus it degrades the breeding process. Consequently, in the present studies, two different types of fuel have been utilized: the Driver Fuel, made of LWRs waste, and the Transmutation Fuel, made of fertile thorium. Since, in the thermal neutron energy range, the microscopic capture cross section of 232Th is about 80-100 times smaller than the fission one of 239Pu, setting thorium in particles with a large kernel and LWRs waste in particles with a small one makes the volume integrated reaction rates better equilibrated. At the light of the above consideration, which drives to load as much thorium as possible, for the Transmutation Fuel they have been selected the JAERI TRISO particles packed 40%; whereas, for the Driver Fuel they have been tested different packing fractions and kernel radii. Since no configuration allowed the reactor to work, the above procedure has been repeated when fertile particles are packed 20%; the latter choice permits over one year of operation, but the build up of 233U represents only a small fraction of the depleted 239Pu. Finally, the previous configuration has been also investigated when the fertile and fissile fuels share the same kernel or when the fertile fuel axially alternates with the fissile one.

Properties of cermet fuels for minor actinides transmutation in ADS

The sub-critical Accelerator Driven System (ADS) is now being considered as a potential means to burn long-lived transuranium nuclides. The preferred fuel for such a fast neutron reactor is uranium-free, highly enriched with plutonium and minor actinides. Requirements for ADS transmuter fuels are linked with the core design and safety parameters, the fuel properties and the ease of reprocessing. This study concerns the properties of metals as matrices, with the particular case of Mo. To improve the neutronic characteristics, enriched molybdenum (Mo-92) is required. To overcome the high enrichment cost, it is proposed to recover the matrix by pellet dissolution, and to recycle it for further use. Irradiation programmes are also planned to examine the in-reactor properties of the material. Based on the current status of the research, the results are promising, but irradiation results are still missing.

Contribution of Phénix to the Development of Transmutation Fuels and Targets in Sodium Fast Breeder Reactors

Studies focusing on different long-lived radioactive waste transmutation scenarios illustrate the relevance of fast breeder reactors (FBRs) vis-à-vis the incineration of minor actinides (MAs) and certain long-lived fission products.This research program evaluates fuels and targets for transmutation, relying mainly on irradiation data from Phénix to experimentally validate and demonstrate the technical feasibility of the envisaged concepts.As regards the homogeneous transmutation of MAs in fast reactors, Phénix clearly demonstrates the good behavior of MA-bearing oxide fuel, at least up to 6.4 at.% of burnup. Similar results on metallic MA-bearing fuels as well as technetium targets will be available very soon. Important knowledge on innovative composite fuels developed for the transmutation of MAs in fast reactors or in accelerator-driven reactors (accelerator-driven systems) is also gained. Inert matrices resistant to neutron and fission product damage have been selected. The role of the microstructure and irradiation conditions on the composite behavior under irradiation is explained.This program also highlights the possibilities of designing and fabricating transmutation targets, obtaining authorization to irradiate these targets in a power reactor—a series of stages to be accomplished in order to demonstrate the technical feasibility of incinerating MAs and technetium in FBRs.

Studies on the Feasibility of the LWRs Waste-Thorium In-Core Fuel Cycle in the Gas Turbine-Modular Helium Reactor

The capability to operate on LWRs waste constitutes one of the major benefits of the Gas Turbine-Modular Helium Reactor; in this paper, it has been evaluated the possibility to incinerate the LWRs waste and to simultaneously breed fissile 233U by fertile thorium. Since a mixture of pure 239Pu-thorium has shown a quite poor neutron economy, the LWRs waste-thorium fuel performance has been also tested when plutonium and thorium are allocated in different TRISO particles. More precisely, when fissile and fertile actinides share the same TRISO kernel, the resonance at 0.29 eV of the fission and capture microscopic cross sections of 239Pu diminishes also the absorption rate of fertile 232Th and thus it degrades the breeding process. Consequently, in the present studies, two different types of fuel have been utilized: the Driver Fuel, made of LWRs waste, and the Transmutation Fuel, made of fertile thorium. Since, in the thermal neutron energy range, the microscopic capture cross section of 232Th is about 80-100 times smaller than the fission one of 239Pu, setting thorium in particles with a large kernel and LWRs waste in particles with a small one makes the volume integrated reaction rates better equilibrated. At the light of the above consideration, which drives to load as much thorium as possible, for the Transmutation Fuel they have been selected the JAERI TRISO particles packed 40%; whereas, for the Driver Fuel they have been tested different packing fractions and kernel radii. Since no configuration allowed the reactor to work, the above procedure has been repeated when fertile particles are packed 20%; the latter choice permits over one year of operation, but the build up of 233U represents only a small fraction of the depleted 239Pu. Finally, the previous configuration has been also investigated when the fertile and fissile fuels share the same kernel or when the fertile fuel axially alternates with the fissile one.

Comparative Studies of ENDF/B-6.8, JEF-2.2 and JENDL-3.2 Data Libraries by Monte Carlo Modeling of High Temperature Reactors on Plutonium Based Fuel Cycles

We performed a numerical comparative analysis of the burnup capability of the Gas Turbine-Modular Helium Reactor (GT-MHR) by the Monte Carlo Continuous Energy Burnup Code (MCB). The MCB code is an extension of MCNP that includes the burnup implementation; it adopts continuous energy cross sections and it evaluates the transmutation trajectories for over 2,400 decaying nuclides. We equipped the MCB code with three different nuclear data libraries: JENDL-3.2, JEF-2.2 and ENDF/B-6.8 processed for temperatures from 300 to 1,800K. The GT-MHR model studied in this paper is fueled by actinides coming from the Light Water Reactors waste, converted into two different types of fuel: Driver Fuel and Transmutation Fuel. The Driver Fuel supplies the fissile nuclides needed to maintain the criticality of the reactor, whereas the Transmutation Fuel depletes non-fissile isotopes and controls reactivity excess. We set the refueling and shuffling period to one year and the in-core fuel residency time to three years. The comparative analysis of the MCB code consists of accuracy and precision studies. In the accuracy studies, we performed the burnup calculation with different nuclear data libraries during the year at which the refueling and shuffling schedule set the equilibrium of the fuel composition. In the precision studies, we repeated the same simulations 20 times with a different pseudorandom number stride and the same nuclear data library.

Key physical parameters and temperature reactivity coefficients of the deep burn modular helium reactor fueled with LWRs waste

We investigated some important neutronic features of the deep burn modular helium reactor (DB-MHR) using the MCNP/MCB codes. Our attention was focused on the neutron flux and its spectrum, capture to fission ratio of 239Pu and the temperature coefficient of fuel and moderator. The DB-MHR is a graphite-moderated helium-cooled reactor proposed by General Atomic to address the need for a fast and efficient incineration of plutonium for non-proliferation purposes as well as the management of light water reactors (LWRs) waste. In fact, recent studies have shown that the use of the DB-MHR coupled to ordinary LWRs would keep constant the world inventory of plutonium for a reactor fleet producing 400 TW e/y. In the present studies, the DB-MHR is loaded with Np–Pu driver fuel (DF) with an isotopic composition corresponding to LWRs spent fuel waste. DF uses fissile isotopes (e.g. 239Pu and 241Pu), previously generated in the LWRs, and maintains criticality conditions in the DB-MHR. After an irradiation of three years, the spent DF is reprocessed and its remaining actinides are manufactured into fresh transmutation fuel (TF). TF mainly contains non-fissile actinides which undergo neutron capture and transmutation during the subsequent three-year irradiation in the DB-MHR. At the same time, TF provides control and negative reactivity feedback to the reactor. After extraction of the spent TF, irradiated for three years, over 94% of 239Pu and 53% of all actinides coming from LWRs waste will have been destroyed in the DB-MHR. In this paper we look at the operation conditions at equilibrium for the DB-MHR and evaluate fluxes and reactivity responses using state of the art 3-D Monte Carlo simulations.

04/00751 Thermophysical properties of inert matrix fuels for actinide transmutation: Ronchi, C. et al. Journal of Nuclear Materials, 2003, 320, (1–2), 54–65

04/00751 Thermophysical properties of inert matrix fuels for actinide transmutation: Ronchi, C. et al. Journal of Nuclear Materials, 2003, 320, (1–2), 54–65

04/01048 Safety aspects of oxide fuels for transmutation and utilization in accelerator driven systems: Maschek W. et al. Journal of Nuclear Material, 2003, 320, (1–2), 147–155

04/01048 Safety aspects of oxide fuels for transmutation and utilization in accelerator driven systems: Maschek W. et al. Journal of Nuclear Material, 2003, 320, (1–2), 147–155

04/00746 Post-irradiation examination of THERMHET composite fuels for transmutation: Noirot, J. et al. Journal of Nuclear Materials, 2003, 320, (1–2), 117–125

04/00746 Post-irradiation examination of THERMHET composite fuels for transmutation: Noirot, J. et al. Journal of Nuclear Materials, 2003, 320, (1–2), 117–125

04/01102 Molecular modelling of transmutation fuels and targets: Petit, T. et al. Journal of Nuclear Materials, 2003, 320, (1–2), 133–137

04/01102 Molecular modelling of transmutation fuels and targets: Petit, T. et al. Journal of Nuclear Materials, 2003, 320, (1–2), 133–137

Comparative Studies of ENDF/B-6.8, JEF-2.2 and JENDL-3.2 Data Libraries by Monte Carlo Modeling of High Temperature Reactors on Plutonium Based Fuel Cycles

We performed a numerical comparative analysis of the burnup capability of the Gas Turbine-Modular Helium Reactor (GT-MHR) by the Monte Carlo Continuous Energy Burnup Code (MCB). The MCB code is an extension of MCNP that includes the burnup implementation; it adopts continuous energy cross sections and it evaluates the transmutation trajectories for over 2,400 decaying nuclides. We equipped the MCB code with three different nuclear data libraries: JENDL-3.2, JEF-2.2 and ENDF/B-6.8 processed for temperatures from 300 to 1,800K. The GT-MHR model studied in this paper is fueled by actinides coming from the Light Water Reactors waste, converted into two different types of fuel: Driver Fuel and Transmutation Fuel. The Driver Fuel supplies the fissile nuclides needed to maintain the criticality of the reactor, whereas the Transmutation Fuel depletes non-fissile isotopes and controls reactivity excess. We set the refueling and shuffling period to one year and the in-core fuel residency time to three years. The co...

The burnup capabilities of the Deep Burn Modular Helium Reactor analyzed by the Monte Carlo Continuous Energy Code MCB

We have investigated the waste actinide burnup capabilities of a Gas Turbine Modular Helium Reactor (GT-MHR, similar to the reactor being designed by General Atomics and Minatom for surplus weapons plutonium destruction) with the Monte Carlo Continuous Energy Burnup Code MCB, an extension of MCNP developed at the Royal Institute of Technology in Stockholm and University of Mining and Metallurgy in Krakow. The GT-MHR is a gas-cooled, graphite-moderated reactor, which can be powered with a wide variety of fuels, like thorium, uranium or plutonium. In the present work, the GT-MHR is fueled with the transuranic actinides contained in Light Water Reactors (LWRs) spent fuel for the purpose of destroying them as completely as possible with minimum reliance on multiple reprocessing steps. After uranium extraction from the LWR spent fuel (UREX), the remaining waste actinides, including plutonium are partitioned into two distinct types of fuel for use in the GT-MHR: Driver Fuel (DF) and Transmutation Fuel (TF). The DF supplies the neutrons to maintain the fission chain reaction, whereas the TF emphasizes neutron capture to induce a deep burn transmutation and provide reactivity control by a negative feedback. When used in this mode, the GT-MHR is called Deep Burn Modular Helium Reactor (DB-MHR). Both fuels are contained in a structure of triple isotropic coated layers, TRISO coating, which has been proven to retain fission products up to 1600 °C and is expected to remain intact for hundreds of thousands of years after irradiation. Other benefits of this reactor consist of: a well-developed technology, both for the graphite-moderated core and the TRISO structure, a high energy conversion efficiency (about 50%), well established passive safety mechanism and a competitive cost. The destruction of more than 94% of 239Pu and the other geologically problematic actinide species makes this reactor a valid proposal for the reduction of nuclear waste and the prevention of proliferation.

The chemistry and physics of modelling nitride fuels for transmutation

Against the background of increased interest in nitride fuel for transmutation of minor actinides, this paper presents the current status of UK modelling of nitride fuel pins. A new equilibrium chemistry model for nitride fuel, suitable for inclusion in fuel pin modelling codes, is described. High-temperature experiments on (U,Zr)N in a sealed capsule indicate that actinide nitrides in a nitrogen atmosphere are stable up to their melting point. The paper also recommends a set of physical and chemical properties for advanced nitride fuels, highlighting the areas where data are presently missing.

Thermophysical properties of inert matrix fuels for actinide transmutation

The thermal transport properties of a set of ‘inert matrix fuels’ (IMFs) for actinide transmutation were investigated and compared. Heat capacity, thermal diffusivity and conductivity were measured in the usable temperature range for MgO- and ZrO 2-based IMFs containing uranium, plutonium and americium, and sintered from co-precipitated or blended oxide powders.

Zirconate pyrochlore as a transmutation target: thermal behaviour and radiation resistance against fission fragment impact

Zirconates with the pyrochlore structure (A 2Zr 2O 7) are investigated at ITU for use as an actinide host in inert matrix fuels for transmutation. Zirconate pyrochlores with A=Nd as an inactive stand in for the trivalent actinides Am and Cm were fabricated and characterised, and their thermal transport properties measured. The low thermal conductivity indicates that zirconate pyrochlore can only be used for transmutation if it is dispersed in a cercer or cermet composite fuel. The temperature profiles of a MgO–An 2Zr 2O 7 composite were calculated using the measured Nd-zirconate thermal conductivity for different concentrations of the included phase. The radiation stability of Nd 2Zr 2O 7 against fission products (FP) was investigated using iodine irradiation (120 MeV). Significant alterations of the implanted regions were observed even at relatively low fluence.

Molecular modelling of transmutation fuels and targets

The stability of point defects and the behaviour of rare gases in uranium dioxide have been studied using electronic structure calculations. Krypton atoms are found to be insoluble in UO 2 whatever the trapping site considered. Their presence induces a swelling of the lattice when they are located in interstitial sites or in oxygen vacancy sites. Due to its smaller atomic size, the predicted helium behaviour is very different. Indeed, helium is found to be soluble in stoichiometric and hyperstoichiometric uranium dioxide in the presence of uranium vacancies or divacancies. Moreover helium atoms induce a lattice contraction except in interstitial sites for which a slight expansion is found. Some preliminary results concerning xenon are also given.