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[en] Literature on the thorium fuel cycle spanning eight decades from the 1940s to the 2010s is identified, categorized, and analyzed. The publications are divided among twelve topical categories, and overall thorium literature trends are evaluated using database analysis techniques. In total, 1449 publications are identified, with the most prevalent topics being Reprocessing and Waste Management, Molten Salt Reactors, Fuels, and Light Water Reactors. In aggregate, reactor-oriented categories (five in all) comprise 45.5% of publications. The US is the most dominant thorium-publishing nation with 916 publications, followed next by India with 82 and then eight other countries having 25 publications or more. National laboratories have contributed 45% of thorium publications, with roughly equal shares of the balance split between government agencies, universities, and corporations/companies. Oak Ridge National Laboratory in the US accounts for more than a quarter of all publications. Specialized criteria are developed and applied to identify some of the most important, or “keystone”, publications for each category. Across the different categories, and for the study of thorium fuel cycles overall, published research reached an intermediate peak in the 1970s followed by a sharp decline in the 1980s and 1990s; however, interest has been revived moving into the 21st century.
[en] Highlights: •Thorium feasibility as an alternative fuel for the Gas-cooled Fast Reactor has been demonstrated. •Thorium core shows acceptable neutronic and safety characteristics. •The depressurization and expansion reactivity effects show improvement. •GFR2400 thorium-based core can work in both open and closed cycles. •GFR2400 thorium-based core can recycle its own MA vector and plutonium of LWRs. •Safety-related parameters of thorium core are degraded in equilibrium cycle. -- Abstract: In this paper thorium fuel feasibility in large scale Gas Cooled Fast Reactor (GCFR) is investigated. The neutronics benchmark used in this study, GFR2400, corresponds to a 2400 MWth GFR concept proposed by the French CEA. MCNPX computational code is used to design a 3D heterogeneous model of the GFR2400 core. A detailed feasibility analysis of the performance of thorium fuel cycle is performed by using thorium as an alternative fertile fuel for the natural uranium vector of the reference core design. The most essential neutronic parameters characterizing the core are determined both for beginning of life (BOL) conditions as well as during burnup. Also, a three-dimensional core cycle-by-cycle simulations is performed to allow explicit characterization of the core behavior and safety-related parameters during both open and equilibrium cycles. The thorium-based core shows favorable neutronic characteristics with an acceptable control and safety parameters for both BOL and open cycle states. The depressurization reactivity effect and core expansion coefficients (axial and radial) show improvement compared to the uranium-based core. However, this improvement is compensated by the deterioration in the effective delayed neutron fraction (β-eff) and the Doppler reactivity Effect. The results of isotopic transmutation and fuel burnup confirm the capability of the core to work in both open and closed cycles and to self-recycle its own MA vector and plutonium of LWRs. However-as is the case in other fast reactors including the uranium-based GFR2400 core, the fuel cycle closure causes safety related parameters to degrade.
[en] The molten salt reactor (MSR) is an attractive breeder reactor. A graphite-moderated MSR can reach breeding because of the online salt processing and refueling. These features give difficulties when the breeding gain (BG) of the MSR is evaluated. The inventory of the core and external stockpiles have to be treated separately in order to quantify the breeding performance of the reactor. In this paper, an improved BG definition is given and it is compared with definitions used earlier. The improved definition was used in an optimization study of the graphite - salt lattice of the core. The aim of the optimization is a passively safe, self-breeder reactor. The fuel channel diameter, graphite volume and thorium concentration were varied while the temperature feedback coefficient of the core, BG - as defined in the paper - and the lifetime of the graphite were calculated. There is a small range of lattices which provide both negative temperature feedback and breeding. Furthermore, breeding is possible only at low power densities in case of the salt processing efficiencies set in this study. In this range of power the lifetime of the graphite is between 12 and 20 years.
[en] Highlights: • Recycle pathways for uranium microspheres. • Evaluation of kinetics and impurity levels. • Impact of impurity levels on the amounts of uranium carbide and uranium dicarbide in the recycled uranium fuel kernels. - Abstract: During the production of uranium oxide microspheres with carbon using internal gelation, a small fraction of microspheres will not meet the required size or sphericity specification. The uranium microspheres with carbon can be rejected after they have been air-dried or converted into uranium carbide and uranium oxide (UCO) kernels. The next step for the rejected spheres was an air oxidation for carbon removal. The air-dried spheres became triuranium octoxide (U3O8) spheres, which were sometimes ground into powder. The UCO kernels became U3O8 powder during the air oxidation. The next recycle step was nitric acid dissolution of the U3O8 spheres or powders to produce acid-deficient uranyl nitrate (ADUN) solutions, which were used to make new uranium oxide microspheres with carbon and subsequently UCO kernels. The kinetics of the acid dissolution process were compared, and the impurity levels in the different ADUN solutions were determined. X-ray diffraction results for the UCO kernels from the initial and recycled ADUN solutions indicate that changes in the impurity levels can impact the uranium carbide to uranium dicarbide ratio in the UCO kernels.
[en] Highlights: • Systematic overview of the MSR advantages and features from reactor physics point of view and from the safety point of view. • Study of the iso-breeding closed Th-cycle parameters as a function of the fuel-to-moderator ratio for single-fluid MSR core. • Analysis of the pump-driven and pump initiated transients in the moderated MSR. - Abstract: Nuclear reactors operated with liquid fuel may have several remarkable advantages and features. The most developed reactor system in this category is the Molten Salt Reactor. It represents an old concept, but its properties are qualifying it for the advanced utilization: inherent safety, excellent neutron economy, continuous or batch reprocessing possibility without fuel fabrication. The focus has currently moved from the graphite moderated MSR studied in the past towards the fast MSR. The aim of this study is to characterize the MSR physics, highlighting its unique fuel cycle advantages using ERANOS-based EQL3D procedure and investigating its specific dynamics features by the dedicated DYN3D-MSR code
[en] Highlights: ► Metric framework determined to compare nuclear fuel cycles. ► Fast and thermal reactors simulated using MATLAB models, including thorium. ► Modelling uses deterministic methods instead of Monte–Carlo for speed. ► Method rapidly identifies relative cycle strengths and weaknesses. ► Significant scope for use in project planning and cycle optimisation. - Abstract: One of the greatest obstacles facing the nuclear industry is that of sustainability, both in terms of the finite reserves of uranium ore and the production of highly radiotoxic spent fuel which presents proliferation and environmental hazards. Alternative nuclear technologies have been suggested as a means of delivering enhanced sustainability with proposals including fast reactors, the use of thorium fuel and tiered fuel cycles. The debate as to which is the most appropriate technology continues, with each fuel system and reactor type delivering specific advantages and disadvantages which can be difficult to compare fairly. This paper demonstrates a framework of performance metrics which, coupled with a first-order lumped reactor model to determine nuclide population balances, can be used to quantify the aforementioned pros and cons for a range of different fuel and reactor combinations. The framework includes metrics such as fuel efficiency, spent fuel toxicity and proliferation resistance, and relative cycle performance is analysed through parallel coordinate plots, yielding a quantitative comparison of disparate cycles.
[en] The development of accelerator-driven sub-critical reactors operating with pure and enriched thorium fuel mixtures has been heralded as delivering a new era in sustainable nuclear power production. Many benefits have been claimed for these systems, particularly with respect to their ability to consume existing plutonium stockpiles and their inability to breed additional plutonium. This paper examines the operation of fast thorium reactors using a lumped model that can demonstrate to first-order accuracy the principles of actinide evolution and equilibrium and allow the identification of trends within the nuclide transformations. The fundamental mechanisms that affect nuclide evolution are demonstrated and the freedoms and constraints bounding this process are shown. Fast reactors operating with a 100% thorium fuel source are shown to generate plutonium and to offer no advantage over enriched thorium fuel in terms of actinide generation in longer-term operation.
[en] Highlights: • Once-through and closed fuel cycles are compared regarding geological disposal. • Amounts and decay heat rates of waste types are determined for each cycle. • Electricity produced per m2 of disposal area is calculated for each cycle. • Disposal-Area Advantage Factor (DAAF) is used for comparing cycles. • Times to decay to NU ingestion-toxicity level are estimated and compared for cycles. - Abstract: The aim of this study is to assess the impact of closing the nuclear fuel cycle on geological disposal of resultant spent fuels and high level wastes. Once-through and closed fuel cycles with different back-end scenarios are compared with regard to geological disposal. The comparison is carried out in terms of two significant parameters for permanent geological repository: disposal densities and long term radiotoxicity indices of wastes generated from fuel cycles. In the first part of the study, fuel cycles are identified and waste characteristics (amount, composition and thermal output) for the burnup values of 33,000, 40,000 and 50,000 MWd/t are estimated by using MONTEBURNS code. Then, areas needed for waste types under consideration are determined by thermal analysis carried out in ANSYS code for a reference repository concept. The second part of the study consists of an assessment of the radiotoxicities (namely ingestion-toxicity indices) of the wastes generated from the fuel cycles considered. According to the results of the disposal density analysis the once-through cycle is the most advantageous one at low burnups. However, at burnups higher than 40,000 MWd/t, the closed cycle with the standard reprocessing is better than the once-through cycle and the other closed fuel cycles. According to the results of the radiotoxicity analysis, the closed cycle with MOX recycling is more advantageous than the once-through cycle and the other closed cycles for the whole burnup range studied.
[en] Highlights: • The limit on the reprocessing is due to the redox potential control. • Alkali and Earth-alkaline elements do not have to be extracted. • Criticality risks have to be studied in the reprocessing unit. • The neutronics properties are not sensitive to chemical data. • The reprocessing chemistry, from a pure numerical point of view, is an issue. - Abstract: Molten Salt Reactors (MSRs) are liquid-fuel reactors, in which the fuel is also the coolant and flows through the core. A particular configuration presented in this paper called the Molten Salt Fast Reactor consists in a Molten Salt Reactor with no moderator inside the core and a salt composition that leads to a fast neutron spectrum. Previous studies showed that this concept (previously called Thorium Molten Salt Reactor – Nonmoderated) has very promising characteristics. The liquid fuel implies a special reprocessing. Each day a small amount of the fuel salt is extracted from the core for on-site reprocessing. To study such a reactor, the materials evolution within the core has to be coupled to the reprocessing unit, since the latter cleans the salt quasi continuously and feeds the reactor. This paper details the issues associated to the numerical coupling of the core and the reprocessing. It presents how the chemistry is introduced inside the classical Bateman equation (evolution of nuclei within a neutron flux) in order to carry a numerical coupled study. To achieve this goal, the chemistry has to be modeled numerically and integrated to the equations of evolution. This paper presents how is it possible to describe the whole concept (reactor + reprocessing unit) by a system of equations that can be numerically solved. Our program is a connection between MCNP and a homemade evolution code called REM. Thanks to this tool; constraints on the fuel reprocessing were identified. Limits are specified to preserve the good neutronics properties of the MSFR. In this paper, we show that the limit rate for the reprocessing is 2.5 l of fuel salt a day, which means that the fuel should be reprocessed within 7000 days approximately if there is a specific way to control the redox potential of the salt. Finally, a last part of this paper analyzes the impact of chemical parameter uncertainties on the reprocessing performance