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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: • 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] 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] The rapid increase in the price of uranium and recent increases in the cost of uranium enrichment and prospective costs of fuel reprocessing suggest that long irradiations of fuel and the postponement of fuel reprocessing would give the lowest cost fuel cycles. The flexibility of using two separate fuels appears to leave the cost advantage with such cycles avoiding any reprocessing of plutonium-bearing fuels and extraction of 233U from thorium fuels only after several years of irradiation. The resulting fuel cycle costs in prospect can be low and very attractive. The analysis employs the LATREP code. (author)
[en] Highlights: •SEU + Thorium fuel in PT-HWRs has a number of benefits to the fuel cycle. •Reduced resource consumption by 23% relative to NU fuel. •Reduced number of DGR used fuel containers by up to 60% •Reduced number of dry storage baskets by up to 32% -- Abstract: Thorium-based fuel cycles offer many potential benefits, including greater long-term energy sustainability and improved waste management, relative to uranium-based fuels. The purpose of this study was to analyze the potential impacts associated with deploying thorium-based fuels in Pressure Tube Heavy Water Reactors (PT-HWRs) in a once-through fuel cycle in Canada, and to compare them with the use of conventional Natural Uranium (NU) fuel. This study analyzed a medium-burnup (∼19.1 MWd/kg) Slightly Enriched Uranium-based fuel augmented by small amounts of thorium (SEU + Th) and a high-burnup (∼40.6 MWd/kg) fuel made with Low Enriched Uranium mixed with thorium (LEU/Th). The deployment of the medium-burnup SEU + Th in Canada reduced resource consumption by 23% relative to the low burnup NU fuel. The medium-burnup fuel required 3% to 60% fewer Deep Geological Repository (DGR) Used Fuel Containers (UFCs) relative to the low burnup fuel, depending on the decay time (10–70 years) of the Used Nuclear Fuel (UNF). Extending the decay duration of UNF decreases its decay power per unit mass, and hence the required number of DGR UFCs per mass of UNF, at the expense of requiring more above-ground dry storage capacity.
[en] The appropriate management of radioactive waste arising from the nuclear fuel cycle is considered to be a key issue in the development of future, more sustainable nuclear energy systems. In this context, the partitioning and transmutation of actinides could play an important role through the achievement of very significant reductions in the actinide content and radiotoxicity of the high-level waste requiring geological disposal. The current paper reports on the results of a detailed physics study carried out to compare the pros and cons of alternative strategies for closure of the nuclear fuel cycle. Different long-term 'steady-state' scenarios have been considered, involving the deployment, to varying degrees, of light water reactors (LWRs) and advanced fast-spectrum systems. The same nuclear data and calculation methods have been used throughout, so that a consistent and reliable comparison of the relative performance of the three basic fuel cycle options (once-through, plutonium recycle, and recycling of all actinides) has been made possible. In addition, with transmutation having been considered employing both critical and accelerator-driven fast-spectrum systems, the study has provided an evaluation of the advantages and disadvantages of these two different advanced system types
[en] Highlights: • Neutronics aspects of multirecycling of Pu, Am and Cm in BWRs have been investigated. • A BWR with uranium-based Pu + MA fuels can manage the MA production of 0.8–2 LWRs. • Thorium-based fuels provide higher support ratios of 1.9–2.8. • Thorium-based fuels exhibit better voiding behavior and stronger Doppler feedback. • Total void worth may be kept negative through all the multirecycling. - Abstract: We have investigated neutronics aspects of multirecycling of Pu, Am and Cm in BWRs, employing three uranium and three thorium-supported transmutation fuels. Our results show, that thorium-based cores allow for higher shares of MA in the fuel and thereby higher MA incineration without encountering a positive total void worth at any point of the multirecycling. In the uranium-based configuration the total void worth sets the limit on the MA share around 2.45%. The thorium-based fuels also exhibit a stronger Doppler feedback and somewhat degraded reactivity as compared to uranium-fuels. The alpha-heating in the fuel reaches equilibrium after six cycles, maintaining values of 24–31 W/kgFUEL in the uranium-based configurations and 32–37 W/kgFUEL in the thorium-based configurations. The neutron emission keeps rising through the multirecycling, the maximum value reached in the XV cycle ranges from 1.4 × 106 to 1.7 × 106 n/s/g for uranium fuels and 2 × 106 n/s/g for the thorium-based fuels