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All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

AN INITIAL ASSESSMENT OF POTENTIAL PRODUCTION TECHNOLOGIES FOR EPSILON-METAL WASTE FORMS

This report examines and ranks a total of seven materials processing techniques that may be potentially utilized to consolidate the undissolved solids from nuclear fuel reprocessing into a low-surface area form. Commercial vendors of processing equipment were contacted and literature researched to gather information for this report. Typical equipment and their operation, corresponding to each of the seven techniques, are described in the report based upon the discussions and information provided by the vendors. Although the report does not purport to describe all the capabilities and issues of various consolidation techniques, it is anticipated that this report will serve as a guide by highlighting the key advantages and disadvantages of these techniques. The processing techniques described in this report were broadly classified into those that employed melting and solidification, and those in which the consolidation takes place in the solid-state. Four additional techniques were examined that were deemed impractical, but were included for completeness. The techniques were ranked based on criteria such as flexibility in accepting wide-variety of feed-stock (chemistry, form, and quantity), ease of long-term maintenance, hot cell space requirements, generation of additional waste streams, cost, and any special considerations. Based on the assumption of ~2.5 L of waste to be consolidated per day, sintering based techniques, namely, microwave sintering, spark plasma sintering and hot isostatic pressing, were ranked as the top-3 choices, respectively. Melting and solidification based techniques were ranked lower on account of generation of volatile phases and difficulties associated with reactivity and containment of the molten metal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Spent Nuclear Fuel Disposition

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNF is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Multiscale models of nuclear waste reprocessing: from the mesoscale to the plant-scale

Nuclear waste reprocessing and nonproliferation models are needed to support the renaissance in nuclear energy. This report summarizes an LDRD project to develop predictive capabilities to aid the next-generation nuclear fuel reprocessing, in SIERRA Mechanics, Sandia’s high performance computing multiphysics code suite and Cantera, an open source software product for thermodynamics and kinetic modeling.

Rao, Rekha Ranjana↗
DOE OSTI2012

Nuclear Fuels: Promise and Limitations

From 1950 through 1980, scientists, engineers and national leaders confidently predicted an early twenty-first century where fast breeder reactors and commercial nuclear fuel reprocessing were commonplace. Such a scenario seemed necessary for a world with the more than 1000 GWe of nuclear energy needed to meet such an ever-increasing thirst for energy. Thirty years later uranium reserves are increasing on pace with consumption, the growth of nuclear power has been slowed, commercial breeder reactors have yet to enter the marketplace, and less than a handful of commercial reprocessing plants operate. As Nobel Laureate Niels Bohr famously said, “Prediction is very difficult, especially if it’s about the future.” The programme for IChemE’s 2012 conference on the nuclear fuel cycle features a graphic of an idealized nuclear fuel cycle that symbolizes the quest for a closed nuclear fuel cycle featuring careful husbanding of precious resources while minimizing the waste footprint. Progress toward achieving this ideal has been disrupted by technology innovations in the mining and petrochemical industries, as well as within the nuclear industry.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Site remediation techniques in India: a review

India is one of the developing countries operating site remediation techniques for the entire nuclear fuel cycle waste for the last three decades. In this paper we intend to provide an overview of remediation methods currently utilized at various hazardous waste sites in India, their advantages and disadvantages. Over the years the site remediation techniques have been well characterized and different processes for treatment, conditioning and disposal are being practiced. Remediation Methods categorized as biological, chemical or physical are summarized for contaminated soils and environmental waters. This paper covers the site remediation techniques implemented for treatment and conditioning of wastelands arising from the operation of nuclear power plant, research reactors and fuel reprocessing units. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Sharp Interface Tracking in Rotating Microflows of Solvent Extraction

The objective of this project is to develop a specialized sharp interface tracking simulation capability for predicting interaction of micron-sized drops and bubbles in rotating flows relevant to optimized design of contactor devices used in solvent extraction processes of spent nuclear fuel reprocessing. The primary outcomes of this project include the capability to resolve drops and bubbles micro-hydrodynamics in solvent extraction contactors, determining from first principles continuum fluid mechanics how micro-drops and bubbles interact with each other and the surrounding shearing fluid for realistic flows. In the near term, this effort will play a central role in providing parameters and insight into the flow dynamics of models that average over coarser scales, say at the millimeter unit length. In the longer term, it will prove to be the platform to conduct full-device, detailed simulations as parallel computing power reaches the exaflop level. The team will develop an accurate simulation tool for flows containing interacting droplets and bubbles with sharp interfaces under conditions that mimic those found in realistic contactor operations. The main objective is to create an off-line simulation capability to model drop and bubble interactions in a domain representative of the averaged length scale. The technical approach is to combine robust interface tracking software, subgrid modeling, validation quality experiments, powerful computational hardware, and a team with simulation modeling, physical modeling and technology integration experience. Simulations will then fully resolve the microflow of drops and bubbles at the microsecond time scale. This approach is computationally intensive but very accurate in treating important coupled physical phenomena in the vicinity of interfaces. The method makes it possible to resolve spatial scales smaller than the typical distance between bubbles and to model some non-equilibrium thermodynamic features such as finite critical tension in cavitating liquids

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Preparation of Waste Fingerprints for the Miscellaneous Beta Gamma Waste Feeds to the Box Encapsulation Plant at Sellafield - 16080

There are a number of legacy pond and silo storage facilities on Sellafield Site that contain a broad range of ILW arising from the early Windscale Pile and Magnox Fuel Reprocessing Programmes. The waste comprises of solid Miscellaneous Beta Gamma Waste (MBGW) from operations (including filters, containerised waste, thermocouple wire, support struts, pumps, soft wastes) and associated sludges that have arisen from the corrosion of the fuel, Magnox Swarf cladding and the MBGW. It is intended to retrieve and package this waste to meet the requirements of interim storage and final disposal in the Box Encapsulation Plant (BEP) at Sellafield. As part of this process each package must be assigned a realistic and justifiable radionuclide and material inventory. These inventories can either be assigned by radiometric measurements including the associated errors or developed by fundamental data underpinning these assessments and a 'chain of custody' approach of knowing where the waste has been and how it has evolved. The subject of this paper is concerned with the development of the waste fingerprints for these legacy waste streams. As the wastes have arisen over a ∼70 year period there have been an assortment of different working practices and developments in equipment design that have led to significant variations in the waste condition and its composition. During storage the waste items have become mixed, corroded and sludge deposits have arisen. Because of this, and that early consignment records are either incomplete or missing, the task of assigning material and radionuclide fingerprints has been very challenging. In addition, to comply with regulatory requirements the inventory assignment has to be sufficiently robust to meet the processing envelope of the encapsulation process i.e. to allow controls to be made on the amounts of reactive materials (e.g. aluminium, Magnox and uranium) that can be processed within each product and allow identification of excluded items from being consigned from the donor plants. Also, as the waste comprises actinides, fission and activation products it was noted that whilst the UK standard inventory code FISPIN10 predicts the actinides and fission products well, there was a significant discrepancy in the prediction of activation products when compared with those predicted by the specialised fusion code FISPACT. It should be noted that the use of FISPACT to predict actinides and fission products is not recommended. To address these issues, - position statements were produced for each of the waste streams that would be consigned to BEP; - gaps within the data were assessed and where possible inactive trial work was identified to address it; - material and radionuclide fingerprint justification documents were developed from the position statements based upon available data, the experimental work and justified assumptions; - where FISPIN runs were required, these were run with a modified activation data library based on the European Activation File 2010 (EAF2010) to model the activation products. This paper summarises the BEP process, discusses examples of how the material and radionuclide waste stream fingerprints have been justified by revisiting the sample data, original consignment records, engineering drawings and recent remotely operated vehicle (ROV) pond surveys and making reasoned assumptions based on operational practices and non-active experimental work. Incorporation of the EAF2010 library within the FISPIN code has offered significant benefits to BEP and indeed other plants processing mixed fission and activated wastes as it allows comparable predictions of activation products compared with that made by FISPACT and for the plant to maintain one data processing stream rather than two if it had to utilise both modelling codes. This work will underpin the inventory submission in support of the Interim Letter of Compliance (ILoC) to Radioactive Waste Management Limited (RWM). The process of examining the original records, plant operation schedules and engineering drawings has enabled removal of some of the pessimisms that have been made in previous assessments. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Sharing Experiences within AREVA D and D Project Portfolio: Four Illustrations - 13049

Over the past ten years, AREVA has performed D and D operations on a wide range of nuclear sites, such as Marcoule and La Hague recycling plants, to Cadarache MOX fuel fabrication plant or Veurey and Annecy metallic Uranium machining plants. Each site is different from the other but some lessons can be shared through this D and D portfolio. In that respect, knowledge management is one of AREVA D and D Technical Department main missions. Four illustrations demonstrate the interest of knowledge share. Waste management is one of the key activities in D and D; It requires a specific characterization methodology, adapted logistics, and optimized waste channels, all of which have been developed over the years by AREVA teams on the site of Marcoule while they are rather new to La Hague, whose main activity remains fuel reprocessing despite the launch of UP2 400 D and D program. The transfer of know how has thus been organized over the past two years. Plasma cutting has been used extensively in Marcoule for years, while prohibited on the site of La Hague following questions raised about the risks associated wit Ruthenium sublimation. La Hague Technical Department has thus developed an experimental protocol to quantify and contain the Ruthenium risk, the result of which will then be applied to Marcoule where the Ruthenium issue has appeared in recent operations. Commissioning and operating fission products evaporators is a rather standard activity on UP2 800 and UP3, while the associated experience has been decreasing in Marcoule following final shutdown in 1998. When the French atomic Energy commission decided to build and operate a new evaporator to concentrate rinsing effluents prior to vitrification in 2009, AREVA La Hague operators were mobilized to test and commission the new equipment, and train local operators. Concrete scabbling is the final stage prior to the free release of a nuclear facility. In the context of Veurey and Annecy final cleanup and declassification, large scale concrete scabbling operations were conducted, and lead to the industrialization of the process and qualification of a new process, NiThrow{sup TM} scabbling technology, developed by AREVA. This experience has now been injected into La Hague D and D scenario and has allowed a significant gain in time and cost for scabbling operations. In short, the variety of experiences and sites under the responsibility of AREVA D and D teams present significant challenges, and yet provide a unique opportunity to innovate and qualify new tools and methods which can then be shared throughout the sites. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Summary of FY-11 Krypton Capture Activities at the Idaho National Laboratory

Dissolution and voloxidation steps performed in the nuclear fuel reprocessing cycle result in the production of radioactive fission products in their respective off-gas streams. Various isotopes of Kr and Xe are included in the composition of the off-gas radioactive constituents. Depending on the fuel cooling times, Kr-85, with a half-life of 10.7 years may need to be captured per EPA regulations for radioactive emissions. This report includes the results of the ongoing investigations into the solid phase adsorption of Kr, specifically using thermal swing operation, as part of the Off-Gas Sigma Team effort. Experimental results produced from sorbent capacity evaluations have been included. Additionally, Kr adsorption isotherms were produced at two test temperatures to assist with modeling efforts. A final task was to define a strategy for legacy samples obtained from historical Kr immobilization experiments performed in the late ‘70’s at the INL. The results reported herein will be included in a compiled final report describing the work completed by the Off-Gas Sigma team members in FY-11.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Electrochemically modulated separations for material accountability measurements

A method for the accurate and timely analysis of accountable materials is critical for safeguards measurements in nuclear fuel reprocessing plants. Non-destructive analysis (NDA) methods, such as gamma spectroscopy, are desirable for their ability to produce near real-time data. However, the high gamma background of the actinides and fission products in spent nuclear fuel limits the use of NDA for real-time online measurements. A simple approach for at-line separation of materials would facilitate the use of at-line detection methods. A promising at-line separation method for plutonium and uranium is electrochemically modulated separations (EMS). Using an electrochemical cell with an anodized glassy carbon electrode, Pu and U oxidation states can be altered by applying an appropriate voltage. Because the affinity of the actinides for the electrode depends on their oxidation states, selective deposition can be turned “on” and “off” with changes in the applied target electrode voltage. A high surface-area cell was designed in house for the separation of Pu from spent nuclear fuel. The cell is shown to capture over 1 µg of material, increasing the likelihood for gamma spectroscopic detection of Pu extracted from dissolver solutions. The large surface area of the electrode also reduces the impact of competitive interferences from some fission products. Flow rates of up to 1 mL min-1 with >50% analyte deposition efficiency are possible, allowing for rapid separations to be effected. Results from the increased surface-area EMS cell are presented, including dilute dissolver solution simulant data.

36 MATERIALS SCIENCE↗
DOE OSTI2011

The Nuclear Energy Advanced Modeling and Simulation Safeguards and Separations Reprocessing Plant Toolkit

This report details the progress made in the development of the Reprocessing Plant Toolkit (RPTk) for the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. RPTk is an ongoing development effort intended to provide users with an extensible, integrated, and scalable software framework for the modeling and simulation of spent nuclear fuel reprocessing plants by enabling the insertion and coupling of user-developed physicochemical modules of variable fidelity. The NEAMS Safeguards and Separations IPSC (SafeSeps) and the Enabling Computational Technologies (ECT) supporting program element have partnered to release an initial version of the RPTk with a focus on software usability and utility. RPTk implements a data flow architecture that is the source of the system's extensibility and scalability. Data flows through physicochemical modules sequentially, with each module importing data, evolving it, and exporting the updated data to the next downstream module. This is accomplished through various architectural abstractions designed to give RPTk true plug-and-play capabilities. A simple application of this architecture, as well as RPTk data flow and evolution, is demonstrated in Section 6 with an application consisting of two coupled physicochemical modules. The remaining sections describe this ongoing work in full, from system vision and design inception to full implementation. Section 3 describes the relevant software development processes used by the RPTk development team. These processes allow the team to manage system complexity and ensure stakeholder satisfaction. This section also details the work done on the RPTk ``black box'' and ``white box'' models, with a special focus on the separation of concerns between the RPTk user interface and application runtime. Section 4 and 5 discuss that application runtime component in more detail, and describe the dependencies, behavior, and rigorous testing of its constituent components.

Enabling Computational Technologies↗
DOE OSTI2019

Advanced Nanomaterials for Nuclear Energy and Nanotechnology

The use of the latest engineered nanomaterials in nuclear energy systems has opened doors for improving the performance and safety of nuclear power. Nuclear nanotechnology (NNT) deals with the use of engineered nanomaterials for future nuclear energy applications. In this paper, the recent progress in research and benefits of using nanomaterials in different areas of nuclear energy like nuclear fuel extraction and fabrication, fission product capturing, creating robust reactor materials, radiation sensing and monitoring, and radioactive waste separation and spent nuclear fuel reprocessing are summarized. Ongoing research around the world in this field is reviewed, including experimental and computational methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

DEVELOPMENT OF CERAMIC WASTE FORMS FOR AN ADVANCED NUCLEAR FUEL CYCLE

A series of ceramic waste forms were developed and characterized for the immobilization of a Cesium/Lanthanide (CS/LN) waste stream anticipated to result from nuclear fuel reprocessing. Simple raw materials, including Al{sub 2}O{sub 3} and TiO{sub 2} were combined with simulated waste components to produce multiphase ceramics containing hollandite-type phases, perovskites (particularly BaTiO{sub 3}), pyrochlores and other minor metal titanate phases. Three fabrication methodologies were used, including melting and crystallizing, pressing and sintering, and Spark Plasma Sintering (SPS), with the intent of studying phase evolution under various sintering conditions. X-Ray Diffraction (XRD) and Scanning Electron Microscopy coupled with Energy Dispersive Spectroscopy (SEM/EDS) results showed that the partitioning of the waste elements in the sintered materials was very similar, despite varying stoichiometry of the phases formed. Identification of excess Al{sub 2}O{sub 3} via XRD and SEM/EDS in the first series of compositions led to a Phase II study, with significantly reduced Al{sub 2}O{sub 3} concentrations and increased waste loadings. The Phase II compositions generally contained a reduced amount of unreacted Al{sub 2}O{sub 3} as identified by XRD. Chemical composition measurements showed no significant issues with meeting the target compositions. However, volatilization of Cs and Mo was identified, particularly during melting, since sintering of the pressed pellets and SPS were performed at lower temperatures. Partitioning of some of the waste components was difficult to determine via XRD. SEM/EDS mapping showed that those elements, which were generally present in small concentrations, were well distributed throughout the waste forms.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Statistical Approaches for Pebble Bed Reactor Operations and Safeguards

The design of pebble bed reactors (PBRs) and their method of operation align more closely with statistical approaches used in manufacturing and process control than traditional safeguards statistical approaches. The reason is PBRs will employ a nondestructive assay (NDA) measurement (burnup measurement system [BUMS]) that is part of the fuel handling system supporting discharge decisions in addition to the reactor code which monitor performance. The integration of these two approaches provides the opportunity to monitor reactor performance statistically for both operations and safeguards in ways not achievable using other reactor designs. For light water reactors (LWRs), knowledge about reactor code performance in predicting irradiated special nuclear material (SNM) content historically was only achieved from special measurement campaigns or from fuel reprocessing. Conversely, through statistical comparison of the BUMS with the reactor code-predicted values, PBRs can achieve this in real time. The resulting SNM distribution is an indicator of reactor performance because factors such as transit time and path of the pebble fuel through the reactor determine the plutonium production and uranium depletion. By analyzing the predicted and measured values, opportunities exist to adjust operating parameters, fuel design, and other characteristics to optimize performance and fuel utilization. From a safeguards perspective, this approach also provides the information necessary to validate declared values and evaluate whether the reactor is being operated as expected. This paper outlines statistical approaches for PBRs that can be used to support both operations and safeguards.

Ball, Cory↗
DOE OSTI2018

Treatment of Problematic Reactive Metal Wastes Using the GeoMelt{sup R} In-Container Vitrification (ICV{sup TM}) Process - 18639

Decommissioning of sodium cooled reactors and fast reactor technologies have generated a number of reactive metal waste configurations that are problematic to treat and typically lack cost effective treatment methods and disposition options. As a result, Veolia Nuclear Solutions, under contract with Idaho National Laboratory (owned by the U.S. Department of Energy and managed and operated by Battelle Energy Alliance, LLC) demonstrated its GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} technology to safely convert sodium metal to a non-reactive vitrified oxide form. The demonstration project, supported by glass formulation and crucible testing, consisted of a series of ICV{sup TM} melts that processed elemental sodium into stable non-reactive glass. Battelle Energy Alliance, LLC is evaluating the GeoMelt{sup R} technology as a means to safely and reliably convert radioactive reactive metal residues that contaminate components from sodium cooled reactors into waste forms that comply with existing disposition pathways. Reactive metal wastes at INL require treatment in order to remove the Resource Conservation and Recovery Act (RCRA) reactivity and ignitability characteristics in order to comply with land disposal restrictions. GeoMelt{sup R}, which is an alternative to other potential treatment approaches, provides a robust approach that chemically converts the reactive metals to an inert oxide while also immobilizing radionuclides in a vitrified waste form with durability equal to or better than vitrified nuclear fuel reprocessing wastes (very robust and inert waste forms). Most other treatment approaches generate hydrogen gas which is problematic. In 2016, Veolia Nuclear Solutions demonstrated the effectiveness of the GeoMelt{sup R} ICV{sup TM} process in deactivating reactive sodium metal. Crucible, bench-scale, and engineering-scale demonstrations were conducted on several surrogate waste configurations with various ratios of sodium metal and glassformers. Each ratio and configuration demonstrated complete deactivation of the surrogate sodium metal. Follow-on work in 2017 demonstrated the deactivation of reactive sodium by GeoMelt{sup R} ICV{sup TM} at a higher waste loading relative to previously demonstrated work performed in 2016. The higher waste loading optimized glass chemistry while enhancing the economical full-scale treatment of reactive metals. The GeoMelt{sup R} technology is a proven radioactive waste treatment technology capable of immobilizing radioactive wastes, including bulk rubble such as drums and other steel vessels usually without pretreatment. Utilizing the GeoMelt{sup R} technology to treat reactive metals eliminates pretreatment steps resulting from having to separate the reactive metal from steel containers or jackets as GeoMelt{sup R} can easily operate at temperatures sufficient to melt the steel and expose the reactive metal for treatment. Eliminating handling steps of reactive metals is a significant safety advantage since reactive metals are pyrophoric. The results of this two-year demonstration program and future planned activities are presented in the paper. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Recent Advances in Cladding Material Extraction from Fuels in Nuclear Fuel Cycles

An improved recycling and recovery process for the cladding material from spent nuclear fuels is very important toward confirming nuclear energy to support ongoing sustainable development of nuclear management by reducing waste and conserving resources. Nuclear spent fuel cladding materials such as zirconium alloys have economic values and can be recovered, and their recovery eliminates problems in waste disposal and conserves valuable resources. Over 110 published reports and journal articles are reviewed and summarized herein, with a main focus on documenting recovery techniques used to recover cladding materials from spent nuclear fuel and recent developments. Several recovery techniques which are used at present times, such as mechanical separation, chemical dissolution, and hydrometallurgical processes have been covered with examples and discussions. Difficulties within the recovery process are also discussed, and most probable areas for future research in improving efficiency and sustainability of recovering cladding material are identified and discussed at the end. Here, this review could be an important document to the field of spent nuclear fuel reprocessing, recovering valuables and thereby offering guidance on how to effectively manage, safely handle, and reduce nuclear waste. In addition to reducing the volume and radiotoxicity of high-level waste, this review also highlights the potential economic benefit of recovering zirconium from spent fuel cladding by relating typical zirconium metal prices to the mass of cladding per tonne of spent fuel, illustrating that the recoverable material value is non-negligible compared with back-end fuel-cycle costs.

Mondal, Kunal [Oak Ridge National Laboratory (ORNL↗
DOE OSTI2013

MSFR TRU-burning potential and comparison with an SFR

The objective of this work is to evaluate the Molten Salt Fast Reactor (MSFR) potential benefits in terms of transuranics (TRU) burning through a comparative analysis with a sodium-cooled FR. The comparison is based on TRU- and MA-burning rates, as well as on the in-core evolution of radiotoxicity and decay heat. Solubility issues limit the TRU-burning rate to 1/3 that achievable in traditional low-CR FRs (low-Conversion-Ratio Fast Reactors). The softer spectrum also determines notable radiotoxicity and decay heat of the equilibrium actinide inventory. On the other hand, the liquid fuel suggests the possibility of using a Pu-free feed composed only of Th and MA (Minor Actinides), thus maximizing the MA burning rate. This is generally not possible in traditional low-CR FRs due to safety deterioration and decay heat of reprocessed fuel. In addition, the high specific power and the lack of out-of-core cooling times foster a quick transition toward equilibrium, which improves the MSFR capability to burn an initial fissile loading, and makes the MSFR a promising system for a quick (i.e., in a reactor lifetime) transition from the current U-based fuel cycle to a novel closed Th cycle. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗