WorldWideScience

Sample records for based msr molten-salt

  1. DYN1D-MSR dynamics code for molten salt reactors

    International Nuclear Information System (INIS)

    This paper reports about the DYN1D-MSR code development and dynamics studies of the molten salt reactors (MSR) - one of the 'Generation IV International Forum' concepts. In this forum the graphite-moderated channel type MSR based on the previous Oak Ridge National Laboratory research is considered. The liquid molten salt serves as a fuel and coolant, simultaneously and causes two physical peculiarities: the fission energy is released predominantly directly into the coolant and the delayed neutrons precursors are drifted by the fuel flow. The drift causes the spread of delayed neutrons distribution to the non-core parts of primary circuit and it can lead to a reactivity loss or gain in the case of fuel flow acceleration or deceleration, respectively. Therefore, specific 3D tool based on in house code DYN3D was developed in FZR. The code DYN3D-MSR is based on the solution of two-group neutron diffusion equation by the help of a nodal expansion method and it includes models of delayed neutrons drift and specific MSR heat release distribution. In this paper the development and verification of 1D version DYN1D-MSR of the code is described. The code has been validated with the experimental data gained from the molten salt reactor experiment performed in the Oak Ridge and after the validation it was applied to several typical transients (overcooling of fuel at the core inlet, reactivity insertion, and the fuel pump trip)

  2. Mass balance analysis of Th-233U based MSR (Molten-Salt Reactor) cycle (THORIMS-NES) transferred from present U-Pu based LWRs (Light Water Reactor)

    International Nuclear Information System (INIS)

    Nuclear power can play a substantial role in countering global warming. There are still unsolved problems such as safety, nuclear proliferation, radioactive-waste under using U-Pu system. Transition from U-Pu LWR (Light Water Reactor) system to Th-233U MSR (Molten-Salt Reactor) system has been analysed in view of the utilization of fissile in form of Pu fuel salt applying the simplified FREGAT process to the spent fuel of LWR. AMSB (Accelerator Molten-Salt Breeder) was also applied as a fissile producer. All fissile in spent fuel can be used by Th-U MSR system so as not to remain storage of spent fuel after retirement of LWR system. The maximum capacity of Th-U MSR system will reach to about 20 x 103 GWe. However storage of spent fuel will remain for the case of rapid growth of Th-U MSR system even though the maximum capacity is large enough. AMSB will start operation about 20 years after the beginning of Th-U MSR system but the timing can be greatly advanced with the scenario of LWR system. Th-U MSR system can be implemented by using the fissile material in spent fuel from LWRs. Detailed assessment of other materials, performance of facilities, strategies of non-proliferation will be needed for the future improvement.

  3. Molten salt reactor for sustainable nuclear power - MSR FUJI. Annex XXX

    International Nuclear Information System (INIS)

    The FUJI is a simplified molten salt reactor (MSR) being designed for operation in a closed thorium-uranium (Th-U) fuel cycle. A direct predecessor of the FUJI is the molten salt breeder reactor (MSBR) based on the concept of a 'single-fluid molten fluoride fuel', developed in the Molten-Salt Reactor Programme (MSRP) at Oak Ridge National Laboratory (ORNL), USA, during 1950-1976. This programme has resulted in the development and demonstration of the basic MSR technology, especially through excellent operation of the experimental molten salt reactor MSRE in 1965-1969; it also produced a conceptual design of the MSBR. The MSBR was a Th-U cycle thermal breeder applying continuous chemical processing of fuel in situ and periodic core graphite replacement to improve breeding performance. The FUJI concept was proposed in connection with the philosophy of the thorium molten salt nuclear energy synergetic system (THORIMS-NES). Different from the MSBR, the FUJI is a concept of a simplified molten salt reactor without continuous chemical processing and periodic core graphite replacement, aimed at attaining near-breeder characteristics in a Th-U closed fuel cycle. Since 1985, conceptual designs of the FUJI for several fuel cycle options have been developed; including the Pu-burning version (FUJI-Pu) designed to incinerate Pu and minor actinides (MA) from spent solid U fuel or weapons-grade Pu. It was suggested that a miniFUJI pilot plant of about 7 MW(e) is constructed first; this construction has been suggested on the site of the Russian Federal Institute of Technical Physics in Snezhinsk. A prototype FUJI-Pu and FUJI-233U of 100-300 MW(e) could then follow as the next logical steps

  4. The molten salt reactors (MSR) pyro chemistry and fuel cycle for innovative nuclear systems

    International Nuclear Information System (INIS)

    In the frame of the studies on next generation nuclear systems, and especially for the molten salt reactors and for the integrated fuel cycle (as IFR), the fuel cycle constraints must be taken into account in the preliminary studies of the system to improve the cycle and reactor optimisation. Among the purposes for next generation nuclear systems, sustainability and waste (radio-toxicity and mass) management are important goals. These goals imply reprocessing and recycling strategies. The objectives of this workshop are to present and to share the different strategies and scenarios, the needs based on these scenarios, the experimental facilities available today or in the future and their capabilities, the needs for demonstration. It aims at: identifying the needs for fuel cycle based on solid fuel or liquid fuel, and especially, the on-line reprocessing or clean up for the molten salt reactors; assessing the state-of-the-art on the pyro-chemistry applied to solid fuel and to present the research activities; assessing the state-of-the-art on liquid fuels (or others), and to present the research activities; expressing the R and D programs for pyro-chemistry, molten salt, and also to propose innovative processes; and proposing some joint activities in the frame of GEDEON and PRACTIS programs. This document brings together the transparencies of 18 contributions dealing with: scenario studies with AMSTER concept (Scenarios, MSR, breeders (Th) and burners); fuel cycle for innovative systems; current reprocessing of spent nuclear fuel (SNF) in molten salts (review of pyro-chemistry processes (non nuclear and nuclear)); high temperature NMR spectroscopies in molten salts; reductive extraction of An from molten fluorides (salt - liquid metal extraction); electrochemistry characterisation; characterisation with physical methods - extraction coefficient and kinetics; electrolytic extraction; dissolution-precipitation of plutonium in the eutectic LiCl-KCl (dissolution and

  5. System design description of forced-convection molten-salt corrosion loops MSR-FCL-3 and MSR-FCL-4

    International Nuclear Information System (INIS)

    Molten-salt corrosion loops MSR-FCL-3 and MSR-FCL-4 are high-temperature test facilities designed to evaluate corrosion and mass transfer of modified Hastelloy N alloys for future use in Molten-Salt Breeder Reactors. Salt is circulated by a centrifugal sump pump to evaluate material compatibility with LiF-BeF2-ThF4-UF4 fuel salt at velocities up to 6 m/s (20 fps) and at salt temperatures from 566 to 7050C (1050 to 13000F). The report presents the design description of the various components and systems that make up each corrosion facility, such as the salt pump, corrosion specimens, salt piping, main heaters, salt coolers, salt sampling equipment, and helium cover-gas system, etc. The electrical systems and instrumentation and controls are described, and operational procedures, system limitations, and maintenance philosophy are discussed

  6. Molten Salt Breeder Reactor Analysis Based on Unit Cell Model

    International Nuclear Information System (INIS)

    Contemporary computer codes like the MCNP6 or SCALE are only good for solving a fixed solid fuel reactor. However, due to the molten-salt fuel, MSR analysis needs some functions such as online reprocessing and refueling, and circulating fuel. J. J. Power of Oak Ridge National Laboratory (ORNL) suggested in 2013 a method for simulating the Molten Salt Breeder Reactor (MSBR) with SCALE, which does not support continuous material processing. In order to simulate MSR characteristics, the method proposes dividing a depletion time into short time intervals and batchwise reprocessing and refueling at each step. We are applying this method by using the MCNP6 and PYTHON and NEWT-TRITON-PYTHON and PYTHON code systems to MSBR. This paper contains various parameters to analyze the MSBR unit cell model such as the multiplication factor, breeding ratio, change of amount of fuel, amount of fuel feeding, and neutron flux distribution. The result of MCNP6 and NEWT module in SCALE show some difference in depletion analysis, but it still seems that they can be used to analyze MSBR. Using these two computer code system, it is possible to analyze various parameters for the MSBR unit cells such as the multiplication factor, breeding ratio, amount of material, total feeding, and neutron flux distribution. Furthermore, the two code systems will be able to be used for analyzing other MSR model or whole core models of MSR

  7. Molten Salt Demonstration Transmuter (comparison of new technical problems with old US MSR plans)

    International Nuclear Information System (INIS)

    A Molten Salt Demonstration Transmuter (MSDT) is required to show the operation and design performance for closing the nuclear spent fuel (NSF) cycle for PWR or WWER reactors operated in the once-through cycle (OTC) mode. The remnant waste (fission products only) would be either permanently stored or held for secondary use. The purpose of this proposal is to establish the design basis for the MSDT and compare contemporary knowledge and demands with that from US plans for MS reactors from 1974, because both technologies are very near (Authors)

  8. The molten salt reactors (MSR) pyro chemistry and fuel cycle for innovative nuclear systems; Congres sur les reacteurs a sels fondus (RSF) pyrochimie et cycles des combustibles nucleaires du futur

    Energy Technology Data Exchange (ETDEWEB)

    Brossard, Ph. [GEDEON, Groupement de Recherche CEA CNRS EDF FRAMATOME (France); Garzenne, C.; Mouney, H. [and others

    2002-07-01

    In the frame of the studies on next generation nuclear systems, and especially for the molten salt reactors and for the integrated fuel cycle (as IFR), the fuel cycle constraints must be taken into account in the preliminary studies of the system to improve the cycle and reactor optimisation. Among the purposes for next generation nuclear systems, sustainability and waste (radio-toxicity and mass) management are important goals. These goals imply reprocessing and recycling strategies. The objectives of this workshop are to present and to share the different strategies and scenarios, the needs based on these scenarios, the experimental facilities available today or in the future and their capabilities, the needs for demonstration. It aims at: identifying the needs for fuel cycle based on solid fuel or liquid fuel, and especially, the on-line reprocessing or clean up for the molten salt reactors; assessing the state-of-the-art on the pyro-chemistry applied to solid fuel and to present the research activities; assessing the state-of-the-art on liquid fuels (or others), and to present the research activities; expressing the R and D programs for pyro-chemistry, molten salt, and also to propose innovative processes; and proposing some joint activities in the frame of GEDEON and PRACTIS programs. This document brings together the transparencies of 18 contributions dealing with: scenario studies with AMSTER concept (Scenarios, MSR, breeders (Th) and burners); fuel cycle for innovative systems; current reprocessing of spent nuclear fuel (SNF) in molten salts (review of pyro-chemistry processes (non nuclear and nuclear)); high temperature NMR spectroscopies in molten salts; reductive extraction of An from molten fluorides (salt - liquid metal extraction); electrochemistry characterisation; characterisation with physical methods - extraction coefficient and kinetics; electrolytic extraction; dissolution-precipitation of plutonium in the eutectic LiCl-KCl (dissolution and

  9. Nickel based alloys compatibility with fuel salts for molten salt reactor with thorium and uranium support

    International Nuclear Information System (INIS)

    R and D on molten salt reactors (MSR) in Europe are concentrated now on fast/intermediate spectrum concepts which were recognised as long-term alternative to solid fuelled fast reactors due to their attractive features: strong negative feedback coefficients, easy in-service inspection, and simplified fuel cycle. For high-temperature MSR corrosion of the metallic container alloy in primary circuit is the primary concern. Key problem receiving current attention include surface fissures in Ni-based alloys probably arising from fission product tellurium attack. This paper summarises results of corrosion tests conducted recently to study effect of oxidation state in selected fuel salts on tellurium attack and to develop means of controlling tellurium cracking in the special Ni - based alloys recently developed for large power units: molten salt actinide recycler and transmuter (MOSART) and molten salt fast reactor (MSFR). Tellurium corrosion of Ni-based alloys was tested in the temperature range from 730 deg. C up to 800 deg. C in stressed and unloaded conditions with fuel LiF-BeF2-UF4 and LiF-BeF2-ThF4-UF4 salt mixtures at different [U(IV)]/[U(III)] ratios from 0.7 up to 500. Following Russian and French Ni-based alloys (in mass%): HN80M-VI (Mo-12, Cr-7.6, Nb-1.5), HN80MTY (Mo-13, Cr-6.8, Al-1.1, Ti-0.9), HN80MTW (Mo-9.4, Cr-7.0, Ti-1.7, W-5.5) and EM-721 (W-25.2, Cr-5.7, Ti-0.17) were used for the study in the corrosion facility. The HN80MTY alloy has shown the best resistance against Te cracking and after test mechanical properties. (authors)

  10. Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

    Science.gov (United States)

    Jaradat, Safwan Qasim Mohammad

    Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

  11. Molten salt reactor related research in Switzerland

    International Nuclear Information System (INIS)

    Switzerland represented by the Paul Scherrer Institute (PSI) is a member of the Generation IV International Forum (GIF). In the past, the research at PSI focused mainly on HTR, SFR, and GFR. Currently, a research program was established also for Molten Salt Reactors (MSR). Safety is the key point and main interest of the MSR research at the Nuclear Energy and Safety (NES) department of PSI. However, it cannot be evaluated without knowing the system design, fuel chemistry, salt thermal-hydraulics features, safety and fuel cycle approach, and the relevant material and chemical limits. Accordingly, sufficient knowledge should be acquired in the other individual fields before the safety can be evaluated. The MSR research at NES may be divided into four working packages (WP): WP1: MSR core design and fuel cycle, WP2: MSR fuel behavior at nominal and accidental conditions, WP3: MSR thermal-hydraulics and decay heat removal system, WP4: MSR safety, fuel stream, and relevant limits. The WPs are proposed so that there are research topics which can be independently studied within each of them. The work plan of the four WPs is based on several ongoing or past national and international projects relevant to MSR, where NES/PSI participates. At the current stage, the program focuses on several specific and design independent studies. The safety is the key point and main long-term interest of the MSR research at NES. (author)

  12. FLUENT-based neutronics and thermal-hydraulics coupling calculation for a liquid-fuel molten salt reactor

    International Nuclear Information System (INIS)

    Molten Salt Reactor (MSR) is the only one using liquid fuel in the six candidate reactors of the Generation IV advanced nuclear power systems with expected remarkable advantages in safety, economics, sustainability, and proliferation resistance. The strong coupling between neutronics and thermal-hydraulics due to fuel movement in the liquid-fuel MSRs induces many new challenges in reactor analyses from the perspective of both theoretical models and solution methods. In this study, the multi-group diffusion theory was adopted to deduce the neutronics model for the liquid-fuel MSRs, in which the salt flow effects on the delayed neutron precursor distributions in space were considered particularly. Since the liquid-fuel salt is a Newton fluid, the single-phase thermal hydraulics model for liquid-fuel MSRs was generally established based on the fundamental laws of the mass, momentum and energy conservation equations as used in the computational fluid dynamic (CFD) method. Since the control equations of the neutronic model can be written in the same form of those solved in the CFD softwares, a neutronics and thermal-hydraulics coupling scheme was proposed and a program was developed based on the FLUENT software by using its user-defined functions and subroutines (UDF and UDS). This program was applied to perform the steady state calculation of the molten salt fast reactor (MSFR), and the main results such as the space distributions of the neutron fluxes, delayed neutron precursors, temperatures, velocities were obtained. The results show that the liquid fuel flow influences the delayed neutron precursors significantly, while slightly affects the neutron fluxes. The flow in the MSFR core generates a vortex near the fertile tank leading the maximal temperature to about 1200 K at the centre of the vortex, which will be optimized in the future core design. (author)

  13. A novel fusion power concept based on molten-salt technology

    International Nuclear Information System (INIS)

    This paper discusses modifications to an old concept for using peaceful nuclear explosions to achieve practical fusion power. With this concept, useful energy and materials are obtained by repetitively setting off nuclear explosions in an underground cavity. This proposal, which is based on molten-salt technology, involves two modifications: line the cavity with steel to make it engineerable and predictable rather than relying on an unsupported earthen cavity such as a cavity excavated in a salt dome; and use molten salt rather than steam. More than 70% of the energy released is then absorbed by liquid-salt evaporation, and the pressure to be contained for a given yield can be reduced by a factor of 3 or more. These modifications result in several improvements in the safety and feasibility of the contained fusion concept which includes: the tritium produced, being insoluble in the molten salt, can easily be pumped away and purified when all the vaporized salt condenses, rather than being mixed with steam; the tritium inventory is substantially reduced, effectively reducing the large hazard in case of accidental venting to the atmosphere; and reducing the yield used in the older studied could reduce the cost of the cavity considerably

  14. Thorium fuel cycle technology for molten salt reactor systems

    International Nuclear Information System (INIS)

    Molten Salt Reactor (MSR) is classified as the non-classical nuclear type based on the specific featured coming out from the use of liquid fuel circulating in the MSR primary circuit. Other uniqueness of the reactor type is based on the fact that the primary or fuel circuit of the reactor in operation for a long run. MSR is the only reactor system, which can be operated with thorium fuel within the pure 232Th - 233U fuel cycle with the breeding factor significantly higher than one. It can bring several advantages, mainly in the radioactive waste management, thanks to minimized production higher actinides. With respects to all these facts and features, the fuel cycle aspects of MSR system are quite complicated, especially if the technology shall guarantee all possible advantages of MSR system concurrently with good economy, technological safety and reliability and inevitable proliferation resistance

  15. Depletion analysis on long-term operation of the conceptual Molten Salt Actinide Recycler and Transmuter (MOSART) by using a special sequence based on SCALE6/TRITON

    International Nuclear Information System (INIS)

    Highlights: ► An automatic computation and control sequence has been developed for MSR neutronics and depletion analyses. ► The method was developed based on a series of stepwise SCALE6/TRITON calculations. ► A detailed reexamination of the MOSART operation in 30 years was performed. ► Clean-up scenarios of fission products have a significant impact on the MOSART operation. - Abstract: A special sequence based on SCALE6/TRITON was developed to perform fuel cycle analysis of the Molten Salt Actinide Recycler and Transmuter (MOSART), with emphasis on the simulation of its dynamic refueling and salt reprocessing scheme during long-term operation. MOSART is one of conceptual designs in the molten salt reactor (MSR) category of the Generation-IV systems. This type of reactors is distinguished by the use of liquid fuel circulating in and out of the core, which offers many unique advantages but complicates the modeling and simulation of core behavior using conventional reactor physics codes. The TRITON control module in SCALE6 can perform reliable depletion and decay analysis for many reactor physics applications due to its problem-dependent cross-section processing and rigorous treatment of neutron transport. In order to accommodate a simulation of on-line refueling and reprocessing scenarios, several in-house programs together with a run script were developed to integrate a series of stepwise TRITON calculations; the result greatly facilitates the neutronics analyses of long-term MSR operation. Using this method, a detailed reexamination of the MOSART operation in 30 years was performed to investigate the neutronic characteristics of the core design, the change of fuel salt composition from start-up to equilibrium, the effects of various salt reprocessing scenarios, the performance of actinide transmutation, and the radiotoxicity reduction

  16. Electrochemical studies of calcium chloride-based molten salt systems

    International Nuclear Information System (INIS)

    Conductance and EMF studies of CaCl2-based melts were performed in the temperature range 790--990 C. Conductivity data collected using magnesia tubes and capillaries showed deviations from the data recommended by the National Bureau of Standards. These deviations are attributed to the slow dissolution of magnesia by the CaCl2-CaO melt. Conductivity data for molten CaCl2 using a pyrolytic boron nitride capillary were in reasonable agreement with the recommended data; however, undissolved CaO in CaCl2 may have caused blockage of the pyrolytic boron nitride capillary, resulting in fluctuations in the measured resistance. The utility of the AgCl/Ag reference electrode in CaCl2-AgCl and CaCl2-CaO-AgCl melts, using asbestos diaphragms and Vycor glass as reference half-cell membranes, was also investigated. Nernstian behavior was observed using both types of reference half-cell membranes in CaCl2-AgCl melts. The AgCl/Ag reference electrode also exhibited Nernstian behavior in CaCl2-CaO-AgCl melts using a Vycor reference half-cell membrane and a magnesia crucible. The use of CaCl2 as a solvent is of interest since it is used in plutonium metal purification, as well as various other commercial applications. 97 refs., 33 figs., 13 tabs

  17. Development of a three dimension multi-physics code for molten salt fast reactor

    International Nuclear Information System (INIS)

    Molten Salt Reactor (MSR) was selected as one of the six innovative nuclear reactors by the Generation IV International Forum (GIF). The circulating-fuel in the can-type molten salt fast reactor makes the neutronics and thermo-hydraulics of the reactor strongly coupled and different from that of traditional solid-fuel reactors. In the present paper: a new coupling model is presented that physically describes the inherent relations between the neutron flux, the delayed neutron precursor, the heat transfer and the turbulent flow. Based on the model, integrating nuclear data processing, CAD modeling, structured and unstructured mesh technology, data analysis and visualization application, a three dimension steady state simulation code system (MSR3DS) for the can-type molten salt fast reactor is developed and validated. In order to demonstrate the ability of the code, the three dimension distributions of the velocity, the neutron flux, the delayed neutron precursor and the temperature were obtained for the simplified MOlten Salt Advanced Reactor Transmuter (MOSART) using this code. The results indicate that the MSR3DS code can provide a feasible description of multi-physical coupling phenomena in can-type molten salt fast reactor. Furthermore, the code can well predict the flow effect of fuel salt and the transport effect of the turbulent diffusion. (authors)

  18. Design of a heterogeneous subcritical nuclear reactor with molten salts based on thorium

    International Nuclear Information System (INIS)

    This paper presents the design of a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a 252Cf source, whose dose levels at the periphery allows its use in teaching and research activities. The design was realized by the Monte Carlo method, where the geometry, dimensions and the fuel was varied in order to obtain the best design. The result was a cubic reactor of 110 cm of side, with graphite moderator and reflector. In the central part having 9 ducts of 3 cm in diameter, eight of them are 110 cm long, which were placed on the Y axis; the separation between each duct is 10 cm. The central duct has 60 cm in length and this contains the 252Cf source, also there are two irradiation channels and the other six contain a molten salt (7LiF - BeF2 - ThF4 - UF4) as fuel. For the design the keff was calculated, neutron spectra and ambient dose equivalent. In the first instance the above was calculated for a virgin fuel, was called case 1; then a percentage of 233U was used and the percentage of Th was decreased and was called case 2. This with the purpose of comparing two different fuels operating within the reactor. For the two irradiation ducts three positions are used: center, back and front, in each duct in order to have different flows. (Author)

  19. Hybrid Molten Salt Reactor (HMSR) System Study

    Energy Technology Data Exchange (ETDEWEB)

    Woolley, Robert D [PPPL; Miller, Laurence F [PPPL

    2014-04-01

    Can the hybrid system combination of (1) a critical fission Molten Salt Reactor (MSR) having a thermal spectrum and a high Conversion Ratio (CR) with (2) an external source of high energy neutrons provide an attractive solution to the world's expanding demand for energy? The present study indicates the answer is an emphatic yes.

  20. Materials considerations for molten salt accelerator-based plutonium conversion systems

    Energy Technology Data Exchange (ETDEWEB)

    DeVan, J.H.; DiStefano, J.R.; Eatherly, W.P.; Keiser, J.R.; Klueh, R.L.

    1994-12-31

    A Molten-Salt Reactor Program for power applications was initiated at the Oak Ridge National Laboratory in 1956. In 1965 the Molten Salt Reactor Experiment (MSRE) went critical and was successfully operated for several years. Operation of the MSRE revealed two deficiencies in the Hastelloy N alloy that had been developed specifically for molten-salt systems. The alloy embrittled at elevated temperatures as a result of exposure to thermal neutrons (radiation damage) and grain boundary embrittlement occurred in materials to fuel salt. Intergranular cracking was found to be associated with fission products, viz. tellurium. An improved Hastelloy N composition was subsequently developed that had better resistance to both of these problems. However, the discovery that fission product cracking could be significantly decreased by making the salt sufficiently reducing offers the prospect of improved compatibility with molten salts containing fission products and resistance to radiation damage in ABC applications. Recommendations are made regarding the types of corrosion tests and mechanistic studies needed to qualify materials for operation with PuF{sub 3}-containing molten salts.

  1. Materials considerations for molten salt accelerator-based plutonium conversion systems

    International Nuclear Information System (INIS)

    Accelerator-driven transmutation technology (ADTT) refers to a concept for a system that uses a blanket assembly driven by a source of neutrons produced when high-energy protons from an accelerator strike a heavy metal target. One application for such a system is called Accelerator-Based Plutonium Conversion, or ABC. Currently, the version of this concept being proposed by the Los Alamos National Laboratory features a liquid lead target material and a blanket fuel of molten fluorides that contain plutonium. Thus, the materials to be used in such a system must have, in addition to adequate mechanical strength, corrosion resistance to molten lead, corrosion resistance to molten fluoride salts, and resistance to radiation damage. In this report the corrosion properties of liquid lead and the LiF-BeF2 molten salt system are reviewed in the context of candidate materials for the above application. Background information has been drawn from extensive past studies. The system operating temperature, type of protective environment, and oxidation potential of the salt are shown to be critical design considerations. Factors such as the generation of fission products and transmutation of salt components also significantly affect corrosion behavior, and procedures for inhibiting their effects are discussed. In view of the potential for extreme conditions relative to neutron fluxes and energies that can occur in an ADTT, a knowledge of radiation effects is a most important factor. Present information for potential materials selections is summarized

  2. Corrosion behavior of Ni-based structural materials for electrolytic reduction in lithium molten salt

    Science.gov (United States)

    Cho, Soo Haeng; Park, Sung Bin; Lee, Jong Hyeon; Hur, Jin Mok; Lee, Han Soo

    2011-05-01

    In this study, the corrosion behavior of new Ni-based structural materials was studied for electrolytic reduction after exposure to LiCl-Li 2O molten salt at 650 °C for 24-216 h under an oxidizing atmosphere. The new alloys with Ni, Cr, Al, Si, and Nb as the major components were melted at 1700 °C under an inert atmosphere. The melt was poured into a preheated metallic mold to prepare an as-cast alloy. The corrosion products and fine structures of the corroded specimens were characterized by scanning electron microscope (SEM), Energy Dispersive X-ray Spectroscope (EDS), and X-ray diffraction (XRD). The corrosion products of as cast and heat treated low Si/high Ti alloys were Cr 2O 3, NiCr 2O 4, Ni, NiO, and (Al,Nb,Ti)O 2; those of as cast and heat treated high Si/low Ti alloys were Cr 2O 3, NiCr 2O 4, Ni, and NiO. The corrosion layers of as cast and heat treated low Si/high Ti alloys were continuous and dense. However, those of as cast and heat treated high Si/low Ti alloys were discontinuous and cracked. Heat treated low Si/high Ti alloy showed the highest corrosion resistance among the examined alloys. The superior corrosion resistance of the heat treated low Si/high Ti alloy was attributed to the addition of an appropriate amount of Si, and the metallurgical evaluations were performed systematically.

  3. Corrosion of chosen Ni-based material exposed to LiF-NaF molten salts

    Czech Academy of Sciences Publication Activity Database

    Král, Lubomír; Čermák, Jiří; Matal, O.; Šimo, T.; Nesvadba, L.

    Ostrava : Tanger s.r.o., 2010, s. 596-600. ISBN 978-80-87294-17-8. [Metal 2010. International Conference on Metallurgy and Materials /19./. Rožnov pod Radhoštěm (CZ), 18.05.2010-20.05.2010] R&D Projects: GA MPO 2A-1TP1/067 Institutional research plan: CEZ:AV0Z20410507 Keywords : molten salts * nickel alloys * corrosion Subject RIV: JF - Nuclear Energetics

  4. Molten salt electrolyte separator

    Science.gov (United States)

    Kaun, Thomas D.

    1996-01-01

    A molten salt electrolyte/separator for battery and related electrochemical systems including a molten electrolyte composition and an electrically insulating solid salt dispersed therein, to provide improved performance at higher current densities and alternate designs through ease of fabrication.

  5. Development of pyro-processing technology for thorium-fuelled molten salt reactor

    International Nuclear Information System (INIS)

    The Molten Salt Reactor (MSR) is classified as the non-classical nuclear reactor type based on the specific features coming out from the use of liquid fuel circulating in the MSR primary circuit. Other uniqueness of the reactor type is based on the fact that the primary circuit of the reactor is directly connected with the on-line reprocessing technology, necessary for keeping the reactor in operation for a long run. MSR is the only reactor system, which can be effectively operated within the 232Th- 233U fuel cycle as thorium breeder with the breeding factor significantly higher than one. The fuel cycle technologies proposed as ford the fresh thorium fuel processing as for the primary circuit fuel reprocessing are pyrochemical and mainly fluoride. Although these pyrochemical processes were never previously fully verified, the present-day development anticipates an assumption for the successful future deployment of the thorium-fuelled MSR technology. (authors)

  6. Gases in molten salts

    CERN Document Server

    Tomkins, RPT

    1991-01-01

    This volume contains tabulated collections and critical evaluations of original data for the solubility of gases in molten salts, gathered from chemical literature through to the end of 1989. Within the volume, material is arranged according to the individual gas. The gases include hydrogen halides, inert gases, oxygen, nitrogen, hydrogen, carbon dioxide, water vapor and halogens. The molten salts consist of single salts, binary mixtures and multicomponent systems. Included also, is a special section on the solubility of gases in molten silicate systems, focussing on slags and fluxes.

  7. Recommendations for a restart of Molten Salt Reactor development

    International Nuclear Information System (INIS)

    The concept of the molten salt reactor (MSR) refuses to go away. The Generation-IV process lists the MSR as one of the six concepts to be considered for extending fuel resources. Good fuel utilization and good economics are required to meet the often cited goal of 10 TWe globally and 1 TWe for the US by non-carbon energy sources in this century by nuclear fission. A strong incentive for the molten salt reactor design is its good fuel utilization, good economics, amazing flexibility and promised large benefits. It can: - use thorium or uranium; o be designed with lots of graphite to have a fairly thermal neutron spectrum or without graphite moderator to have a fast neutron spectrum reactor; - fission uranium isotopes and plutonium isotopes; - operate with non-weapon grade fissile fuel, or in suitable sites it can operate with enrichment between reactor-grade and weapon-grade fissile fuel; - be a breeder or near breeder; - operate at temperature >1100 degree C if carbon composites are successfully employed. Enhancing 232U content in the uranium to over 500 pm makes the fuel undesirable for weapons, but it should not detract from its economic use in liquid fuel reactors: a big advantage in nonproliferation. Economics of the MSR is enhanced by operating at low pressure and high temperature and may even lead to the preferred route to hydrogen production. The cost of the electricity produced from low enriched fuel averaged over the life of the entire process, has been predicted to be about 10% lower than that from LWRs, and 20% lower for high enriched fuel, with uncertainties of about 10%. The development cost has been estimated at about 1 B$ (e.g., a 100 M$/y base program for ten years) not including construction of a series of reactors leading up to the deployment of multiple commercial units at an assumed cost of 9 B$ (450 M$/y over 20 years). A benefit of liquid fuel is that smaller power reactors can faithfully test features of larger reactors, thereby reducing the

  8. Summary of the Workshop on Molten Salt Reactor Technologies Commemorating the 50th Anniversary of the Startup of the Molten Salt Reactor Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Betzler, Benjamin R [ORNL; Mays, Gary T [ORNL

    2016-01-01

    A workshop on Molten Salt Reactor (MSR) technologies commemorating the 50th anniversary of the Molten Salt Reactor Experiment (MSRE) was held at Oak Ridge National Laboratory on October 15 16, 2015. The MSRE represented a pioneering experiment that demonstrated an advanced reactor technology: the molten salt eutectic-fueled reactor. A multinational group of more than 130 individuals representing a diverse set of stakeholders gathered to discuss the historical, current, and future technical challenges and paths to deployment of MSR technology. This paper provides a summary of the key messages from this workshop.

  9. Diffusion Welding of Alloys for Molten Salt Service - Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Denis Clark; Ronald Mizia

    2012-05-01

    The present work is concerned with heat exchanger development for molten salt service, including the proposed molten salt reactor (MSR), a homogeneous reactor in which the fuel is dissolved in a circulating fluid of molten salt. It is an outgrowth of recent work done under the Next Generation Nuclear Plant (NGNP) program; what the two reactor systems have in common is an inherently safe nuclear plant with a high outlet temperature that is useful for process heat as well as more conventional generation The NGNP program was tasked with investigating the application of a new generation of nuclear power plants to a variety of energy needs. One baseline reactor design for this program is a high temperature, gas-cooled reactor (HTGR), which provides many options for energy use. These might include the conventional Rankine cycle (steam turbine) generation of electricity, but also other methods: for example, Brayton cycle (gas turbine) electrical generation, and the direct use of the high temperatures characteristic of HTGR output for process heat in the chemical industry. Such process heat is currently generated by burning fossil fuels, and is a major contributor to the carbon footprint of the chemical and petrochemical industries. The HTGR, based on graphite fuel elements, can produce very high output temperatures; ideally, temperatures of 900 C or even greater, which has significant energy advantages. Such temperatures are, of course, at the frontiers of materials limitations, at the upper end of the performance envelope of the metallic materials for which robust construction codes exist, and within the realm of ceramic materials, the fabrication and joining of which, on the scale of large energy systems, are at an earlier stage of development. A considerable amount of work was done in the diffusion welding of materials of interest for HTGR service with alloys such as 617 and 800H. The MSR output temperature is also materials limited, and is projected at about 700 C

  10. Diffusion Welding of Alloys for Molten Salt Service - Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Denis Clark; Ronald Mizia; Piyush Sabharwall

    2012-09-01

    The present work is concerned with heat exchanger development for molten salt service, including the proposed molten salt reactor (MSR), a homogeneous reactor in which the fuel is dissolved in a circulating fluid of molten salt. It is an outgrowth of recent work done under the Next Generation Nuclear Plant (NGNP) program; what the two reactor systems have in common is an inherently safe nuclear plant with a high outlet temperature that is useful for process heat as well as more conventional generation The NGNP program was tasked with investigating the application of a new generation of nuclear power plants to a variety of energy needs. One baseline reactor design for this program is a high temperature, gas-cooled reactor (HTGR), which provides many options for energy use. These might include the conventional Rankine cycle (steam turbine) generation of electricity, but also other methods: for example, Brayton cycle (gas turbine) electrical generation, and the direct use of the high temperatures characteristic of HTGR output for process heat in the chemical industry. Such process heat is currently generated by burning fossil fuels, and is a major contributor to the carbon footprint of the chemical and petrochemical industries. The HTGR, based on graphite fuel elements, can produce very high output temperatures; ideally, temperatures of 900 °C or even greater, which has significant energy advantages. Such temperatures are, of course, at the frontiers of materials limitations, at the upper end of the performance envelope of the metallic materials for which robust construction codes exist, and within the realm of ceramic materials, the fabrication and joining of which, on the scale of large energy systems, are at an earlier stage of development. A considerable amount of work was done in the diffusion welding of materials of interest for HTGR service with alloys such as 617 and 800H. The MSR output temperature is also materials limited, and is projected at about 700

  11. Molten salt-based growth of bulk GaN and InN for substrates.

    Energy Technology Data Exchange (ETDEWEB)

    Waldrip, Karen Elizabeth

    2007-08-01

    An atmospheric pressure approach to growth of bulk group III-nitrides is outlined. Native III-nitride substrates for optoelectronic and high power, high frequency electronics are desirable to enhance performance and reliability of these devices; currently, these materials are available in research quantities only for GaN, and are unavailable in the case of InN. The thermodynamics and kinetics of the reactions associated with traditional crystal growth techniques place these activities on the extreme edges of experimental physics. The novel techniques described herein rely on the production of the nitride precursor (N{sup 3-}) by chemical and/or electrochemical methods in a molten halide salt. This nitride ion is then reacted with group III metals in such a manner as to form the bulk nitride material. The work performed during the period of funding (February 2006-September 2006) focused on establishing that mass transport of GaN occurs in molten LiCl, the construction of a larger diameter electrochemical cell, the design, modification, and installation of a made-to-order glove box (required for handling very hygroscopic LiCl), and the feasibility of using room temperature molten salts to perform nitride chemistry experiments.

  12. Parametric analyses of single-zone thorium-fueled molten salt reactor fuel cycle options

    International Nuclear Information System (INIS)

    Analyses of fuel cycle options based on thorium-fueled Molten Salt Reactors (MSRs) have been performed in support of fuel cycle screening and evaluation activities for the United States Department of Energy. The MSR options considered are based on thermal spectrum MSRs with 3 different separations levels: full recycling, limited recycling, and 'once-through' operation without active separations. A single-fluid, single-zone 2250 MWth (1000 MWe) MSR concept consisting of a fuel-bearing molten salt with graphite moderator and reflectors was used as the basis for this study. Radiation transport and isotopic depletion calculations were performed using SCALE 6.1 with ENDF/B-VII nuclear data. New methodology developed at Oak Ridge National Laboratory (ORNL) enables MSR analysis using SCALE, modeling material feed and removal by taking user-specified parameters and performing multiple SCALE/TRITON simulations to determine the resulting equilibrium operating conditions. Parametric analyses examined the sensitivity of the performance of a thorium MSR to variations in the separations efficiency for protactinium and fission products. Results indicate that self-sustained operation is possible with full or limited recycling but once-through operation would require an external neutron source. (authors)

  13. Parametric study of natural circulation flow in molten salt fuel in molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pauzi, Anas Muhamad, E-mail: Anas@uniten.edu.my [Centre of Nuclear Energy, Universiti Tenaga Nasional (UNITEN), Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia); Cioncolini, Andrea; Iacovides, Hector [School of Mechanical, Aerospace, and Civil Engineering (MACE), University of Manchester, Oxford Road, M13 9PL Manchester (United Kingdom)

    2015-04-29

    The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software called FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.

  14. Parametric study of natural circulation flow in molten salt fuel in molten salt reactor

    International Nuclear Information System (INIS)

    The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software called FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015

  15. Thermodynamic characterization of salt components for Molten Salt Reactor fuel

    NARCIS (Netherlands)

    Capelli, E.

    2016-01-01

    The Molten Salt Reactor (MSR) is a promising future nuclear fission reactor technology with excellent performance in terms of safety and reliability, sustainability, proliferation resistance and economics. For the design and safety assessment of this concept, it is extremely important to have a thor

  16. Study of fluid fuel influence on delayed neutron in Molten Salt Reactor

    International Nuclear Information System (INIS)

    Background: The Molten Salt Reactor (MSR) is one of the six advanced reactor types for future nuclear energy systems in the Generation IV International Forum (GIF), possessing the advantages of good neutron economics, inherent safety, processing online, nuclear nonproliferation, etc. Due to the fluid fuel, a part of delayed neutron precursors drifts out of the reactor and decays in the loop, which is different from the solid-fuel reactor. Purpose: The distribution of delayed neutrons with different fuel velocity for Molten Salt Reactor Experiment (MSRE) is analyzed to provide reference for safety design of MSR. Methods: Based on the neutron kinetics and approximate method of homogeneous reactor, a distribution model of delayed neutron was presented. This model is applied to the distribution study of delayed neutrons with different fuel velocity. Results: The number of delayed neutrons in a core radius of 20 cm was more than that in a core radius of 30 cm. When the flow velocity of fuel decreased by a half, the discrepancy of delayed neutrons between the fifth-group and sixth-group precursors is within 6%. Conclusion: The number of delayed neutrons is increasing toward the center of reactor core. The group of delayed neutron precursors with shorter half-time has little effect on the flow velocity of molten salt. (authors)

  17. Thorium conversion optimization in two-fluid molten-salt reactor

    International Nuclear Information System (INIS)

    Molten-Salt Reactors (MSR) are an attractive reactor system for various purposes. They can be designed to be operated in a fast neutron spectrum for spent fuel transmutation or in a thermal spectrum. Thermal MSRs provide an ideal platform for conversion of thorium to 233U. Flowing salt can be continuously reprocessed to minimize neutron losses due to neutron absorption in fission products. This study deals with a static neutronic optimization of a Two-Fluid MSR concept. Such a reactor features two separated molten-salt streams in the reactor core. One salt contains fissile material 233U, the other thorium. Separation of these streams improves the conversion capabilities of MSRs. Such a design was analysed for Molten-Salt Breeder Reactor (MSBR) development. This reactor was not realized, but it is used as a reference for this study. Monte-Carlo code MCNP5 was used to model a simplified MSBR core and for calculation of the breeding capabilities of this design. Several basic geometric parameters were selected for evaluation of their effect on characteristics of the reactor. Based on this analysis, an improved designed was prepared with shorter fissile material doubling time. The whole analysis was carried out for fresh fuel composition. It is possible to expect that on-line fuel reprocessing will limit fuel composition changes during reactor operation. Only effect of 233Pa accumulation on the thorium conversion was studied for several fuel reprocessing rates. (author)

  18. Performance of a 200 MWe molten-salt reactor operated in thorium-uranium fuel-cycle

    International Nuclear Information System (INIS)

    An improved small Molten-Salt Reactor MSR (FUJI-U3) is proposed here. The basic improvement is to introduce the 3-region core design concept in order to remove graphite replacement by reducing the maximum neutron flux. Based on the calculations using the nuclear analyses code SRAC95 and the burnup analysis code ORIGEN2, it is concluded that there is no need to replace graphite moderator for 30 years operation of FUJI-U3. (author)

  19. Thorium molten-salt nuclear energy synergetics

    International Nuclear Information System (INIS)

    One of the most practical and rational approaches for establishing the idealistic Thorium resource utilization program has been presented, which might be effective to solve the principal energy problems, concerning safety, proliferation and terrorism, resource, power size and fuel cycle economy, for the next century. The first step will be the development of Small Molten-Salt Reactors as a flexible power station, which is suitable for early commercialization of Th reactors not necessarily competing with proven Large Solid-Fuel Reactors. Therefore, the more detailed design works and practical R and D planning should be performed under the international cooperations soon, soundly depending on the basic technology established by ORNL already. R and D cost would be surprisingly low. This reactor(MSR) seems to be idealistic not only in power-size, siting, safety, safeguard and economy, but also as an effective partner of Molten-Salt Fissile Breeders(MSB) in order to establish the simplest and economical Thorium molten-salt breeding fuel cycle named THORIMS-NES in all over the world including the developing countries and isolated areas. This would be one of the most practical replies to the Lilienthal's appeal of 'A NEW START' in Nuclear Energy. (author)

  20. Characteristics on the SAP-based wasteform containing radioactive molten salt waste - 16137

    International Nuclear Information System (INIS)

    This study investigated a unique wasteform containing molten salt wastes which are generated from the pyro-process for the spent fuel treatment. Using a conventional sol-gel process, SiO2-Al2O3-P2O5 (SAP) inorganic material reactive to metal chlorides were prepared. By using this inorganic composite, a monolithic wasteform were successfully fabricated via a simple process, reaction at 650 deg. C and sintering at 1100 deg. C. This unique wasteform should be qualified if it meets the requirements for final disposal. For this reasons, this paper characterized its chemical durability, physical properties, morphology and etc. In the SAP, there are three kinds of chains, Si-O-Si as a main chain, Si-O-Al as a side chain and Al-O-P/P-O-P as a reactive chain. Alkali metal chlorides were converted into metal aluminosilicate (LixAlxSi1-xO2-x) and metal phosphate(Li3PO4 and Cs2AlP3O10) while alkali earth and rare earth chlorides were changed into only metal phosphates (Sr5(PO4)3Cl and CePO4). These reaction products were compatible to borosilicate glasses which were functioned as a chemical binder for metal aluminosilicate and a physical binder for metal phosphates. By these phenomena, the wasteform was formed homogeneously above μm scale. This would affect the leaching behaviors of each radionuclides or component of binder. The leach rates of Cs and Sr under the PCT-A test condition were about 10-3g/m2day. The physical properties (Cp, k, ρ, Hv, and etc) were very reasonable. Other leaching tests (ISO, MCC-1P) are on-going. From these results, it could be concluded that SAP can be considered as an effective stabilizer on metal chlorides and the method using SAP will give a chance to minimize the waste volume for the final disposal of salt wastes through further researches. (authors)

  1. Physics research for molten salt reactor with different core boundaries

    International Nuclear Information System (INIS)

    Background: Unlike the traditional solid fuel reactor with fixed boundary conditions, the inlet and outlet pipe and the core of molten salt reactor fuel is connected so that the flowing liquid fuel can travel freely between the pipe and core. Purpose: This article has made systematic study of the influence of different molten salt fuel regions on reactor physics, including the top and bottom fuel of the core vessel and the pipe fuel. The core physics was researched under different boundary conditions, and the region of the effective core was indicated subsequently. Methods: MSRE was taken as the reference reactor and the calculation was completed based on the Monte Carlo Code MCNP. Results: Results show that the fuel in the top and bottom of vessel impacts on keff and energy spectrum obviously. The influence of outlet pipe on keff was negligible when pipe radius less than 25 cm, and the perturbation of outlet pipe on the keff could be neglected when its length more than 20 cm. Conclusions: Results provide rational theory for the design of MSR and the development of computation code. (authors)

  2. Fast Spectrum Molten Salt Reactor Options

    Energy Technology Data Exchange (ETDEWEB)

    Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Patton, Bruce W [ORNL; Howard, Rob L [ORNL; Harrison, Thomas J [ORNL

    2011-07-01

    During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

  3. An innovative approach to dynamics modeling and simulation of the molten salt rector experiment

    International Nuclear Information System (INIS)

    The Molten Salt Reactor Experiment (MSRE) was a circulating fuel thermal reactor built and operated in the sixties. As the only Molten Salt Reactor (MSR) testing facility for which extensive experimental data are available, it can be considered as a reference for the development of modeling approaches for the studies related to the Gen-IV MSR. In this work, a geometric multi-scale approach has been adopted for the simulation of the MSRE plant. The data and the experimental results relative to the U-233 fuelled reactor are considered. The neutronic parameters have been determined using the Monte Carlo code Serpent. The reactor core is divided into three radial regions, each one described by a 3D channel in which Navier-Stokes and energy conservation equations plus delayed neutron precursors (DNP) balance equations are solved. Determination of the generated power is obtained employing a point kinetics like equation, fed with importance weighted values of temperatures and DNP concentrations. The remaining part of the plant, that includes the primary and secondary cooling circuits, is modeled by means of zero-dimensional components. The results attained with such modeling approach are compared with experimental data both in time and frequency domain, showing good agreement. The adopted approach, thanks to the punctual, coupled solution of the governing equations in the core, gives better insights into the thermal behavior of the graphite and its effects on MSR dynamics than commonly used correlation-based solvers. (author)

  4. Design of a passive residual heat removal system for the FUJI-233Um molten salt reactor system

    International Nuclear Information System (INIS)

    Highlights: • A passive decay heat removal system for a small molten salt reactor is analyzed. • The system uses water as a coolant and an air cooler as final heat sink. • If the diameter of the coolant pipes is chosen correctly, problems of thermal shock can be avoided. • Safe cooling for a period of at least 10 days can be achieved. • The possibility of draining the primary system is an important safety feature of molten salt reactors. - Abstract: This paper discusses the design and analysis of a passive decay heat removal system for a Molten Salt Reactor (MSR) of 450 MWth. Following the disaster at the Fukushima-1 nuclear power station, it is clear that the public will demand improved safety performance if nuclear power is to be accepted as a sustainable source of CO2-free energy. In this scope, thorium-based MSRs have very promising properties in the area of passive safety, resource availability and proliferation resistance. Molten Salt Reactor (MSR) systems can be equipped with an emergency salt drain tank. Under any severe accident, all the fuel salt can be drained by gravity into the drain tank, thus, the primary system can be safely emptied of fissile materials and fission products. The ultimate safety can be assured by the integrity of the fuel salt in the drain tank or in other words, the capability of residual heat removal from the fuel salt in the drain tank. From this point of view, we investigated the feasibility of a passive residual heat removal system for the drain tank of an MSR (FUJI-233Um of 450 MWth). We concluded that a system comprising a large drain tank and 60, large-diameter coolant tubes can withstand the thermal shock due to the hot fuel salt, and therefore we conclude that system is feasible

  5. Molten salt fueled nuclear facility with steam-and gas turbine cycles of heat transformation

    International Nuclear Information System (INIS)

    The molten salt fueled nuclear facilities with fuel circulating in the primary circuit have a series of the potential advantages in comparison with the traditional thermal and fast reactors with solid fuel elements. These advantages are ensured by the possibility to receive effective neutron balance in the core, minimum margin reactivity, more deep fuel burnup, unbroken correctness of the fuel physical and chemical properties and by low prices of the fuel cycle. The neutron and thermal-physical calculations of the various variants of the MSFNF with steam-water and gas turbine power circuits and their technical and economical comparison are carried out in this article. Calculations of molten salt nuclear power plant with gas turbine power circuit have been carried out using chemically reacting working body ''nitrin'' (N304 + 1%NO). The molten salt fueled reactors with the thermal power near of 2300 MW with two fuel compositions have been considered. The base variant has been taken the design of NPP with VVER NP-1000 when comparing the results of the calculations. Its economical performances are presented in prices of 1990. The results of the calculations show that it is difficult to determine the advantages of any one of the variants considered in a unique fashion. But NPP with MSR possesses large reserves in the process of optimization of cycle and energy equipment parameters that can improve its technical and economical performances sufficiently. (author)

  6. Thermodynamic characterization of salt components for Molten Salt Reactor fuel

    OpenAIRE

    Capelli, E.

    2016-01-01

    The Molten Salt Reactor (MSR) is a promising future nuclear fission reactor technology with excellent performance in terms of safety and reliability, sustainability, proliferation resistance and economics. For the design and safety assessment of this concept, it is extremely important to have a thorough knowledge of the physico-chemical properties of molten fluorides salts, which are one of the best options for the reactor fuel. This dissertation presents the thermodynamic description of the ...

  7. Molten salt reactor type

    International Nuclear Information System (INIS)

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. Emphasize is put essentially on the fuel salt of the primary circuit inside which fission reactions occur. The reasons why the (LiF-BeF2-ThF4-UF4) salt was chosen for the M.S.B.R. concept are examined; the physical, physicochemical and chemical properties of this salt are discussed with its interactions with the structural materials and its evolution in time. An important part of this volume is devoted to the continuous reprocessing of the active salt, the project designers having deemed advisable to take advantage at best from the availability of a continuous purification, in a thermal breeding. The problem of tritium formation and distribution inside the reactor is also envisaged and the fundamentals of the chemistry of the secondary coolant salt are given. The solutions proposed are: the hydrogen scavenging of the primary circuit, a reduction in metal permeability by an oxyde layer deposition on the side in contact with the vapor, and tritium absorption through an isotope exchange with the hydroxifluoroborate

  8. Molten Salt Breeder Reactor Analysis Methods

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jinsu; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    Utilizing the uranium-thorium fuel cycle shows considerable potential for the possibility of MSR. The concept of MSBR should be revised because of molten salt reactor's advantage such as outstanding neutron economy, possibility of continuous online reprocessing and refueling, a high level of inherent safety, and economic benefit by keeping off the fuel fabrication process. For the development of MSR research, this paper provides the MSBR single-cell, two-cell and whole core model for computer code input, and several calculation results including depletion calculation of each models. The calculations are carried out by using MCNP6, a Monte Carlo computer code, which has CINDER90 for depletion calculation using ENDF-VII nuclear data. From the calculation results of various reactor design parameters, the temperature coefficients are all negative at the initial state and MTC becomes positive at the equilibrium state. From the results of core rod worth, the graphite control rod alone cannot makes the core subcritical at initial state. But the equilibrium state, the core can be made subcritical state only by graphite control rods. Through the comparison of the results of each models, the two-cell method can represent the MSBR core model more accurately with a little more computational resources than the single-cell method. Many of the thermal spectrum MSR have adopted a multi-region single-fluid strategy.

  9. Steady thermal hydraulic analysis for a molten salt reactor

    Institute of Scientific and Technical Information of China (English)

    ZHANG Dalin; QIU Suizheng; LIU Changliang; SU Guanghui

    2008-01-01

    The Molten Salt Reactor (MSR) can meet the demand of transmutation and breeding. In this study, theoretical calculation of steady thermal hydraulic characteristics of a graphite-moderated channel type MSR is conducted. The DRAGON code is adopted to calculate the axial and radial power factor firstly. The flow and heat transfer model in the fuel salt and graphite are developed on basis of the fundamental mass, momentum and energy equations. The results show the detailed flow distribution in the core, and the temperature profiles of the fuel salt, inner and outer wall in the nine typical elements along the axial flow direction are also obtained.

  10. Development of pyro-separation technology based on molten salt electrolysis

    International Nuclear Information System (INIS)

    In order to effectively recover uranium, rotation speed of solid cathode was examined, and effect of uranium concentration and current density on electrodeposition were confirmed. And the potentiostatic and galvanostatic electrorefining experiments were conducted. Element used in the experiments were Zr, Nd, La chlorides. The reduction potentials of chlorides metals on liquid Cd cathode were measured by cyclic voltammetry experiments. The electrowinning experiments were performed in order to recover small amounts of uranium in salt. Experimental set-up for the batch type reductive extraction experiments were developed and installed. On the base of experimental results of batch type, multi-stage extraction equipment was set-up, and optimum number of stage and recover yield were measured. In the oxidative extraction study is examine selective separation behavior of the rare earth metals from alloy composed of actinide and lanthanide metals to determine the effective separation condition. Before distillation experiment of salt and liquid cadmium were carried out, basic operation conditions such as temperature and pressure were measured by thermal analysis in advance. Finally, key information such as process building volume, cost of facility and operation, and amounts of waste generated was reviewed and compared in order to estimate economy for wet and dry process

  11. Design of a heterogeneous subcritical nuclear reactor with molten salts based on thorium; Diseno de un reactor nuclear subcritico heterogeneo con sales fundidas a base de torio

    Energy Technology Data Exchange (ETDEWEB)

    Medina C, D.; Hernandez A, P.; Letechipia de L, C.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Sajo B, L., E-mail: dmedina_c@hotmail.com [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. Postal 89000, Caracas 1080-A (Venezuela, Bolivarian Republic of)

    2015-09-15

    This paper presents the design of a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a {sup 252}Cf source, whose dose levels at the periphery allows its use in teaching and research activities. The design was realized by the Monte Carlo method, where the geometry, dimensions and the fuel was varied in order to obtain the best design. The result was a cubic reactor of 110 cm of side, with graphite moderator and reflector. In the central part having 9 ducts of 3 cm in diameter, eight of them are 110 cm long, which were placed on the Y axis; the separation between each duct is 10 cm. The central duct has 60 cm in length and this contains the {sup 252}Cf source, also there are two irradiation channels and the other six contain a molten salt ({sup 7}LiF - BeF{sub 2} - ThF{sub 4} - UF{sub 4}) as fuel. For the design the k{sub eff} was calculated, neutron spectra and ambient dose equivalent. In the first instance the above was calculated for a virgin fuel, was called case 1; then a percentage of {sup 233}U was used and the percentage of Th was decreased and was called case 2. This with the purpose of comparing two different fuels operating within the reactor. For the two irradiation ducts three positions are used: center, back and front, in each duct in order to have different flows. (Author)

  12. Development status and potential program for development of proliferation-resistant molten-salt reactors

    International Nuclear Information System (INIS)

    Preliminary studies of existing and conceptual molten-salt reactor (MSR) designs have led to the identification of conceptual systems that are technologically attractive when operated with denatured uranium as the principal fissile fuel. These denatured MSRs would also have favorable resource-utilization characteristics and substantial resistance to proliferation of weapons-usable nuclear materials. The report presents a summary of the current status of technology and a discussion of the major technical areas of a possible base program to develop commercial denatured MSRs. The general areas treated are (1) reactor design and development, (2) safety and safety related technology, (3) fuel-coolant behavior and fuel processing, and (4) reactor materials. A substantial development effort could lead to authorization for construction of a molten-salt test reactor about 5 years after the start of the program and operation of the unit about 10 years later. A prototype commercial denatured MSR could be expected to begin operating 25 years from the start of the program. The postulated base program would extend over 32 years and would cost about $700 million (1978 dollars, unescalated). Additional costs to construct the MSTR, $600 million, and the prototype commercial plant, $1470 million, would bring the total program cost to about $2.8 billion. Additional allowances probably should be made to cover contingencies and incidental technology areas not explicitly treated in this preliminary review

  13. Development status and potential program for development of proliferation-resistant molten-salt reactors

    Energy Technology Data Exchange (ETDEWEB)

    Engel, J.R.; Bauman, H.F.; Dearing, J.F.; Grimes, W.R.; McCoy, H.E. Jr.

    1979-03-01

    Preliminary studies of existing and conceptual molten-salt reactor (MSR) designs have led to the identification of conceptual systems that are technologically attractive when operated with denatured uranium as the principal fissile fuel. These denatured MSRs would also have favorable resource-utilization characteristics and substantial resistance to proliferation of weapons-usable nuclear materials. The report presents a summary of the current status of technology and a discussion of the major technical areas of a possible base program to develop commercial denatured MSRs. The general areas treated are (1) reactor design and development, (2) safety and safety related technology, (3) fuel-coolant behavior and fuel processing, and (4) reactor materials. A substantial development effort could lead to authorization for construction of a molten-salt test reactor about 5 years after the start of the program and operation of the unit about 10 years later. A prototype commercial denatured MSR could be expected to begin operating 25 years from the start of the program. The postulated base program would extend over 32 years and would cost about $700 million (1978 dollars, unescalated). Additional costs to construct the MSTR, $600 million, and the prototype commercial plant, $1470 million, would bring the total program cost to about $2.8 billion. Additional allowances probably should be made to cover contingencies and incidental technology areas not explicitly treated in this preliminary review.

  14. Advanced heat exchanger development for molten salts

    International Nuclear Information System (INIS)

    Highlights: • Hastelloy N and 242, shows corrosion resistance to molten salt at nominal operating temperatures. • Both diffusion welds and sheet material in Hastelloy N were corrosion tested in at 650, 700, and 850 °C for 200, 500, and 1000 h. • Thermal gradients and galvanic couples in the molten salts enhance corrosion rates. • Corrosion rates found were typically <10 mils per year. - Abstract: This study addresses present work concerned with advanced heat exchanger development for molten salt in nuclear and non-nuclear thermal systems. The molten salt systems discussed herein use alloys, such as Hastelloy N and 242, that show good corrosion resistance in molten salt at nominal operating temperatures up to 700 °C. These alloys were diffusion welded, and the corresponding information is presented. Test specimens were prepared for exposing diffusion welds to molten salt environments. Hastelloy N and 242 were found to be weldable by diffusion welding, with ultimate tensile strengths about 90% of base metal values. Both diffusion welds and sheet material in Hastelloy N were corrosion tested in 58 mol% KF and 42 mol% ZrF4 at 650, 700, and 850 °C for 200, 500, and 1000 h. Corrosion rates were similar between welded and nonwelded materials, typically <100 μm per year after 1000 h of corrosion tests. No catastrophic corrosion was observed in the diffusion welded regions. For materials of construction, nickel-based alloys and alloys with dense nickel coatings are effectively inert to corrosion in fluorides, but not so in chlorides. Hence, additional testing of selected alloys for resistance to intergranular corrosion is needed, as is a determination of corrosion rate as a function of the type of salt impurity and alloy composition, with respect to chromium and carbon, to better define the best conditions for corrosion resistance. Also presented is the division of the nuclear reactor and high-temperature components per American Society of Mechanical Engineers

  15. Potentialities of the molten salt reactor concept for a sustainable nuclear power production based on thorium cycle in epithermal spectrum

    International Nuclear Information System (INIS)

    In the case of a significant nuclear contribution to world energy needs, the problem of present nuclear waste management pose the sustainability of the PWR fuel cycle back into question. Studies on storage and incineration of these wastes should therefore go hand in hand with studies on innovative systems dedicated to a durable nuclear energy production, as reliable, clean and safe as possible. We are here interested in the concept of molten salt reactor, whose fuel is liquid. This particularity allows an online pyrochemical reprocessing which gives the possibility to overcome some neutronic limits. In the late sixties, the MSBR (Molten Salt Breeder Reactor) project of a graphite-moderated fluoride molten salt reactor proved thus that breeding is attainable with thorium in a thermal spectrum, provided that the online reprocessing is appropriate. By means of simulation tools developed around the Monte Carlo code MCNP, we first re-evaluate the performance of a reference system, which is inspired by the MSBR project. The complete study of the pre-equilibrium transient of this 2,500 MWth reactor, started with 232Th/233U fuel, allows us to validate our reference choices. The obtained equilibrium shows an important reduction of inventories and induced radio-toxicities in comparison with the other possible fuel cycles. The online reprocessing is efficient enough to make the system breed, with a doubling time of about thirty years at equilibrium. From the reference system, we then test different options in terms of neutron economy, transmutation and control of reactivity. We find that the online reprocessing brings most of its flexibility to this system, which is particularly well adapted to power generation with thorium. The study of transition scenarios to this fuel cycle quantifies the limits of a possible deployment from the present French power stock, and finally shows that a rational management of the available plutonium would be necessary in any case. (author)

  16. 50 years of searching for means of energy technology revolution. Thorium molten-salt reactors without nuclear proliferation

    International Nuclear Information System (INIS)

    The reduction of CO2 emission is a global concern for all industrialized countries. Nuclear energy can play an important role in this context since it does not emit CO2 for power generation. However, there are still some concerns and difficulties with current nuclear power such as nuclear proliferation, radioactive waste, safety, economy, usability etc. Therefore, there is a need for a novel global scale nuclear industry aiming to supply half of all primary energy by the middle of this century. During the 1960s, the US developed the molten salt breeder reactor (MSBR) and a small prototype was operated at Oak Ridge National Laboratory (ORNL). There is now renewed interest in its concept by several countries. Strong incentives for the molten salt reactor (MSR) design are its good fuel utilization, good economics, amazing fuel flexibility, and promising benefits. In Japan, the nuclear energy system 'THORIMS-NES (Thorium Molten-Salt Nuclear Energy Synergetic System)' is a new concept, designed such as to realize a novel enormous nuclear industry without requiring investment, solving the developmental difficulties of MSBR. Following several R and D years and a pilot scale prototype (miniFUJI), a new commercial project has just been launched aiming to realize a thorium-based molten salt reactor (FUJI). The present paper will outline this system and its strategy development, high-lighting the advantages of the Thorium-MSR such as (1) a conclusive measure on green-house-gas elimination, (2) nuclear proliferation and terrorism resistance (no plutonium production with its effective burning), (3) little high level nuclear waste production, (4) high safety (no severe accidents principally and practically), (5) rich thorium resources (distributed on large global scale), (6) low electricity generation cost, (7) easy deployment (small/simple economical reactors). (author)

  17. Study of trans-uranian incineration in molten salt reactor

    International Nuclear Information System (INIS)

    With the emergence of new options for nuclear power, molten salt reactors are envisaged for waste management. The aim of this thesis is to show how molten salt reactors can help to solve the transuranics issue. Their high versatility regarding to isotopic vector allows to accommodate large fractions of minor actinides as compared to solid fuel system. In this thesis, a neutronics study of molten salt reactors, MSR, has been conducted. For this purpose, two reference systems were considered, TIER1 and AMSTER. In the case of TIER1, an optimisation was made to reach an equilibrium. The analysis of both systems showed the main characteristics of MSR: their link to chemistry and on line reprocessing. In this work, several methods to drive the system to a state of equilibrium have been implemented and compared. During this process the isotopic composition and neutron spectrum, thus the nuclear reaction cross sections, vary tremendously. It is essential to take these evolutions into account in order to accurately estimate the equilibrium state. This has been accomplished inside the multi-recycling procedure we set with ERANOS. A dedicated calculation schema has been realized to simulate superthermal systems with this computation code. These results were checked through a benchmark against other computer codes. Then, with multi-recycling method, several molten salt systems have been compared in order to define the optimal reactor for transuranics incineration. Nevertheless, a final choice can not only be done using only neutronics characteristics since chemistry and thermal-hydraulics constraints are really important for MSR. Moreover, a complete safety study would be required. (author)

  18. Dynamic modelling and simulation of linear Fresnel solar field model based on molten salt heat transfer fluid

    Science.gov (United States)

    Hakkarainen, Elina; Tähtinen, Matti

    2016-05-01

    Demonstrations of direct steam generation (DSG) in linear Fresnel collectors (LFC) have given promising results related to higher steam parameters compared to the current state-of-the-art parabolic trough collector (PTC) technology using oil as heat transfer fluid (HTF). However, DSG technology lacks feasible solution for long-term thermal energy storage (TES) system. This option is important for CSP technology in order to offer dispatchable power. Recently, molten salts have been proposed to be used as HTF and directly as storage medium in both line-focusing solar fields, offering storage capacity of several hours. This direct molten salt (DMS) storage concept has already gained operational experience in solar tower power plant, and it is under demonstration phase both in the case of LFC and PTC systems. Dynamic simulation programs offer a valuable effort for design and optimization of solar power plants. In this work, APROS dynamic simulation program is used to model a DMS linear Fresnel solar field with two-tank TES system, and example simulation results are presented in order to verify the functionality of the model and capability of APROS for CSP modelling and simulation.

  19. The molten salt reactor adventure

    International Nuclear Information System (INIS)

    A personal history of the development of molten salt reactors in the United States is presented. The initial goal was an aircraft propulsion reactor, and a molten fluoride-fueled Aircraft Reactor Experiment was operated at Oak Ridge National Laboratory in 1954. In 1956, the objective shifted to civilian nuclear power, and reactor concepts were developed using a circulating UF4-ThF4 fuel, graphite moderator, and Hastelloy N pressure boundary. The program culminated in the successful operation of the Molten Salt Reactor Experiment in 1965 to 1969. By then the Atomic Energy Commission's goals had shifted to breeder development; the molten salt program supported on-site reprocessing development and study of various reactor arrangements that had potential to breed. Some commercial and foreign interest contributed to the program which, however, was terminated by the government in 1976. The current status of the technology and prospects for revived interest are summarized

  20. Optimization of temperature coefficient and breeding ratio for a graphite-moderated molten salt reactor

    International Nuclear Information System (INIS)

    Highlights: • The temperature feedback coefficient with different moderation ratios for TMSR in thermal neutron region is optimized. • The breeding ratio and doubling time of a thermal TMSR with three different reprocessing schemes are analyzed. • The smaller hexagon size and larger salt fraction with more negative feedback coefficient can better satisfy the safety demands. • A shorter reprocessing time can achieve a better breeding ratio in a thermal TMSR. • The graphite moderator lifespan is compared with other MSRs and discussed. - Abstract: Molten salt reactor (MSR) has fascinating features: inherent safety, no fuel fabrication, online fuel reprocessing, etc. However, the graphite moderated MSR may present positive feedback coefficient which has severe implications for the transient behavior during operation. In this paper, the feedback coefficient and the breeding ratio are optimized based on the fuel-to-graphite ratio variation for a thorium based MSR (TMSR). A certain thermal core with negative feedback coefficient and relative high initial breeding ratio is chosen for the reprocessing scheme analysis. The breeding performances for the TMSR under different online fuel reprocessing efficiencies and frequencies are evaluated and compared with other MSR concepts. The results indicate that the thermal TMSR can get a breeding ratio greater than 1.0 with appropriate reprocessing scheme. The low fissile inventory in thermal TMSR leads to a short doubling time and low transuranic (TRU) inventory. The lifetime of graphite used for the TMSR is also discussed

  1. Open problems in reprocessing of a molten salt reactor fuel

    International Nuclear Information System (INIS)

    The study of fuel cycle in a molten salt reactor (MSR) needs deeper understanding of chemical methods used for reprocessing of spent nuclear fuel and preparation of MSR fuel, as well as of the methods employed for reprocessing of MSR fuel itself. Assuming that all the reprocessing is done on the basis of electrorefining, we formulate some open questions that should be answered before a flow sheet diagram of the reactor is designed. Most of the questions concern phenomena taking place in the vicinity of an electrode, which influence the efficiency of the reprocessing and sensibility of element separation. Answer to these questions would be an important step forward in reactor set out. (Authors)

  2. Study of the pyrochemical treatment-recycling process of the Molten Salt Reactor fuel

    International Nuclear Information System (INIS)

    The Separation Processes Studies Laboratory (Commissariat a l'energie Atomique) has made a preliminary assessment of the reprocessing system associated with Molten Salt Fast Reactor (MSFR). The scheme studied in this paper is based on the principle of reductive extraction and metal transfer that constituted the core process designed for the Molten Salt Breeder Reactor (MSBR), although the flow diagram has been adapted to the current needs of the Molten Salt Reactor Fast (MSFR).

  3. Molten-salt synthesis and composition-dependent luminescent properties of barium tungsto-molybdate-based solid solution phosphors

    Science.gov (United States)

    Xiang-Hong, He; Zhao-Lian, Ye; Ming-Yun, Guan; Ning, Lian; Jian-Hua, Sun

    2016-02-01

    Pr3+-activated barium tungsto-molybdate solid solution phosphor Ba(Mo1-zWz)O4:Pr3+ is successfully fabricated via a facile molten-salt approach. The as-synthesized microcrystal is of truncated octahedron and exhibits deep-red-emitting upon blue light excitation. Powder x-ray diffraction and Raman spectroscopy techniques are utilized to investigate the formation of solid solution phosphor. The luminescence behaviors depend on the resulting composition of the microcrystals with fixed Pr3+-doping concentration, while the host lattices remain in a scheelite structure. The forming solid solution via the substitution of [WO4] for [MoO4] can significantly enhance its luminescence, which may be due to the fact that Ba(Mo1-zWz)O4:Pr3+ owns well-defined facets and uniform morphologies. Owing to its properties of high phase purity, well-defined facets, highly uniform morphologies, exceptional chemical and thermal stabilities, and stronger emission intensity, the resulting solid solution phosphor is expected to find potential applications in phosphor-converted white light-emitting diodes (LEDs). Project supported by the Construction Fund for Science and Technology Innovation Group from Jiangsu University of Technology, China, the Key Laboratory of Atmospheric Environment Monitoring and Pollution Control, China (Grant No. KHK1409), the Priority Academic Program Development of Jiangsu Higher Education Institutions, China, and the National Natural Science Foundation of China (Grant No. 21373103).

  4. A new method to evaluate the sealing reliability of the flanged connections for Molten Salt Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Li, Qiming, E-mail: liqiming@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Tian, Jian; Zhou, Chong [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Wang, Naxiu, E-mail: wangnaxiu@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China)

    2015-06-15

    Highlights: • We novelly valuate the sealing reliability of the flanged connections for MSRs. • We focus on the passive decrease of the leak impetus in flanged connections. • The modified flanged connections are acquired a sealing ability of self-adjustment. • Effects of redesigned flange configurations on molten salt leakage are discussed. - Abstract: The Thorium based Molten Salt Reactor (TMSR) project is a future Generation IV nuclear reactor system proposed by the Chinese Academy of Sciences with the strategic goal of meeting the growing energy needs in the Chinese economic development and social progress. It is based on liquid salts served as both fuel and primary coolant and consequently great challenges are brought into the sealing of the flanged connections. In this study, an improved prototype flange assembly is performed on the strength of the Freeze-Flange initially developed by Oak Ridge National Laboratory (ORNL). The calculation results of the finite element model established to analyze the temperature profile of the Freeze-Flange agree well with the experimental data, which indicates that the numerical simulation method is credible. For further consideration, the ideal-gas thermodynamic model, together with the mathematical approximation, is novelly borrowed to theoretically evaluate the sealing performance of the modified Freeze-Flange and the traditional double gaskets bolted flange joint. This study focuses on the passive decrease of the leak driving force due to multiple gaskets introduced in flanged connections for MSR. The effects of the redesigned flange configuration on molten salt leakage resistance are discussed in detail.

  5. A new method to evaluate the sealing reliability of the flanged connections for Molten Salt Reactors

    International Nuclear Information System (INIS)

    Highlights: • We novelly valuate the sealing reliability of the flanged connections for MSRs. • We focus on the passive decrease of the leak impetus in flanged connections. • The modified flanged connections are acquired a sealing ability of self-adjustment. • Effects of redesigned flange configurations on molten salt leakage are discussed. - Abstract: The Thorium based Molten Salt Reactor (TMSR) project is a future Generation IV nuclear reactor system proposed by the Chinese Academy of Sciences with the strategic goal of meeting the growing energy needs in the Chinese economic development and social progress. It is based on liquid salts served as both fuel and primary coolant and consequently great challenges are brought into the sealing of the flanged connections. In this study, an improved prototype flange assembly is performed on the strength of the Freeze-Flange initially developed by Oak Ridge National Laboratory (ORNL). The calculation results of the finite element model established to analyze the temperature profile of the Freeze-Flange agree well with the experimental data, which indicates that the numerical simulation method is credible. For further consideration, the ideal-gas thermodynamic model, together with the mathematical approximation, is novelly borrowed to theoretically evaluate the sealing performance of the modified Freeze-Flange and the traditional double gaskets bolted flange joint. This study focuses on the passive decrease of the leak driving force due to multiple gaskets introduced in flanged connections for MSR. The effects of the redesigned flange configuration on molten salt leakage resistance are discussed in detail

  6. Impact on breeding rate of different Molten Salt reactor core structures

    International Nuclear Information System (INIS)

    Background: Molten Salt Reactor (MSR) has several advantages over the other Generation IV reactor. Referred to the French CNRS research and compared to the fast reactor, super epithermal neutron spectrum reactor type is slightly lower and beading rate reaches 1.002. Purpose: The aim is to explore the best conversion zone layout scheme in the super epithermal neutron spectrum reactor. This study can make nuclear fuel as one way to solve the energy problems of mankind in future. Methods: Firstly, SCALE program is used for molten salt reactor graphite channel, molten salt core structure, control rods, graphite reflector and layer cladding structure. And the SMART modules are used to record the important actinides isotopes and their related reaction values of each reaction channel. Secondly, the thorium-uranium conversion rate is calculated. Finally, the better molten salt reactor core optimum layout scheme is studied comparing with various beading rates. Results: Breading zone layout scheme has an important influence on the breading rate of MSR. Central graphite channels in the core can get higher neutron flux irradiation. And more 233Th can convert to 233Pa, which then undergoes beta decay to become 233U. The graphite in the breading zone gets much lower neutron flux irradiation, so the life span of this graphite can be much longer than that of others. Because neutron flux irradiation in the uranium molten salt graphite has nearly 10 times higher than the graphite in the breading zone, it has great impact on the thorium-uranium conversion rates. For the super epithermal neutron spectrum molten salt reactors, double salt design cannot get higher thorium-uranium conversion rates. The single molten salt can get the same thorium-uranium conversion rate, meanwhile it can greatly extend the life of graphite in the core. Conclusions: From the analysis of calculation results, Blanket breeding area in different locations in the core can change the breeding rates of thorium

  7. Kinetics, dynamics and neutron noise in Molten Salt Reactors

    International Nuclear Information System (INIS)

    Reactor kinetic and dynamic properties of Molten Salt Reactors (MSR) are investigated in a simple model, which allows closed compact analytical solutions to be obtained. The goal is to gain insight, rather than to produce high-quality quantitative data. Through an interpretation of the different terms in the basic equations, and by means of analytical solutions, various approximations are introduced and their validity discussed. The dynamical behaviour of MSRs and their response to small stationary perturbations is described and discussed in comparison with traditional systems. (author)

  8. Thorium cycle implementation through plutonium incineration by thorium molten-salt nuclear energy synergetics

    International Nuclear Information System (INIS)

    Considering the increasing world energy demand and the urgent necessity of replacement of fossil-fuel by nuclear energy for survival of the global environmental crisis, we urgently need to prepare a more rational and a huge nuclear industry. As an improved alternative of present technology, the utilization of U is strongly recommended. ORNL proposed an idealistic MSBR since 1970. We modified it to the world-wide applicable system: THORIMS-NES [Thorium Molten-Salt Nuclear Energy Synergetic System], which is composed of simple thermal fission power stations (FUJI) and fissile-producing Accelerator Molten-Salt Breeder (AMSB). FUJI is a size-flexible NEAR BREEDER even not using continuous chemical processing and core-graphite exchange, and AMSB is based on a single-fluid molten-salt target/blanket concept, the technological development of which is easy and simple except for the high-current proton accelerator. THORIMS-NES has many advantages, and here the issues of safety, nuclear-proliferation and social/philosophical acceptance is mostly explained. In practice, the shift to THORIMS-NES from the present U-Pu cycle era will be smoothly implemented by converting Pu and TRU in weapons and spent-fuels into molten fluoride salt by a drying process (such as the Russian FREGATE project) which was established by the French, Russians and Czechs. Pilot plant 'mini FUJI', 7MW(e) might be commissioned after 7 years depending on the result of successful 4 years operation of MSRE in ORNL, and Small Demonstration Reactor 'FUJI-Pu', 150MW(e) can probably be in operation 12 years from now utilizing the world ability of Na-Reactor Technology. Depending on such MSR-technology development, AMSB-Pu might be able to industrialize 20 years from now. (author)

  9. Behavior study on Na heat pipe in passive heat removal system of new concept molten salt reactor

    International Nuclear Information System (INIS)

    The high temperature Na heat pipe is an effective device for transporting heat, which is characterized by remarkable advantages in conductivity, isothermally and passively working. The application of Na heat pipe on passive heat removal system of new concept molten salt reactor (MSR) is significant. The transient performance of high temperature Na heat pipe was simulated by numerical method under the MSR accident. The model of the Na heat pipe was composed of three conjugate heat transfer zones, i.e. the vapor, wick and wall. Based on finite element method, the governing equations were solved by making use of FORTRAN to acquire the profiles of the temperature, velocity and pressure for the heat pipe transient operation. The results show that the high temperature Na heat pipe has a good performance on operating characteristics and high heat transfer efficiency from the frozen state. (authors)

  10. Electrochemistry of molten salt systems within context of modern nuclear fuel cycles

    International Nuclear Information System (INIS)

    Among other applications, electrochemical based separation of actinides and lanthanides from molten salt media seems to be suitable method for reprocessing schemes in several fuel cycles of modern reactor types which are currently under development. Within this work, electrochemical properties of selected actinides (U,Th), lanthanides (Eu, Sm, Nd, Pr, Gd, La) and other elements (Mo, Zr) were studied in several molten fluoride systems (LiF-BeF2, LiF-CaF2, LiF-NaF, LiF-NaF-KF). Based on obtained results, electrolytic experiments to deposit some of the elements on inert and reactive electrode were performed. LiF-BeF2 (FLiBe) melt is considered to be a crucial system for most of the molten salt reactor (MSR) concepts. In FLiBe, basic electrochemical properties of selected actinides and lanthanides were studied. Due to FLiBe's narrow electrochemical window, only uranium can be directly deposited on the cathode. Electrolytic deposition of uranium on Mo and Ni electrodes was achieved. In BeF2-free melts, full spectrum of available elements was studied by variety of electrochemical methods. Electrolytic experiments were performed with special attention given to pulsed-current electrolysis and reactive electrode (Ni) on which the deposits are in the form of alloys and successful separation of uranium from gadolinium was demonstrated. (author)

  11. A new approach for modeling and analysis of molten salt reactors using SCALE

    International Nuclear Information System (INIS)

    The Office of Fuel Cycle Technologies (FCT) of the DOE Office of Nuclear Energy is performing an evaluation and screening of potential fuel cycle options to provide information that can support future research and development decisions based on the more promising fuel cycle options. [1] A comprehensive set of fuel cycle options are put into evaluation groups based on physics and fuel cycle characteristics. Representative options for each group are then evaluated to provide the quantitative information needed to support the valuation of criteria and metrics used for the study. Included in this set of representative options are Molten Salt Reactors (MSRs), the analysis of which requires several capabilities that are not adequately supported by the current version of SCALE or other neutronics depletion software packages (e.g., continuous online feed and removal of materials). A new analysis approach was developed for MSR analysis using SCALE by taking user-specified MSR parameters and performing a series of SCALE/TRITON calculations to determine the resulting equilibrium operating conditions. This paper provides a detailed description of the new analysis approach, including the modeling equations and radiation transport models used. Results for an MSR fuel cycle option of interest are also provided to demonstrate the application to a relevant problem. The current implementation is through a utility code that uses the two-dimensional (2D) TRITON depletion sequence in SCALE 6.1 but could be readily adapted to three-dimensional (3D) TRITON depletion sequences or other versions of SCALE. (authors)

  12. Three-dimensional thermalhydraulic analysis of molten salt reactor concepts

    International Nuclear Information System (INIS)

    Partitioning and transmutation is expected to be a promising option to extend the possibilities of nuclear energy and give a good solution for the problem of high level radwaste. Several liquid-fueled reactor concepts or accelerator driven subcritical systems (ADS) were proposed as transmutors. Many of these consider fluoride based molten salts as the liquid fuel and coolant medium. The thermalhydraulic behavior of these systems is expected to be fundamentally different than the behavior of widely used water-cooled reactors with solid fuel. Considering large flow domains three-dimensional thermalhydraulic analysis seems to be applicable. Since the fuel is the coolant medium as well, one can expect a stronger coupling between reactor physics and thermalhydraulics, too. In the present paper the application of Computational Fluid Dynamics (CFD) for three-dimensional thermalhydraulics simulations of molten salt reactor concepts is introduced. First a homogenous single region molten salt reactor concept is studied and optimized. In this model the heat carrier/fuel salt is circulated through the core by external pumps. The nominal thermal output is 2500 MW. Another single region reactor concept is introduced as well. This concept has internal heat exchangers in the flow domain and the molten salt is circulated by natural convection. In the paper the results of CFD calculations with these concepts are presented. In the further work our objective is to investigate the thermalhydraulics of the multi-region molten salt reactor. (author)

  13. Molten salts safety and hazards: an annotated bibliography

    International Nuclear Information System (INIS)

    The present information was compiled as a guide to good practice and as an aid in discovering potentially dangerous situations with molten salts in research and technology, and is based on a survey of the primary, secondary, and tertiary scientific literature to December, 1976. The format of an annotated bibliography was selected for this communication; the results are reported with reference to nine categories: hazards; reactive chemical hazards; dangerous mixtures of inorganic compounds; potentially hazardous metal-molten salts mixtures; precautions; applications of molten salts in process design and technology; water in melts; water solubility data; and reviews. A systems index is included. Titles of selected publications are included in the list of references as an additional aid to the user. (Auth.)

  14. Development of a Minichannel Compact Primary Heat Exchanger for a Molten Salt Reactor

    OpenAIRE

    Lippy, Matthew Stephen

    2011-01-01

    The first Molten Salt Reactor (MSR) was designed and tested at Oak Ridge National Laboratory (ORNL) in the 1960â s, but recent technological advancements now allow for new components, such as heat exchangers, to be created for the next generation of MSRâ s and molten salt-cooled reactors. The primary (fuel salt-to-secondary salt) heat exchanger (PHX) design is shown here to make dramatic improvements over traditional shell-and-tube heat exchangers when changed to a compact heat exchanger de...

  15. Fuels and fission products clean up for molten salt reactor of the incinerator type

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V.; Gorbunov, V.; Zakirov, R. [RRC-Karchatov Institute, Moscow (Russian Federation)

    2000-07-01

    The objective of this paper is to discuss the feasibility of molten salt reactor technology for treatment of plutonium, minor actinides and fission products, when the reactor and fission product cleanup unit are planned as an integral system. This contribution summarizes the reasons which led to selection of the salt compositions for the molten salt reactor of the TRU incinerator type (MSB). Special characteristics of behavior of TRUs and fission products during power operation of MSB concepts are presented. The present paper briefly reviews the processing developments underlying the prior molten salt reactor (MSR) programs and relates then to the separation requirements for the MSB concept. Status and development needs in the thermodynamic properties of fluorides and fission product cleanup methods (with emphasis on actinides-lanthanides separation) are discussed. (authors)

  16. Fuels and fission products clean up for molten salt reactor of the incinerator type

    International Nuclear Information System (INIS)

    The objective of this paper is to discuss the feasibility of molten salt reactor technology for treatment of plutonium, minor actinides and fission products, when the reactor and fission product cleanup unit are planned as an integral system. This contribution summarizes the reasons which led to selection of the salt compositions for the molten salt reactor of the TRU incinerator type (MSB). Special characteristics of behavior of TRUs and fission products during power operation of MSB concepts are presented. The present paper briefly reviews the processing developments underlying the prior molten salt reactor (MSR) programs and relates then to the separation requirements for the MSB concept. Status and development needs in the thermodynamic properties of fluorides and fission product cleanup methods (with emphasis on actinides-lanthanides separation) are discussed. (authors)

  17. Corrosion of metallic materials for molten salt reactors

    International Nuclear Information System (INIS)

    Fluoride salts are contemplated for innovative nuclear applications such as primary or secondary coolant in the next generation systems, solvent for the processing of spent fuel or fuel of the Molten Salt Reactor (MSR). Considering the elevated temperatures, compatibility of structural materials with the molten environment is a key factor for the feasibility of those processes. Determining properties for metallic structures are mechanical strength and resistance against corrosion. Corrosion of metals in molten fluorides has been investigated in support of the MSRs' development. Various alloys were tested in mixtures of LiF, BeF2, NaF, ZrF4, ThF4, NaBF4, etc up to about 815degC using static experiments, convection loops and in-pile expositions. Low chromium (approx. 6 wt% Cr), molybdenum-strengthened nickel-base alloys showed sufficient mechanical properties together with appropriate corrosion resistance. However, mass transfer occurs in circuits where a thermal gradient operates. Corrosion by mass transfer is certainly a complex process involving several elementary or coupled steps of solid diffusion, chemical and electrochemical reaction, liquid diffusion, convection... To elucidate such an intricate system, it seems worth investigating fundamentals of metal/salt interactions, especially the influence of environmental and material factors. Electrochemistry was shown to be an efficient tool to study the charge transfer and liquid diffusion steps. This extended abstract gives an insight on basic electrochemical studies on corrosion of metallic materials in molten fluorides. In a first stage, LiF-NaF at 900degC was selected as a reference medium. Electrochemical techniques were proved to be practicable in fluoride melts and were efficient tools for studying reactivity of metal and assessing kinetics. Electrochemistry was carried out in a graphite crucible under argon atmosphere using a platinum wire as inert comparison electrode. First linear and cyclic voltammetries

  18. Reactivity insertion accident in a small molten salt reactor

    International Nuclear Information System (INIS)

    Molten Salt Reactors (MSRs) have a long history with the first design studies beginning in the 1950's at the Oak Ridge National Laboratory (ORNL). MSRs have many advantages such as improved potential in safety, proliferation resistance, resource sustainability and waste reduction. But MSR developmental activities have slowed considerably except for some small scale efforts, mostly in Russia, France, Japan and a few other places and there are few data available to support detailed analyses. Recently, a conceptual design of a small MSR, name Fuji-12 has been proposed. Fuji-12 operates with the same fuel salt as the Molten Salt Breeder Reactor (MSBR) designed by ORNL. But it differs from the ORNL design in several ways, such as no on-site chemical processing plant and a low rated power. The authors are interested in the MSR concept due to its high potential in the areas of safety, proliferation resistance, resource sustainability and waste reduction, all necessary requirements for the generation IV nuclear power systems. Therefore the MSR concept has been selected as one of the more promising candidates for future consideration. The authors believe that additional investigations are necessary for future study. From this point of view, the authors have analyzed various reactivity insertion accidents due to control rod malfunctions in a MSR named FUJI-12. The MSR can be operated with a small excess reactivity and also the control rods for power adjustment consist of graphite, which has buoyancy in the fuel salt. Thus the reactivity addition could be limited by design. However at the same time the delayed neutron fraction is quite small due to the usage of U-233 as fissile material and the circulation of the fuel salt. Therefore reactivity insertion accident should be qualitatively evaluated. The reactor transients were analyzed without scram in order to evaluate the severity of such accidents against the safety. The severity of the accident was discussed for the outlet

  19. Alloys compatibility in molten salt fluorides: Kurchatov Institute related experience

    Science.gov (United States)

    Ignatiev, Victor; Surenkov, Alexandr

    2013-10-01

    In the last several years, there has been an increased interest in the use of high-temperature molten salt fluorides in nuclear power systems. For all molten salt reactor designs, materials selection is a very important issue. This paper summarizes results, which led to selection of materials for molten salt reactors in Russia. Operating experience with corrosion thermal convection loops has demonstrated good capability of the “nickel-molybdenum alloys + fluoride salt fueled by UF4 and PuF3 + cover gas” system up to 750 °C. A brief description is given of the container material work in progress. Tellurium corrosion of Ni-based alloys in stressed and unloaded conditions studies was also tested in different molten salt mixtures at temperatures up to 700-750 °C, also with measurement of the redox potential. HN80MTY alloy with 1% added Al is the most resistant to tellurium intergranular cracking of Ni-base alloys under study.

  20. Probability safety assessment of LOOP accident to molten salt reactor

    International Nuclear Information System (INIS)

    Background: Loss of offsite power (LOOP) is a possible accident to any type of reactor, and this accident can reflect the main idea of reactor safety design. Therefore, it is very important to conduct a study on probabilistic safety assessment (PSA) of the molten salt reactor that is under LOOP circumstance. Purpose: The aim is to calculate the release frequency of molten salt radioactive material to the core caused by LOOP, and find out the biggest contributor to causing the radioactive release frequency. Methods: We carried out the PSA analysis of the LOOP using the PSA process risk spectrum, and assumed that the primary circuit had no valve and equipment reliability data based on the existing mature power plant equipment reliability data. Results: Through the PSA analysis, we got the accident sequences of the release of radioactive material to the core caused by LOOP and its frequency. The results show that the release frequency of molten salt radioactive material to the core caused by LOOP is about 2×10-11/(reactor ·year), which is far below that of the AP1000 LOOP. In addition, through the quantitative analysis, we obtained the point estimation and interval estimation of uncertainty analysis, and found that the biggest contributor to cause the release frequency of radioactive material to the core is the reactor cavity cooling function failure. Conclusion: This study provides effective help for the design and improvement of the following molten salt reactor system. (authors)

  1. The Integral Molten Salt Reactor (IMSR)

    Energy Technology Data Exchange (ETDEWEB)

    Leblanc, D. [Terrestrial Energy, Mississauga, Ontario (Canada)

    2014-12-15

    The Integral Molten Salt Reactor is a simple burner or converter design that seeks to maximize passive and inherent safety features in order to minimize development time and achieve true cost innovation. Its integration of all primary systems into a unit sealed for the design life of the reactor will be reviewed with focus on the unique design aspects that make this a pragmatic approach. The IMSR is being developed by Terrestrial Energy in a range of power outputs with initial focus on an 80 MWth (32.5 MWe) unit primarily for remote energy needs. Similar units of modestly larger dimension and up to 600 MWth (291 MWe) are planned that remain truck transportable and able to compete in base load electricity markets worldwide. (author)

  2. The Integral Molten Salt Reactor (IMSR)

    International Nuclear Information System (INIS)

    The Integral Molten Salt Reactor is a simple burner or converter design that seeks to maximize passive and inherent safety features in order to minimize development time and achieve true cost innovation. Its integration of all primary systems into a unit sealed for the design life of the reactor will be reviewed with focus on the unique design aspects that make this a pragmatic approach. The IMSR is being developed by Terrestrial Energy in a range of power outputs with initial focus on an 80 MWth (32.5 MWe) unit primarily for remote energy needs. Similar units of modestly larger dimension and up to 600 MWth (291 MWe) are planned that remain truck transportable and able to compete in base load electricity markets worldwide. (author)

  3. Molten salt synthesis of potassium hexatitanate

    Science.gov (United States)

    Zaremba, T.

    2012-09-01

    Potassium hexatitanate fibrous crystals have been synthesized by a conventional solid-state reaction and via molten salt process. The molten salt process has been shown to be effective in preparing fine and non-agglomerated K2Ti6O13 whiskers. The type of molten salt (KCl, NaCl-KCl) has a significant effect on the chemical composition of the whiskers. By using a eutectic mixture of NaCl and KCl, the replacement of potassium ions in solid potassium hexatitanate by smaller sodium ions from the chloride flux can be achieved. The characterization of the samples was carried out by means of XRD, SEM, EDX and WDX.

  4. Analysis of reactivity initiated transient from control rod failure events of a molten salt reactor

    International Nuclear Information System (INIS)

    In a molten salt reactor (MSR), the fuel is dissolved in fluoride salt. In this paper, the reactivity worth and reactivity initiated transient of Molten-Salt Reactor Experiment (MSRE) in the control rod failure events are analyzed. The point kinetic coupling heat-transfer model with decay character of six-group delayed neutron precursors due to the fuel motion is applied. The relative power and temperature transient under reactivity step and ramp initiated at different power levels are studied. The results show that the reactor power and temperature increase to a maximum, where they begin to decrease to stable values. Comparing with full power level, the transient result at low power level is more serious. The results are of help in our study on safety characteristics of an MSR system. (authors)

  5. Effect of Halide Flux on Physicochemical Properties of MgCl2-Based Molten Salts for Accelerating Zirconium Production: Thermodynamic Assessment

    Science.gov (United States)

    Shin, Jae Hong; Park, Joo Hyun

    2016-07-01

    The effective halide flux additive for increasing the density of MgCl2 mixture and for decreasing the activity of MgCl2 was investigated in order to improve the reaction efficiency between gaseous ZrCl4 and fresh Mg melt to produce zirconium sponge. Thermochemical computation using FactSageTM software was primarily carried out, followed by the experimental confirmation. The addition of CaCl2, BaCl2, MgF2, and CaF2 to the molten MgCl2 increases the density of the melts, indicating that these halide additives can be a candidate to increase the density of the MgCl2-based molten salts. Among them, BaCl2, MgF2, and CaF2 are the useful additives. The activity of MgCl2 can be reduced by the addition of BaCl2, KCl, NaCl, MgF2, and CaF2, among which the CaF2 is the most effective additive to reduce the activity of MgCl2 with the strongest negative deviation from an ideality. Thus, the addition of CaF2 to the MgCl2, forming the MgCl2-CaF2 binary melt, is the most effective way not only to increase the density of the melt but also to decrease the activity of MgCl2, which was experimentally confirmed. Consequently, the production rate of zirconium sponge by magnesiothermic reduction process can be accelerated by the addition of CaF2.

  6. Library dependency of effective multiplication factor in thorium molten salt reactor

    International Nuclear Information System (INIS)

    Molten salt reactors (MSRs) have been selected as one of the generation IV reactor systems for its high potential in the areas of safety, proliferation resistance, resource sustainability and waste reduction. One of the disadvantages of thermal MSRs is the replacement of graphite moderator due to neutron irradiation damage. From this point of view, we are interested in fast neutron spectrum MSR concepts without moderator. Now, we are trying to evaluate the performance of a fast MSR that uses the 'Japanese recycled plutonium' as fissile material. We chose Thorium Molten Salt Reactor 'TMSR' as our target fast MSR. TMSR is studied by LPSC/IN2P3/CNRS laboratories in France, and has single fluid molten salt fuel surrounded by fertile blanket salt. First of all, we tried to reproduce the results of TMSR research in France. As a result, we found that at BOC, effective multiplication factor which we calculated is too large. Consulting with original researchers, it is suggested that this difference may be caused by the difference of nuclear data libraries. While original researchers use ENDF/B-VI as the nuclear data library, we use JENDL3.3. In fact, the result of effective multiplication factor in SRAC2006 calculation using ENDF/B-VI is smaller than that using JENDL3.3. In this paper, we will discuss in more detail about the library dependency in TMSR using nine nuclear data libraries. (author)

  7. Cooling molten salt reactors using “gas-lift”

    Energy Technology Data Exchange (ETDEWEB)

    Zitek, Pavel, E-mail: zitek@kke.zcu.cz, E-mail: klimko@kke.zcu.cz; Valenta, Vaclav, E-mail: zitek@kke.zcu.cz, E-mail: klimko@kke.zcu.cz; Klimko, Marek, E-mail: zitek@kke.zcu.cz, E-mail: klimko@kke.zcu.cz [University of West Bohemia in Pilsen, Univerzitní 8, 306 14 Pilsen (Czech Republic)

    2014-08-06

    This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a “Two-phase flow demonstrator” (TFD) used for experimental study of the “gas-lift” system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for “gas-lift” (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

  8. Cooling molten salt reactors using "gas-lift"

    Science.gov (United States)

    Zitek, Pavel; Valenta, Vaclav; Klimko, Marek

    2014-08-01

    This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a "Two-phase flow demonstrator" (TFD) used for experimental study of the "gas-lift" system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for "gas-lift" (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

  9. Cooling molten salt reactors using “gas-lift”

    International Nuclear Information System (INIS)

    This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a “Two-phase flow demonstrator” (TFD) used for experimental study of the “gas-lift” system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for “gas-lift” (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors

  10. Preparation and Molten Salt as Performances of Room Electrolyte carbon Capacitor Based on Trifluoroacetamide n CarbonLiPF6 and

    Institute of Scientific and Technical Information of China (English)

    2012-01-01

    A novel room moRen salt with an eutectic temperature of about -62℃ is prepared using LiPF6 and trifluoroacetamide as precursors. And then its performance is evaluated in carbon-carbon electrochemical double layerdifferent molar ratios are characterized and then the liquid-solid phase diagram is presented. The electrochemical performance tests show that the as-prepared LiPF6/trifluoroacetamide molten salt is a promising electrolyte candidate for carboncarbon EDLCs.

  11. Feasibility study of a pilot scale molten salt reactor demonstration

    International Nuclear Information System (INIS)

    Energy Process Developments Ltd. was awarded a grant by Innovate UK in July 2014 to undertake a year-long project to determine the feasibility of developing a pilot scale molten salt reactor in the UK. The study looked at six current available proposed MSR configurations and proposed the immediate next steps for design and build of a chosen demonstrator reactor project. Tremendous knowledge growth in the 60 years of the first nuclear era has not seen substantial advances in nuclear fission technology much beyond the Pressurised Water Reactor, initially a hastily adopted device for military and civil applications, and essentially comprising water cooling of solid fuel elements. The imminent second nuclear era requires introduction of inherently more efficient, safer, cheaper, nuclear power obtainable with liquid-fuelled - namely Molten Salt Reactor (MSR) technology - the best out of the six Gen IV options. This Gen IV option, when considered in 2002, was believed to be decades away from readiness. This study reviews more recent work. The evidence is that the MSR is ready now. In the immediate urgency of the present, this liquid-fuelled reactor technology can be seen as highly innovative, necessary and rewarding. It is ready to form a key part of any affordable policy proposals for the UK energy supply. This feasibility study is seen as the first step towards full scale implementation of the technology. MSRs are passively safe, operate at atmospheric pressure, at higher efficiencies than PWRs and can be load following. Thorium is the ultimate fuel of choice which can provide the world with a near limitless supply of energy. A demonstration reactor will show the media, public and investors that this technology exists as a clean source of cheap sustainable power. The project reviewed the status of all MSR activity internationally, the regulatory regime in the UK and potential sites. Nuclear insurers were consulted on their insurability and the outlook of an energy

  12. Proceedings of the conference on molten salts in nuclear technology

    International Nuclear Information System (INIS)

    The third stage of the Indian nuclear programme envisages extended use of -fuelled reactors with thorium as the fertile material. India is pursuing many concepts of thorium based reactors. One of them is the 300 MWe Advanced Heavy Water Reactor (AHWR). High temperature reactor design options are also being worked out to supply process heat to produce hydrogen by water splitting as a substitute to petroleum based transport fuel. Another option, which is very attractive for large scale deployment and effective thorium utilization, is the molten salt-breeder reactor (MSBR), which can be configured to give significant breeding within thermal energy spectrum of neutrons. DAE is also pursuing metallic fuelled fast breeder reactors with co-located pyrochemical reprocessing plants as an option to increase the breeding gain, reduce the doubling time of the fuel production and reprocess short-cooled and high burnup fuel. In both of these developmental aspects, molten salts play very important role and Bhabha Atomic Research Centre and its sister organization Indira Gandhi Centre of Atomic Research have initiated extensive molten salt research programs. Extensive investigations have been carried out and literature made available by ORNL in the open domain on MSRE and MSBR. These investigations cover almost all aspects of reactor design including, fuel identification, composition, materials of construction, compatibility issues, reactor physics aspects, expected thermal power output etc. Even though the MSBR programme was terminated during 1970, research continued up to as long as 1976. India did contribute to this molten salt reactor concept during early period of 1970s. The research activities were basically focused on solubility studies of actinide fluorides in different salt mixtures. After G-IV initiative, MSBR concept has triggered a re-look and research in this direction has also been revived worldwide. Recent conceptual developments on the design of fast neutron

  13. Molten salts and energy related materials.

    Science.gov (United States)

    Fray, Derek

    2016-08-15

    Molten salts have been known for centuries and have been used for the extraction of aluminium for over one hundred years and as high temperature fluxes in metal processing. This and other molten salt routes have gradually become more energy efficient and less polluting, but there have been few major breakthroughs. This paper will explore some recent innovations that could lead to substantial reductions in the energy consumed in metal production and in carbon dioxide production. Another way that molten salts can contribute to an energy efficient world is by creating better high temperature fuel cells and novel high temperature batteries, or by acting as the medium that can create novel materials that can find applications in high energy batteries and other energy saving devices, such as capacitors. Carbonate melts can be used to absorb carbon dioxide, which can be converted into C, CO and carbon nanoparticles. Molten salts can also be used to create black silicon that can absorb more sunlight over a wider range of wavelengths. Overall, there are many opportunities to explore for molten salts to play in an efficient, low carbon world. PMID:27276650

  14. Molten salt converter reactors: from DMSR to SmAHTR

    International Nuclear Information System (INIS)

    Molten salt reactors were developed extensively from the 1950s to 1970s as a thermal breeder alternative on the Thorium-233U cycle. Simplified designs running as fluid fuel converters without salt processing as well as TRISO fueled, salt cooled reactors both hold much promise as potential small modular reactors and as larger base load producers. A background will be presented along with the most likely routes forward for a Canadian development program. (author)

  15. Characters of neutron noise in full-size molten salt reactor

    International Nuclear Information System (INIS)

    Highlights: • The larger system size makes full-size MSR deviate from point kinetic behavior. • The increasing velocity has non-monotonic effect on the effective delayed neutron fraction. • The amplitude of Green’s function at low frequencies is inversely proportional to the effective delayed neutron fraction. • The range of plateau region is smaller due to the more prominent point kinetic effect. - Abstract: In the present paper, the frequency-dependent and space-dependent behavior of the neutron noise in a full-size Molten Salt Reactor (MSR) is investigated. The theoretical models considering the fuel circulation are established based on one-group neutron diffusion theory. Green’s function of the neutron noise induced by a propagating perturbation is introduced with linear noise theory, due to the small perturbation. The equations are numerically calculated by developing a code, in which the eigenfunction expansion method is adopted. The static results show that the effective delayed neutron fraction changes non-monotonically with the increasing fuel velocity. In the dynamic case, the results are compared to those obtained in 1D MSR and a traditional reactor, in order to figure out the effects of both the fuel circulation and the system size. It is found that there is no difference in 1D and 3D MSR systems from the view of fuel circulation, i.e., the fuel circulation enhances the spatial neutronic coupling and leads to the stronger point kinetic effect. The more prominent space-dependent effect founded in 3D traditional reactors is also observed in the MSR, due to the looser neutronic coupling and the unique singularity of Green’s function in the location of the perturbation. Another interesting finding is that Green’s function for low frequencies changes non-monotonically with increasing velocity. The largest magnitude of Green’s function is observed at the velocity where the effective delayed neutron fraction reaches its minimum. Finally, the

  16. Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, Pavel N.; Balanin, Andrey L.; Dudnikov, Anatoly A.; Fomichenko, Petr A.; Nevinitsa, Vladimir A.; Frolov, Aleksey A.; Lubina, Anna S.; Sedov, Aleksey A.; Subbotin, Aleksey S.; Blandinsky, Viktor Yu. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    A subcritical molten salt reactor is proposed for minor actinides (separated from spent fuel VVER-1000 light water reactor) incineration and for {sup 233}U conversion from {sup 232}Th. Here the subcritical molten salt reactor with fuel composition of heavy nuclide fluorides in molten LiF - NaF - KF salt and with external neutron source, based on 1 GeV proton accelerator and molten salt cooled tungsten target is considered. The paper presents the results of parametrical analysis of equilibrium nuclide composition of molten salt reactor with minor actinides feed in dependence of core dimensions, average neutron flux and external neutron source intensity. Reactor design is defined; requirements to external neutron source are posed; heavy nuclides equilibrium and fuel cycle main parameters are calculated.

  17. Design of the distributed control system for HTS molten salt test loop

    International Nuclear Information System (INIS)

    Background: Experimental Physics and Industrial Control System (EPICS) is the distributed control system software which is commonly used in large-scale experimental physics facilities. Purpose: We wish to apply it into the field of molten salt reactor relevant process control, e.g. HTS 1st test loop of thorium-based molten salt reactor (TMSR), which is characterized by heating, circulation, cooling and other process control. Methods: During the development of EPICS based control system, the Simense S7-300 PLC and Yokogawa FA-M3 PLC hardware drivers specification format transformation has been done, to support the Autosave and devIocState plug, and ASYN serial communication between the DTI-1000 reference digital temperature indicator and the EPICS IOC has been developed with the StreamDevice packages, also EDM interface program was modified to support PV control of the 'molten salt flow', 'fans', 'molten salt pump rotation', etc. by dynamic symbols. Results: EPICS based control system achieved the standardized communication of three hardwares with different types, the monitoring and storage of nearly 500 process variables, and the dynamic monitoring of control process. Conclusions: The EPICS-based control system can achieve the molten salt heating, feeding, cooling and other process control of the molten salt test loop. Safety analysis and research of control system requires further efforts to implement. (authors)

  18. Conceptual design of Indian molten salt breeder reactor

    Indian Academy of Sciences (India)

    P K Vijayan; A Basak; I V Dulera; K K Vaze; S Basu; R K Sinha

    2015-09-01

    The third stage of Indian nuclear power programme envisages the use of thorium as the fertile material with 233U, which would be obtained from the operation of Pu/Th-based fast reactors in the later part of the second stage. Thorium-based reactors have been designed in many configurations, from light water-cooled designs to high-temperature liquid metal-cooled options. Another option, which holds promise, is the molten salt-fuelled reactor, which can be configured to give significant breeding ratios. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian molten salt breeder reactor (IMSBR). Presently, various design options and possibilities are being studied from the point of view of reactor physics and thermal hydraulic design. In parallel, fundamental studies on natural circulation and corrosion behaviour of various molten salts have also been initiated.

  19. Review of R and D of thorium molten-salt reactor

    International Nuclear Information System (INIS)

    Nuclear power will be used continuously as low-carbon energy for a sustainable society. On the other hand, prime attention is inclined to thorium molten-salt reactor (MSR) in recent years due to necessity for radioactive waste disposal, nuclear non-proliferation ('world without nuclear weapon') and higher safety of nuclear power. Thorium MSR is a type of liquid-fueled reactor utilizing thorium as fertile with some fissile materials contained in molten-salt. Research and development (R and D) of MSR started in the 1950s at Oak-Ridge National Laboratory (ORNL) for aircraft reactors. R and D activities were later extended to develop civilian power reactors using thorium, a concept which was completed in the 1970s. Nonetheless, technical problems still remained, such as temperature reactivity coefficient, material corrosion, and tritium permeation at heat exchanger. Since the 1980s several R and D activities to circumvent these problems have been done inside and outside ORNL. The problems relating to thorium MSR and recent R and D activities will be introduced in this paper. (author)

  20. Neutronics Assessment of Molten Salt Breeding Blanket Design Options

    International Nuclear Information System (INIS)

    Neutronics assessment has been performed for molten salt breeding blanket design options that can be utilized in fusion power plants. The concepts evaluated are a self-cooled Flinabe blanket with Be multiplier and dual-coolant blankets with He-cooled FW and structure. Three different molten salts were considered including the high melting point Flibe, a low melting point Flibe, and Flinabe. The same TBR can be achieved with a thinner self-cooled blanket compared to the dual-coolant blanket. A thicker Be zone is required in designs with Flinabe. The overall TBR will be ∼1.07 based on 3-D calculations without breeding in the divertor region. Using Be yields higher blanket energy multiplication than obtainable with Pb. A modest amount of tritium is produced in the Be (∼3 kg) over the blanket lifetime of ∼3 FPY. Using He gas in the dual-coolant blanket results in about a factor of 2 lower blanket shielding effectiveness. We show that it is possible to ensure that the shield is a lifetime component, the vacuum vessel is reweldable, and the magnets are adequately shielded. We conclude that molten salt blankets can be designed for fusion power plants with neutronics requirements such as adequate tritium breeding and shielding being satisfied

  1. Al/Cl2 molten salt battery

    Science.gov (United States)

    Giner, J.

    1972-01-01

    Molten salt battery has been developed with theoretical energy density of 5.2 j/kg (650 W-h/lb). Battery, which operates at 150 C, can be used in primary mode or as rechargeable battery. Battery has aluminum anode and chlorine cathode. Electrolyte is mixture of AlCl3, NaCl, and some alkali metal halide such as KCl.

  2. Surface functionalization by molten salt electrolytic processes

    International Nuclear Information System (INIS)

    The attention has been paid to surface functionalization by molten salt electrolytic processes. Three topics on the experimental results obtained by the authors are described: the electrochemical formation of zirconium metal film and zirconium alloy film on ceramic, surface nitriding of titanium by electrochemical process and an anodic oxide film formation on nickel. (author)

  3. Preliminary Study for Inventories of Minor Actinides in Thorium Molten Salt Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Choong Wie; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    It has different characteristic with the conventional reactors which use a solid fuel. It can continually supply the fuel by online refueling and reprocessing of minor actinides so that those can be separated and eliminated from the reactor. The MSR maintains steady state except initial stage and the reactor becomes stable. In this research, considering online refueling, bubbling and reprocessing, the basic concept for evaluation of the inventory of minor actinide in the molten salt reactor is driven using the Bateman equation. The simulation results, where REM and MCNP code from CNRS (Centre National de la Recherche Scientifique) applied to the concept equation are analyzed. The analysis of the basic concept was carried out for evaluation of the inventory of the minor actinides in MSR. It was thought that the inventories of the minor actinides should be evaluated by solving the modified Bateman equation due to the MSR characteristic of online refueling, chemical reprocessing and bubbling.

  4. Preliminary Study for Inventories of Minor Actinides in Thorium Molten Salt Reactor

    International Nuclear Information System (INIS)

    It has different characteristic with the conventional reactors which use a solid fuel. It can continually supply the fuel by online refueling and reprocessing of minor actinides so that those can be separated and eliminated from the reactor. The MSR maintains steady state except initial stage and the reactor becomes stable. In this research, considering online refueling, bubbling and reprocessing, the basic concept for evaluation of the inventory of minor actinide in the molten salt reactor is driven using the Bateman equation. The simulation results, where REM and MCNP code from CNRS (Centre National de la Recherche Scientifique) applied to the concept equation are analyzed. The analysis of the basic concept was carried out for evaluation of the inventory of the minor actinides in MSR. It was thought that the inventories of the minor actinides should be evaluated by solving the modified Bateman equation due to the MSR characteristic of online refueling, chemical reprocessing and bubbling

  5. Preparation of biomorphic silicon carbide–mullite ceramics using molten salt synthesis

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Wei, E-mail: wwchem@126.com [Department of Chemical Engineering, College of Environment Science and Engineering, Key Laboratory of Subsurface Hydrology and Ecology in Arid Areas, Ministry of Education, Chang’an University, 126# Yanta Road, Xi' an 710054, Shaanxi (China); Hou, Guangya [College of Chemical Engineering and Materials Science, Zhejiang University of Technology, Hangzhou 310014 (China); Wang, Boya; Deng, Shunxi [Department of Chemical Engineering, College of Environment Science and Engineering, Key Laboratory of Subsurface Hydrology and Ecology in Arid Areas, Ministry of Education, Chang’an University, 126# Yanta Road, Xi' an 710054, Shaanxi (China)

    2014-09-15

    Biomorphic silicon carbide–mullite ceramics were prepared from beech wood using liquid Si infiltration and molten salts synthesis. The resulting mullite whiskers coating, as well as the growth mechanism in molten Al{sub 2}(SO{sub 4}){sub 3}–Na{sub 2}SO{sub 4} environment, have been investigated using scanning electron microscopy (SEM), X-ray diffraction (XRD), thermogravimetric analysis (TGA) and Fourier transform infrared spectroscopy (FTIR) techniques. The biomorphic SiC ceramics derived from the beech wood template have coarse pore walls consisting of β-SiC grains with diameters ranging from 5 μm to 20 μm. After the molten salts reactions between biomorphic SiC substrate and mixture molten salts (Al{sub 2}(SO{sub 4}){sub 3}–Na{sub 2}SO{sub 4}), porous Silicon carbide–mullite ceramics with cilia-like microstructure were obtained. This unique structure has potential application in hot gases filters. An oxidation–dissolution cycle was proposed to explain the mullite whiskers growth in molten salts environment. - Graphical abstract: Biomorphic silicon carbide–mullite ceramics with cilia-like microstructure prepared from beech wood using liquid Si infiltration (LSI) and molten salts reactions (MSR) processes. Mullite whiskers with nanometer-sized diameters and micrometer-sized lengths grow on the surface of SiC substrate, and the biomorphic silicon carbide–mullite ceramics inherit the porous microstructure originated from biomorphic SiC ceramics and beech wood. The mullite whiskers grow on the pores' surface of biomorphic SiC to form cilia-like surface, and this special structure can be used for hot gases filter. - Highlights: • Biomorphic silicon carbide–mullite ceramics were prepared. • An oxidation–dissolution mechanism was proposed to explain the coating formation. • The unique structure has potential application in hot gases filter.

  6. Preparation of biomorphic silicon carbide–mullite ceramics using molten salt synthesis

    International Nuclear Information System (INIS)

    Biomorphic silicon carbide–mullite ceramics were prepared from beech wood using liquid Si infiltration and molten salts synthesis. The resulting mullite whiskers coating, as well as the growth mechanism in molten Al2(SO4)3–Na2SO4 environment, have been investigated using scanning electron microscopy (SEM), X-ray diffraction (XRD), thermogravimetric analysis (TGA) and Fourier transform infrared spectroscopy (FTIR) techniques. The biomorphic SiC ceramics derived from the beech wood template have coarse pore walls consisting of β-SiC grains with diameters ranging from 5 μm to 20 μm. After the molten salts reactions between biomorphic SiC substrate and mixture molten salts (Al2(SO4)3–Na2SO4), porous Silicon carbide–mullite ceramics with cilia-like microstructure were obtained. This unique structure has potential application in hot gases filters. An oxidation–dissolution cycle was proposed to explain the mullite whiskers growth in molten salts environment. - Graphical abstract: Biomorphic silicon carbide–mullite ceramics with cilia-like microstructure prepared from beech wood using liquid Si infiltration (LSI) and molten salts reactions (MSR) processes. Mullite whiskers with nanometer-sized diameters and micrometer-sized lengths grow on the surface of SiC substrate, and the biomorphic silicon carbide–mullite ceramics inherit the porous microstructure originated from biomorphic SiC ceramics and beech wood. The mullite whiskers grow on the pores' surface of biomorphic SiC to form cilia-like surface, and this special structure can be used for hot gases filter. - Highlights: • Biomorphic silicon carbide–mullite ceramics were prepared. • An oxidation–dissolution mechanism was proposed to explain the coating formation. • The unique structure has potential application in hot gases filter

  7. Characterisations of HVOF sprayed NiCrBSi coatings on Ni- and Fe-based superalloys and evaluation of cyclic oxidation behaviour of some Ni-based superalloys in molten salt environment

    International Nuclear Information System (INIS)

    Microstructure plays a predominant role in determining material behaviour. Increasing microstructure uniformity has long been considered a fruitful means of improving thermal, chemical and mechanical properties of the materials. High velocity oxy-fuel (HVOF) is one of the emerging technologies among the thermal spraying techniques, for producing uniform and dense coatings, having high hardness and good adhesion values. In this study, HVOF technique was used to deposit NiCrBSi coatings, approximately 250-300 μm thick, on the Ni- and Fe-based superalloys for hot corrosion applications. The coatings were characterised in relation to coating thickness, porosity, microhardness and microstructure. The hot corrosion behaviour of the coatings deposited on nickel-based superalloys after exposure to molten salt (Na2SO4-60% V2O5) at 900 deg. C under cyclic conditions was also studied. The techniques used in the present investigation include X-ray diffraction, optical microscopy, scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDAX) and electron probe microanalysis (EPMA). The thermogravimetric technique was used to establish kinetics of corrosion. The structure of the as sprayed NiCrBSi coating mainly consisted of γ-nickel solid solution containing small fraction of Cr7C3 and Ni3B phases. Very weak peaks of NiCr2O4 spinel oxides were also formed during spraying of the coatings. Some porosity (less than 1.4%) and inclusions were observed in the structure of the coatings. Coating microhardness values were found to be in the range of 750-930 Hv (Vickers Hardness) on different substrates. The NiCrBSi coating was found to be very effective in decreasing the corrosion rate in the given molten salt environment at 900 deg. C. The hot corrosion resistance imparted by NiCrBSi coatings may be attributed to the formation of oxides of silicon, chromium, nickel and spinels of nickel and chromium

  8. Chemistry and technology of Molten Salt Reactors - history and perspectives

    International Nuclear Information System (INIS)

    Molten Salt Reactors represent one of promising future nuclear reactor concept included also in the Generation IV reactors family. This reactor type is distinguished by an extraordinarily close connection between the reactor physics and chemical technology, which is given by the specific features of the chemical form of fuel, representing by molten fluoride salt and circulating through the reactor core and also by the requirements of continuous 'on-line' reprocessing of the spent fuel. The history of Molten Salt Reactors reaches the period of fifties and sixties, when the first experimental Molten Salt Reactors were constructed and tested in ORNL (US). Several molten salt techniques dedicated to fresh molten salt fuel processing and spent fuel reprocessing were studied and developed in those days. Today, after nearly thirty years of discontinuance, a renewed interest in the Molten Salt Reactor technology is observed. Current experimental R and D activities in the area of Molten Salt Reactor technology are realized by a relatively small number of research institutions mainly in the EU, Russia and USA. The main effort is directed primarily to the development of separation processes suitable for the molten salt fuel processing and reprocessing technology. The techniques under development are molten salt/liquid metal extraction processes, electrochemical separation processes from the molten salt media, fused salt volatilization techniques and gas extraction from the molten salt medium

  9. Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

    Institute of Scientific and Technical Information of China (English)

    ZHANG Da-Lin; QIU Sui-Zheng; LIU Chang-Liang; SU Guang-Hui

    2008-01-01

    The Molten Salt Reactor (MSR),one of the‘Generation Ⅳ'concepts,is a liquid-fuel reactor,which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt.The study on its neutronice considering the fuel salt flow,which is the base of the thermal-hydraulic calculation and safety analysis,must be done.In this paper,the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method.The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes,and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method,and the discretization equations are computed by the source iteration method.The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained.The numerical calculated results show that,the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor;however,it affects the distribution of the delayed neutron precursors significantly,especially the long-lived one.In addition,it could be found that the delayed neutron precursors influence the nentronics slightly under the steady condition.

  10. Physical and chemical feasibility of fueling molten salt reactors with TRU's trifluorides

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V.; Feinberg, O.; Konakov, S.; Subbotine, S.; Surenkov, A.; Zakirov, R. [Kurchatov Institute, RRC, Moscow (Russian Federation)

    2001-07-01

    The molten salt reactor (MSR) concept is very important for consideration as an element of future nuclear energy systems. These reactor systems are unique in many ways. Particularly, the MSRs appear to have substantial promise not only as advanced TRU free system operating in U-Th cycle, but also as transmuter of TRU. Physical and chemical feasibility of fueling MSR with TRU trifluorides is examined. Solvent compositions with and without U-Th as fissile / fertile addition are considered. The principle reactor and fuel cycle variables available for optimizing the performance of MSR as TRU transmuting system are discussed. These efforts led to the definition in minimal TRU mass flow rate, reduced total losses to waste and maximum possible burn up rate for the molten salt transmuter. The current status of technology and prospects for revisited interest are summarized. Significant chemical problems are remain to be resolved at the end of prior MSRs programs, notably, graphite life durability, tritium control, fate of noble metal fission products. Questions arising from plutonium and minor actinide fueling include: corrosion and container chemistry, new redox buffer for systems without uranium, analytical chemistry instrumentation, adequate constituent solubilities, suitable fuel processing and waste form development. However these problems appear to be soluble. (author)

  11. Molten-Salt Depleted-Uranium Reactor

    OpenAIRE

    Dong, Bao-Guo; Dong, Pei; Gu, Ji-Yuan

    2015-01-01

    The supercritical, reactor core melting and nuclear fuel leaking accidents have troubled fission reactors for decades, and greatly limit their extensive applications. Now these troubles are still open. Here we first show a possible perfect reactor, Molten-Salt Depleted-Uranium Reactor which is no above accident trouble. We found this reactor could be realized in practical applications in terms of all of the scientific principle, principle of operation, technology, and engineering. Our results...

  12. Electromagnetic process to purify nuclear molten salts

    OpenAIRE

    Mestre Molist, Marta

    2011-01-01

    The separation between actinides and fission products could be realized by a pyrometallurgical liquid/liquid extraction process. Thanks to an oxide-reductive reaction, a mass exchange occurs at the interface between a polluted molten salt layer and a containment loquid metal layer. The aims of this project is to develop the diffusion process which occurs and the thermodynamics process associated to this phenomena. This work consist in two parts: -An experimental prototype industrial type inst...

  13. Molten salt battery having inorganic paper separator

    Science.gov (United States)

    Walker, Jr., Robert D.

    1977-01-01

    A high temperature secondary battery comprises an anode containing lithium, a cathode containing a chalcogen or chalcogenide, a molten salt electrolyte containing lithium ions, and a separator comprising a porous sheet comprising a homogenous mixture of 2-20 wt.% chrysotile asbestos fibers and the remainder inorganic material non-reactive with the battery components. The non-reactive material is present as fibers, powder, or a fiber-powder mixture.

  14. Development of structural materials to enable the electrochemical reduction of spent oxide nuclear fuel in a molten salt electrolyte

    International Nuclear Information System (INIS)

    For the development of the advanced spent fuel management process based on the molten salt technology, it is essential to choose the optimum material for the process equipment handling a molten salt. In this study, corrosion behavior of Fe-base superalloy, Ni-base superalloy, non-metallic material and surface modified superalloy were investigated in the hot molten salt under oxidation atmosphere. These experimental data will suggest a guideline for the selection of corrosion resistant materials and help to find the operation criteria of each equipment in aspects of high temperature characteristics and corrosion retardation

  15. Thermal Characterization of Molten Salt Systems

    Energy Technology Data Exchange (ETDEWEB)

    Toni Y. Gutknecht; Guy L. Fredrickson

    2011-09-01

    The phase stability of molten salts in an electrorefiner (ER) may be adversely affected by the buildup of sodium, fission products, and transuranics in the electrolyte. Potential situations that need to be avoided are the following: (1) salt freezing due to an unexpected change in the liquidus temperature, (2) phase separation or non-homogeneity of the molten salt due to the precipitation of solids or formation of immiscible liquids, and (3) any mechanism that can result in the separation and concentration of fissile elements from the molten salt. Any of these situations would result in an off-normal condition outside the established safety basis for electrorefiner (ER) operations. The stability (and homogeneity) of the phases can potentially be monitored through the thermal characterization of the salts, which can be a function of impurity concentration. This report describes the experimental results of typical salts compositions, which consist of chlorides of potassium, lithium, strontium, samarium, praseodymium, lanthanum, barium, cerium, cesium, neodymium, sodium and gadolinium chlorides as a surrogate for both uranium and plutonium, used for the processing of used nuclear fuels.

  16. Candidate molten salt investigation for an accelerator driven subcritical core

    OpenAIRE

    SOOBY Elizabeth; Baty, Austin; BENES ONDREJ; McIntyre, Peter; Pogue, Nathaniel; Salanne, Mathieu; Sattarov, Akhdiyor

    2013-01-01

    We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. Molecular dynamics (MD) codes have been used to estimate properties of the molt...

  17. Renewable energy and the role of molten salts and carbon

    Directory of Open Access Journals (Sweden)

    Fray D.

    2013-01-01

    Full Text Available Molten carbonate fuel cells have been under development for a number of years and reliable units are successfully working at 250kW scale and demonstration units have produced up to 2 MW. Although these cells cannot be considered as renewable as the fuel, hydrogen or carbon monoxide is consumed and not regenerated, the excellent reliability of such a cell can act as a stimulus to innovative development of similar cells with different outcomes. Molten salt electrolytes based upon LiCl - Li2O can be used to convert carbon dioxide, either drawn from the output of a conventional thermal power station or from the atmosphere, to carbon monoxide or carbon. Recently, dimensionally stable anodes have been developed for molten salt electrolytes, based upon alkali or alkaline ruthenates which are highly electronically conducting and these may allow the concept of high temperature batteries to be developed in which an alkali or alkaline earth element reacts with air to form oxides when the battery is discharging and the oxide decomposes when the battery is being recharged. Batteries using these concepts may be based upon the Hall-Heroult cell, which is used worldwide for the production of aluminium on an industrial scale, and could be used for load levelling. Lithium ion batteries are, at present, the preferred energy source for cars in 2050 as there are sufficient lithium reserves to satisfy the world’s energy needs for this particular application. Graphite is used in lithium ion batteries as the anode but the capacity is relatively low. Silicon and tin have much higher capacities and the use of these materials, encapsulated in carbon nanotubes and nanoparticles will be described. This paper will review these interesting developments and demonstrate that a combination of carbon and molten salts can offer novel ways of storing energy and converting carbon dioxide into useful products.

  18. Renewable energy and the role of molten salts and carbon

    OpenAIRE

    Fray D.

    2013-01-01

    Molten carbonate fuel cells have been under development for a number of years and reliable units are successfully working at 250kW scale and demonstration units have produced up to 2 MW. Although these cells cannot be considered as renewable as the fuel, hydrogen or carbon monoxide is consumed and not regenerated, the excellent reliability of such a cell can act as a stimulus to innovative development of similar cells with different outcomes. Molten salt electrolytes based upon LiCl - L...

  19. On an optimized neutron shielding for an advanced molten salt fast reactor design

    International Nuclear Information System (INIS)

    The molten salt reactor technology has gained renewed interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner core vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all core internal structures. On the basis of this new geometry a model for neutron physics calculation is presented and applied for a shielding optimization. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system has to be significantly increased and will finally be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem. (author)

  20. Helium-cooled molten-salt fusion breeder

    International Nuclear Information System (INIS)

    We present a new conceptual design for a fusion reactor blanket that is intended to produce fissile material for fission power plants. Fast fission is suppressed by using beryllium instead of uranium to multiply neutrons. Thermal fission is suppressed by minimizing the fissile inventory. The molten-salt breeding medium (LiF + BeF2 + ThF4) is circulated through the blanket and to the on-line processing system where 233U and tritium are continuously removed. Helium cools the blanket and the austenitic steel tubes that contain the molten salt. Austenitic steel was chosen because of its ease of fabrication, adequate radiation-damage lifetime, and low corrosion by molten salt. We estimate that a breeder having 3000 MW of fusion power will produce 6500 kg of 233U per year. This amount is enough to provide makeup for 20 GWe of light-water reactors per year or twice that many high-temperature gas-cooled reactors or Canadian heavy-water reactors. Safety is enhanced because the afterheat is low and blanket materials do not react with air or water. The fusion breeder based on a pre-MARS tandem mirror is estimated to cost $4.9B or 2.35 times a light-water reactor of the same power. The estimated cost of the 233U produced is $40/g for fusion plants costing 2.35 times that of a light-water reactor if utility owned or $16/g if government owned

  1. Development of computer code packages for molten salt reactor core analysis

    International Nuclear Information System (INIS)

    This paper presents the implementations of the Oak Ridge National Laboratory (ORNL) approach for Molten Salt Reactor (MSR) core analysis with two nuclear reactor core analysis computer code systems. The first code system has been set up with the MCNP6 Monte Carlo code, its depletion module CINDER90 and the PYTHON script language. The second code system has been set up with the NEWT transport calculation module and ORIGEN depletion module connected by TRITON sequence in SCALE code, and the PYTHON script language. The PYTHON script language is used for implementing the online reprocessing of molten-salt fuel, and feeding new fertile material in the computer code simulations. In this paper, simplified nuclear reactor core models of a Molten Salt Breeder Reactor (MSBR), designed by ORNL in the 1960's, and FUJI-U3 designed by Toyohashi University of Technology (TUT) in the 2000's, were analyzed by the two code systems. Using these, various reactor design parameters of the MSRs were compared, such as the multiplication factor, breeding ratio, amount of material, total feeding, neutron flux distribution, and temperature coefficient. (author)

  2. The effects of core zoning on optimization of design analysis of molten salt reactor

    International Nuclear Information System (INIS)

    Highlights: • 1/8 of core is simulated by MCNP and thermal-hydraulic code simultaneously. • Effects of core zoning are studied by dividing the core into two regions. • Both the neutronics and thermal-hydraulic behavior are investigated. • The flat flux distribution is achieved in the optimization analysis. • The flat flux can lead to worse thermal-hydraulic behavior occasionally. - Abstract: The molten salt reactor (MSR) is one of six advanced reactor types in the frame of the Generation 4 International Forum. In this study, a multiple-channel analysis code (MAC) is developed to analyze thermal-hydraulics behavior and MCNP4c is used to study the neutronics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, namely temperature distribution, flow distribution and pressure drop. The MCNP4c performs the analysis of effective multiplication factor, neutron flux, power distribution and conversion ratio. In this work, the modification of core configuration is achieved by different core zoning and various fuel channel diameters, contributing to flat flux distribution. Specifically, the core is divided into two regions and the effects of different core zoning on the both neutronics and thermal-hydraulic behavior of moderated molten salt reactor are investigated. We conclude that the flat flux distribution cannot always guarantee better performance in thermal-hydraulic perspective and can decreases the graphite lifetime significantly

  3. Use of thorium in the generation IV Molten Salt reactors and perspectives for Brazil

    International Nuclear Information System (INIS)

    Interest in thorium stems mainly from the fact that it is expected a substantial increase in uranium prices over the next fifty years. The reactors currently in operation consume 65,500 tons of uranium per year. Each electrical gigawatt (GWe) additional need about 200 tU mined per year. So advanced fuel cycles, which increase the reserves of nuclear materials are interesting, particularly the use of thorium to produce the fissile isotope 233U. It is important to mention some thorium advantages. Thorium is three to five times more abundant than uranium in the earth's crust. Thorium has only one oxidation state. Additionally, thoria produces less radiotoxicity than the UO2 because it produces fewer amounts of actinides, reducing the radiotoxicity of long life nuclear waste. ThO2 has higher corrosion resistance than UO2, besides being chemically stable due to their low water solubility. The burning of Pu in a reactor based in thorium also decreases the inventories of Pu from the current fuel cycles, resulting in lower risks of material diversion for use in nuclear weapons. There are some ongoing projects in the world, taking into consideration the proposed goals for Generation IV reactors, namely: sustainability, economics, safety and reliability, proliferation resistance and physical protection. Some developments on the use of thorium in reactors are underway, with the support of the IAEA and some governs. Can be highlighted some reactor concepts using thorium as fuel: CANDU; ADTR -Accelerator Driven Thorium Reactor; AHWR -Advanced Heavy Water Reactor proposed by India (light water cooled and moderated by heavy water) and the MSR -Molten Salt Reactor. The latter is based on a reactor concept that has already been successfully tested in the U.S. in the 50s, for use in aircrafts. In this paper, we discuss the future importance of thorium, particularly for Brazil, which has large mineral reserves of this strategic element, the characteristics of the molten salt reactor

  4. Fast molten salt reactor-transmuter for closing nuclear fuel cycle on minor actinides

    International Nuclear Information System (INIS)

    Creation fast critical molten salt reactor for burning-out minor actinides and separate long-living fission products in the closed nuclear fuel cycle is the most perspective and actual direction. The reactor on melts salts - molten salt homogeneous reactor with the circulating fuel, working as burner and transmuter long-living radioactive nuclides in closed nuclear fuel cycle, can serve as an effective ecological cordon from contamination of the nature long-living radiotoxic nuclides. High-flux fast critical molten-salt nuclear reactors in structure of the closed nuclear fuel cycle of the future nuclear power can effectively burning-out / transmute dangerous long-living radioactive nuclides, make radioisotopes, partially utilize plutonium and produce thermal and electric energy. Such reactor allows solving the problems constraining development of large-scale nuclear power, including fueling, minimization of radioactive waste and non-proliferation. Burning minor actinides in molten salt reactor is capable to facilitate work solid fuel power reactors in system NP with the closed nuclear fuel cycle and to reduce transient losses at processing and fabrications fuel pins. At substantiation MSR-transmuter/burner as solvents fuel nuclides for molten-salt reactors various salts were examined, for example: LiF - BeF2; NaF - LiF - BeF2; NaF-LiF ; NaF-ZrF4 ; LiF-NaF -KF; NaCl. RRC 'Kurchatov institute' together with other employees have developed the basic design reactor installations with molten salt reactor - burner long-living nuclides for fluoride fuel composition with the limited solubility minor actinides (MAF3 10 mol %) allows to develop in some times more effective molten salt reactor with fast neutron spectrum - burner/ transmuter of the long-living radioactive waste. In high-flux fast reactors on melts salts within a year it is possible to burn ∼300 kg minor actinides per 1 GW thermal power of reactor. The technical and economic estimation given power

  5. The synthesis of cementitious compounds in molten salts

    OpenAIRE

    Sheikh, R. A.

    2016-01-01

    This thesis describes an investigation into the synthesis of cementitious compounds in molten salts. These compounds are produced in energy-intensive industries (EIIs), such as the cement process, and are responsible for emitting significant quantities of carbon dioxide (CO2) emissions. Molten salt synthesis (MSS) involves dissolving compounds in a molten salt and reacting in solution. If the MSS of cementitious compounds can occur at lower temperatures than EIIs, this could lead to fewer qua...

  6. Conceptual design of loop-in-tank type Indian molten salt breeder reactor concept

    International Nuclear Information System (INIS)

    The third stage of Indian nuclear power programme envisages use of thorium as fertile material with 233U, which is proposed to be obtained from reprocessing of spent fuel of Pu/Th based fast reactors in the later part of the second stage of the programme. In India, thorium based reactors have been designed in many configurations, from light water cooled designs to high temperature liquid metal and molten salt cooled options. Another option, which holds promise, is the molten salt-fuelled reactor, which can be configured to give significant breeding ratios. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian Molten Salt Breeder Reactor (IMSBR). (author)

  7. Accelerator molten-salt breeding and thorium fuel cycle

    International Nuclear Information System (INIS)

    The recent efforts at the development of fission energy utilization have not been successful in establishing fully rational technology. A new philosophy should be established on the basis of the following three principles: (1) thorium utilization, (2) molten-salt fuel concept, and (3) separation of fissile-breeding and power-generating functions. Such philosophy is called 'Thorium Molten-Salt Nuclear Energy Synergetics [THORIMS-NES]'. The present report first addresses the establishment of 233U breeding fuel cycle, focusing on major features of the Breeding and Chemical Processing Centers and a small molten-salt power station (called FUJI-II). The development of fissile producing breeders is discussed in relation to accelerator molten-salt breeder (AMSB), impact fusion molten-salt breeder, and inertial-confined fusion hybrid molten-salt breeder. Features of the accelerator molten-salt breeder are described, focusing on technical problems with accelerator breeders (or spallators), design principle of the accelerator molten-salt breeder, selection of molten salt compositions, and nuclear- and reactor-chemical aspects of AMSB. Discussion is also made of further research and development efforts required in the future for AMSB. (N.K.)

  8. Experimental studies of actinides in molten salts

    Energy Technology Data Exchange (ETDEWEB)

    Reavis, J.G.

    1985-06-01

    This review stresses techniques used in studies of molten salts containing multigram amounts of actinides exhibiting intense alpha activity but little or no penetrating gamma radiation. The preponderance of studies have used halides because oxygen-containing actinide compounds (other than oxides) are generally unstable at high temperatures. Topics discussed here include special enclosures, materials problems, preparation and purification of actinide elements and compounds, and measurements of various properties of the molten volts. Property measurements discussed are phase relationships, vapor pressure, density, viscosity, absorption spectra, electromotive force, and conductance. 188 refs., 17 figs., 6 tabs.

  9. Experimental studies of actinides in molten salts

    International Nuclear Information System (INIS)

    This review stresses techniques used in studies of molten salts containing multigram amounts of actinides exhibiting intense alpha activity but little or no penetrating gamma radiation. The preponderance of studies have used halides because oxygen-containing actinide compounds (other than oxides) are generally unstable at high temperatures. Topics discussed here include special enclosures, materials problems, preparation and purification of actinide elements and compounds, and measurements of various properties of the molten volts. Property measurements discussed are phase relationships, vapor pressure, density, viscosity, absorption spectra, electromotive force, and conductance. 188 refs., 17 figs., 6 tabs

  10. Preliminary study on weapon grade uranium utilization in molten salt reactor miniFUJI

    Energy Technology Data Exchange (ETDEWEB)

    Aji, Indarta Kuncoro [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Waris, A., E-mail: awaris@fi.itb.ac.id [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa No. 10 Bandung 40132 (Indonesia)

    2014-09-30

    Preliminary study on weapon grade uranium utilization in 25MWth and 50MWth of miniFUJI MSR (molten salt reactor) has been carried out. In this study, a very high enriched uranium that we called weapon grade uranium has been employed in UF{sub 4} composition. The {sup 235}U enrichment is 90 - 95 %. The results show that the 25MWth miniFUJI MSR can get its criticality condition for 1.56 %, 1.76%, and 1.96% of UF{sub 4} with {sup 235}U enrichment of at least 93%, 90%, and 90%, respectively. In contrast, the 50 MWth miniFUJI reactor can be critical for 1.96% of UF{sub 4} with {sup 235}U enrichment of at smallest amount 95%. The neutron spectra are almost similar for each power output.

  11. Numerical simulation and experimental investigation of natural convection heat transfer of molten salt around fine wire

    Institute of Scientific and Technical Information of China (English)

    LU; YuanWei; LI; XiaoLi; LI; Qiang; WU; YuTing; MA; ChongFang

    2013-01-01

    In order to get the natural convection heat transfer mechanism of molten salt, the experimental investigation of natural convective heat transfer of LiNO3was studied after it was simulated by numerical calculation. Experiment was carried out on the nat-ural convection heat transfer of air and water around the fine wire using the method of Joule heating. The results showed that the natural convection heat transfer of air and water around the fine wire agreed well with Fand’s correlation. Based on the aforementioned experiment, the natural convection heat transfer of molten salt LiNO3was studied by experiment and the same results were got. Therefore, the natural convection heat transfer of molten salt can be calculated by Fand’s correlation, which takes into consideration the effect of viscosity dissipation.

  12. Transient simulation of molten salt central receiver

    Science.gov (United States)

    Doupis, Dimitri; Wang, Chuan; Carcorze-Soto, Jorge; Chen, Yen-Ming; Maggi, Andrea; Losito, Matteo; Clark, Michael

    2016-05-01

    Alstom is developing concentrated solar power (CSP) utilizing 60/40wt% NaNO3-KNO3 molten salt as the working fluid in a tower receiver for the global renewable energy market. In the CSP power generation cycle, receivers undergo a daily cyclic operation due to the transient nature of solar energy. Development of robust and efficient start-up and shut-down procedures is critical to avoiding component failures due to mechanical fatigue resulting from thermal transients, thus maintaining the performance and availability of the CSP plant. The Molten Salt Central Receiver (MSCR) is subject to thermal transients during normal daily operation, a cycle that includes warmup, filling, operation, draining, and shutdown. This paper describes a study to leverage dynamic simulation and finite element analysis (FEA) in development of start-up, shutdown, and transient operation concepts for the MSCR. The results of the FEA also verify the robustness of the MSCR design to the thermal transients anticipated during the operation of the plant.

  13. Heterogeneous structure effect on molten salt blanket neutronics

    Energy Technology Data Exchange (ETDEWEB)

    Grebyonkin, K.F.; Kandiev, Ya.Z.; Malyshkin, G.N.; Orlov, A.I. [Inst. of Technical Pysics, Chelyabinsk (Russian Federation). Dept. of Physics

    1997-09-01

    The report presents the results of the molten salt blanket neutronics calculations performed for researchers of a facility for accelerator-driven transmutation of long-lived radioactive wastes and plutonium conversion. Heterogeneous structure effect on molten salt blanket neutronics was studied through computation. 4 refs., 1 fig., 1 tab.

  14. An inventory analysis of thermal-spectrum thorium-fueled molten salt reactor concepts

    International Nuclear Information System (INIS)

    Inventory analyses of thermal-spectrum, thorium-fueled molten salt reactors (MSRs) have been performed to support US Department of Energy fuel cycle screening and evaluation activities within the. A single-fluid, single-zone 2250 MWth (1000 MWe) MSR concept with a fuel-bearing molten fluoride salt moderated by graphite was used as the basis for this work. Depletion calculations were performed using SCALE 6.1.1 with ENDF/B-VII.0 nuclear data. Equilibrium conditions were evaluated for several design parameter sets using a methodology developed at Oak Ridge National Laboratory (ORNL) that enables MSR analysis by performing multiple SCALE/TRITON depletion calculations with material flow modeling calculations between time steps. Adequate modeling approximations were identified by comparing results obtained from calculations that used different modeling choices and levels of fidelity. Parametric analyses examined the performance sensitivity of a thorium MSR to different separations approaches and elemental removal efficiencies. Finally, an inventory analysis for a thorium-fueled MSR with full recycling demonstrated how these insights can be applied and showed that such a system appears feasible from a mass flow and reactivity basis. (author)

  15. Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

    International Nuclear Information System (INIS)

    Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor

  16. Molten-Salt Depleted-Uranium Reactor

    CERN Document Server

    Dong, Bao-Guo; Gu, Ji-Yuan

    2015-01-01

    The supercritical, reactor core melting and nuclear fuel leaking accidents have troubled fission reactors for decades, and greatly limit their extensive applications. Now these troubles are still open. Here we first show a possible perfect reactor, Molten-Salt Depleted-Uranium Reactor which is no above accident trouble. We found this reactor could be realized in practical applications in terms of all of the scientific principle, principle of operation, technology, and engineering. Our results demonstrate how these reactors can possess and realize extraordinary excellent characteristics, no prompt critical, long-term safe and stable operation with negative feedback, closed uranium-plutonium cycle chain within the vessel, normal operation only with depleted-uranium, and depleted-uranium high burnup in reality, to realize with fission nuclear energy sufficiently satisfying humanity long-term energy resource needs, as well as thoroughly solve the challenges of nuclear criticality safety, uranium resource insuffic...

  17. Molten salts database for energy applications

    CERN Document Server

    Serrano-López, Roberto; Cuesta-López, Santiago

    2013-01-01

    The growing interest in energy applications of molten salts is justified by several of their properties. Their possibilities of usage as a coolant, heat transfer fluid or heat storage substrate, require thermo-hydrodynamic refined calculations. Many researchers are using simulation techniques, such as Computational Fluid Dynamics (CFD) for their projects or conceptual designs. The aim of this work is providing a review of basic properties (density, viscosity, thermal conductivity and heat capacity) of the most common and referred salt mixtures. After checking data, tabulated and graphical outputs are given in order to offer the most suitable available values to be used as input parameters for other calculations or simulations. The reviewed values show a general scattering in characterization, mainly in thermal properties. This disagreement suggests that, in several cases, new studies must be started (and even new measurement techniques should be developed) to obtain accurate values.

  18. Development of viscometers for molten salts

    International Nuclear Information System (INIS)

    Viscometers specially designed for molten salts were made. One is a oscillating cup type and the other is a capillary type. In the case of the oscillating cup viscometer, the viscosity is determined absolutely through the period and the logarithmic decrement of oscillation with other physical parameters. The period and the logarithmic decrement are calculated from the time intervals between two photo-detectors' intercepts of the reflected laser beam. The capillary viscometer used is made of quartz and the sample is sealed under vacuum, which is placed in a transparent furnace. Efflux time is measured by direct visual observation. Cell constants are determined with distilled water as a calibrating liquid. Viscosities of molten KCl are measured with each viscometer. The differences between measured and standard values of molten KCl at several temperatures are within 5% for the oscillating cup viscometer and within 3% for the capillary viscometer. (author)

  19. Molten salt reactors and the oil sands: odd couple or key to north american energy independence?

    International Nuclear Information System (INIS)

    The use of nuclear power to aid oil sands development has often been proposed largely due to the virtual elimination of natural gas use and thus a large reduction in GHG emissions. Nuclear power can replace natural gas for process steam production (SAGD) and electricity generation but also potentially for hydrogen production to upgrade bitumen for pipeline transit, synthetic crude production and even at the final refinery stage. Prior candidates included CANDU and gas cooled Pebble Bed Reactors. The case for CANDU use can be shown to be marginally economic with a proven technology but with an uncertainty of current construction costs and too large a unit size (~2400 MWth). PBRs offered modest theoretical cost savings, smaller unit size and the ability to offer higher temperatures needed for thermochemical hydrogen production from water. Interest in PBRs however has greatly waned with the cancellation of their major South African development program which highlighted the severe challenges of helium as a coolant and TRISO fuel manufacturing. More recently, Small Modular Reactors based on scaled down light water reactor technology have attracted interest but are unlikely to compete economically outside of niche applications. However, a 'new' reactor option, the Molten Salt Reactor, has been rapidly gaining momentum over the past decade. This 'new' technology was actually developed over 50 years ago as a thorium breeder reactor to compete with the sodium cooled fast breeder reactor (U-Pu cycle). During this time two molten salt test reactors were constructed. A modern version however would likely be a simpler converter design using Low Enriched Uranium but needing only a small fraction the uranium resources of LWRs or CANDUs. Besides resource sustainability, these unique designs offer large potential improvements in the areas of capital costs, safety and nuclear waste. This presentation will explain the unique attributes and advantages of these

  20. Molten salt reactors and the oil sands: odd couple or key to north american energy independence?

    Energy Technology Data Exchange (ETDEWEB)

    LeBlanc, D., E-mail: d_leblanc@rogers.com [Ottawa Valley Research Associates Ltd., Ottawa, Ontario (Canada); Quesada, M.; Popoff, C.; Way, D. [Penumbra Energy, Calgary, Alberta (Canada)

    2012-07-01

    The use of nuclear power to aid oil sands development has often been proposed largely due to the virtual elimination of natural gas use and thus a large reduction in GHG emissions. Nuclear power can replace natural gas for process steam production (SAGD) and electricity generation but also potentially for hydrogen production to upgrade bitumen for pipeline transit, synthetic crude production and even at the final refinery stage. Prior candidates included CANDU and gas cooled Pebble Bed Reactors. The case for CANDU use can be shown to be marginally economic with a proven technology but with an uncertainty of current construction costs and too large a unit size (~2400 MWth). PBRs offered modest theoretical cost savings, smaller unit size and the ability to offer higher temperatures needed for thermochemical hydrogen production from water. Interest in PBRs however has greatly waned with the cancellation of their major South African development program which highlighted the severe challenges of helium as a coolant and TRISO fuel manufacturing. More recently, Small Modular Reactors based on scaled down light water reactor technology have attracted interest but are unlikely to compete economically outside of niche applications. However, a 'new' reactor option, the Molten Salt Reactor, has been rapidly gaining momentum over the past decade. This 'new' technology was actually developed over 50 years ago as a thorium breeder reactor to compete with the sodium cooled fast breeder reactor (U-Pu cycle). During this time two molten salt test reactors were constructed. A modern version however would likely be a simpler converter design using Low Enriched Uranium but needing only a small fraction the uranium resources of LWRs or CANDUs. Besides resource sustainability, these unique designs offer large potential improvements in the areas of capital costs, safety and nuclear waste. This presentation will explain the unique attributes and advantages of these

  1. Electrochemical studies on plutonium in molten salts

    Energy Technology Data Exchange (ETDEWEB)

    Bourges, G. [CEA-Centre d' etudes de Valduc, 21 120 Is sur Tille (France)], E-mail: gilles.bourges@cea.fr; Lambertin, D.; Rochefort, S. [CEA-Centre d' etudes de Valduc, 21 120 Is sur Tille (France); Delpech, S.; Picard, G. [Laboratoire d' Electrochimie et de Chimie Analytique (UMR7575, CNRS), ENSCP, 11 rue Pierre et Marie Curie, 75231 Paris (France)

    2007-10-11

    Electrochemical studies on plutonium have been supporting the development of pyrochemical processes involving plutonium at CEA. The electrochemical properties of plutonium have been studied in molten salts - ternary eutectic mixture NaCl-KCl-BaCl{sub 2}, equimolar mixture NaCl-KCl and pure CaCl{sub 2} - and in liquid gallium at 1073 K. The formal, or apparent, standard potential of Pu(III)/Pu redox couple in eutectic mixture of NaCl-KCl-BaCl{sub 2} at 1073 K determined by potentiometry is equal to -2.56 V (versus Cl{sub 2}, 1 atm/Cl{sup -} reference electrode). In NaCl-KCl eutectic mixture and in pure CaCl{sub 2} the formal standard potentials deduced from cyclic voltammetry are respectively -2.54 V and -2.51 V. These potentials led to the calculation of the activity coefficients of Pu(III) in the molten salts. Chronoamperometry on plutonium in liquid gallium using molten chlorides - CaCl{sub 2} and equimolar NaCl/KCl - led to the determination of the activity coefficient of Pu in liquid Ga, log {gamma} = -7.3. This new data is a key parameter to assess the thermodynamic feasibility of a process using gallium as solvent metal. By comparing gallium with other solvent metals - cadmium, bismuth, aluminum - gallium appears to be, with aluminum, more favorable for the selectivity of the separation at 1073 K of plutonium from cerium. In fact, compared with a solid tungsten electrode, none of these solvent liquid metals is a real asset for the selectivity of the separation. The role of a solvent liquid metal is mainly to trap the elements.

  2. Electrochemical studies on plutonium in molten salts

    International Nuclear Information System (INIS)

    Electrochemical studies on plutonium have been supporting the development of pyrochemical processes involving plutonium at CEA. The electrochemical properties of plutonium have been studied in molten salts - ternary eutectic mixture NaCl-KCl-BaCl2, equimolar mixture NaCl-KCl and pure CaCl2 - and in liquid gallium at 1073 K. The formal, or apparent, standard potential of Pu(III)/Pu redox couple in eutectic mixture of NaCl-KCl-BaCl2 at 1073 K determined by potentiometry is equal to -2.56 V (versus Cl2, 1 atm/Cl- reference electrode). In NaCl-KCl eutectic mixture and in pure CaCl2 the formal standard potentials deduced from cyclic voltammetry are respectively -2.54 V and -2.51 V. These potentials led to the calculation of the activity coefficients of Pu(III) in the molten salts. Chronoamperometry on plutonium in liquid gallium using molten chlorides - CaCl2 and equimolar NaCl/KCl - led to the determination of the activity coefficient of Pu in liquid Ga, log γ = -7.3. This new data is a key parameter to assess the thermodynamic feasibility of a process using gallium as solvent metal. By comparing gallium with other solvent metals - cadmium, bismuth, aluminum - gallium appears to be, with aluminum, more favorable for the selectivity of the separation at 1073 K of plutonium from cerium. In fact, compared with a solid tungsten electrode, none of these solvent liquid metals is a real asset for the selectivity of the separation. The role of a solvent liquid metal is mainly to trap the elements

  3. Application of coupled neutronics/thermal-hydraulics computational method for steady-state analysis of molten salt reactor

    International Nuclear Information System (INIS)

    The three-dimensional power distribution of the molten salt reactor (MSR) designed by Oak Ridge National Laboratory (ORNL) was analyzed by using MCNP code. Besides, a multiple-channel analysis code (MAC) was developed for the special geometry of graphite moderator design and coupled with MCNP code to analyze the neuron behavior as well as thermal-hydraulics of MSR experiment. The feasibility and accuracy of the analysis with coupled neutronics/thermal-hydraulics were validated by relevant results from the ORNL technical report. The results show that the coupling method can obtain accurately the power distribution, temperature distribution, pressure drop and mass flow distribution. This work is helpful for further design analysis and operation of MSR. (authors)

  4. A Novel Modeling of Molten-Salt Heat Storage Systems in Thermal Solar Power Plants

    Directory of Open Access Journals (Sweden)

    Rogelio Peón Menéndez

    2014-10-01

    Full Text Available Many thermal solar power plants use thermal oil as heat transfer fluid, and molten salts as thermal energy storage. Oil absorbs energy from sun light, and transfers it to a water-steam cycle across heat exchangers, to be converted into electric energy by means of a turbogenerator, or to be stored in a thermal energy storage system so that it can be later transferred to the water-steam cycle. The complexity of these thermal solar plants is rather high, as they combine traditional engineering used in power stations (water-steam cycle or petrochemical (oil piping, with the new solar (parabolic trough collector and heat storage (molten salts technologies. With the engineering of these plants being relatively new, regulation of the thermal energy storage system is currently achieved in manual or semiautomatic ways, controlling its variables with proportional-integral-derivative (PID regulators. This makes the overall performance of these plants non optimal. This work focuses on energy storage systems based on molten salt, and defines a complete model of the process. By defining such a model, the ground for future research into optimal control methods will be established. The accuracy of the model will be determined by comparing the results it provides and those measured in the molten-salt heat storage system of an actual power plant.

  5. The Thorium Molten Salt Reactor Moving on from the MSBR

    CERN Document Server

    Mathieu, L; Brissot, R; Le Brun, C; Liatard, E; Loiseaux, J M; Méplan, O; Merle-Lucotte, E; Nuttin, A; Wilson, J; Garzenne, C; Lecarpentier, D; Walle, E

    2006-01-01

    A re-evaluation of the Molten Salt Breeder Reactor concept has revealed problems related to its safety and to the complexity of the reprocessing considered. A reflection is carried out anew in view of finding innovative solutions leading to the Thorium Molten Salt Reactor concept. Several main constraints are established and serve as guides to parametric evaluations. These then give an understanding of the influence of important core parameters on the reactor's operation. The aim of this paper is to discuss this vast research domain and to single out the Molten Salt Reactor configurations that deserve further evaluation.

  6. Control strategies in a thermal oil - Molten salt heat exchanger

    Science.gov (United States)

    Roca, Lidia; Bonilla, Javier; Rodríguez-García, Margarita M.; Palenzuela, Patricia; de la Calle, Alberto; Valenzuela, Loreto

    2016-05-01

    This paper presents a preliminary control scheme for a molten salt - thermal oil heat exchanger. This controller regulates the molten salt mass flow rate to reach and maintain the desired thermal oil temperature at the outlet of the heat exchanger. The controller architecture has been tested using an object-oriented heat exchanger model that has been validated with data from a molten salt testing facility located at CIEMAT-PSA. Different simulations are presented with three different goals: i) to analyze the controller response in the presence of disturbances, ii) to demonstrate the benefits of designing a setpoint generator and iii) to show the controller potential against electricity price variations.

  7. Transient analysis of a molten salt central receiver (MSCR) in a solar power plant

    Science.gov (United States)

    Joshi, A.; Wang, C.; Akinjiola, O.; Lou, X.; Neuschaefer, C.; Quinn, J.

    2016-05-01

    Alstom is developing solar power tower plants utilizing molten salt as the working fluid. In solar power tower, the molten salt central receiver (MSCR) atop of the tower is constructed of banks of tubes arranged in panels creating a heat transfer surface exposed to the solar irradiation from the heliostat field. The molten salt heat transfer fluid (HTF), in this case 60/40%wt NaNO3-KNO3, flows in serpentine flow through the surface collecting sensible heat thus raising the HTF temperature from 290°C to 565°C. The hot molten salt is stored and dispatched to produce superheated steam in a steam generator, which in turn produces electricity in the steam turbine generator. The MSCR based power plant with a thermal energy storage system (TESS) is a fully dispatchable renewable power plant with a number of opportunities for operational and economic optimization. This paper presents operation and controls challenges to the MSCR and the overall power plant, and the use of dynamic model computer simulation based transient analyses applied to molten salt based solar thermal power plant. This study presents the evaluation of the current MSCR design, using a dynamic model, with emphasis on severe events affecting critical process response, such as MS temperature deviations, and recommend MSCR control design improvements based on the results. Cloud events are the scope of the transient analysis presented in this paper. The paper presents results from a comparative study to examine impacts or effects on key process variables related to controls and operation of the MSCR plant.

  8. An experimental test plan for the characterization of molten salt thermochemical properties in heat transport systems

    Energy Technology Data Exchange (ETDEWEB)

    Pattrick Calderoni

    2010-09-01

    Molten salts are considered within the Very High Temperature Reactor program as heat transfer media because of their intrinsically favorable thermo-physical properties at temperatures starting from 300 C and extending up to 1200 C. In this context two main applications of molten salt are considered, both involving fluoride-based materials: as primary coolants for a heterogeneous fuel reactor core and as secondary heat transport medium to a helium power cycle for electricity generation or other processing plants, such as hydrogen production. The reference design concept here considered is the Advanced High Temperature Reactor (AHTR), which is a large passively safe reactor that uses solid graphite-matrix coated-particle fuel (similar to that used in gas-cooled reactors) and a molten salt primary and secondary coolant with peak temperatures between 700 and 1000 C, depending upon the application. However, the considerations included in this report apply to any high temperature system employing fluoride salts as heat transfer fluid, including intermediate heat exchangers for gas-cooled reactor concepts and homogenous molten salt concepts, and extending also to fast reactors, accelerator-driven systems and fusion energy systems. The purpose of this report is to identify the technical issues related to the thermo-physical and thermo-chemical properties of the molten salts that would require experimental characterization in order to proceed with a credible design of heat transfer systems and their subsequent safety evaluation and licensing. In particular, the report outlines an experimental R&D test plan that would have to be incorporated as part of the design and operation of an engineering scaled facility aimed at validating molten salt heat transfer components, such as Intermediate Heat Exchangers. This report builds on a previous review of thermo-physical properties and thermo-chemical characteristics of candidate molten salt coolants that was generated as part of the

  9. Study of the pyrochemical treatment-recycling process of the Molten Salt Reactor fuel; Estudio de sistema de un proceso de tratamiento-reciclaje piroquimico del combustible de un reactor de sales fundidas

    Energy Technology Data Exchange (ETDEWEB)

    Boussier, H.; Heuer, D.

    2010-07-01

    The Separation Processes Studies Laboratory (Commissariat a l'energie Atomique) has made a preliminary assessment of the reprocessing system associated with Molten Salt Fast Reactor (MSFR). The scheme studied in this paper is based on the principle of reductive extraction and metal transfer that constituted the core process designed for the Molten Salt Breeder Reactor (MSBR), although the flow diagram has been adapted to the current needs of the Molten Salt Fast Reactor (MSFR).

  10. System Requirements Document for the Molten Salt Reactor Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Aigner, R.D.

    2000-04-01

    The purpose of the conversion process is to convert the {sup 233}U fluoride compounds that are being extracted from the Molten Salt Reactor Experiment (MSRE) equipment to a stable oxide for long-term storage at Bldg. 3019.

  11. Boric Ester-Type Molten Salt via Dehydrocoupling Reaction

    Directory of Open Access Journals (Sweden)

    Noriyoshi Matsumi

    2014-11-01

    Full Text Available Novel boric ester-type molten salt was prepared using 1-(2-hydroxyethyl-3-methylimidazolium chloride as a key starting material. After an ion exchange reaction of 1-(2-hydroxyethyl-3-methylimidazolium chloride with lithium (bis-(trifluoromethanesulfonyl imide (LiNTf2, the resulting 1-(2-hydroxyethyl-3-methylimidazolium NTf2 was reacted with 9-borabicyclo[3.3.1]nonane (9-BBN to give the desired boric ester-type molten salt in a moderate yield. The structure of the boric ester-type molten salt was supported by 1H-, 13C-, 11B- and 19F-NMR spectra. In the presence of two different kinds of lithium salts, the matrices showed an ionic conductivity in the range of 1.1 × 10−4–1.6 × 10−5 S cm−1 at 51 °C. This was higher than other organoboron molten salts ever reported.

  12. Main Experimental Results of ISTC-1606 for Recycling and Transmutation in Molten Salt Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, Victor; Feynberg, Olga; Merzlyakov, Aleksandr; Surenkov, Aleksandr [Russian Research Center - Kurchatov Institute, Kurchatov sq. 1, Moscow, RF, 123182 (Russian Federation); Subbotin, Vladimir; Zakirov, Raul; Toropov, Andrey; Panov, Aleksandr [Russian Federal Nuclear Center - Institute of Technical Physics, Snezhinsk (Russian Federation); Afonichkin, Valery [Institute of High-Temperature Electrochemistry, Ekaterinburg (Russian Federation)

    2008-07-01

    To examine and demonstrate the feasibility of molten salt reactors (MSR) to reduce long lived waste toxicity and to produce efficiently electricity in closed fuel cycle, some national and international studies were initiated last years. In this paper main focus is placed on experimental evaluation of single stream Molten Salt Actinide Recycler and Transmuter (MOSART) system fuelled with different compositions of plutonium plus minor actinide trifluoride (AnF{sub 3}) from LWR spent nuclear fuel without U-Th support. This paper summarizes main experimental results of ISTC-1606 related to physical and chemical properties of fuel salt, container materials for fuel circuit, and fuel salt clean up of MOSART system. As result of ISTC-1606 studies claim is made, that the {sup 7}Li,Na,Be/F and {sup 7}Li,Be/F solvents selected for primary system appear to resolve main reactor physics, thermal hydraulics, materials compatibility, fuel salt clean up and safety problems as applied to the MOSART concept development. The created experimental facilities and the database on properties of fuel salt mixtures and container materials are used for a choice and improvement fuel salts and coolants for new applications of this high temperature technology for sustainable nuclear power development. (authors)

  13. Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

    International Nuclear Information System (INIS)

    Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF2-ThF4-233UF4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155

  14. Main Experimental Results of ISTC-1606 for Recycling and Transmutation in Molten Salt Systems

    International Nuclear Information System (INIS)

    To examine and demonstrate the feasibility of molten salt reactors (MSR) to reduce long lived waste toxicity and to produce efficiently electricity in closed fuel cycle, some national and international studies were initiated last years. In this paper main focus is placed on experimental evaluation of single stream Molten Salt Actinide Recycler and Transmuter (MOSART) system fuelled with different compositions of plutonium plus minor actinide trifluoride (AnF3) from LWR spent nuclear fuel without U-Th support. This paper summarizes main experimental results of ISTC-1606 related to physical and chemical properties of fuel salt, container materials for fuel circuit, and fuel salt clean up of MOSART system. As result of ISTC-1606 studies claim is made, that the 7Li,Na,Be/F and 7Li,Be/F solvents selected for primary system appear to resolve main reactor physics, thermal hydraulics, materials compatibility, fuel salt clean up and safety problems as applied to the MOSART concept development. The created experimental facilities and the database on properties of fuel salt mixtures and container materials are used for a choice and improvement fuel salts and coolants for new applications of this high temperature technology for sustainable nuclear power development. (authors)

  15. Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

    Energy Technology Data Exchange (ETDEWEB)

    Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia); Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik [Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia)

    2015-09-30

    Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

  16. Development of High-temperature Molten Salt Transport Technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S. H.; Park, G. I.; Park, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The pyrochemical process, which has been developed by Korea Atomic Energy Research Institute (KAERI) since 1997, consists of processes such as pretreatment, oxide reduction, electrorefining, electrowinning, and waste salt treatment. In pyroprocessing, high-temperature molten salt transport technologies are required because the molten salt used in an electrorefiner should be transported to next process, the electrowinning process to recover U/TRU/RE after the electrorefining process is finished. However, there have been few transport studies on high temperature molten salt. Therefore, in pyrometallurgical processing, the development of high-temperature molten salt transport technology is a crucial prerequisite. In this study, three different salt transport technologies (gravity, suction pump, and centrifugal pump) were investigated. In addition, the performance test of the apparatus in the system was then carried out. After the electrorefining process, the molten salt used is transported to an electrowinning system to recover U/TRU/RE, and a high temperature molten salt transfer technology by suction is now being developed. To develop engineering-scale salt transport technology, a PRIDE salt transport system was designed and installed a Ar cell, 2{sup nd} of the PRIDE facility for engineering-scale salt transport demonstration, and its performance was confirmed from blank and performance tests for the PRIDE salt transport system.

  17. New breeding gain definitions and their application to the optimization of a molten salt reactor design

    International Nuclear Information System (INIS)

    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.

  18. Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

    Energy Technology Data Exchange (ETDEWEB)

    Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui [Xi' an Jiaotong Univ. (China). State Key Laboratory of Multiphase Flow in Power Engineering

    2016-05-15

    Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

  19. Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

    International Nuclear Information System (INIS)

    Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

  20. Rational Pu-disposition for U-production by THORIMS-NES (Thorium Molten-Salt Nuclear Energy Synergetics)

    International Nuclear Information System (INIS)

    Plutonium should not be simply burned out and be used to produce a new fissile 233U, because for solving the global energy/environmental problems in the next century a huge-size Fission Industry should be established preparing huge fissile materials. It seems to be established by THORIMS-NES [Thorium Molten-Salt Nuclear Energy Synergetics] composed of simple Molten-Salt Fission Reactors (MSR: FUJI-series] and 233U producing Accelerator Molten-Salt Breeders [AMSB]. These new Th facilities will be developed by utilizing/eliminating the ''excess Pu'' separated from Pu weapon-heads and the spent-fuel of U-Pu reactors. A semi-real scenario for proceedings such work in the first half of the next century is presented. THORIMS-NES will have big advantages not only in safety and radiowaste issues, but also in nuclear-proliferation/terrorism and economy guaranteeing low R and D cost. (author). 18 refs, 4 figs, 3 tabs

  1. Preliminary Neutronics Design Studies for a Molten Salt Blanket LIFE Engine

    International Nuclear Information System (INIS)

    looking at fast ignition and hot spot ignition fusion options are documented, along with limited scoping studies performed to investigate other options of interest that surfaced during the main design effort. Lastly, side studies that were not part of the main design effort but may alter future work performed on LIFE engine designs are shown. The majority of all work reported in this document was performed during the Molten Salt Fast Ignition Moderator Study (MSFIMS) which sought to optimize the amount of moderator mixed into the molten salt region in order to produce the most compelling design. The studies in this report are of a limited scope and are intended to provide a preliminary neutronics analysis of the design concepts described herein to help guide decision processes and explore various options that a LIFE engine with a molten salt blanket might enable. None of the designs shown in this report, even reference cases selected for detailed description and analysis, have been fully optimized. The analyses were performed primarily as a neutronics study, though some consultation was made regarding thermal-hydraulic and structural concerns during both scoping out an initial model and subsequent to identifying a neutronics-based reference case to ensure that the design work contained no glaring mechanical or thermal issues that would preclude its feasibility. Any analyses and recommendations made in this report are either primarily or solely from the point of view of LIFE neutronics and ignore other fundamental issues related to molten salt fuel blankets such as chemical processing feasibility and political feasibility of a molten salt system

  2. Study of trans-uranian incineration in molten salt reactor; Etude de l'incineration des transuraniens en reacteur a sel fondu

    Energy Technology Data Exchange (ETDEWEB)

    Valade, M

    2000-10-27

    With the emergence of new options for nuclear power, molten salt reactors are envisaged for waste management. The aim of this thesis is to show how molten salt reactors can help to solve the transuranics issue. Their high versatility regarding to isotopic vector allows to accommodate large fractions of minor actinides as compared to solid fuel system. In this thesis, a neutronics study of molten salt reactors, MSR, has been conducted. For this purpose, two reference systems were considered, TIER1 and AMSTER. In the case of TIER1, an optimisation was made to reach an equilibrium. The analysis of both systems showed the main characteristics of MSR: their link to chemistry and on line reprocessing. In this work, several methods to drive the system to a state of equilibrium have been implemented and compared. During this process the isotopic composition and neutron spectrum, thus the nuclear reaction cross sections, vary tremendously. It is essential to take these evolutions into account in order to accurately estimate the equilibrium state. This has been accomplished inside the multi-recycling procedure we set with ERANOS. A dedicated calculation schema has been realized to simulate superthermal systems with this computation code. These results were checked through a benchmark against other computer codes. Then, with multi-recycling method, several molten salt systems have been compared in order to define the optimal reactor for transuranics incineration. Nevertheless, a final choice can not only be done using only neutronics characteristics since chemistry and thermal-hydraulics constraints are really important for MSR. Moreover, a complete safety study would be required. (author)

  3. Fission product removal from molten salt using zeolite

    International Nuclear Information System (INIS)

    Spent nuclear fuel (SNF) can be treated in a molten salt electrorefiner for conversion into metal and mineral waste forms for geologic disposal. The fuel is dissolved in molten chloride salt. Non-transuranic fission products in the molten salt are ion-exchanged into zeolite A, which is subsequently mixed with glass and consolidated. Zeolite was found to be effective in removing fission product cations from the molten salt. Breakthrough of cesium and the alkaline earths occurred more rapidly than was observed for the rare earths. The effluent composition as a function of time is presented, as well as results for the distribution of fission products along the length of the column. Effects of temperature and salt flow rate are also discussed

  4. Establishing the design basis for a Molten Salt Demonstration Transmuter

    International Nuclear Information System (INIS)

    A Molten Salt Demonstration Transmuter is required to show the operation and design performance for closing the nuclear spent fuel cycle for PWR or WWER reactors operated in the once-through cycle mode. The remnant waste would be either permanently stored or held for secondary use. The purpose of this proposal is to establish the design basis for the Molten Salt Demonstration Transmuter. It is supposed that once-through-cycle nuclear spent fuel would be delivered to the Molten Salt Demonstration Transmuter in the standard transportable container includes 84 WWER-440 SNF assemblies each weighing 250 kg and containing 120 kg U, and about 1.2 kg of Pu and minor actinides. One assembly at a time will be withdrawn from the container and chemically processed to supply Pu and minor actinides at the rate necessary for burn-up compensation. (Authors)

  5. Injector nozzle for molten salt destruction of energetic waste materials

    Science.gov (United States)

    Brummond, William A.; Upadhye, Ravindra S.

    1996-01-01

    An injector nozzle has been designed for safely injecting energetic waste materials, such as high explosives, propellants, and rocket fuels, into a molten salt reactor in a molten salt destruction process without premature detonation or back burn in the injection system. The energetic waste material is typically diluted to form a fluid fuel mixture that is injected rapidly into the reactor. A carrier gas used in the nozzle serves as a carrier for the fuel mixture, and further dilutes the energetic material and increases its injection velocity into the reactor. The injector nozzle is cooled to keep the fuel mixture below the decomposition temperature to prevent spontaneous detonation of the explosive materials before contact with the high-temperature molten salt bath.

  6. Electrochemistry of actinide and lanthanide in molten salt system

    International Nuclear Information System (INIS)

    In the partition and transmutation processes of reprocessing of spent fuel or radioactive waste in nuclear power plant, the dry type reprocessing method using molten salt and liquid metal as a solvent is studied. Most especially researches on the electrolysis of the actinide nitride in the molten salts corresponding to reprocessing of nitride fuel cannot be found. This report is a research result about the electro-chemical behavior of actinide and lanthanide on the electrode in molten LiCL-KCL eutectic system. When anode potential was less than -0.4V in recovery of U metal by the molten salt electrolysis of UN, the electrolysis efficiency of the recovery is not influenced by the generation of UNCL and the oxidation-reduction reaction of U4+/U3+. Moreover, generation of a chlorination nitride was not seen in the case where PuN and NpN are used. (H. Katsuta)

  7. Study on application of molten salt oxidation technology (MSO) for PVC wastes treatment

    International Nuclear Information System (INIS)

    The project 'Study on application of molten salt oxidation (MSO) for PVC plastic wastes treatment' aims at three followings: 1) Installation of lab-scale MSO unit with essential compositions builds up foundation for the 2) estimation of waste destruction efficiency of the technology. 3) Based on the results of testing PVC - the chlorinated organic wastes on the lab-scale unit, the ability of the technology application at pilot-scale level will be primary estimated. The adjustment and correction of some compositions in the lab-scale unit theoretically designed during experiment overcame the shortages by design and fabrication such as heat distribution regime, feeding wastes and draining spent salt. These solutions adapt to the technical requirement of operation as well as scientific requirement of the research on MSO process. PVC waste treatment was tested on the MSO lab-scale unit in different conditions of operation temperature, superficial air velocity related to air/oxygen feeding rate, waste feeding rate. The testing results showed that destruction efficiency of chlorine in MSO technology was almost absolute. HCl and Cl2 emission were insignificant in different operation conditions. HCl and Cl2 emission depend on resident time and nature of molten salt. However, with inherent attributes of MSO technology emission of CO is not avoided in processing waste treatment. Therefore, finding active solutions for reduction CO emission is essential to complete the technology. The experiments also were carried in conditions of single molten salt (Na2CO3) and molten (Na2CO3 - K2CO3) eutectic. The comparison of efficiency of these tests gives idea of using molten salt eutectic to reduce operation cost in MSO technology. Based on operation parameters and scientific verification results during experiments, the introductory procedure of waste treatment by MSO process was built up. Thereby, primary estimation of development of the technology in pilot-scale is given. (author)

  8. Molten Salt Promoting Effect in Double Salt CO2 Absorbents

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Keling; Li, Xiaohong S.; Chen, Haobo; Singh, Prabhakar; King, David L.

    2016-01-01

    The purpose of this paper is to elaborate on the concept of molten salts as catalysts for CO2 absorption by MgO, and extend these observations to the MgO-containing double salt oxides. We will show that the phenomena involved with CO2 absorption by MgO and MgO-based double salts are similar and general, but with some important differences. This paper focuses on the following key concepts: i) identification of conditions that favor or disfavor participation of isolated MgO during double salt absorption, and investigation of methods to increase the absorption capacity of double salt systems by including MgO participation; ii) examination of the relationship between CO2 uptake and melting point of the promoter salt, leading to the recognition of the role of pre-melting (surface melting) in these systems; and iii) extension of the reaction pathway model developed for the MgO-NaNO3 system to the double salt systems. This information advances our understanding of MgO-based CO2 absorption systems for application with pre-combustion gas streams.

  9. LIFE Materails: Molten-Salt Fuels Volume 8

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R; Brown, N; Caro, A; Farmer, J; Halsey, W; Kaufman, L; Kramer, K; Latkowski, J; Powers, J; Shaw, H; Turchi, P

    2008-12-11

    The goals of the Laser Inertial Fusion Fission Energy (LIFE) is to use fusion neutrons to fission materials with no enrichment and minimum processing and have greatly reduced wastes that are not of interest to making weapons. Fusion yields expected to be achieved in NIF a few times per day are called for with a high reliable shot rate of about 15 per second. We have found that the version of LIFE using TRISO fuel discussed in other volumes of this series can be modified by replacing the molten-flibe-cooled TRISO fuel zone with a molten salt in which the same actinides present in the TRISO particles are dissolved in the molten salt. Molten salts have the advantage that they are not subject to radiation damage, and hence overcome the radiation damage effects that may limit the lifetime of solid fuels such as TRISO-containing pebbles. This molten salt is pumped through the LIFE blanket, out to a heat exchanger and back into the blanket. To mitigate corrosion, steel structures in contact with the molten salt would be plated with tungsten or nickel. The salt will be processed during operation to remove certain fission products (volatile and noble and semi-noble fission products), impurities and corrosion products. In this way neutron absorbers (fission products) are removed and neutronics performance of the molten salt is somewhat better than that of the TRISO fuel case owing to the reduced parasitic absorption. In addition, the production of Pu and rare-earth elements (REE) causes these elements to build up in the salt, and leads to a requirement for a process to remove the REE during operation to insure that the solubility of a mixed (Pu,REE)F3 solid solution is not exceeded anywhere in the molten salt system. Removal of the REE will further enhance the neutronics performance. With molten salt fuels, the plant would need to be safeguarded because materials of interest for weapons are produced and could potentially be removed.

  10. Ionic liquids: the link to high-temperature molten salts?

    Science.gov (United States)

    El Abedin, Sherif Zein; Endres, Frank

    2007-11-01

    Due to their wide thermal windows, ionic liquids can be regarded as the missing link between aqueous/organic solutions and high-temperature molten salts. They can be employed efficiently for the coating of other metals with thin layers of tantalum, aluminum, and presumably many others at reasonable temperatures by electrochemical means. The development of ionic liquids, especially air and water stable ones, has opened the door for the electrodeposition of reactive elements such as, for example, Al, Ta, and Si, which in the past were only accessible using high-temperature molten salts or, in part, organic solvents. PMID:17521159

  11. A Possible Regenerative, Molten-Salt, Thermoelectric Fuel Cell

    Science.gov (United States)

    Greenberg, Jacob; Thaller, Lawrence H.; Weber, Donald E.

    1964-01-01

    Molten or fused salts have been evaluated as possible thermoelectric materials because of the relatively good values of their figures of merit, their chemical stability, their long liquid range, and their ability to operate in conjunction with a nuclear reactor to produce heat. In general, molten salts are electrolytic conductors; therefore, there will be a transport of materials and subsequent decomposition with the passage of an electric current. It is possible nonetheless to overcome this disadvantage by using the decomposition products of the molten-salt electrolyte in a fuel cell. The combination of a thermoelectric converter and a fuel cell would lead to a regenerative system that may be useful.

  12. Launching the Thorium fuel cycle with the molten salt fast reactor

    International Nuclear Information System (INIS)

    Starting from the Molten Salt Breeder Reactor project of Oak-Ridge, an innovative concept called Molten Salt Fast Reactor or MSFR based on a fast neutron spectrum has been proposed, resulting from extensive parametric studies in which various core arrangements, reprocessing performances and salt compositions were investigated to adapt the reactor in the framework of the deployment of a thorium based reactor fleet on a worldwide scale. In the MSFR, the liquid fuel processing is part of the reactor where a small side stream of the molten salt is processed for fission product removal and then returned to the reactor. Because of this characteristic, the MSFR can operate with widely varying fuel compositions. Thanks to this fuel composition flexibility, the MSFR concept may use as initial fissile load, 233U or uranium or also the transuranic elements currently produced by light water reactors. This paper addresses the characteristics of these different launching modes of the MSFR and the Thorium fuel cycle, in terms of safety, proliferation, breeding, and deployment capacities of these reactor configurations. To illustrate the deployment capacities of the MSFR concept, a French nuclear deployment scenario is finally presented, demonstrating that launching the Thorium fuel cycle is easily feasible while closing the current fuel cycle and optimizing the long-term waste management. (authors)

  13. Thermo-mechanical and optical optimization of the molten salt receiver for a given heliostat field

    Science.gov (United States)

    Augsburger, Germain; Das, Apurba K.; Boschek, Erik; Clark, Michael M.

    2016-05-01

    The tower type molten salt solar thermal power plant has proven to be advantageous over other utility scale solar power plant configurations due to its scalability and provision of storage, thereby improving the dispatchability. The configuration consists of a molten salt central receiver (MSCR) located atop an optimally located tower within a heliostat field with thousands of mirrors. The MSCR receives the concentrated energy from the heliostat field which heats a molten salt heat transfer fluid for thermal storage and utilization in producing steam as and when required for power generation. The MSCR heat transfer surface consists of banks of tangent tubes arranged in panels. The combined cost of the heliostat field and the receiver is 40%-50% of the total plant cost, which calls for optimization to maximize their utilization. Several previous studies have looked into the optimum solar power plant size based on various site conditions. However, the combined optimization of the receiver and the heliostat field has not been reported before. This study looks into the optimum configuration of the receiver for a given heliostat field. An in-house tool has been developed to select and rank a few receiver surface configurations (typically preference is presented based on the performance and various other practical considerations (e.g. total surface area, cost of material, ability of aiming strategies to distribute the flux). The methodology thus provided can be used as a guideline to arrive at an optimum receiver configuration for a given heliostat field.

  14. Safety criteria and guidelines for MSR accident analysis

    International Nuclear Information System (INIS)

    Accident analysis for Molten Salt Reactor (MSR) has been investigated at ORNL for MSRE in 1960s. Since then, safety criteria or guidelines have not been defined for MSR accident analysis. Regarding the safety criteria, the authors showed one proposal in this paper. In order to establish guidelines for MSR accident analysis, we have to investigate all possible accidents. In this paper, the authors describe the philosophy for accident analysis, and show 40 possible accidents. They are at first classified as external cause accidents and internal cause accidents. Since the former ones are generic accidents, we investigate only the latter ones, and categorize them to 4 types, such as power excursion accident, flow decrease accident, fuel-salt leak accident, and other accidents mostly specific to MSR. Each accident is described briefly, with some numerical results by the authors. (author)

  15. Design study of small molten-salt fission power station suitable for coupling with accelerator molten-salt breeder

    International Nuclear Information System (INIS)

    A design study of /sup 233/U fueled 350 MWth(150MWe) molten-salt fission reactor was proceeded as an example of the economical utility facilities improving excellent inherent safety and easy operation and maintenance as follows (1) no exchange of core graphite resulting a sealed reactor vessel, (2) 99% removal of fission gases only and no continuous chemical processing, (3) very high conversion ratio such as 1.00 (fuel self-sufficient), (4) usefulness for the Trans-U incineration and the non-nuclear proliferation. Its low concentration of /sup 233/UF/sub 4/ will be significant for the symbiotic molten-salt fuel cycle with Accelerator Molten-Salt Breeder or the similiars

  16. Castable cements to prevent corrosion of metals in molten salts

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Vidal, J. C.; Morton, E.

    2016-08-01

    Castable cements on metals form a protective barrier that is able to prevent permeation of molten salts towards metallic surfaces. Silica-based castable cements are capable of protecting containment metallic alloys from the corrosive attack of molten chlorides at temperatures as high as 650 degrees C. Boron nitride (BN) blocking the pores in the cured cement prevents permeation of the molten chloride towards the metal surface. The cements tested are not chemically stable in molten carbonates, because the bonding components dissolved into molten carbonates salt. The corrosion rate is 7.72+/-0.32 mm/year for bare stainless steel 347 in molten eutectic NaCl - 65.58 wt% LiCl at 650 degrees C, which is the baseline used for determining how well the cement protects the metallic surfaces from corrosion. In particular the metal fully encapsulated with Aremco 645-N with pores filled with boron nitride immersed in molten eutectic NaCl - 65.58 wt% LiCl at 650 degrees C shows a corrosion rate of 9E-04 mm/year. The present study gives initial corrosion rates. Long-term tests are required to determine if Aremco 645-N with BN coating on metal has long term chemical stability for blocking salt permeation through coating pores.

  17. Molten salt extraction at the Rocky Flats Plant

    International Nuclear Information System (INIS)

    This paper reports on the Rocky Flats Plant which uses dicesium hexachloroplutonate (DCHP) as an oxidant and calcium chloride as a solvent in the molten salt extraction (MSE) of americium from plutonium metal. This process was implemented into production using the following operating parameters: 0.20 salt/metal (g/g) ratio and 2.5 salt/DCHP (g/g). The results from the technology transfer show a 95.6% americium extraction efficiency. On-going research is directed at reducing the amount of DCHP required and minimizing the amount of plutonium metal lost to the salt. A study was performed to profile the neutron and gamma radiation associated with the MSE process. The DCHP MSE process was broken down into fifty steps: the neutron and gamma radiation was measured at each step. Based on the analysis of the radiation profiles, recommendations to reduce the radiation exposure to the operator have been made. These recommendations primarily have focused on the feed metal preparation area and the material handling outside of the glovebox, especially during bag cut operations

  18. An evaluation of possible next-generation high temperature molten-salt power towers.

    Energy Technology Data Exchange (ETDEWEB)

    Kolb, Gregory J.

    2011-12-01

    Since completion of the Solar Two molten-salt power tower demonstration in 1999, the solar industry has been developing initial commercial-scale projects that are 3 to 14 times larger. Like Solar Two, these initial plants will power subcritical steam-Rankine cycles using molten salt with a temperature of 565 C. The main question explored in this study is whether there is significant economic benefit to develop future molten-salt plants that operate at a higher receiver outlet temperature. Higher temperatures would allow the use of supercritical steam cycles that achieve an improved efficiency relative to today's subcritical cycle ({approx}50% versus {approx}42%). The levelized cost of electricity (LCOE) of a 565 C subcritical baseline plant was compared with possible future-generation plants that operate at 600 or 650 C. The analysis suggests that {approx}8% reduction in LCOE can be expected by raising salt temperature to 650 C. However, most of that benefit can be achieved by raising the temperature to only 600 C. Several other important insights regarding possible next-generation power towers were also drawn: (1) the evaluation of receiver-tube materials that are capable of higher fluxes and temperatures, (2) suggested plant reliability improvements based on a detailed evaluation of the Solar Two experience, and (3) a thorough evaluation of analysis uncertainties.

  19. Treatment of plutonium process residues by molten salt oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Stimmel, J.; Wishau, R.; Ramsey, K.B.; Montoya, A.; Brock, J. [Los Alamos National Lab., NM (United States); Heslop, M. [Naval Surface Warfare Center (United States). Indian Head Div.; Wernly, K. [Molten Salt Oxidation Corp. (United States)

    1999-04-01

    Molten Salt Oxidation (MSO) is a thermal process that can remove more than 99.999% of the organic matrix from combustible {sup 238}Pu material. Plutonium processing residues are injected into a molten salt bed with an excess of air. The salt (sodium carbonate) functions as a catalyst for the conversion of the organic material to carbon dioxide and water. Reactive species such as fluorine, chlorine, bromine, iodine, sulfur, phosphorous and arsenic in the organic waste react with the molten salt to form the corresponding neutralized salts, NaF, NaCl, NaBr, NaI, Na{sub 2}SO{sub 4}, Na{sub 3}PO{sub 4} and NaAsO{sub 2} or Na{sub 3}AsO4. Plutonium and other metals react with the molten salt and air to form metal salts or oxides. Saturated salt will be recycled and aqueous chemical separation will be used to recover the {sup 238}Pu. The Los Alamos National Laboratory system, which is currently in the conceptual design stage, will be scaled down from current systems for use inside a glovebox.

  20. Treatment of plutonium process residues by molten salt oxidation

    International Nuclear Information System (INIS)

    Molten Salt Oxidation (MSO) is a thermal process that can remove more than 99.999% of the organic matrix from combustible 238Pu material. Plutonium processing residues are injected into a molten salt bed with an excess of air. The salt (sodium carbonate) functions as a catalyst for the conversion of the organic material to carbon dioxide and water. Reactive species such as fluorine, chlorine, bromine, iodine, sulfur, phosphorous and arsenic in the organic waste react with the molten salt to form the corresponding neutralized salts, NaF, NaCl, NaBr, NaI, Na2SO4, Na3PO4 and NaAsO2 or Na3AsO4. Plutonium and other metals react with the molten salt and air to form metal salts or oxides. Saturated salt will be recycled and aqueous chemical separation will be used to recover the 238Pu. The Los Alamos National Laboratory system, which is currently in the conceptual design stage, will be scaled down from current systems for use inside a glovebox

  1. Conceptual design of a pool type molten salt breeder reactor

    International Nuclear Information System (INIS)

    The renewed interest in molten salt coolant technology is backed by the 50 years history of molten salt nuclear technology development, mainly in Oak Ridge National Laboratory (ORNL). In Indian context MSBR is found to be one of the options for sustainable nuclear energy generation, especially in the third stage of the nuclear programme. The system can be operated at high temperature which makes high efficiency power conversion and efficient hydrogen generation through thermo-chemical reactions possible. At present development is in progress in BARC on two molten salt reactor concepts, one is pool type and the other is loop type. Here the design of pool type concept with 850MWe power is described. The core is designed to operate in the fast spectrum region so the conversion of 233U breeding is possible from thorium. Preliminary thermal hydraulic analysis is carried out with LiF-ThF4-UF4 as the primary fuel and coolant. The blanket material is also a molten salt, LiF-ThF4. Reactor physics calculations are also carried out for the feasibility studies of the core design of the reactor. FLiNaK is used as the secondary coolant for the calculations. Both forced circulation and natural circulation options are evaluated. (author)

  2. Automated ''float'' method for determination of densities of molten salts

    DEFF Research Database (Denmark)

    Andreasen, Helge A.; Bjerrum, Niels; Foverskov, Carl Erik

    1977-01-01

    A new system for measuring densities of molten salt systems is described. The system consists of an accurate metal block furnace, the temperature of which can be changed linearly in time, a fused quartz tube containing quartz floats loaded with a ferromagnetic material, a differential transformer....... Review of Scientific Instruments is copyrighted by The American Institute of Physics....

  3. Boric ester-type molten salt via dehydrocoupling reaction.

    Science.gov (United States)

    Matsumi, Noriyoshi; Toyota, Yoshiyuki; Joshi, Prerna; Puneet, Puhup; Vedarajan, Raman; Takekawa, Toshihiro

    2014-01-01

    Novel boric ester-type molten salt was prepared using 1-(2-hydroxyethyl)-3-methylimidazolium chloride as a key starting material. After an ion exchange reaction of 1-(2-hydroxyethyl)-3-methylimidazolium chloride with lithium (bis-(trifluoromethanesulfonyl) imide) (LiNTf2), the resulting 1-(2-hydroxyethyl)-3-methylimidazolium NTf2 was reacted with 9-borabicyclo[3.3.1]nonane (9-BBN) to give the desired boric ester-type molten salt in a moderate yield. The structure of the boric ester-type molten salt was supported by 1H-, 13C-, 11B- and 19F-NMR spectra. In the presence of two different kinds of lithium salts, the matrices showed an ionic conductivity in the range of 1.1 × 10⁻⁴-1.6 × 10⁻⁵ S cm⁻¹ at 51 °C. This was higher than other organoboron molten salts ever reported. PMID:25405738

  4. Towards the thorium fuel cycle with molten salt fast reactors

    International Nuclear Information System (INIS)

    Highlights: • Neutronic calculations for fast spectrum molten salt reactor. • Evaluation of the fissile matter to be used in such reactor as initial fissile load. • Capabilities to transmute transuranic elements. • Deployment scenarios of the Thorium fuel cycle. • Waste management optimization with molten salt fast reactor. - Abstract: There is currently a renewed interest in molten salt reactors, due to recent conceptual developments on fast neutron spectrum molten salt reactors (MSFRs) using fluoride salts. It has been recognized as a long term alternative to solid-fueled fast neutron systems with a unique potential (large negative temperature and void coefficients, lower fissile inventory, no initial criticality reserve, simplified fuel cycle, wastes reduction etc.) and is thus one of the reference reactors of the Generation IV International Forum. In the MSFR, the liquid fuel processing is part of the reactor where a small side stream of the molten salt is processed for fission product removal and then returned to the reactor. Because of this characteristic, the MSFR can operate with widely varying fuel compositions, so that the MSFR concept may use as initial fissile load, 233U or enriched uranium or also the transuranic elements currently produced by light water reactors. This paper addresses the characteristics of these different launching modes of the MSFR and the Thorium fuel cycle, in terms of safety, proliferation, breeding, and deployment capacities of these reactor configurations. To illustrate the deployment capacities of the MSFR concept, a French nuclear deployment scenario is finally presented, demonstrating that launching the Thorium fuel cycle is easily feasible while closing the current fuel cycle and optimizing the long-term waste management via stockpile incineration in MSRs

  5. Hydrocracking of coal using molten salts as catalysts

    Science.gov (United States)

    Kikkawa, S.; Nomura, M.; Sakashita, H.; Nishimura, M.; Miyake, M.

    1981-10-01

    Characteristics of the reactions during coal liquefaction and the hydrocracking of coal and coal-related materials using ZnCl2-transition metal chloride or ZnCl2-alkaline metal chloride are discussed. The studies involve development of a molten salt catalyst for hydrocracking heavy residual oils or coals, including hydrocarbons containing many heteroatoms. It was found that ZnCl2 shows higher activity for hydrocracking of anthracene and phenanthrene, and experiments with Yubari coal using the binary metal catalysts ZnCl2-MoCl5 and ZnCl2-CrCl3 are described. The use of molten salts in the desulphurization of heavy residual oils is also explored, specifically for the hydrocracking of benziophene, and the possibility that a coal-like polymer structure containing an oxygen surplus might depolymerize above ternary melts is suggested.

  6. Electrowinning Al from Al2S3 in Molten Salt

    OpenAIRE

    Xiao, Y; Van der Plas, D.W.; Bohte, J.; Lans, S.C.; Van Sandwijk, A.; Reuter, M.A.

    2007-01-01

    In order to investigate an alternative process for the production of primary aluminum via a sulfide intermediate, the electrochemical behavior of Al2S3 in molten salt has been studied on a laboratory scale. The effects of electrolyte composition, temperature, and cell design on the cell performance have been investigated. Temperature and cryolite addition have positive effects on the current density. Increasing the anode-to-cathode surface area (closer to unity) and shortening the interelectr...

  7. Simulation of neutron diffusion and transient analysis of MSR

    International Nuclear Information System (INIS)

    Molten salt reactor (MSR) is a potential nuclear power reactor of Generation IV. The working process of the primary loop of an MSR is studied in this paper. A physical model is established to describe the coupled heat transfer for the MSR core channels, the temperature negative feedback and the neutron characteristics. The simulation code, NDPID, has been developed with the object-oriented method, conducting the neutron diffusion and transient analysis in a parallel way. The simulation data and diagrams of neutron, power, flow rate and temperature can be obtained via graphical user interface. The simulation results can be used for further study on MSRs of larger dimensions and more complicated geometry. (authors)

  8. Molten salt processing of mixed wastes with offgas condensation

    International Nuclear Information System (INIS)

    We are developing an advanced process for treatment of mixed wastes in molten salt media at temperatures of 700--1000 degrees C. Waste destruction has been demonstrated in a single stage oxidation process, with destruction efficiencies above 99.9999% for many waste categories. The molten salt provides a heat transfer medium, prevents thermal surges, and functions as an in situ scrubber to transform the acid-gas forming components of the waste into neutral salts and immobilizes potentially fugitive materials by a combination of particle wetting, encapsulation and chemical dissolution and solvation. Because the offgas is collected and assayed before release, and wastes containing toxic and radioactive materials are treated while immobilized in a condensed phase, the process avoids the problems sometimes associated with incineration processes. We are studying a potentially improved modification of this process, which treats oxidizable wastes in two stages: pyrolysis followed by catalyzed molten salt oxidation of the pyrolysis gases at ca. 700 degrees C. 15 refs., 5 figs., 1 tab

  9. Optical absorption of dilute solutions of metals in molten salts

    International Nuclear Information System (INIS)

    The F-centre model for the bound state and the first optical transition of an electron in a metal-molten salt solution is examined in the high dilution limit appropriate for comparison with optical absorption data. It is first argued that the model is consistent with recent neutron diffraction and computer simulation data on the structure of pure molten salts, and not incompatible with an Anderson localization model for the electronic conductivity of the solution at higher concentration of metal. A detailed evaluation of the model is presented for the case of a molten salt of equi-sized ions simulating molten KCl. The treatment of the electronic states is patterned after semicontinuum approximations previously applied to the F-centre in ionic crystals, but the equilibrium radius of the electronic cavity and its fluctuations are determined self-consistently from the free energy of the solution. The detailed analysis of this case and the agreement of the results with experiment allow the construction of a simple parametrization scheme, which is then applied to explore the trends of the optical absorption spectrum and of the volume of mixing through the whole family of M-MX solutions, where M is an alkali and X a halogen. Similarities and differences of the electronic bound state in the crystal and in the liquid are underlined. (author)

  10. Molten Salts for High Temperature Reactors: University of Wisconsin Molten Salt Corrosion and Flow Loop Experiments -- Issues Identified and Path Forward

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Matt Ebner; Manohar Sohal; Phil Sharpe; Thermal Hydraulics Group

    2010-03-01

    Considerable amount of work is going on regarding the development of high temperature liquid salts technology to meet future process needs of Next Generation Nuclear Plant. This report identifies the important characteristics and concerns of high temperature molten salts (with lesson learned at University of Wisconsin-Madison, Molten Salt Program) and provides some possible recommendation for future work

  11. Thorium cycle and molten salt reactors: field parameters and field constraints investigations toward 'thorium molten salt reactor' definition

    International Nuclear Information System (INIS)

    Producing nuclear energy in order to reduce the anthropic CO2 emission requires major technological advances. Nuclear plants of 4. generation have to respond to several constraints, as safety improvements, fuel breeding and radioactive waste minimization. For this purpose, it seems promising to use Thorium Cycle in Molten Salt Reactors. Studies on this domain have already been carried out. However, the final concept suffered from serious issues and was discontinued. A new reflection on this topic is being led in order to find acceptable solutions, and to design the Thorium Molten Salt Reactor concept. A nuclear reactor is simulated by the coupling of a neutron transport code with a materials evolution code. This allows us to reproduce the reactor behavior and its evolution all along its operation. Thanks to this method, we have studied a large number of reactor configurations. We have evaluated their efficiency through a group of constraints they have to satisfy. This work leads us to a better understanding of many physical phenomena controlling the reactor behavior. As a consequence, several efficient configurations have been discovered, allowing the emergence of new points of view in the research of Molten Salt Reactors. (author)

  12. Experimental modelling and numerical analysis of a molten salt fast reactor

    International Nuclear Information System (INIS)

    In this paper experimental and numerical investigation of the MSFR (Molten Salt Fast Reactor) concept will be presented. This homogeneous, single region liquid fuelled fast reactor concept uses fluoride-based molten salts with fissile uranium and thorium and other heavy nuclei content with the purpose of applying the thorium cycle and the burn-up of transuranic elements. Molten salt reactors with liquid fuel have a unique safety related property that needs clear understanding. In the core neutron flux and fission distribution is determined by the flow field through distribution and transport of fissile material and delayed neutron precursors. Since the MSFR concept has a single region homogeneous core without internal structures, it is a difficult task to ensure stable flow field, which is also strongly coupled to the volumetric heat generation. These considerations suggest that experimental and numerical modelling (including the option of coupled neutronics-thermal-hydraulics) would be needed to better understand the flow phenomena in such geometry. A scaled and segmented experimental mock-up of MSFR was designed and built at BME NTI with the purpose of investigating the flow behavior inside the core region using particle image velocimetry. Not only the basic flow behavior inside the core region can be investigated but measurement data can also provide resource for the validation of computational fluid dynamics models, specific problems or phenomena (for example inlet geometry, optional internal structures, mixing) may be studied as well. Measurement results of steady state conditions will be presented with comparison of measurement data and results of numerical analyses. (author)

  13. Degradation of organochloride pesticides by molten salt oxidation at IPEN: spin-off nuclear activities

    International Nuclear Information System (INIS)

    Nuclear spin-off has at least two dimensions. It may provide benefits to the society such as enlarge knowledge base, strengthen infrastructure and benefit technology development. Besides this, to emphasize that some useful technologies elapsed from nuclear activities can affect favorably the public opinion about nuclear energy. In this paper is described a technology developed initially by the Rockwell Int. company in the USA more than thirty years ago to solve some problems of nuclear fuel cycle wastes. For different reasons the technology was not employed. In the last years the interest in the technology was renewed and IPEN has developed his version of the method applicable mainly to the safe degradation of hazardous wastes. This study was motivated by the world interest in the development of advanced processes of waste decomposition, due to the need of safer decomposition processes, particularly for the POPs - persistent organic pollutants and particularly for the organ chlorides. A tendency observed at several countries is the adoption of progressively more demanding legislation for the atmospheric emissions, resultants of the waste decomposition processes. The suitable final disposal of hazardous organic wastes such as PCBs (polychlorinated biphenyls), pesticides, herbicides and hospital residues constitutes a serious problem. In some point of their life cycles, these wastes should be destroyed, in reason of the risk that they represent for the human being, animals and plants. The process involves using a chemical reactor containing molten salts, sodium carbonate or some alkaline carbonates mixtures to decompose the organic waste. The decomposition is performed by submerged oxidation and the residue is injected below the surface of a turbulent salt bath along with the oxidizing agent. Decomposition of halogenated compounds, among which some pesticides, is particularly effective in molten salts. The process presents properties such as intrinsically safe

  14. Recent advances in the molten salt technology for the destruction of energetic materials

    International Nuclear Information System (INIS)

    The DOE has thousands of pounds of energetic materials which result from dismantlement operations at the Pantex Plant. The authors have demonstrated the Molten Salt Destruction (MSD) Process for the treatment of explosives and explosive-containing wastes on a 1.5 kilogram of explosive per hour scale and are currently building a 5 kilogram per hour unit. MSD converts the organic constituents of the waste into non-hazardous substances such as carbon dioxide, nitrogen and water. Any inorganic constituents of the waste, such as binders and metallic particles, are retained in the molten salt. The destruction of energetic material waste is accomplished by introducing it, together with air, into a crucible containing a molten salt, in this case a eutectic mixture of Na, K, and Li carbonates. The following pure component DOE and DoD explosives have been destroyed in LLNL's experimental unit at their High Explosives Applications Facility (HEAF): ammonium picrate, HMX, K-6, NQ, NTO, PETN, RDX, TATB, and TNT. In addition, the following formulations were also destroyed: Comp B, LX-10, LX-16, LX-17, PBX-9404, and XM46, a US Army liquid gun propellant. In this 1.5 kg/hr unit, the fractions of carbon converted to CO and of chemically bound nitrogen converted to NOx were found to be well below 1T. In addition to destroying explosive powders and molding powders the authors have also destroyed materials that are typical of real world wastes. These include shavings from machined pressed parts of plastic bonded explosives and sump waste containing both explosives and non-explosive debris. Based on the information obtained on the smaller unit, the authors have constructed a 5 kg/hr MSD unit, incorporating LLNL's advanced chimney design. This unit is currently under shakedown tests and evaluation

  15. Recent Research of Thorium Molten-Salt Reactor from a Sustainability Viewpoint

    Directory of Open Access Journals (Sweden)

    Takashi Kamei

    2012-09-01

    Full Text Available The most important target of the concept “sustainability” is to achieve fairness between generations. Its expanding interpolation leads to achieve fairness within a generation. Thus, it is necessary to discuss the role of nuclear power from the viewpoint of this definition. The history of nuclear power has been the control of the nuclear fission reaction. Once this is obtained, then the economy of the system is required. On the other hand, it is also necessary to consider the internalization of the external diseconomy to avoid damage to human society caused by the economic activity itself, due to its limited capacity. An extreme example is waste. Thus, reducing radioactive waste resulting from nuclear power is essential. Nuclear non-proliferation must be guaranteed. Moreover, the FUKUSHIMA accident revealed that it is still not enough that human beings control nuclear reaction. Further, the most essential issue for sustaining use of one technology is human resources in manufacturing, operation, policy-making and education. Nuclear power will be able to satisfy the requirements of sustainability only when these subjects are addressed. The author will review recent activities of a thorium molten-salt reactor (MSR as a cornerstone for a sustainable society and describe its objectives and forecasts.

  16. Oxidation of hydrogen halides to elemental halogens with catalytic molten salt mixtures

    Science.gov (United States)

    Rohrmann, Charles A.

    1978-01-01

    A process for oxidizing hydrogen halides by means of a catalytically active molten salt is disclosed. The subject hydrogen halide is contacted with a molten salt containing an oxygen compound of vanadium and alkali metal sulfates and pyrosulfates to produce an effluent gas stream rich in the elemental halogen. The reduced vanadium which remains after this contacting is regenerated to the active higher valence state by contacting the spent molten salt with a stream of oxygen-bearing gas.

  17. Structure and dynamics of molten salts

    International Nuclear Information System (INIS)

    Modern techniques of liquid state physics have been successfully used over the last decade to probe the microscopic structure and dynamics of a variety of multicomponent liquids in which relative ordering of the species is present near freezing. The alkali halides are prototypes for this specific type of short range order in relation to the nature of bonding, but the systems in question include also other monovalent and polyvalent metal-ion halides, alkali-based intermetallic compounds, and chalcogen-based alloys. A viewpoint is taken in this review which gives attention to relations between liquid and solid phase properties across melting for compound systems at stoichiometric composition. In addition, large deviations from stoichiometry can be realized in the liquid phase, to display trends of evolution of structure, bonding and electronic states with composition. (author)

  18. Technical review of Molten Salt Oxidation

    International Nuclear Information System (INIS)

    The process was reviewed for destruction of mixed low-level radioactive waste. Results: extensive development work and scaleup has been documented on coal gasification and hazardous waste which forms a strong experience base for this MSO process; it is clearly applicable to DOE wastes such as organic liquids and low-ash wastes. It also has potential for processing difficult-to-treat wastes such as nuclear grade graphite and TBP, and it may be suitable for other problem waste streams such as sodium metal. MSO operating systems may be constructed in relatively small units for small quantity generators. Public perceptions could be favorable if acceptable performance data are presented fairly; MSO will likely require compliance with regulations for incineration. Use of MSO for offgas treatment may be complicated by salt carryover. Figs, tabs, refs

  19. Synthesis and characterization of silicide coating on niobium alloy produced using molten salt method

    International Nuclear Information System (INIS)

    Nb based alloys are promising structural materials for high temperature reactors due to their strength at higher temperatures. However Nb based alloys undergoes substantial oxidation at high temperatures. In order to improve its oxidation resistance property at high temperatures (>400 °C) a protective layer must be provided to avoid direct contact of the component to atmospheric oxygen. In the present work, attempts have been made to obtain silicide coatings on Nb alloy using molten salt method. In this method, deposition of silicon is a multistep process. Metallic Si produced by the subsequent reactions in the molten salt diffuses and an oxidation resistant silicide coating forms on the surface of substrate. To study the variation in the thickness of coated layer on the Nb alloy, experiments were carried out at different temperature and time periods. These silicide coated samples were characterized using optical, SEM and XRD techniques. Based on these results mechanism of silicide coating on Nb alloys has been discussed in detail. (author)

  20. Plastic properties of tunsten produced by electrolysis of molten salts

    International Nuclear Information System (INIS)

    A study is made into bend ductility, microhardness and texture of tungsten produced by electrolysis of CsCl and KCl-NaF molten salts. The influence of texture and high temperature annealing on ductile-brittle transition temperature was determined using specimens of electrolytic tungsten coatings 0.3 mm thick. For tungsten specimens of perfect texture [111] transition temperature constitutes 250 deg C. High temperature annealing (1400 deg C 4 h) raises the temperature of ductile-brittle transition. This fact may be related to the process of impurity concentration on grain boundaries which length decreases on annealing. 8 refs., 3 figs.2 tabs

  1. Fission product behavior in the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Essentially all the fission product data for numerous and varied samples taken during operation of the Molten Salt Reactor Experiment or as part of the examination of specimens removed after particular phases of operation are reported, together with the appropriate inventory or other basis of comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct chemical groups. Evidence for fission product behavior during operation over a period of 26 months with 235U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using 233U fuel over a period of about 15 months (more than 5100 effective full-power hours)

  2. Metallic materials corrosion problems in molten salt reactors

    International Nuclear Information System (INIS)

    The USA forecastings concerning the molten salt reactors are reviewed (mixtures of fluorides containing the fuel, operating between 560 and 7000C). Corrosion problems are important in these reactors. The effects of certain characteristic factors on corrosion are analyzed: humidity and metallic impurities in the salts, temperature gradients, speed of circulation of salts, tellurium from fission products, coupling. In the molten fluorides and experimental conditions, the materials with high Ni content are particularly corrosion resistant alloys (hastelloy N). The corrosion of this material is about 2.6 mg.cm-2 at 7000C

  3. Novel ternary molten salt electrolytes for intermediate-temperature sodium/nickel chloride batteries

    Science.gov (United States)

    Li, Guosheng; Lu, Xiaochuan; Coyle, Christopher A.; Kim, Jin Y.; Lemmon, John P.; Sprenkle, Vincent L.; Yang, Zhenguo

    2012-12-01

    The sodium-nickel chloride (ZEBRA) battery is operated at relatively high temperature (250-350 °C) to achieve adequate electrochemical performance. Reducing the operating temperature in the range of 150200 °C can not only lead to enhanced cycle life by suppressing temperature-related degradations, but also allow the use of lower cost materials for construction. To achieve adequate electrochemical performance at lower operating temperatures, reduction in ohmic losses is required, including the reduced ohmic resistance of β″-alumina solid electrolyte (BASE) and the incorporation of low melting point secondary electrolytes. In present work, planar-type Na/NiCl2 cells with a thin BASE (600 μm) and low melting point secondary electrolyte were evaluated at reduced temperatures. Molten salts used as secondary electrolytes were fabricated by the partial replacement of NaCl in the standard secondary electrolyte (NaAlCl4) with other lower melting point alkali metal salts such as NaBr, LiCl, and LiBr. Electrochemical characterization of these ternary molten salts demonstrated improved ionic conductivity and sufficient electrochemical window at reduced temperatures. Furthermore, Na/NiCl2 cells with 50 mol% NaBr-containing secondary electrolyte exhibited reduced polarizations at 175 °C compared to the cell with the standard NaAlCl4 catholyte. The cells also exhibited stable cycling performance even at 150 °C.

  4. Searching for an idealistic nuclear energy system. The thorium molten-salt nuclear energy synergetic system

    International Nuclear Information System (INIS)

    The solar-based energy should become a global major one at the end of this century. The intermediate between fossil and solar energies has to be filled by the fission energy. For the decisive improvement of its safety, radioactive-waste and plutonium issues, the Thorium Molten-Salt Nuclear Energy Synergetic System should be established by the use of fluid-fuel of 7LiF-BeF2-based molten fluoride salt, as a triple functional medium for nuclear-reaction, heat-transfer and chemical-process without radiation-damage. This system is composed of simple and economical fuel-self-sustaining small power reactors and Accelerator Molten-Salt Breeders for spallation breeding. The economical extinction of radioactive-wastes will be achieved by using excess fuel (neutron) at the phase-out period of this system near the end of this century. This technology will be essential for solving global environmental and poverty issues and for the complete abolition of nuclear weapons. (author)

  5. Nuclear analysis and optimization of the molten-salt fusion hybrid reactor

    International Nuclear Information System (INIS)

    An improved method of studying the neutronic characteristics of fusion hybrid reactor blankets has been developed. Two major improvements over previous analysis methods have been accomplished. The first of these improvements is the introduction of one-dimensional, homogenized-region blanket neutronic models in which resonance and spatial self-shielding effects are treated explicitly. The second improvement involves the application of an iterative gradient-ascent based optimization scheme. In this method, key blanket dimensions and concentrations are automatically varied in a search for a configuration which maximizes neutronic performance. The specific fusion hybrid blanket design to which these new methods of analysis are applied in an evolution of the U-233 producing molten-salt-in-tubes concept studied by Lawrence Livermore National Laboratory (LLNL). Optimistic analysis techniques initially predicted the fissile fuel production capacity of this blanket to be 6400 kg of U-233 per year when driven by a 3000 MW tandem mirror fusion driver. The improved and more realistic analysis techniques employed in this study predict that an optimized molten-salt blanket design will produce over 6700 kg of U-233 per year when driven by the same tandem mirror device. Finally, the techniques and data base developed in this study have been designed to be easily extended to the task of performing future, more extensive analysis. Such an analysis might involve the minimization of fuel costs in an entire fusion hybrid reactor complex. 20 refs., 10 figs., 14 tabs

  6. Molten Salt Power Tower Cost Model for the System Advisor Model (SAM)

    Energy Technology Data Exchange (ETDEWEB)

    Turchi, C. S.; Heath, G. A.

    2013-02-01

    This report describes a component-based cost model developed for molten-salt power tower solar power plants. The cost model was developed by the National Renewable Energy Laboratory (NREL), using data from several prior studies, including a contracted analysis from WorleyParsons Group, which is included herein as an Appendix. The WorleyParsons' analysis also estimated material composition and mass for the plant to facilitate a life cycle analysis of the molten salt power tower technology. Details of the life cycle assessment have been published elsewhere. The cost model provides a reference plant that interfaces with NREL's System Advisor Model or SAM. The reference plant assumes a nominal 100-MWe (net) power tower running with a nitrate salt heat transfer fluid (HTF). Thermal energy storage is provided by direct storage of the HTF in a two-tank system. The design assumes dry-cooling. The model includes a spreadsheet that interfaces with SAM via the Excel Exchange option in SAM. The spreadsheet allows users to estimate the costs of different-size plants and to take into account changes in commodity prices. This report and the accompanying Excel spreadsheet can be downloaded at https://sam.nrel.gov/cost.

  7. Determination and evaluation of the thermophysical properties of an alkali carbonate eutectic molten salt.

    Science.gov (United States)

    An, Xuehui; Cheng, Jinhui; Zhang, Peng; Tang, Zhongfeng; Wang, Jianqiang

    2016-08-15

    The thermal physical properties of Li2CO3-Na2CO3-K2CO3 eutectic molten salt were comprehensively investigated. It was found that the liquid salt can remain stable up to 658 °C (the onset temperature of decomposition) by thermal analysis, and so the investigations on its thermal physical parameters were undertaken from room temperature to 658 °C. The density was determined using a self-developed device, with an uncertainty of ±0.00712 g cm(-3). A cooling curve was obtained from the instrument, giving the liquidus temperature. For the first time, we report the obtainment of the thermal diffusivity using a laser flash method based on a special crucible design and establishment of a specific sample preparation method. Furthermore, the specific heat capacity was also obtained by use of DSC, and combined with thermal diffusivity and density, was used to calculate the thermal conductivity. We additionally built a rotating viscometer with high precision in order to determine the molten salt viscosity. All of these parameters play an important part in the energy storage and transfer calculation and safety evaluation for a system. PMID:27203821

  8. Thermodynamic analysis on the direct preparation of metallic vanadium from NaVO3 by molten salt electrolysis☆

    Institute of Scientific and Technical Information of China (English)

    Wei Weng; Mingyong Wang; Xuzhong Gong; Zhi Wang; Zhancheng Guo

    2016-01-01

    A novel and environmentally friendly route to directly prepare metallic vanadium from NaVO3 by molten salt electrolysis is proposed. The feasibility about the direct electro-reduction of NaVO3 to metallic vanadi-um is analyzed based on the thermodynamic calculations and experimental verifications. The theoretical decomposition voltage of NaVO3 to metallic vanadium is only 0.47 V at 800 °C and much lower than that of the alkali and alkali earth metal chloride salts. The value is slightly higher than that of low-valence vanadium oxides such as V2O3, V3O5 and VO. However, the low-valence vanadium oxides can be further electro-reduced to metallic vanadium thermodynamically. The thermodynamic analysis is verified by the experimental results. The direct preparation of metallic vanadium from NaVO3 by molten salt electrolysis is feasible.

  9. Recommendations for a demonstrator of Molten Salt Fast Reactor

    International Nuclear Information System (INIS)

    There is currently a renewed interest in molten salt reactors, due to recent conceptual developments on fast neutron spectrum molten salt reactors (MSFRs) using fluoride salts. This concept, operated in the Thorium fuel cycle, may be started either with 233U, enriched U and/or TRU elements as initial fissile load. It has been recognized as a long term alternative to solid fuelled fast neutron systems with a unique potential (such as large negative temperature and void coefficients, lower fissile inventory, no initial criticality reserve, simplified fuel cycle, wastes reduction…) and is thus one of the reference reactors of the Generation IV International Forum. This paper will focus on recommendations to define a demonstrator representing the key points of the reference MSFR power reactor (3000 MWth, fuel salt volume of 18 m3). The MSFR demonstrator is designed to assess the technological choices of this innovative system (fuel salt, structural materials, fuel heat exchangers…). It seems finally possible to slightly modify such a demonstrator which could then be a self-breeder modular reactor. (author)

  10. A three-dimensional simulation of a parabolic trough solar collector system using molten salt as heat transfer fluid

    International Nuclear Information System (INIS)

    Investigations on the thermal physics mechanisms of the parabolic trough collector systems (PTCs) play a vital role in the utilization of solar energy. In this paper, a three-dimensional simulation based on Finite Element Method (FVM) is established to solve the complex problem coupling with radiation, heat conduction and convection in the PTCs. The performances of the PTCs using molten salt as the heat transfer fluid were numerically studied, and the influences of the key operating parameters on the PTCs were investigated. As a result, it can be found that the circumferential temperature difference (CTD) of the absorber increases with the rising of the direct normal irradiance (DNI) and decreases with the increase of heat transfer fluid (HTF) inlet temperature and inlet velocity. With the velocity of the molten salt in the range of 1 m/s–4 m/s, the DNIs of 500 W/m2–1250 W/m2 and the inlet temperature of 623 K–825 K, the CTD of the absorber can reach 12 K–42 K. Furthermore, the numerical results indicate the non-uniform distribution of the solar energy flux affects the CTD of receiver while has a little influence on the thermal efficiency. The promising results will provide a reference for the design of the novel parabolic trough solar collectors. - Highlights: • A coupled three-dimensional simulation is established. • The performance of the PTCs with molten salt as HTF was obtained. • The effects of key parameters on the PTCs with molten salt were achieved. • The coupling characteristics of thermal and fluid of the receiver were disclosed

  11. Effects of Different Molten-salt on the Synthesis of Hexagonal Barium Ferrite

    Institute of Scientific and Technical Information of China (English)

    WANG Jing-ping; LIU Ying; HU Min; ZHANG Mi-lin

    2006-01-01

    Hexagonal-plate BaFe12O19 nano-particles were prepared by combining the co-precipitation and molten-salt method. The effects of molten-salt NaCl on the products have been compared with those of KCl. The X-ray diffraction analysis,the environment scanning electron microscopy and the vibrating sample magnetometer were used to measure their characteristics. The results show that the molten-salt KCl is helpful to obtain pure phase while the molten-salt NaCl will prompt the crystallinity of particles to increase. It is also found that, given the molten-salt of NaCl and that of KCl of the same quantity, the particles formed are also the same size. The size of particles slightly increases as the quantity of the salt increases. Moreover, the sample prepared under 200% NaCl salt system has magnetization of 71.5 emu/g.

  12. Pyrochemical process in molten salt for the nuclear engineering field. Focus on the electrorefining process

    International Nuclear Information System (INIS)

    The pyrochemical process in molten salt for spent fuel reprocessing and radioactive waste treatment has recently gained attention in the Japanese nuclear engineering field. The electrorefining or electrowinning process in molten salt is suitable for the recovery of U and transuranium elements (TRUs) which are difficult to recover from spent fuel in aqueous solution because of the negative redox potentials. Initial work on metallic fuel reprocessing by a pyrochemical process began in Japan in 1987 and is almost established at an engineering stage. The TRUs recovery process from high-level radioactive liquid waste has been fundamentally established. The Long-lived fission products (LLFPs) recovery process by electromigration in molten salt has been studied in fundamental data. However, application of the pyrochemical process in molten salt for radioactive waste treatment has just started. This paper reviews pyrochemical processes in molten salt for application in the nuclear engineering field from the view point of process development. (author)

  13. Molten salt destruction of rubber and chlorinated solvents

    International Nuclear Information System (INIS)

    Acceptable methods for the treatment of mixed wastes are not currently available. The authors have investigated Molten Salt Destruction (MSD) as an alternative to incineration of mixed wastes. MSD differs from incineration in several ways: there is no evidence of open flames in MSD, the containment of actinides is accomplished by chemical means (wetting and dissolution), the operating temperature of MSD is much lower (700--590 C vs 1,000--1,200 C) thus lowering the volatility of actinides. Furthermore, no acid gases are released from MSD. These advantages provide the main incentive for developing MSD as an alternative to incineration. The authors have demonstrated the viability of the MSD process to cleanly destroy rubber and chlorinated solvents

  14. Dismantled weapons fuel burning in molten salt reactors

    International Nuclear Information System (INIS)

    The advantages of burning fissile material from dismantled weapons in molten salt reactors (MSRs) are described. The fluid fuel MSRs with some, or full, processing are nondedicated reactors that generate energy and completely burn the fissile material on a continuous basis. No fuel fabrication is needed, and the entire dismantling can be done in a secure facility. Shipments are made in small, safe, and secure quantities. Denaturing, spiking, or mixing can be done at the source for added safety. MSRs are very safe reactors that help close the fuel cycle and simplify waste treatment, thereby contributing to acceptability. Additionally, MSRs are expected to be economically competitive as electric power stations. The safety, security, simplicity, economy, and proliferation resistant properties support the deployment in countries that have the need

  15. Heats of Mixing in Binary Systems of Molten Salts

    International Nuclear Information System (INIS)

    The heat of mixing is an important thermodynamic property in binary mixtures. As a result of the recent development of high-temperature calorimetry we have been able to determine directly the heat of mixing in binary systems of molten salts. In this work we present the results of thermochemical measurements carried out in our laboratories for the systems (Rb-K)Cl; (Rb-Na)Cl; (Ag-Na)Cl; (Na-K)Br and(Br-Cl)Na for different concentrations and temperatures. In our view, the most significant components of the heat of mixing are the ionic contribution and the polarization energy of the ions. Consequently, use could be made of a relation of the form: ΔHM = Qi - Qp. The heat of mixing can then have either positive or negative values depending on the sign and the preponderance of the Qi and Qp energies. (author)

  16. Quantitative analysis of cesium in synthetic lithium molten salts

    International Nuclear Information System (INIS)

    An analytical technique for fission products in lithium molten salts of spent PWR (Pressurized Water Reactor) fuels has been studied for the establishment of optimum chemical engineering process and the evaluation of process material balance in developing Direct Oxide Reduction Process with lithium metal. As part of the basic research, synthetic dissolver solutions of lithium chloride containing trace amounts of fission product elements (La, Ce, Pr, Nd, Sm, Eu, Gd, Y, Cs, Ru, Rh, Pd, Mo, Zr, Cd, Ba, Sr, Te and Se) was prepared and used in establishing the selective separation technique of cesium from lithium chloride matrix using cation exchange chromatography. Its recovery was measured by flame atomic absorption spectrometry and the reliability of this technique was evaluate

  17. Filbe molten salt research for tritium breeder applications

    International Nuclear Information System (INIS)

    This paper presents an overview of Flibe (2Lif·BeF2) molten salt research activities conducted at the INEEL as part of the Japan-US JUPITER-II joint research program. The research focuses on tritium/chemistry issues for self-cooled Flibe tritium breeder applications and includes the following activities: (1) Flibe preparation, purification, characterization and handling, (2) development and testing of REDOX strategies for containment material corrosion control, (3) tritium behavior and management in Flibe breeder systems, and (4) safety testing (e.g., mobilization of Flibe during accident scenarios). This paper describes the laboratory systems developed to support these research activities and summarizes key results of this work to date. (author)

  18. Electrodeposition of aluminum on aluminum surface from molten salt

    Institute of Scientific and Technical Information of China (English)

    Wenmao HUANG; Xiangyu XIA; Bin LIU; Yu LIU; Haowei WANG; Naiheng MA

    2011-01-01

    The surface morphology,microstructure and composition of the aluminum coating of the electrodeposition plates in AlC13-NaC1-KC1 molten salt with a mass ratio of 8:1:1 were investigated by SEM and EDS.The binding force was measured by splat-cooling method and bending method.The results indicate that the coatings with average thicknesses of 12 and 9 μm for both plates treated by simple grinding and phosphating are compacted,continuous and well adhered respectively. Tetramethylammonium chloride (TMAC) can effectively prevent the growth of dendritic crystal,and the anode activation may improve the adhesion of the coating. Binding force analysis shows that both aluminum coatings are strongly adhered to the substrates.

  19. Electrochemical synthesis of niobium-hafnium coatings in molten salts

    International Nuclear Information System (INIS)

    Graphite is widely used in technology because of its unique properties. A drawback of graphite is its low heat resistance in oxidizing atmospheres. To increase its heat resistance, Nb-Hf protective coatings were synthesized. Electrodeposition of niobium coatings on graphite with subsequent precise surface alloying of niobium with hafnium was studied. Electrochemical synthesis of Nb-Hf coatings from molten salt systems containing compounds of niobium and hafnium was used too. It was shown that Nb-Hf coatings with a planar growing front can be obtained if the concentration and therefore the limiting current density of the more electropositive component Nb is kept low. Nb-Hf coatings with a thickness of 20 - 30 μm have been obtained in this way from an NaCl-KCl-K2NbF7 (1 wt%)-K2HfF6 (10 wt%)-NaF (5 wt%) melt, above the limiting current density of niobium deposition. (orig.)

  20. Electrochemical Synthesis of Niobium-Hafnium Coatings in Molten Salts

    Science.gov (United States)

    Kuznetsov, Sergey A.; Kuznetsova, Svetlana V.

    2007-08-01

    Graphite is widely used in technology because of its unique properties. A drawback of graphite is its low heat resistance in oxidizing atmospheres. To increase its heat resistance, Nb-Hf protective coatings were synthesized. Electrodeposition of niobium coatings on graphite with subsequent precise surface alloying of niobium with hafnium was studied. Electrochemical synthesis of Nb-Hf coatings from molten salt systems containing compounds of niobium and hafnium was used too. It was shown that Nb-Hf coatings with a planar growing front can be obtained if the concentration and therefore the limiting current density of the more electropositive component Nb is kept low. Nb-Hf coatings with a thickness of 20 - 30 μm have been obtained in this way from an NaCl-KCl-K2NbF7 (1 wt%)-K2HfF6 (10 wt%)-NaF (5 wt%) melt, above the limiting current density of niobium deposition.

  1. Steady-state and dynamic behavior of a moderated molten salt reactor

    International Nuclear Information System (INIS)

    Highlights: • Steady-state and transient coupled calculation scheme. • Study of impact of the substance properties on the operating conditions and on the reactivity feedback coefficients. • Several pump-driven and temperature induced full power transients calculated and discussed. - Abstract: The moderated Molten Salt Reactor (MSR) is an attractive breeder reactor. However, the temperature feedback coefficient of such a system can be positive due to the contribution of the moderator, an effect that can only be avoided with special measures. A previous study (Nagy et al., 2010) aimed to find a core design that is a breeder and has negative overall temperature feedback coefficient. In this paper, a coupled calculation scheme, which includes the reactor physics, heat transfer and fluid dynamics calculations is introduced. It is used both for steady-state and for dynamic calculations to evaluate the safety of the core design which was selected from the results of the previous study. The calculated feedback coefficients on the salt and graphite temperatures, power and uranium concentration prove that the core design derived in the previous optimization study is safe because the temperature feedback coefficient of the core and of the power is sufficiently negative. Transient calculations are performed to show the inherent safety of the reactor in case of reactivity insertion. As it is shown, the response of the reactor to these transients is initially dominated by the strong negative feedback of the salt. In all the presented transients, the reactor power stabilizes and the temperature of the salt never approaches its boiling point

  2. Nuclear Hybrid Energy System: Molten Salt Energy Storage (Summer Report 2013)

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Michael George mckellar; Su-Jong Yoon

    2013-11-01

    Effective energy use is a main focus and concern in the world today because of the growing demand for energy. The nuclear hybrid energy system (NHES) is a valuable technical concept that can potentially diversify and leverage existing energy technologies. This report considers a particular NHES design that combines multiple energy systems including a nuclear reactor, energy storage system (ESS), variable renewable generator (VRG), and additional process heat applications. Energy storage is an essential component of this particular NHES because its design allows the system to produce peak power while the nuclear reactor operates at constant power output. Many energy storage options are available, but this study mainly focuses on a molten salt ESS. The primary purpose of the molten salt ESS is to enable the nuclear reactor to be a purely constant heat source by acting as a heat storage component for the reactor during times of low demand, and providing additional capacity for thermo-electric power generation during times of peak electricity demand. This report will describe the rationale behind using a molten salt ESS and identify an efficient molten salt ESS configuration that may be used in load following power applications. Several criteria are considered for effective energy storage and are used to identify the most effective ESS within the NHES. Different types of energy storage are briefly described with their advantages and disadvantages. The general analysis to determine the most efficient molten salt ESS involves two parts: thermodynamic, in which energetic and exergetic efficiencies are considered; and economic. Within the molten salt ESS, the two-part analysis covers three major system elements: molten salt ESS designs (two tank direct and thermocline), the molten salt choice, and the different power cycles coupled with the molten salt ESS. Analysis models are formulated and analyzed to determine the most effective ESS. The results show that the most

  3. Corrosion behaviour of Ti3SiC2 with LiF-NaF-KF molten salt

    International Nuclear Information System (INIS)

    Background: Recently, the molten salt reactor (MSR), a generation IV fission reactor candidate, has drawn much attention because of its intrinsic safety. However, the harsh service environment of the MSR, especially the corrosion environment, raises many challenges in terms of applying structural materials, such as intergranular cracking and embrittlement. MAX phases materials are promising structural materials that can be used in MSRs. However, the corrosion behaviour of these materials in molten LiF-NaF-KF (FLiNaK) is yet to be evaluated. Purpose: The present work is a preliminary investigation of the corrosion behaviour of MAX phase materials in molten fluoride salts and aims to understand the corrosion mechanism of MAX phases to enable their application in next generation MSR. Methods: We choose two common MAX phases: Ti3SiC2 and Ti3AlC2 as experiment subject. The corrosion tests were performed at 850℃ for 144 h in airtight graphite crucibles under an argon cover gas. Results and Conclusion: The corrosion of these two MAX phases in molten FLiNaK salt mainly showed as the corrosion of element A and then left us mostly cubic TiCx. The difference was that Ti3AlC2 lost Al entirely, however, the loss of Si in Ti3SiC2 occurred only 150 μm depth below the surface. The weight loss data showed that Ti3SiC2 had a much better corrosion resistance than Ti3AlC2. (authors)

  4. Electrodeposition of tin from EMI⋅BF4⋅Cl room temperature molten salts

    Directory of Open Access Journals (Sweden)

    Morimitsu M.

    2003-01-01

    Full Text Available The electrochemistry of Sn(II was investigated with cyclic voltammetry and chronoamperometry in the 1-ethyl-3-methylimidazolium tetrafluoroborate molten salt containing free chloride ions (EMI ⋅BF4 ⋅Cl originated from the mixture of EMIC and NaBF4 (60:40 mol%. The well defined redox waves for the electro deposition and dissolution of tin were observed on a platinum electrode at 303 K. The deposition of tin proceeded through a quasi-reversible step with two electron transfer, and the deposited tin was sufficiently recovered during oxidation. The experimental current-time transient coincided with the theory based on one-dimensional diffusion control.

  5. Characterization of the molten salt reactor experiment fuel and flush salts

    International Nuclear Information System (INIS)

    Wise decisions about the handling and disposition of spent fuel from the Molten Salt Reactor Experiment (MSRE) must be based upon an understanding of the physical, chemical, and radiological properties of the frozen fuel and flush salts. These open-quotes staticclose quotes properties can be inferred from the extensive documentation of process history maintained during reactor operation and the knowledge gained in laboratory development studies. Just as important as the description of the salt itself is an understanding of the dynamic processes which continue to transform the salt composition and govern its present and potential physicochemical behavior. A complete characterization must include a phenomenological characterization in addition to the typical summary of properties. This paper reports on the current state of characterization of the fuel and flush salts needed to support waste management decisions

  6. Magneto-hydrodynamic detection of vortex shedding for molten salt flow sensing.

    Energy Technology Data Exchange (ETDEWEB)

    Kruizenga, Alan Michael; Crocker, Robert W.

    2012-09-01

    High temperature flow sensors must be developed for use with molten salts systems at temperatures in excess of 600%C2%B0C. A novel magneto-hydrodynamic sensing approach was investigated. A prototype sensor was developed and tested in an aqueous sodium chloride solution as a surrogate for molten salt. Despite that the electrical conductivity was a factor of three less than molten salts, it was found that the electrical conductivity of an electrolyte was too low to adequately resolve the signal amidst surrounding noise. This sensor concept is expected to work well with any liquid metal application, as the generated magnetic field scales proportionately with electrical conductivity.

  7. Synthesis of TiNi/Ti2Ni Composite Particles in Molten Salts

    Institute of Scientific and Technical Information of China (English)

    YANG Rui-song; CUI Li-shan; ZHENG Yan-jun

    2006-01-01

    A new process of synthesizing TiNi/Ti2Ni composite particles, high temperature molten salts method, is introduced. This method uses molten salts as a reaction medium that does not take part in the chemical reaction and can be easily dissolved in rinsing water. According this method, the composite particles were prepared in molten salts at 700 ℃-900 ℃. By means of differential scanning calorimetry (DSC), the reversible martensitic transformation of TiNi particles in these composite particles was confirmed.

  8. Very Efficient Nucleophilic Aromatic Fluorination Reaction in Molten Salts: A Mechanistic Study

    International Nuclear Information System (INIS)

    We report a quantum chemical study of an extremely efficient nucleophilic aromatic fluorination in molten salts. We describe that the mechanism involves solvent anion interacting with the ion pair nucleophile M+F- (M = Na, K, Rb, Cs) to accelerate the reaction. We show that our proposed mechanism may well explain the excellent efficiency of molten salts for SNAr reactions, the relative efficacy of the metal cations, and also the observed large difference in rate constants in two molten salts (n-C4H9)4N+ CX3SO3-, (X=H, F) with slightly different sidechain (-CH3 vs. -CF3)

  9. Molten salt reactors. Synthesis of studies realized between 1973 and 1983. Experimental loop file

    International Nuclear Information System (INIS)

    Four test loops were developed for the experimental study of a molten salt reactor with lead salt direct contact. A molten salt loop, completely in graphite, including the pump, showed that this material is convenient for salt containment and circulation. Reactor components like flowmeters, electromagnetic pumps, pressure gauge, valves developed for liquid sodium, were tested with liquid lead. A water-mercury loop was built for lead-molten salt simulation studies. Finally a lead-salt loop (COMPARSE) was built to study the behaviour of salt particles carried by lead in the heat exchanger

  10. Corrosion effects between molten salts and thermal storage material for concentrated solar power plants

    International Nuclear Information System (INIS)

    Highlights: ► Thermal energy storage combining sensible heat and phase change materials (PCMs). ► Asbestos-containing wastes valorisation in concentrated solar power plants. ► Interactions between inertized asbestos-containing waste and molten salts. ► High temperature NMR characterization of corrosion. -- Abstract: Today, thermal energy storage (TES) is a key issue for concentrated solar power plants (CSPs). The available and mature technologies of TES do not mach all the actualised criteria for those properties. Alternative approaches have to be identified and developed to guaranty the expected extension of CSP implementations with respect to the IEA 2050 scenario. In this context, promising hybrid TES systems based upon the combination of sensible heat and liquid/solid phase change material (PCM) sub-systems are considered. For the sensible heat stage, a recycled refractory ceramic made of inertised asbestos-containing waste (IACW) is proposed. For the PCM stage, high temperature inorganic salts are considered. One major aspect of the hybrid TES is the integration of the two stages together. Therefore, the present study is focussed on the needed assessment concerning the compatibility between the IACW and the molten salts in terms of corrosion. Sulphates, phosphates, carbonates and nitrates salts have been experimented and the corrosion effects characterised by in situ NMR, ex situ X-ray diffraction and Scanning Electron Microscopy. Among those available salts, only the nitrates have shown good compatibility with IACW materials. For higher temperature levels, other salts or eutectics will have to be considered to allow hybrid TES with direct contact.

  11. Transient analyses for a molten salt fast reactor with optimized core geometry

    International Nuclear Information System (INIS)

    Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential

  12. Incorporating supercritical steam turbines into molten-salt power tower plants : feasibility and performance.

    Energy Technology Data Exchange (ETDEWEB)

    Pacheco, James Edward; Wolf, Thorsten [Siemens Energy, Inc., Orlando, FL; Muley, Nishant [Siemens Energy, Inc., Orlando, FL

    2013-03-01

    Sandia National Laboratories and Siemens Energy, Inc., examined 14 different subcritical and supercritical steam cycles to determine if it is feasible to configure a molten-salt supercritical steam plant that has a capacity in the range of 150 to 200 MWe. The effects of main steam pressure and temperature, final feedwater temperature, and hot salt and cold salt return temperatures were determined on gross and half-net efficiencies. The main steam pressures ranged from 120 bar-a (subcritical) to 260 bar-a (supercritical). Hot salt temperatures of 566 and 600%C2%B0C were evaluated, which resulted in main steam temperatures of 553 and 580%C2%B0C, respectively. Also, the effects of final feedwater temperature (between 260 and 320%C2%B0C) were evaluated, which impacted the cold salt return temperature. The annual energy production and levelized cost of energy (LCOE) were calculated using the System Advisory Model on 165 MWe subcritical plants (baseline and advanced) and the most promising supercritical plants. It was concluded that the supercritical steam plants produced more annual energy than the baseline subcritical steam plant for the same-size heliostat field, receiver, and thermal storage system. Two supercritical steam plants had the highest annual performance and had nearly the same LCOE. Both operated at 230 bar-a main steam pressure. One was designed for a hot salt temperature of 600%C2%B0C and the other 565%C2%B0C. The LCOEs for these plants were about 10% lower than the baseline subcritical plant operating at 120 bar-a main steam pressure and a hot salt temperature of 565%C2%B0C. Based on the results of this study, it appears economically and technically feasible to incorporate supercritical steam turbines in molten-salt power tower plants.

  13. Pyrochemical reprocessing of molten salt fast reactor fuel: focus on the reductive extraction step

    Directory of Open Access Journals (Sweden)

    Rodrigues Davide

    2015-12-01

    Full Text Available The nuclear fuel reprocessing is a prerequisite for nuclear energy to be a clean and sustainable energy. In the case of the molten salt reactor containing a liquid fuel, pyrometallurgical way is an obvious way. The method for treatment of the liquid fuel is divided into two parts. In-situ injection of helium gas into the fuel leads to extract the gaseous fission products and a part of the noble metals. The second part of the reprocessing is performed by ‘batch’. It aims to recover the fissile material and to separate the minor actinides from fission products. The reprocessing involves several chemical steps based on redox and acido-basic properties of the various elements contained in the fuel salt. One challenge is to perform a selective extraction of actinides and lanthanides in spent liquid fuel. Extraction of actinides and lanthanides are successively performed by a reductive extraction in liquid bismuth pool containing metallic lithium as a reductive reagent. The objective of this paper is to give a description of the several steps of the reprocessing retained for the molten salt fast reactor (MSFR concept and to present the initial results obtained for the reductive extraction experiments realized in static conditions by contacting LiF-ThF4-UF4-NdF3 with a lab-made Bi-Li pool and for which extraction efficiencies of 0.7% for neodymium and 14.0% for uranium were measured. It was concluded that in static conditions, the extraction is governed by a kinetic limitation and not by the thermodynamic equilibrium.

  14. Transient analyses for a molten salt fast reactor with optimized core geometry

    Energy Technology Data Exchange (ETDEWEB)

    Li, R., E-mail: rui.li@kit.edu [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Wang, S.; Rineiski, A.; Zhang, D. [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Merle-Lucotte, E. [Laboratoire de Physique Subatomique et de Cosmologie – IN2P3 – CNRS/Grenoble INP/UJF, 53, rue des Martyrs, 38026 Grenoble (France)

    2015-10-15

    Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.

  15. Determination of Stoichiometry of Solutes in Molten Salt Solvents by Correlations of Relative Raman Band Intensities

    DEFF Research Database (Denmark)

    Boghosian, Soghomon; Berg, Rolf W.

    1999-01-01

    Raman spectroscopy has been used to determine the stoichiometry of solute complexes in molten salts at high temperatures under static equilibrium conditions, A simple formalism is derived for correlating relative Raman band intensities with stoichiometric coefficients. The experimental procedures...

  16. Molten-salt reactor program. Semiannual progress report for period ending February 29, 1976

    Energy Technology Data Exchange (ETDEWEB)

    McNeese, L.E.

    1976-08-01

    Separate abstracts and indexing were prepared for sections dealing with MSBR design and development; chemistry of fuel-salt and coolant-salt systems and analytical methods; materials development; fuel processing for molten-salt reactors; and salt production. (DG)

  17. Molten salt reactors. Synthesis of studies realized between 1973 and 1983. Chemistry file

    International Nuclear Information System (INIS)

    The chemistry of molten salt reactors was first acquired by foreign literature and developed by experimental studies. Salt preparation, analysis, chemical and electrochemical properties, interaction with metals or graphites and use of molten lead for direct cooling are examined

  18. Development of High-Temperature Transport System for Molten Salt in Pyroprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sung Ho; Kim, In Tae; Park, Sung Bin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The electrorefining process, which is a key process in pyroprocessing, is composed of two parts, electrorefining to deposit a uranium with a solid cathode and electrowinning to co-deposit TRU and RE with a liquid cadmium cathode (LCC). As the electrorefining operation proceedes, TRU and RE are accumulated in electrolyte LiCl-KCl salt, and after the electrorefining process, the molten salt used in an electrorefining reactor should by transported to the next process, the electrowinning process, to recover U/TRU/RE; Thus, a molten salt transfer system by suction is now being developed. An apparatus for suction transport experiments was designed and constructed for the development of high- temperature molten salt transport technology. Suction transport experiments were performed using LiC-KCl eutectic salt. The feasibility of pyro-reprocessing has been demonstrated through many laboratory-scale experiments. In pyroprocessing, a eutectic LiCl-KCl salt was used as a liquid elextrolyte for a recovery of actinides. However, reliable transport technologies for these high temperature liquids have not yet been developed. A preliminary study on high-temperature transport technology for molten salt by suction is now being carried out. In this study, three different salt transport technologies (gravity, suction pump, and centrifugal pump) were investigated to select the most suitable method for molten salt transport. An apparatus for suction transport experiments was designed and installed for the development of high-temperature molten salt transport technology. Basic preliminary suction transport experiments were carried out using the prepared LiC-KCl eutectic salt at 500 .deg. C to observe the transport behavior of LiCl-KCl molten salt. In addition, a PRIDE salt transport system was designed and installed for an engineering-scale salt transport demonstration. Several types of suction transport experiments using molten salt (LiCl-KCl eutectics) for the development of a high

  19. Advanced Thermal Storage System with Novel Molten Salt: December 8, 2011 - April 30, 2013

    Energy Technology Data Exchange (ETDEWEB)

    Jonemann, M.

    2013-05-01

    Final technical progress report of Halotechnics Subcontract No. NEU-2-11979-01. Halotechnics has demonstrated an advanced thermal energy storage system with a novel molten salt operating at 700 degrees C. The molten salt and storage system will enable the use of advanced power cycles such as supercritical steam and supercritical carbon dioxide in next generation CSP plants. The salt consists of low cost, earth abundant materials.

  20. Thermal Energy Storage in Molten Salts: Overview of Novel Concepts and the DLR Test Facility (TESIS)

    OpenAIRE

    Breidenbach, Nils; Martin, Claudia; Jockenhöfer, Henning; Bauer, Thomas

    2016-01-01

    At present, two-tank molten salt storage systems are the established commercially available concept for solar thermal power plants. Due to their very low vapour pressure and comparatively high thermal stability, molten salts are preferred as the heat transfer fluid and storage medium. Therefore, the development of alternative, more cost-effective concepts is an important step in making thermal energy storage more competitive for industrial processes and solar thermal applications. The pape...

  1. Deposition of niobium plate on niobium-titanium from molten salts

    International Nuclear Information System (INIS)

    A possibility of using Nb-Ti alloys (50 and 34 mas.% of Ti) as substrates for deposition of niobium coating of chloride-fluoride and fluoride molten salts is studied. Corrosion behaviour of alloys indicates in the electrolytic bath within 970-1070 K interval, coating structure and state of coating-substrate boundary are investigated. Chloride-fluoride molten salt usefullness for making products with niobium coatings is shown

  2. Recovery of plutonium from molten salt extraction residues

    International Nuclear Information System (INIS)

    Savannah River Laboratory (SRL), Savannah River Plant (SRP), and Rocky Flats Plant (RFP) are jointly developing a process to recover plutonium from molten salt extraction residues. These NaCl, KCl, MgCl2 residues, which are generated in the pyrochemical extraction of 241Am from aged plutonium metal, contain up to 25 wt % dissolved PUCl3 and up to 2 wt % AmCl3. The objective is to develop a process to convert these residues to plutonium metal product and discardable waste. The first step of the conceptual process is to convert the actinides to a heterogenous scrub alloy with aluminum and magnesium. This step, performed at RFP, effectively separates the actinides from the bulk of the chloride. This scrub alloy will then be dissolved in a HNO3-HF solution at SRP. Residual chloride will be removed by precipitation with Hg2(NO3)2 followed by centrifugation. Plutonium and americium will be separated using the Purex solvent extraction process. The 241Am will be diverted to the solvent extraction waste stream where it can either be discarded to the waste farm or recovered. The plutonium will be finished via PuF3 precipitation, oxidation to a mixture of PUF4 and PuO2, followed by reduction to plutonium metal with calcium

  3. Optimized molten salt receivers for ultimate trough solar fields

    Science.gov (United States)

    Riffelmann, Klaus-J.; Richert, Timo; Kuckelkorn, Thomas

    2016-05-01

    Today parabolic trough collectors are the most successful concentrating solar power (CSP) technology. For the next development step new systems with increased operation temperature and new heat transfer fluids (HTF) are currently developed. Although the first power tower projects have successfully been realized, up to now there is no evidence of an all-dominant economic or technical advantage of power tower or parabolic trough. The development of parabolic trough technology towards higher performance and significant cost reduction have led to significant improvements in competitiveness. The use of molten salt instead of synthetic oil as heat transfer fluid will bring down the levelized costs of electricity (LCOE) even further while providing dispatchable energy with high capacity factors. FLABEG has developed the Ultimate TroughTM (UT) collector, jointly with sbp Sonne GmbH and supported by public funds. Due to its validated high optical accuracy, the collector is very suitable to operate efficiently at elevated temperatures up to 550 °C. SCHOTT will drive the key-innovations by introducing the 4th generation solar receiver that addresses the most significant performance and cost improvement measures. The new receivers have been completely redesigned to provide a product platform that is ready for high temperature operation up to 550 °C. Moreover distinct product features have been introduced to reduce costs and risks in solar field assembly and installation. The increased material and design challenges incurred with the high temperature operation have been reflected in sophisticated qualification and validation procedures.

  4. Decommissioning of the Molten Salt Reactor Experiment: A technical evaluation

    International Nuclear Information System (INIS)

    This report completes a technical evaluation of decommissioning planning for the former Molten Salt Reactor Experiment, which was shut down in December, 1969. The key issues revolve around the treatment and disposal of some five tons of solid fuel salt which contains over 30 kg of fissionable uranium-233 plus fission products and higher actinides. The chemistry of this material is complicated by the formation of elemental fluorine via a radiolysis reaction under certain conditions. Supporting studies carried out as part of this evaluation include (a) a broad scope analysis of possible options for storage/disposal of the salts, (b) calculation of nuclide decay in future years, (c) technical evaluation of the containment facility and hot cell penetrations, (d) review and update of surveillance and maintenance procedures, (e) measurements of facility groundwater radioactivity and sump pump operation, (f) laboratory studies of the radiolysis reaction, and (g) laboratory studies which resulted in finding a suitable getter for elemental fluorine. In addition, geologic and hydrologic factors of the surrounding area were considered, and also the implications of entombment of the fuel in-place with concrete. The results of this evaluation show that the fuel salt cannot be left in its present form and location permanently. On the other hand, extended storage in its present form is quite acceptable for 20 to 30 years, or even longer. For continued storage in-place, some facility modifications are recommended. 30 refs., 5 figs., 9 tabs

  5. Molten salt synthesis of mullite nanowhiskers using different silica sources

    Institute of Scientific and Technical Information of China (English)

    Tao Yang; Peng-long Qiu; Mei Zhang; Kuo-Chih Chou; Xin-mei Hou; Bai-jun Yan

    2015-01-01

    Mullite nanowhiskers with Al-rich structure were prepared by molten salt synthesis at 1000°C for 3 h in air using silica, amor-phous silica, and ultrafine silica as the silica sources. The phase and morphology of the synthesized products were investigated by X-ray dif-fraction, scanning electron microscopy, energy dispersive spectroscopy, and transmission electron microscopy. A thermogravimetric and differential thermal analysis was carried out to determine the reaction mechanism. The results reveal that the silica sources play an important role in determining the morphology of the obtained mullite nanowhiskers. Clusters and disordered arrangements are obtained using common silica and amorphous silica, respectively, whereas the use of ultrafine silica leads to highly ordered mullite nanowhiskers that are 80−120 nm in diameter and 20−30μm in length. Considering the growth mechanisms, mullite nanowhiskers in the forms of clusters and highly ordered arrangements can be attributed to heterogeneous nucleation, whereas disordered mullite nanowhiskers are obtained by homogenous nuclea-tion.

  6. A study on the corrosion-control test of material for molten salt handling(II)

    International Nuclear Information System (INIS)

    On this technical report, corrosion behaviors of Fe-Ni binary alloys in molten salts were investigated in the temperature range of 650∼850 deg C. In a molten salt of LiCl, the internal oxidation occurred in the alloys studied and the corrosion rate followed the parabolic kinetics and it increased with an increase of Fe content. In a mixed molten salt of LiCl-Li2O, an internal oxidation occurred in Fe-rich alloy and uniformal corrosion in Ni-rich alloy. Corrosion behaviors of Fe-Ni-Cr alloys in molten salts were investigated in the temperature range of 650∼850 deg C. In a molten salt of LiCl, an internal oxidation of Fe of the alloy without Cr occurred, and a dense protective oxide scale of LiCrO2 of the alloy with Cr formed. In a mixed molten salt of LiCl-Li2O, an internal oxidation of Fe of the alloy without Cr and an internal oxidation of Cr of the alloy with Cr occurred

  7. Application of MSR MA Burner for Multi-Component NP with Various Nuclear Fuel Cycles

    International Nuclear Information System (INIS)

    In the paper 3-component nuclear power (NP) systems were considered. They include: Thermal Reactors (TR), Fast Breeder Reactors (FBR) and Molten Salt Reactors (MSR). Three types of NP systems with different nuclear fuel cycles (NFCs) were investigated: U-Pu, U-Th and U-Pu-Th. For all NP systems equilibrium states and corresponding system characteristics were defined: contribution of different reactors in total power production and minor actinides (MA) equilibriums. The investigation of system parameters of NP shows advantages and disadvantages of different NFCs. The considered U-Pu system in the equilibrium state has a smaller FBR contribution, larger MSR contribution, and TRU equilibrium is two orders greater than for the U-Th system. The considered U-Pu-Th system is a sort of compromise option. It has minimal FBR contribution, relatively small MSR contribution and almost three times less TRU equilibrium in comparison with the U-Pu system. (authors)

  8. Simulation of radiation dose distribution and thermal analysis for the bulk shielding of an optimized molten salt reactor

    Institute of Scientific and Technical Information of China (English)

    张志宏; 夏晓彬; 蔡军; 王建华; 李长园; 葛良全; 张庆贤

    2015-01-01

    The Chinese Academy of Science has launched a thorium-based molten-salt reactor (TMSR) research project with a mission to research and develop a fission energy system of the fourth generation. The TMSR project intends to construct a liquid fuel molten-salt reactor (TMSR-LF), which uses fluoride salt as both the fuel and coolant, and a solid fuel molten-salt reactor (TMSR-SF), which uses fluoride salt as coolant and TRISO fuel. An optimized 2 MWth TMSR-LF has been designed to solve major technological challenges in the Th-U fuel cycle. Preliminary conceptual shielding design has also been performed to develop bulk shielding. In this study, the radiation dose and temperature distribution of the shielding bulk due to the core were simulated and analyzed by performing Monte Carlo simulations and computational fluid dynamics (CFD) analysis. The MCNP calculated dose rate and neutron and gamma spectra indicate that the total dose rate due to the core at the external surface of the concrete wall was 1.91 µSv/h in the radial direction, 1.16 µSv/h above and 1.33 µSv/h below the bulk shielding. All the radiation dose rates due to the core were below the design criteria. Thermal analysis results show that the temperature at the outermost surface of the bulk shielding was 333.86 K, which was below the required limit value. The results indicate that the designed bulk shielding satisfies the radiation shielding requirements for the 2 MWth TMSR-LF.

  9. The preliminary analysis on the steady-state and kinetic features of the molten salt pebble-bed reactor

    International Nuclear Information System (INIS)

    A novel design concept of molten salt pebble-bed reactor with an ultra-simplified integral primary circuit called 'Nuclear Hot Spring' has been proposed, featured by horizontal coolant flow in a deep pool pebble-bed reactor, providing 'natural safety' features with natural circulation under full power operation and less expensive primary circuit arrangement. In this work, the steady-state physical properties of the equilibrium state of the molten salt pebble-bed reactor are calculated by using the VSOP code, and the steady-state thermo-hydraulic analysis is carried out based on the approximation of absolutely horizontal flow of the coolant through the core. A new concept of 2-dimensional, both axial and radial, multi-pass on-line fuelling scheme is presented. The result reveals that the radial multi-pass scheme provides more flattened power distribution and safer temperature distribution than the one-pass scheme. A parametric analysis is made corresponding to different pebble diameters, the key parameter of the core resistance and the temperature at the pebble center. It is verified that within a wide range of pebble diameters, the maximum pebble center temperatures are far below the safety limit of the fuel, and the core resistance is considerably less than the buoyant force, indicating that the natural circulation under full power operation is achievable and the ultra-simplified integral primary circuit without any pump is possible. For the kinetic properties, it is verified that the negative temperature coefficient is achieved in sufficient under-moderated condition through the preliminary analysis on the temperature coefficients of fuel, coolant and moderator. The requirement of reactivity compensation at the shutdown stages of the operation period is calculated for the further studies on the reactivity control. The molten salt pebble-bed reactor with horizontal coolant flow can provide enhanced safety and economical features. (authors)

  10. Molten Salt Test Loop (MSTL) system customer interface document.

    Energy Technology Data Exchange (ETDEWEB)

    Gill, David Dennis; Kolb, William J.; Briggs, Ronald D.

    2013-09-01

    The National Solar Thermal Test Facility at Sandia National Laboratories has a unique test capability called the Molten Salt Test Loop (MSTL) system. MSTL is a test capability that allows customers and researchers to test components in flowing, molten nitrate salt. The components tested can range from materials samples, to individual components such as flex hoses, ball joints, and valves, up to full solar collecting systems such as central receiver panels, parabolic troughs, or linear Fresnel systems. MSTL provides realistic conditions similar to a portion of a concentrating solar power facility. The facility currently uses 60/40 nitrate %E2%80%9Csolar salt%E2%80%9D and can circulate the salt at pressure up to 40 bar (600psi), temperature to 585%C2%B0C, and flow rate of 44-50kg/s(400-600GPM) depending on temperature. The purpose of this document is to provide a basis for customers to evaluate the applicability to their testing needs, and to provide an outline of expectations for conducting testing on MSTL. The document can serve as the basis for testing agreements including Work for Others (WFO) and Cooperative Research and Development Agreements (CRADA). While this document provides the basis for these agreements and describes some of the requirements for testing using MSTL and on the site at Sandia, the document is not sufficient by itself as a test agreement. The document, however, does provide customers with a uniform set of information to begin the test planning process.

  11. Plutonium and minor actinides utilization in Thorium molten salt reactor

    Science.gov (United States)

    Waris, Abdul; Aji, Indarta K.; Novitrian, Kurniadi, Rizal; Su'ud, Zaki

    2012-06-01

    FUJI-12 reactor is one of MSR systems that proposed by Japan. The original FUJI-12 design considers Th/233U or Th/Pu as main fuel. In accordance with the currently suggestion to stay away from the separation of Pu and minor actinides (MA), in this study we evaluated the utilization of Pu and MA in FUJI-12. The reactor grade Pu was employed in the present study as a small effort of supporting THORIMS-NES scenario. The result shows that the reactor can achieve its criticality with the Pu & MA composition in the fuel of 5.96% or more.

  12. Future fuel cycle closing by molten salt reactors

    International Nuclear Information System (INIS)

    In this paper the situations, reasons and expected time table concerning future nuclear fuel cycle closing and influences of fossil raw materials deficiencies, expected during the coming century are discussed. Asymptotic states in which amount of fission minor actinides is in equilibrium with corresponding fission products amount has been demonstrated. Mass and other properties of chosen isotopes from Czech Republic NPPs (four VVER-440 units of the Dukovany NPP and two VVER-1000 units of the Temelin NPP in the year 2040) are presented. Demo-calculation was chosen to demonstrate principally feasibility of MSR burner to serve about 20 standard PWR 1000MWe. Basic salt composition: 0.4115·LiF+0.103·NaF+0.3755·KF+0.11·MAF3 (molar percent). Geometry: R-Z cylinder R=135cm; height = 310cm; surrounded by graphite reflector. Actinides in power reactors spent fuel, g/kg, Actinide masses in annual MSR feed, Equilibrium masses of heavy metals and Fission products effect on neutron multiplication factor Keff are also presented. In core total volume 17.5 m3 the heavy metal mass was about 19.9 tons, total power release ∼ 2.86 GWth and average power density ∼160.9 MW/m3. About 990 kg/year fission products are generated. Fission products equilibrium amounts in core depend on fission product disposal rate and vary from ∼ 100 kg to 1000 kg

  13. Design of new molten salt thermal energy storage material for solar thermal power plant

    International Nuclear Information System (INIS)

    Highlights: ► New quaternary reciprocal system (K, Na/NO2, Cl, NO3) is prepared. ► This molten salt has a lower melting point. ► This new salt has excellent thermal stability. ► This salt mixture has a reduced cost. - Abstract: In order to obtain molten salt with lower melting point, higher thermal stability and reduced cost relative to previously available materials, a variety of molten salt mixtures of alkali nitrates are investigated by experimental methods. However, since measurements are generally expensive and time-consuming, it is of interest to be able to predict melting point and the component of multi-component systems by using the numerical methods. In this paper, eutectic point and component of a new kind of the quaternary reciprocal system (K, Na/NO2, Cl, NO3) are determined firstly by conformal ionic solution theory. Then thermal stability of the mixtures that show a lower melting point is measured by thermogravimetric analysis device. Experimental results show the agreement between measurements and calculations is found to be very good. This kind of molten salt has a lower melting point, 140 °C. It is thermally stable at temperatures up to 500 °C, and may be used up to 550 °C for short periods. Besides, this molten salt has a reduced cost relative to previous low-melting nitrate mixtures due to the elimination of cesium nitrate and lithium nitrate

  14. Electron beam-driven subcritical cascade molten salt unit for three component conception of nuclear industry

    International Nuclear Information System (INIS)

    The preliminary sketch of one of the possible conceptions to develop the beam-driven subcritical molten salt reactor have been described earlier. The essence of the concept is the use of the molten salt reactor cascade system to reduce the driving power of the linac units. Following the new concept, prerequisities and motivations will be discussed in the frame of the existing world nuclear power system. In order to avoid a deep technological nuclear industrial reorganization for a short-range plan we would like to examine a concept which consists of three types of nuclear reactors. They are thermal, fast, and burner reactor types. A role and advantage of the use of the molten salt subcritical reactors as a burner reactor type have been shown for the existing nuclear technology structure. The discussed nuclear power system with a closed fuel cycle could be supplemented with dry fluoride volatility reprocessing of spent fuel and molten salt centrifugation, thermal diffusion, and electrochemical methods for separation. The general approach of the discussed conception relates to the third component of the nuclear industry that includes accelerator-driven molten salt subcritical cascade burners, cyclotrons or system of electron linacs and non-purex fission product/actinide separation systems. 10 refs., 3 tabs

  15. Space Molten Salt Reactor Concept for Nuclear Electric Propulsion and Surface Power

    Science.gov (United States)

    Eades, M.; Flanders, J.; McMurray, N.; Denning, R.; Sun, X.; Windl, W.; Blue, T.

    Students at The Ohio State University working under the NASA Steckler Grant sought to investigate how molten salt reactors with fissile material dissolved in a liquid fuel medium can be applied to space applications. Molten salt reactors of this kind, built for non-space applications, have demonstrated high power densities, high temperature operation without pressurization, high fuel burn up and other characteristics that are ideal for space fission systems. However, little research has been published on the application of molten salt reactor technology to space fission systems. This paper presents a conceptual design of the Space Molten Salt Reactor (SMSR), which utilizes molten salt reactor technology for Nuclear Electric Propulsion (NEP) and surface power at the 100 kWe to 15 MWe level. Central to the SMSR design is a liquid mixture of LiF, BeF2 and highly enriched U235F4 that acts as both fuel and core coolant. In brief, some of the positive characteristics of the SMSR are compact size, simplified core design, high fuel burn up percentages, proliferation resistant features, passive safety mechanisms, a considerable body of previous research, and the possibility for flexible mission architecture.

  16. Heat Transfer and Latent Heat Storage in Inorganic Molten Salts for Concentrating Solar Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Mathur, Anoop [Terrafore Inc.

    2013-08-14

    A key technological issue facing the success of future Concentrating Solar Thermal Power (CSP) plants is creating an economical Thermal Energy Storage (TES) system. Current TES systems use either sensible heat in fluids such as oil, or molten salts, or use thermal stratification in a dual-media consisting of a solid and a heat-transfer fluid. However, utilizing the heat of fusion in inorganic molten salt mixtures in addition to sensible heat , as in a Phase change material (PCM)-based TES, can significantly increase the energy density of storage requiring less salt and smaller containers. A major issue that is preventing the commercial use of PCM-based TES is that it is difficult to discharge the latent heat stored in the PCM melt. This is because when heat is extracted, the melt solidifies onto the heat exchanger surface decreasing the heat transfer. Even a few millimeters of thickness of solid material on heat transfer surface results in a large drop in heat transfer due to the low thermal conductivity of solid PCM. Thus, to maintain the desired heat rate, the heat exchange area must be large which increases cost. This project demonstrated that the heat transfer coefficient can be increase ten-fold by using forced convection by pumping a hyper-eutectic salt mixture over specially coated heat exchanger tubes. However,only 15% of the latent heat is used against a goal of 40% resulting in a projected cost savings of only 17% against a goal of 30%. Based on the failure mode effect analysis and experience with pumping salt at near freezing point significant care must be used during operation which can increase the operating costs. Therefore, we conclude the savings are marginal to justify using this concept for PCM-TES over a two-tank TES. The report documents the specialty coatings, the composition and morphology of hypereutectic salt mixtures and the results from the experiment conducted with the active heat exchanger along with the lessons learnt during

  17. On-line measurements of UV-VIS spectra in high-temperature molten salt media : development of measuring systems

    International Nuclear Information System (INIS)

    Recently, ionic melts have become attractive reaction media in many fields. Molten salt based electrochemical processes have been proposed as a promising method for future nuclear programs and more specifically for spent fuel processing. Molten alkaline chloride based melts are considered as a promising reaction media. For this, it is interesting to understand the chemical nature of the actinides and lanthanides in high-temperature melt. Some spectroscopy provides essential information on the exact nature of f-block elements LiCl-KCl melt system. High temperature electronic absorption spectroscopy challenges researchers to design and build specific apparatus/equipment and maintain certain strict physicochemical conditions. First of all, to reach to that goal, it is necessary to setup special apparatus and measuring equipment. Here, we report the details of the design of the reaction system combined with the instrumentation of the spectrometer system. Also, application result of the measuring system to U(III) involved chemical reaction in molten salt media was introduced

  18. Modeling Solute Thermokinetics in LiCI-KCI Molten Salt for Nuclear Waste Separation

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, Dane; Eapen, Jacob

    2013-10-01

    Recovery of actinides is an integral part of a closed nuclear fuel cycle. Pyrometallurgical nuclear fuel recycling processes have been developed in the past for recovering actinides from spent metallic and nitride fuels. The process is essentially to dissolve the spent fuel in a molten salt and then extract just the actinides for reuse in a reactor. Extraction is typically done through electrorefining, which involves electrochemical reduction of the dissolved actinides and plating onto a cathode. Knowledge of a number of basic thermokinetic properties of salts and salt-fuel mixtures is necessary for optimizing present and developing new approaches for pyrometallurgical waste processing. The properties of salt-fuel mixtures are presently being studied, but there are so many solutes and varying concentrations that direct experimental investigation is prohibitively time consuming and expensive (particularly for radioactive elements like Pu). Therefore, there is a need to reduce the number of required experiments through modeling of salt and salt-fuel mixture properties. This project will develop first-principles-based molecular modeling and simulation approaches to predict fundamental thermokinetic properties of dissolved actinides and fission products in molten salts. The focus of the proposed work is on property changes with higher concentrations (up to 5 mol%) of dissolved fuel components, where there is still very limited experimental data. The properties predicted with the modeling will be density, which is used to assess the amount of dissolved material in the salt; diffusion coefficients, which can control rates of material transport during separation; and solute activity, which determines total solubility and reduction potentials used during electrorefining. The work will focus on La, Sr, and U, which are chosen to include the important distinct categories of lanthanides, alkali earths, and actinides, respectively. Studies will be performed using LiCl-KCl salt

  19. Potentiometric Sensor for Real-Time Remote Surveillance of Actinides in Molten Salts

    Energy Technology Data Exchange (ETDEWEB)

    Natalie J. Gese; Jan-Fong Jue; Brenda E. Serrano; Guy L. Fredrickson

    2012-07-01

    A potentiometric sensor is being developed at the Idaho National Laboratory for real-time remote surveillance of actinides during electrorefining of spent nuclear fuel. During electrorefining, fuel in metallic form is oxidized at the anode while refined uranium metal is reduced at the cathode in a high temperature electrochemical cell containing LiCl-KCl-UCl3 electrolyte. Actinides present in the fuel chemically react with UCl3 and form stable metal chlorides that accumulate in the electrolyte. This sensor will be used for process control and safeguarding of activities in the electrorefiner by monitoring the concentrations of actinides in the electrolyte. The work presented focuses on developing a solid-state cation conducting ceramic sensor for detecting varying concentrations of trivalent actinide metal cations in eutectic LiCl-KCl molten salt. To understand the basic mechanisms for actinide sensor applications in molten salts, gadolinium was used as a surrogate for actinides. The ß?-Al2O3 was selected as the solid-state electrolyte for sensor fabrication based on cationic conductivity and other factors. In the present work Gd3+-ß?-Al2O3 was prepared by ion exchange reactions between trivalent Gd3+ from GdCl3 and K+-, Na+-, and Sr2+-ß?-Al2O3 precursors. Scanning electron microscopy (SEM) was used for characterization of Gd3+-ß?-Al2O3 samples. Microfocus X-ray Diffraction (µ-XRD) was used in conjunction with SEM energy dispersive X-ray spectroscopy (EDS) to identify phase content and elemental composition. The Gd3+-ß?-Al2O3 materials were tested for mechanical and chemical stability by exposing them to molten LiCl-KCl based salts. The effect of annealing on the exchanged material was studied to determine improvements in material integrity post ion exchange. The stability of the ß?-Al2O3 phase after annealing was verified by µ-XRD. Preliminary sensor tests with different assembly designs will also be presented.

  20. On the performance of CSP oil-cooled plants, with and without heat storage in tanks of molten salts

    International Nuclear Information System (INIS)

    The most-used thermodynamic CSP (concentrating solar plants) in the world, provided with linear parabolic collectors cooled by oil, have been analyzed in the two configurations employed: with heat storage in two tanks filled with molten salts and without heat storage. The performances and the costs of the plants have been analyzed in the paper according to solar multiple (ranging between 1 and 3) and to storage capacity (ranging between 0 and 24 h), in terms of annual electrical energy, average annual plant efficiency, charge factor, capital cost and levelized cost of energy (LCOE). Also a method of economic optimization, based on the evaluation of the minimum value of the levelized cost of energy is presented. The minimum LCOE value, in the case of heat storage, is obtained for a solar multiple of 2.2 and a storage capacity of 16 h. In the plants without storage, minimum LCOE is achieved for SM (solar multiple) equal to 1.2. - Highlights: • A model to analyze the performance of oil thermodynamic solar plant is presented. • Plants without heat storage and with storage in molten salts are considered. • Annual electricity production, efficiency, capital cost, CF and LCOE are estimated. • Storage capacity and solar multiple values which minimize LCOE have been found

  1. Thermodynamic Assessment of Hot Corrosion Mechanisms of Superalloys Hastelloy N and Haynes 242 in Eutectic Mixture of Molten Salts KF and ZrF4

    Energy Technology Data Exchange (ETDEWEB)

    Michael V. Glazoff

    2012-02-01

    The KF - ZrF4 system was considered for the application as a heat exchange agent in molten salt nuclear reactors (MSRs) beginning with the work carried out at ORNL in early fifties. Based on a combination of excellent properties such as thermal conductivity, viscosity in the molten state, and other thermo-physical and rheological properties, it was selected as one of possible candidates for the nuclear reactor secondary heat exchanger loop.

  2. Application of annexation principle to the study of thermodynamic properties of ternary molten salts CaCl2-MgCl2-NaCl

    Institute of Scientific and Technical Information of China (English)

    ZHANG Jian

    2004-01-01

    Based on the practical basis of measured activities and phase diagrams as well as in the light of the mass action law, the model of inseparable cations and anions of molten salts and mattes, and the annexation principle of two kinds of solutions in binary melts, the calculating model of mass action concentrations of molten salts CaC12-MgCl2-NaCl was formulated. The results of calculation not only agree with experimental values, but also obey the mass action law, testifying that the model formulated can embody the sauctural characteristics of these temary salts, and that the model of inseparable cations and anions as well as the annexation principle of two kinds of solutions in binary melts are also applicable to these ternary salts.

  3. Physical chemistry of molten-salt batteries. Final report, 1 October 1980-September 1981. Current-induced composition gradients in molten LiCl-KCl

    International Nuclear Information System (INIS)

    Current-induced composition gradients have been predicted in mixed molten salt battery electrolytes. Composition shifts, if large enough, can produce significant deleterious effects, such as solid phase precipitation in or near the electrodes of molten salt batteries, including the LiAl/LiCl-KCl/FeS/sub x/ battery. Quantitative measurements are needed to determine the extent of the gradients and to find means to reduce them. This report presents the first quantitative SEM/EDX measurements with high distance resolution (<50 μm) of the shape of the composition profile in LiCl-KCl electrolyzed between LiAl electrodes. Also, current-induced precipitation of LiCl in a porous LiAl anode is indicated by SEM/EDX examination. The measured compositions are consistent with predictions from mass transport models based on the electrode reactions, migrational and diffusional mobilities. 5 figures, 4 tables

  4. Reversible Electro-Optic Device Employing Aprotic Molten Salts And Method

    Science.gov (United States)

    Warner, Benjamin P.; McCleskey, T. Mark; Burrell, Anthony K.; Hall, Simon B.

    2005-03-01

    A single-compartment reversible mirror device having a solution of aprotic molten salt, at least one soluble metal-containing species comprising metal capable of being electrodeposited, and at least one anodic compound capable of being oxidized was prepared. The aprotic molten salt is liquid at room temperature and includes lithium and/or quaternary ammonium cations, and anions selected from trifluoromethylsulfonate (CF.sub.3 SO.sub.3.sup.-), bis(trifluoromethylsulfonyl)imide ((CF.sub.3 SO.sub.2).sub.2 N.sup.-), bis(perfluoroethylsulfonyl)imide ((CF.sub.3 CF.sub.2 SO.sub.2).sub.2 N.sup.-) and tris(trifluoromethylsulfonyl)methide ((CF.sub.3 SO.sub.2).sub.3 C.sup.-). A method for preparing substantially pure molten salts is also described.

  5. Potentiometric Sensor for Real-Time Monitoring of Multivalent Ion Concentrations in Molten Salt

    Energy Technology Data Exchange (ETDEWEB)

    Peter A. Zink; Jan-Fong Jue; Brenda E. Serrano; Guy L. Fredrickson; Ben F. Cowan; Steven D. Herrmann; Shelly X. Li

    2010-07-01

    Electrorefining of spent metallic nuclear fuel in high temperature molten salt systems is a core technology in pyroprocessing, which in turn plays a critical role in the development of advanced fuel cycle technologies. In electrorefining, spent nuclear fuel is treated electrochemically in order to effect separations between uranium, noble metals, and active metals, which include the transuranics. The accumulation of active metals in a lithium chloride-potassium chloride (LiCl-KCl) eutectic molten salt electrolyte occurs at the expense of the UCl3-oxidant concentration in the electrolyte, which must be periodically replenished. Our interests lie with the accumulation of active metals in the molten salt electrolyte. The real-time monitoring of actinide concentrations in the molten salt electrolyte is highly desirable for controlling electrochemical operations and assuring materials control and accountancy. However, real-time monitoring is not possible with current methods for sampling and chemical analysis. A new solid-state electrochemical sensor is being developed for real-time monitoring of actinide ion concentrations in a molten salt electrorefiner. The ultimate function of the sensor is to monitor plutonium concentrations during electrorefining operations, but in this work gadolinium was employed as a surrogate material for plutonium. In a parametric study, polycrystalline sodium beta double-prime alumina (Na-ß?-alumina) discs and tubes were subject to vapor-phase exchange with gadolinium ions (Gd3+) using a gadolinium chloride salt (GdCl3) as a precursor to produce gadolinium beta double-prime alumina (Gd-ß?-alumina) samples. Electrochemical impedance spectroscopy and microstructural analysis were performed on the ion-exchanged discs to determine the relationship between ion exchange and Gd3+ ion conductivity. The ion-exchanged tubes were configured as potentiometric sensors in order to monitor real-time Gd3+ ion concentrations in mixtures of gadolinium

  6. Fluoride partitioning R and D programme for molten salt transmutation reactor systems in the Czech Republic

    Energy Technology Data Exchange (ETDEWEB)

    Uhlir, J. [Nuclear Research Institute Rez plc, CZ (Czech Republic); Priman, V.; Vanicek, J. [Czech Power Company, Praha (Czech Republic)

    2001-07-01

    The transmutation of spent nuclear fuel is considered a prospective alternative conception to the current conception based on the non-reprocessed spent fuel disposal into underground repository. The Czech research and development programme in the field of partitioning and transmutation is founded on the Molten Salt Transmutation Reactor system concept with fluoride salts based liquid fuel, the fuel cycle of which is grounded on pyrochemical / pyrometallurgical fluoride partitioning of spent fuel. The main research activities in the field of fluoride partitioning are oriented mainly towards technological research of Fluoride Volatility Method and laboratory research on electro-separation methods from fluoride melts media. The Czech national conception in the area of P and T research issues from the national power industry programme and from the Czech Power Company intentions of the extensive utilization of nuclear power in our country. The experimental R and D work is concentrated mainly in the Nuclear Research Institute Rez plc that plays a role of main nuclear research workplace for the Czech Power Company. (author)

  7. Fluoride partitioning R and D programme for molten salt transmutation reactor systems in the Czech Republic

    International Nuclear Information System (INIS)

    The transmutation of spent nuclear fuel is considered a prospective alternative conception to the current conception based on the non-reprocessed spent fuel disposal into underground repository. The Czech research and development programme in the field of partitioning and transmutation is founded on the Molten Salt Transmutation Reactor system concept with fluoride salts based liquid fuel, the fuel cycle of which is grounded on pyrochemical / pyrometallurgical fluoride partitioning of spent fuel. The main research activities in the field of fluoride partitioning are oriented mainly towards technological research of Fluoride Volatility Method and laboratory research on electro-separation methods from fluoride melts media. The Czech national conception in the area of P and T research issues from the national power industry programme and from the Czech Power Company intentions of the extensive utilization of nuclear power in our country. The experimental R and D work is concentrated mainly in the Nuclear Research Institute Rez plc that plays a role of main nuclear research workplace for the Czech Power Company. (author)

  8. Gaseous and volatile fission product release from molten salt nuclear fuel

    International Nuclear Information System (INIS)

    Molten salt nuclear fuel can be used in reactors simply by replacing the solid fuel pellets in conventional fuel assemblies with molten salts. One issue with this approach is the fate of gaseous fission products. If the tubes are sealed then the pressure will build up (as happens in metallic fuels). Chemical thermodynamic calculations have been carried out on the nature of the volatile species released and show that, providing fuel salt redox potential is properly controlled, only relatively radiochemically benign isotopes are volatile. Gases can therefore be simply vented from the fuel tubes into the coolant avoiding pressure increases.(author)

  9. Computation fluid dynamic modelling of natural convection heat flow in unpumped molten salt fuel tubes

    International Nuclear Information System (INIS)

    Use of static molten salt nuclear fuel in simple tubes was discarded in 1949 without considering how convection could affect its utility. This poster describes CFD studies showing that such tubes are practical as fuel elements in essentially conventional fuel assemblies. They can achieve power densities above 250kW per liter of fuel salt (higher than PWR's) and do so without causing the tube wall to heat to dangerous levels. This discovery enables the achievement of the many benefits of molten salt fuel while utilizing the highly developed technology, regulatory, non proliferation and safety benefits of current fuel assembly technology. (author)

  10. Proceedings of the 4th workshop on molten salts technology and computer simulation

    International Nuclear Information System (INIS)

    This report is the Proceedings of the 4th Workshop on Molten Salts Technology and Computer Simulation, which was held on December 20, 2004, at Tokai Research Establishment of Japan Atomic Energy Research Institute (JAERI). The purpose of this workshop is to exchange information and views on molten salts technology and computer simulation among the specialists from domestic organizations, and to discuss the recent and future research status for this research field. The intensive discussion was made among approximately 55 participants. The 14 of the presented papers are indexed individually. (J.P.N.)

  11. Domestic Material Content in Molten-Salt Concentrating Solar Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Turchi, Craig [National Renewable Energy Lab. (NREL), Golden, CO (United States); Kurup, Parthiv [National Renewable Energy Lab. (NREL), Golden, CO (United States); Akar, Sertac [National Renewable Energy Lab. (NREL), Golden, CO (United States); Flores, Francisco [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2015-08-26

    This study lists material composition data for two concentrating solar power (CSP) plant designs: a molten-salt power tower and a hypothetical parabolic trough plant, both of which employ a molten salt for the heat transfer fluid (HTF) and thermal storage media. The two designs have equivalent generating and thermal energy storage capacities. The material content of the saltHTF trough plant was approximately 25% lower than a comparably sized conventional oil-HTF parabolic trough plant. The significant reduction in oil, salt, metal, and insulation mass by switching to a salt-HTF design is expected to reduce the capital cost and LCOE for the parabolic trough system.

  12. Electrodeposition of Ca Metal in CaCl2-CaO Molten Salt

    Institute of Scientific and Technical Information of China (English)

    GUO; Jun-kang; WANG; Chang-shui; CAO; Long-hao; OUYANG; Ying-gen

    2013-01-01

    To realize the continuouscalciothermic reduction in molten salts,the electrodeposition behavior of Ca metal in CaCl2-CaO molten salt was investigated by cylic voltammetry.The cyclic voltammograms at the scan rate of 100 mV/s are shown in Fig.1.As is shown,the electrodeposition potential of Ca deviated from-1.66 V to-0.97 V after CaO was added to molten CaCl2 and the decomposition of CaO

  13. Methane-steam reforming by molten salt - membrane reactor using concentrated solar thermal energy

    International Nuclear Information System (INIS)

    By utilization of concentrated solar thermal energy for steam reforming of natural gas, which is an endothermic reaction, the chemical energy of natural gas can be up-graded. The chemical system for steam reforming of natural gas with concentrated solar thermal energy was studied to produce hydrogen by using the thermal storage with molten salt and the membrane reactor. The original steam reforming module with hydrogen permeable palladium membrane was developed and fabricated. Steam reforming of methane proceeded with the original module with palladium membrane below the decomposition temperature of molten salt (around 870 K). (authors)

  14. Molten salt CO2 capture and electro-transformation (MSCC-ET) into capacitive carbon at medium temperature: effect of the electrolyte composition.

    Science.gov (United States)

    Deng, Bowen; Chen, Zhigang; Gao, Muxing; Song, Yuqiao; Zheng, Kaiyuan; Tang, Juanjuan; Xiao, Wei; Mao, Xuhui; Wang, Dihua

    2016-08-15

    Electrochemical transformation of CO2 into functional materials or fuels (i.e., carbon, CO) in high temperature molten salts has been demonstrated as a promising way of carbon capture, utilisation and storage (CCUS) in recent years. In a view of continuous operation, the electrolysis process should match very well with the CO2 absorption kinetics. At the same time, in consideration of the energy efficiency, a molten salt electrochemical cell running at lower temperature is more beneficial to a process powered by the fluctuating renewable electricity from solar/wind farms. Ternary carbonates (Li : Na : K = 43.5 : 31.5 : 25.0) and binary chlorides (Li : K = 58.5 : 41.5), two typical kinds of eutectic melt with low melting points and a wide electrochemical potential window, could be the ideal supporting electrolyte for the molten salt CO2 capture and electro-transformation (MSCC-ET) process. In this work, the CO2 absorption behaviour in Li2O/CaO containing carbonates and chlorides were investigated on a home-made gas absorption testing system. The electrode processes as well as the morphology and properties of carbon obtained in different salts are compared to each other. It was found that the composition of molten salts significantly affects the absorption of CO2, electrode processes and performance of the product. Furthermore, the relationship between the absorption and electro-transformation kinetics are discussed based on the findings. PMID:27193751

  15. Preparation of Superconductor YBCO-123/Ag Composite Through Urea Molten Salt

    International Nuclear Information System (INIS)

    Superconductor YBCO-123/Ag composite has been prepared through Urea molten salt by mixing salt nitrate of Yttrium, Barium, Copper and Silver. The weight of Silver content varied from 0 % - 50 %. After pyrolysis process the powder was subjected to calcination at 300, 500 and 700 oC subsequently for 1 hour. The calcined powders was pelletized into a disk of 1.0 cm in diameter and thickness of 2-3 mm. Sintering of pellet samples was done at 900 oC for 16 hours. Meissner effect on all samples displayed superconductivity phenomena. Samples were examined by XRD, SEM, measurement of critical temperature by using susceptibility magnet vs temperature, and critical current density measurement by using four point probe. Based on orthorhombic structure of YBCO-123 the result of the lattice crystal calculation were is a = 3.8167 - 3.8241 Ao; b = 3.8561 - 3.8895 Ao and c = 11.6518 - 11.7104 Ao, this showed that silver did not influence the structure of YBCO. Superconductor YBCO-123/Ag composite was prepared. This was proved by critical current density. Jc data which showed increased with increasing of silver content and the highest result was 9.71 x 105 Amp/m2

  16. Fuel reprocessing of the fast molten salt reactor: actinides et lanthanides extraction

    International Nuclear Information System (INIS)

    The fuel reprocessing of the molten salt reactor (Gen IV concept) is a multi-steps process in which actinides and lanthanides extraction is performed by a reductive extraction technique. The development of an analytic model has showed that the contact between the liquid fuel LiF-ThF4 and a metallic phase constituted of Bi-Li provide firstly a selective and quantitative extraction of actinides and secondly a quantitative extraction of lanthanides. The control of this process implies the knowledge of saline phase properties. Studies of the physico-chemical properties of fluoride salts lead to develop a technique based on potentiometric measurements to evaluate the fluoro-acidity of the salts. An acidity scale was established in order to classify the different fluoride salts considered. Another electrochemical method was also developed in order to determine the solvation properties of solutes in fluoride F- environment (and particularly ThF4 by F-) in reductive extraction technique, a metallic phase is also involved. A method to prepare this phase was developed by electro-reduction of lithium on a bismuth liquid cathode in LiCl-LiF melt. This technique allows to accurately control the molar fraction of lithium introduced into the liquid bismuth, which is a main parameter to obtain an efficient extraction. (author)

  17. Molten salt corrosion resistance of FeAl alloy with additions of Li, Ce and Ni

    International Nuclear Information System (INIS)

    The corrosion performance of FeAl intermetallic alloys with additions of (1 at.%)Li, Ce, Ni and combinations (Ce + Li and Ce + Ni) in molten salts have been studied using the weight loss technique. Salts included Na2SO4 and NaVO3 and testing temperatures included 600, 650 and 700 deg. C for NaVO3, and 900, 950 and 1000 deg. C for Na2SO4 during 100 h. The corroded specimens were studied in the scanning electronic microscope (SEM) and the corrosion products analyzed with an X-ray energy dispersive analyzer (EDX) attached to it. The corrosion resistance in NaVO3 increases as the temperature increased, whereas in Na2SO4 decreased. The effect of the different alloying elements depended upon the salt used. In NaVO3, for instance, the FeAl + Ce + Li alloy was one with the highest corrosion rates but in Na2SO4 it had the lowest corrosion rate. The addition of these elements most of times increased the corrosion rate of the FeAl-base alloy, whereas in Na2SO4 most of times decreased the corrosion rate. The results are discussed in terms of the degree of protectiveness that the external Al2O3 layer gives to the alloys depending on the testing temperature

  18. Integrated demonstration of molten salt oxidation with salt recycle for mixed waste treatment

    International Nuclear Information System (INIS)

    Molten Salt Oxidation (MSO) is a thermal, nonflame process that has the inherent capability of completely destroying organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. Lawrence Livermore National Laboratory (LLNL) has prepared a facility and constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are performed under carefully controlled (experimental) conditions. The system consists of a MSO processor with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. This integrated system was designed and engineered based on laboratory experience with a smaller engineering-scale reactor unit and extensive laboratory development on salt recycle and final forms preparation. In this paper we present design and engineering details of the system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is identification of the most suitable waste streams and waste types for MSO treatment

  19. Development of molten salt electrorefining for separating transuranium elements. Plutonium collection efficiency lowered by neodymium reduction

    International Nuclear Information System (INIS)

    Molten salt electrorefining using a solid cathode has been studied for collecting plutonium with rare-earth elements remaining in the salt. Scandium, a surrogate for plutonium, and neodymium were used for the experiments. The difference in the standard potential between the Sc/Sc(3) and the Pu/Pu(3) couples was no more than 0.025 volts at soot, and the diffusion coefficient of Sc (3) ions appeared to be nearly equal to that of Pu (3) ions. Neodymium could decrease the current efficiency of plutonium collection due to the reduction of Nd (3) to Nd (2) at the cathode. The electrodeposition tests for collecting scandium metal onto solid cathodes were carried out in LiCl-KCl eutectic melt containing ScCl3and NdCl3. The current efficiency was more than 70% when the concentration of ScCl3 was higher, and then decreased to less than 5% as the concentration decreased. The separation factor of scandium from neodymium varied between 102 and 104. The experimental results were consistent with the calculated results obtained by using the data for a stationary current and the Nernst equation. Preliminary estimations of the electrowinning process to collect plutonium and uranium was carried out based on the Sc/Nd separation studies. (author)

  20. Electrochemical separation of actinides and fission products in molten salt electrolyte

    Science.gov (United States)

    Gay, R. L.; Grantham, L. F.; Fusselman, S. P.; Grimmett, D. L.; Roy, J. J.

    1995-09-01

    Molten salt electrochemical separation may be applied to accelerator-based conversion (ABC) and transmutation systems by dissolving the fluoride transport salt in LiCl-KCl eutectic solvent. The resulting fluoride-chloride mixture will contain small concentrations of fission product rare earths (La, Nd, Gd, Pr, Ce, Eu, Sm, and Y) and actinides (U, Np, Pu, Am, and Cm). The Gibbs free energies of formation of the metal chlorides are grouped advantageously such that the actinides can be deposited on a solid cathode with the majority of the rare earths remaining in the electrolyte. Thus, the actinides are recycled for further transmutation. Rockwell and its partners have measured the thermodynamic properties of the metal chlorides of interest (rare earths and actinides) and demonstrated separation of actinides from rare earths in laboratory studies. A model is being developed to predict the performance of a commercial electrochemical cell for separations starting with PUREX compositions. This model predicts excellent separation of plutonium and other actinides from the rare earths in metal-salt systems.

  1. Validation of electro-thermal simulation with experimental data to prepare online operation of a molten salt target at ISOLDE for the Beta Beams

    CERN Document Server

    Cimmino, S; Marzari, S; Stora, T

    2013-01-01

    The main objective of the Beta Beams is to study oscillation property of pure electrons neutrinos. It produces high energy beams of pure electron neutrinos and anti-neutrinos for oscillation experiments by beta decay of He-6 and Ne-18 radioactive ion beams, stored in a decay ring at gamma = 100. The production of He-6 beam has already been accomplished using a thick beryllium oxide target. However, the production of the needed rate of Ne-18 has proven to be more challenging. In order to achieve the requested yield for Ne-18 a new high power target design based on a circulating molten salt loop has been proposed. To verify some elements of the design, a static molten salt target prototype has been developed at ISOLDE and operated successfully. This paper describes the electro-thermal study of the molten salt target taking into account the heat produced by Joule effect, radiative heat exchange, active water cooling due to forced convection and air passive cooling due to natural convection. The numerical results...

  2. High-Temperature Viscosity Measurement of LiCl-KCl Molten Salts Comprising Actinides and Lanthanides

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jongyun; Bae, Sangeun; Kim, Daehyun; Choi, Yong Suk; Yeon, Jeiwon; Song, Kyuseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-11-15

    Physical properties of molten salt such as viscosity, density, solubility, surface tension, solubility, diffusivity, electrical/thermal conductivity, etc., are important for the design and the early fault detection of each unit processes in the pyroprocess. Among them, the viscosities of the high-temperature molten salts should be considered much more carefully when designing the plant as well as handling and transferring the molten fluid containing radioactive actinides and lanthanides from one unit process to the other in the pyroprocess. The viscosity is also related to the electrical properties and the structure of the liquids. However, there are only viscosity data of pure molten LiCl-KCl eutectic in a limited range of temperatures (890-1080 K). Therefore, interpolation and extrapolation are required to determine the viscosity at a specific process condition, for example, at ca. 773 K for the transfer of molten salt in the pyroprocess. In addition, there have been no previous reports on the viscosity of LiCl-KCl eutectic containing a wide concentration range of actinides and lanthanides such as uranium (U), neodymium (Nd), cerium (Ce), and lanthanum (La), which are of major interests regarding the recycling of LiCl-KCl molten salts.

  3. Application of Molten Salt Reactor Technology to MMW In-Space NEP and Surface Power Missions

    International Nuclear Information System (INIS)

    Anticipated manned nuclear electric propulsion (NEP) and planetary surface power missions will require multi-megawatt nuclear reactors that are lightweight, operationally robust, and sealable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional multi-megawatt gas-cooled and liquid metal concepts. (authors)

  4. Application of Molten Salt Reactor Technology to MMW In-Space NEP and Surface Power Missions

    Science.gov (United States)

    Patton, Bruce; Sorensen, Kirk; Rodgers, Stephen (Technical Monitor)

    2002-01-01

    Anticipated manned nuclear electric propulsion (NEP) and planetary surface power missions will require multimegawatt nuclear reactors that are lightweight, operationally robust, and scalable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional multimegawatt gas-cooled and liquid metal concepts.

  5. Thorium Molten-Salt Nuclear Energy Synergetics system for the global energy demand

    International Nuclear Information System (INIS)

    In this century, we have to solve the two antithetical energy problems. The one is a huge and steep energy demand by the population explosion and the other is a CO2 emission reduction for preventing global warming. It is clear that innovative nuclear system satisfying the following conditions can only solve these problems: adequate safety; nuclear proliferation resistance and safeguards; economic competitiveness; improved radio-waste management; resource utilization, and flexible applicability. To meet these conditions, a new concept THORIMS-NES (Thorium Molten-Salt Nuclear Energy Synergetics) has been proposed, which is composed of simple molten-salt reactors (FUJI) and Accelerator Molten-Salt Breeders (AMSB) which produce the fissile 233U from Th and feed it to the nearly self-sustaining molten-salt reactor FUJI. A number of FUJIs can use 239Pu as part of their fissile fuel along with the fertile Th thereby eliminating the excess Pu in the world. A nuclear option for Brazil using this system is presented. (author)

  6. Molten salt e.m.f. cell measurements on U-Ga alloys

    International Nuclear Information System (INIS)

    The Gibbs free energy of formation of intermetallic compounds, UGa3, UGa2 and U2Ga3 were determined by using high temperature molten salt galvanic cell measurements in the temperature range of 644-988 K, 751-947 K and 800-950 K, respectively. (author)

  7. Magnetohydrodynamic pumps for molten salts in cooling loops of high-temperature nuclear reactors

    Czech Academy of Sciences Publication Activity Database

    Doležel, Ivo; Kotlan, V.; Ulrych, B.

    2011-01-01

    Roč. 87, č. 5 (2011), s. 28-33. ISSN 0033-2097 Grant ostatní: GA MŠk(CZ) MEB051041 Institutional research plan: CEZ:AV0Z20570509 Keywords : magnetohydrodynamic pump * molten salt * electric field Subject RIV: JA - Electronics ; Optoelectronics, Electrical Engineering Impact factor: 0.244, year: 2011 http://pe.org.pl/

  8. Global nuclear energy system-- thorium molten-salt nuclear energy synergetics

    International Nuclear Information System (INIS)

    A simple rational thorium (Th) breeding fuel cycle system, the Thorium Molten-Salt Nuclear Energy Synergetics [THORIMS-NES], composed of simple power stations [utility facility] and fissile producers [process facility], is proposed to establish the essential improvement in issues of safety, power-size flexibility, anti-terrorism and radio-waste, economy, etc. securing the simple operation, maintenance and chemical processing. The design study of small Molten-Salt Reactor, FUJI-II, was performed from 1983, and it showed that it would have a fuel self-sustaining character even not only in a small size, 350MWth-155MWe, but also in the conditions of no core-graphite exchange and no continuous chemical processing, except Kr, Xe and T removal. FUJI-II will be effectively able to incinerate the trans-U elements in the rate of ca.50 kg/y, by replacing about one-third of standard 233U fissile. The fissile producers were studied since 1980 and proposed the Accelerator Molten-Salt Breeder [AMSB], as one of the most promising one, which held a target salt containing 0.5 mol % 233UF4 produced by the spallation of Th or U nuclei in salt and 1 GeV proton. The target salt can be supplied straightly to the fuel tank of Molten-Salt Reactors (FUJI). (author). 20 refs., 6 figs., 3 tabs

  9. Ethanol steam reforming heated up by molten salt CSP: Reactor assessment

    NARCIS (Netherlands)

    Falco, De Marcello; Gallucci, Fausto

    2010-01-01

    In this paper hydrogen production via reforming of ethanol has been studied in a novel hybrid plant consisting in a ethanol reformer and a concentrating solar power (CSP) plant using molten salt as heat carrier fluid. The heat needed for the reforming of ethanol has been supplied to the system by mo

  10. Development of the neutron kinetics code for thermal molten salt reactor

    International Nuclear Information System (INIS)

    In the molten salt reactor, the delayed neutron precursors continuously change their position both in the core and in the external loop due to the flow of fuel, which lead to the loss of reactivity and decrease of effective delayed neutron fraction. Therefore, the neutron kinetics of molten salt reactor is significantly different from that of conventional reactors using solid fuels. In this study, a 3D neutron kinetics model of molten salt reactor considering the flow effects of delayed neutron precursors was established. The analytic basis functions expansion nodal method for arbitrary triangular-z node was performed to solve the three dimensional neutron diffusion equations and the method of characteristics was used to find the solution of delayed neutron precursor equations within the whole primary circuit. To verify the code, calculations were preformed on a homogeneous reactor model. The results were in good agreement with the reference results. Besides, the influence of fuel circulation on the kinetic characteristic of reactor was investigated. The results showed some special phenomenon in the molten salt reactor. (author)

  11. Modified ADS molten salt processes for back-end fuel cycle of PWR spent fuel

    International Nuclear Information System (INIS)

    The back-end fuel cycle concept for PWR spent fuel is explained. This concept is adequate for Korea, which has operated both PWR and CANDU reactors. Molten salt processes for accelerator driven system (ADS) were modified both for the transmutation of long-lived radioisotopes and for the utilisation of the remained fissile uranium in PWR spent fuels. Prior to applying molten salt processes to PWR fuel, hydrofluorination and fluorination processes are applied to obtain uranium hexafluoride from the spent fuel pellet. It is converted to uranium dioxide and fabricated into CANDU fuel. From the remained fluoride compounds, transuranium elements can be separated by the molten salt technology such as electrowinning and reductive extraction processes for transmutation purpose without weakening the proliferation resistance of molten salt technology. The proposed fuel cycle concept using fluorination processes is thought to be adequate for our nuclear program and can replace DUPIC (Direct Use of spent PWR fuel in CANDU reactor) fuel cycle. Each process for the proposed fuel cycle concept was evaluated in detail

  12. Nuclear power technology system with molten salt reactor for transuranium nuclides burning in closed fuel cycle

    International Nuclear Information System (INIS)

    A concept of nuclear power technology system with homogeneous molten salt reactors for burning and transmutation of long-lived radioactive toxic nuclides is considered in the paper. Disposition of such reactors in enterprises of fuel cycle allows to provide them with power and facilitate solution of problems with rad waste with minimal losses. (Authors)

  13. A comparison of conventional and prototype nondestructive measurements on molten salt extraction residues

    International Nuclear Information System (INIS)

    Fourteen molten salt extraction residues were assayed by conventional and prototype nondestructive assay (NDA) techniques to be compared with destructive chemical analysis in an effort to identify acceptable NDA measurement methods for this matrix. NDA results on seven samples and destructive results on four samples are presented

  14. Certain properties of thin-film niobium carbide coatings on carbon steels obtained in molten salts

    International Nuclear Information System (INIS)

    Niobium carbide coatings have been deposited by means of a currentless transfer of electronegative niobium metal to a more electropositive substratum made of carbon steel in molten salts containing niobium compounds. Corrosion resistance of niobium carbide coated products is studied, wear resistance and tribological characteristics of the coatings are determined

  15. Impact analysis of criticality safety for 10-MWt solid thorium-based molten salt reactor spent nuclear fuel storage system%10-MWt固态钍基熔盐堆乏燃料贮存系统临界安全影响分析

    Institute of Scientific and Technical Information of China (English)

    田金; 夏晓彬; 彭超; 张志宏

    2015-01-01

    10-MWt固态钍基熔盐堆(Thorium-based Molten Salt Reactor-Solid Fuel,TMSR-SF)使用TRISO(Tri-structural isotropic)颗粒燃料元件,并采用熔融氟盐作为一回路冷却剂,附着在燃料元件上的熔盐有可能影响系统反应性.因此,需要分析在燃料元件的贮存过程中熔盐附着燃料元件对贮存临界安全的影响.使用SCALE6.1的TRITON (Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion)模块对TMSR-SF堆芯建模并进行燃耗计算,使用MCNP对乏燃料贮存系统进行临界计算.分别考虑熔盐浸渗球形燃料元件和熔盐包覆在球形燃料元件表面两种典型情况下,熔盐附着对贮存系统反应性的影响.针对乏燃料贮存系统,以浸渗最大量,即熔盐体积是石墨体积的13.9%为前提,临界计算结果表明,熔盐浸渗入石墨基体贮存系统的反应性比熔盐包覆在球形燃料元件表面的贮存系统的反应性要大5%;与没有熔盐附着的情况相比,有熔盐附着的情况下贮存系统反应性要大15%.对乏燃料贮存系统的临界安全分析可知,两种典型的熔盐附着模型对贮存系统的反应性存在一定的影响,但无论是熔盐浸渗还是包覆,贮存系统仍处于次临界,意味着贮存系统在正常工况下是安全的.

  16. Novel Molten Salts Thermal Energy Storage for Concentrating Solar Power Generation

    Energy Technology Data Exchange (ETDEWEB)

    Reddy, Ramana G. [The University of Alabama

    2013-10-23

    The explicit UA program objective is to develop low melting point (LMP) molten salt thermal energy storage media with high thermal energy storage density for sensible heat storage systems. The novel Low Melting Point (LMP) molten salts are targeted to have the following characteristics: 1. Lower melting point (MP) compared to current salts (<222ºC) 2. Higher energy density compared to current salts (>300 MJ/m3) 3. Lower power generation cost compared to current salt In terms of lower power costs, the program target the DOE's Solar Energy Technologies Program year 2020 goal to create systems that have the potential to reduce the cost of Thermal Energy Storage (TES) to less than $15/kWh-th and achieve round trip efficiencies greater than 93%. The project has completed the experimental investigations to determine the thermo-physical, long term thermal stability properties of the LMP molten salts and also corrosion studies of stainless steel in the candidate LMP molten salts. Heat transfer and fluid dynamics modeling have been conducted to identify heat transfer geometry and relative costs for TES systems that would utilize the primary LMP molten salt candidates. The project also proposes heat transfer geometry with relevant modifications to suit the usage of our molten salts as thermal energy storage and heat transfer fluids. The essential properties of the down-selected novel LMP molten salts to be considered for thermal storage in solar energy applications were experimentally determined, including melting point, heat capacity, thermal stability, density, viscosity, thermal conductivity, vapor pressure, and corrosion resistance of SS 316. The thermodynamic modeling was conducted to determine potential high temperature stable molten salt mixtures that have thermal stability up to 1000 °C. The thermo-physical properties of select potential high temperature stable (HMP) molten salt mixtures were also experimentally determined. All the salt mixtures align with the

  17. Amster: a molten-salt reactor concept generating its own 233U and incinerating transuranium elements

    International Nuclear Information System (INIS)

    In the coming century, sustainable development of atomic energy will require the development of new types of reactors able to exceed the limits of the existing reactor types, be it in terms of optimum use of natural fuel resources, reduction in the production of long-lived radioactive waste, or economic competitiveness. Of the various candidates with the potential to meet these needs, molten-salt reactors are particularly attractive, in the light of the benefits they offer, arising from two fundamental features: - A liquid fuel does away with the constraints inherent in solid fuel, leading to a drastic simplification of the fuel cycle, in particular making in possible to carry out on-line pyrochemical reprocessing; - Thorium cycle and thermal spectrum breeding. The MSBR concept proposed by ORNL in the 1970's thus gave a breeding factor of 1.06, with a doubling time of about 25 years. However, given the tight neutron balance of the thorium cycle (the η of 233U is about 2.3), MSBR performance is only possible if there are strict constraints set on the in-line reprocessing unit: all the 233Pa must be removed from the core so that it can decay on the 233U in no more than about ten days (or at least 15 tonnes of salt to be extracted from the core daily), and the absorbing fission products, in particular the rare earths, must be extracted in about fifty days. With the AMSTER MSR concept, which we initially developed for incinerating transuranium elements, we looked to reduce the mass of salt to be reprocessed in order to minimise the size and complexity of the reprocessing unit coupled to the reactor, and the quantity of transuranium elements sent for disposal, as this is directly proportional to the mass of salt reprocessed for extraction of the fission products. Given that breeding was not an absolute necessity, because the reactor can be started by incinerating the transuranium elements from the spent fuel assemblies of current reactors, or if necessary by loading

  18. Comparative studies on plutonium and 233U utilization in miniFUJI MSR

    International Nuclear Information System (INIS)

    Molten salt reactor (MSR) has many merits such as safety enhancement and capability to be used for hydrogen production. A comparative evaluation of plutonium and 233U utilization in miniFUJI MSR has been performed. Reactor grade plutonium (RGPu), weapon grade plutonium (WGPu), and super grade plutonium (SGPu) have been utilized in the present study. The reactors can obtain their criticality condition with the 233U concentration in the Th-233U fuel, RGPu concentration in Th-RGPu fuel, WGPu concentration in Th-WGPu fuel, and SGPu concentration in Th-SGPu fuel of 0.52%, 5.76%, 2.16%, and 1.96%, respectively. The Th-233U fuel results in the soft neutron spectra of miniFUJI reactor. The neutron spectra turn into harder with the enlarging of plutonium concentration in loaded fuel where Th-RGPu fuel gives the hardest neutron spectra. (author)

  19. Measurements and Analysis of Oxygen Bubble Distributions in LiCl-KCl Molten Salt

    Energy Technology Data Exchange (ETDEWEB)

    Ryan W. Bezzant; Supathorn Phongikaroon; Michael F. Simpson

    2013-03-01

    Transparent system experimental studies have been performed to provide measurement and analysis of oxygen bubble distributions and mass transfer coefficients at different sparging rates ranging from 0.05 to 0.20 L/min in LiCl-KCl molten salt at 500 degrees C using a high-speed digital camera and an oxygen sensor. The results reveal that bubble sizes and rise velocities increased with an increase in oxygen sparging rate. The bubbles observed were ellipsoidal in shape, and an equivalent diameter based on the ellipsoid volume was calculated. The average equivalent bubble diameters at 500 degrees C and these oxygen sparging rates range from 2.63 to 4.07 mm. Results show that the bubble equivalent diameters at each respective sparging rate are normally distributed. A Fanning friction factor correlation was produced to predict a bubble’s rise velocity based on its equivalent diameter. The oxygen mass transfer coefficients for four sparging rates were calculated using the oxygenation model. These calculated values were within the order of magnitude of 10-2 cm/sec and followed a decreasing trend corresponding to an increasing bubble size and sparging rate. The diffusivities were calculated based on two different types of mechanisms, one based on physics of the bubbles and the other on systematic properties. The results reveal that diffusivity values calculated from bubble physics are 1.65 to 8.40 x 10-5 cm2/sec, which are within the range suggested by literature for gases in liquids of a similar viscosity.

  20. Engineering development studies for molten-salt breeder reactor processing No. 19

    International Nuclear Information System (INIS)

    Fabrication and assembly of carbon steel vessels for metal transfer experiment MTE-3B was continued. Examination of the vessels and analysis of the salt and metal phases from the previously operated experiment MTE-3 was completed. Internal surfaces exposed to salts and bismuth appeared in excellent condition. Failure of the oxidation-resistant protective coating on the external surfaces allowed significant oxidation of these surfaces at the 6500C operating temperature, but was not extensive enough to affect the vessel integrity. A different protective coating with superior air-oxidation resistance was applied to the MTE-3B vessels. X-ray fluorescence analyses of the Li-Bi phase from the rare-earth stripper at the LiCl--Li-Bi interface contained significant amounts of iron and thorium. A 6-in. diam low-carbon steel stirred interface contactor was installed in the Salt-Bismuth Flowthrough Facility. Results from the first six runs using 97Zr and 237U tracers indicate that the salt-phase mass transfer coefficient based on 237U counting data is 37 +- 3 percent of the value predicted by the Lewis correlation for runs 1, 2, 3, and 5, and is 116 +- 10 percent of the Lewis value for runs 4 and 6. The mass transfer coefficients based on 97Zr counting data are felt to be less reliable than those based on 237U because of the inability to correct for self absorption of the 743.37 keV β- in the solid bismuth samples. Reaction of gaseous UF6 with UF4 dissolved in molten salt and the subsequent reduction with hydrogen of the resultant UF5 will be a flowthrough operation, and the main vessels will consist of a 36-liter feed tank, a UF6 absorption vessel, a hydrogen reduction column, and a receiver vessel. (U.S.)

  1. Development of electrolytic process in molten salt media for light rare-earth metals production. The metallic cerium electrodeposition

    International Nuclear Information System (INIS)

    The development of molten salt process and the respective equipment aiming rare-earth metals recovery was described. In the present case, the liquid cerium metal electrodeposition in a molten electrolytes of cerium chloride and an equimolar mixture of sodium and potassium chlorides in temperatures near 800C was studied. Due the high chemical reactivity of the rare-earth metals in the liquid state and their molten halides, an electrolytic cell was constructed with controlled atmosphere, graphite crucibles and anodes and a tungsten cathode. The electrolytic process variables and characteristics were evaluated upon the current efficiency and metallic product purity. Based on this evaluations, were suggested some alterations on the electrolytic reactor design and upon the process parameters. (author). 90 refs, 37 figs, 20 tabs

  2. Complex formation in molten salts: association constants of cadmium-iodo complexes in molten potassium nitrate-barium nitrate eutectic

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, R.K.; Gaur, H.C. (Delhi Univ. (India). Dept. of Chemistry)

    1981-06-01

    Association constants for the formation of CdI/sup +/ and CdI/sub 2/ in molten KNO/sub 3/-Ba(NO/sub 3/)/sub 2/(87.6:12.4 mol%) eutectic in the temperature range 568.2 - 608.2 K have been evaluated from activity coefficient by measurement of emf of the molten salt concentration cell, Ag, AgI(s)/KNO/sub 3/ -Ba(NO/sub 3/)/sub 2/, KI/KNO/sub 3/ - Ba(NO/sub 3/)/sub 2/, KI, Cd(NO/sub 3/)/sub 2/ or CdI/sub 2//AgI(s), Ag. Data do not suggest the formation of polynuclear species under the experimental conditions employed. The temperature coefficients of the association constants are predictable from equations based on quasi-lattice model.

  3. Pore Scale Thermal Hydraulics Investigations of Molten Salt Cooled Pebble Bed High Temperature Reactor with BCC and FCC Configurations

    Directory of Open Access Journals (Sweden)

    Shixiong Song

    2014-01-01

    CFD results and empirical correlations’ predictions of pressure drop and local Nusselt numbers. Local pebble surface temperature distributions in several default conditions are investigated. Thermal removal capacities of molten salt are confirmed in the case of nominal condition; the pebble surface temperature under the condition of local power distortion shows the tolerance of pebble in extreme neutron dose exposure. The numerical experiments of local pebble insufficient cooling indicate that in the molten salt cooled pebble bed reactor, the pebble surface temperature is not very sensitive to loss of partial coolant. The methods and results of this paper would be useful for optimum designs and safety analysis of molten salt cooled pebble bed reactors.

  4. Solid oxide membrane-assisted controllable electrolytic fabrication of metal carbides in molten salt.

    Science.gov (United States)

    Zou, Xingli; Zheng, Kai; Lu, Xionggang; Xu, Qian; Zhou, Zhongfu

    2016-08-15

    Silicon carbide (SiC), titanium carbide (TiC), zirconium carbide (ZrC), and tantalum carbide (TaC) have been electrochemically produced directly from their corresponding stoichiometric metal oxides/carbon (MOx/C) precursors by electrodeoxidation in molten calcium chloride (CaCl2). An assembled yttria stabilized zirconia solid oxide membrane (SOM)-based anode was employed to control the electrodeoxidation process. The SOM-assisted controllable electrochemical process was carried out in molten CaCl2 at 1000 °C with a potential of 3.5 to 4.0 V. The reaction mechanism of the electrochemical production process and the characteristics of these produced metal carbides (MCs) were systematically investigated. X-ray diffraction, scanning electron microscopy, and transmission electron microscopy analyses clearly identify that SiC, TiC, ZrC, and TaC carbides can be facilely fabricated. SiC carbide can be controlled to form a homogeneous nanowire structure, while the morphologies of TiC, ZrC, and TaC carbides exhibit porous nodular structures with micro/nanoscale particles. The complex chemical/electrochemical reaction processes including the compounding, electrodeoxidation, dissolution-electrodeposition, and in situ carbonization processes in molten CaCl2 are also discussed. The present results preliminarily demonstrate that the molten salt-based SOM-assisted electrodeoxidation process has the potential to be used for the facile and controllable electrodeoxidation of MOx/C precursors to micro/nanostructured MCs, which can potentially be used for various applications. PMID:27195950

  5. Potentiometric Sensor for Real-Time Monitoring of Multivalent Ion Concentrations in Molten Salt

    International Nuclear Information System (INIS)

    Electrorefining of spent metallic nuclear fuel in high temperature molten salt systems is a core technology in pyroprocessing, which in turn plays a critical role in the development of advanced fuel cycle technologies. In electrorefining, spent nuclear fuel is treated electrochemically in order to effect separations between uranium, noble metals, and active metals, which include the transuranics. The accumulation of active metals in a lithium chloride-potassium chloride (LiCl-KCl) eutectic molten salt electrolyte occurs at the expense of the UCl3-oxidant concentration in the electrolyte, which must be periodically replenished. Our interests lie with the accumulation of active metals in the molten salt electrolyte. The real-time monitoring of actinide concentrations in the molten salt electrolyte is highly desirable for controlling electrochemical operations and assuring materials control and accountancy. However, real-time monitoring is not possible with current methods for sampling and chemical analysis. A new solid-state electrochemical sensor is being developed for real-time monitoring of actinide ion concentrations in a molten salt electrorefiner. The ultimate function of the sensor is to monitor plutonium concentrations during electrorefining operations, but in this work gadolinium was employed as a surrogate material for plutonium. In a parametric study, polycrystalline sodium beta double-prime alumina (Na-β(doubleprime)-alumina) discs and tubes were subject to vapor-phase exchange with gadolinium ions (Gd3+) using a gadolinium chloride salt (GdCl3) as a precursor to produce gadolinium beta double-prime alumina (Gd-β(doubleprime)-alumina) samples. Electrochemical impedance spectroscopy and microstructural analysis were performed on the ion-exchanged discs to determine the relationship between ion exchange and Gd3+ ion conductivity. The ion-exchanged tubes were configured as potentiometric sensors in order to monitor real-time Gd3+ ion concentrations in

  6. Tungsten coatings electro-deposited on CFC substrates from oxide molten salt

    International Nuclear Information System (INIS)

    Tungsten is considered as plasma facing material in fusion devices because of its high melting point, its good thermal conductivity, its low erosion rate and its benign neutron activation properties. On the other hand, carbon based materials like C/C fiber composites (CFC) have been used for plasma facing materials (PFMs) due to their high thermal shock resistance, light weight and high strength. Tungsten coatings on CFC substrates are used in the JET divertor in the frame of the JET ITER-like wall project, and have been prepared by plasma spray (PS) and other techniques. In this study, tungsten coatings were electro-deposited on CFC from Na2WO4–WO3 molten salt under various deposition parameters at 900 °C in air. In order to obtain tungsten coatings with excellent performance, the effects of pulse duration ratio and pulse current density on microstructures and crystal structures of tungsten coatings were investigated by X-ray diffraction (XRD, Rigaku Industrial Co., Ltd., D/MAX-RB) and a scanning electron microscope (SEM, JSM 6480LV). It is found that the pulsed duration ratio and pulse current density had a significant influence on tungsten nucleation and electro-crystallization phenomena. SEM observation revealed that intact, uniform and dense tungsten coatings formed on the CFC substrates. Both the average grain size and thickness of the coating increased with the pulsed current density. The XRD results showed that the coatings consisted of a single phase of tungsten with the body centered cubic (BCC) structure. The oxygen content of electro-deposited tungsten coatings was lower than 0.05%, and the micro-hardness was about 400 HV

  7. Uncertainty studies of real anode surface area in computational analysis for molten salt electrorefining

    International Nuclear Information System (INIS)

    Highlights: → Numerical electrochemo-fluid modeling of pyrochemical electrorefining in cross comparison with 2D and 3D analysis models. → Benchmark study on cell potential of molten LiCl-KCl electrorefining with Mark-IV electrorefiner containing EBR-II spent fuel. → Determination of real anode surface area profile model governing electrorefining performance. → Identification of uncertainty factors in electrorefining causing disagreements between simulation and experiment. → Fully transient performance analysis of 80 hours Mark-IV electrorefining with multi-species multi-reaction 1D model. - Abstract: This study examines how much cell potential changes with five differently assumed real anode surface area cases. Determining real anode surface area is a significant issue to be resolved for precisely modeling molten salt electrorefining. Based on a three-dimensional electrorefining model, calculated cell potentials compare with an experimental cell potential variation over 80 h of operation of the Mark-IV electrorefiner with driver fuel from the Experimental Breeder Reactor II. We succeeded to achieve a good agreement with an overall trend of the experimental data with appropriate selection of a mode for real anode surface area, but there are still local inconsistencies between theoretical calculation and experimental observation. In addition, the results were validated and compared with two-dimensional results to identify possible uncertainty factors that had to be further considered in a computational electrorefining analysis. These uncertainty factors include material properties, heterogeneous material distribution, surface roughness, and current efficiency. Zirconium's abundance and complex behavior have more impact on uncertainty towards the latter period of electrorefining at given batch of fuel. The benchmark results found that anode materials would be dissolved from both axial and radial directions at least for low burn-up metallic fuels after active

  8. Molten salt corrosion resistance of FeAl alloy with additions of Li, Ce and Ni

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez-Rodriguez, J.G. [Universidad Autonoma del Estado de Morelos, FCQI-CIICAP, 62210 Cuernavaca, Mor (Mexico); Luna-Ramirez, A. [Instituto de Investigaciones Electricas, Cuernavaca, Mor (Mexico); CIMAV, Miguel de Cervantes 120, Complejo Industrial Chihuahua, 31109 Chihuahua, Chih (Mexico); Salazar, M. [Instituto de Investigaciones Electricas, Cuernavaca, Mor (Mexico); Instituto Mexicano del Petroleo, Eje Central Lazaro Cardenas, Mexico, D.F (Mexico); Porcayo-Calderon, J. [Instituto de Investigaciones Electricas, Cuernavaca, Mor (Mexico); Rosas, G. [Universidad Michoacana de San Nicolas de Hidalgo, Morelia, Mich (Mexico); Martinez-Villafane, A. [CIMAV, Miguel de Cervantes 120, Complejo Industrial Chihuahua, 31109 Chihuahua, Chih (Mexico)]. E-mail: martinez.villafane@cimav.edu.mx

    2005-06-15

    The corrosion performance of FeAl intermetallic alloys with additions of (1 at.%)Li, Ce, Ni and combinations (Ce + Li and Ce + Ni) in molten salts have been studied using the weight loss technique. Salts included Na{sub 2}SO{sub 4} and NaVO{sub 3} and testing temperatures included 600, 650 and 700 deg. C for NaVO{sub 3}, and 900, 950 and 1000 deg. C for Na{sub 2}SO{sub 4} during 100 h. The corroded specimens were studied in the scanning electronic microscope (SEM) and the corrosion products analyzed with an X-ray energy dispersive analyzer (EDX) attached to it. The corrosion resistance in NaVO{sub 3} increases as the temperature increased, whereas in Na{sub 2}SO{sub 4} decreased. The effect of the different alloying elements depended upon the salt used. In NaVO{sub 3}, for instance, the FeAl + Ce + Li alloy was one with the highest corrosion rates but in Na{sub 2}SO{sub 4} it had the lowest corrosion rate. The addition of these elements most of times increased the corrosion rate of the FeAl-base alloy, whereas in Na{sub 2}SO{sub 4} most of times decreased the corrosion rate. The results are discussed in terms of the degree of protectiveness that the external Al{sub 2}O{sub 3} layer gives to the alloys depending on the testing temperature.

  9. Engineering Evaluation of Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiement for the Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Carlberg, Jon A.; Roberts, Kenneth T.; Kollie, Thomas G.; Little, Leslie E.; Brady, Sherman D.

    2009-09-30

    This evaluation was performed by Pro2Serve in accordance with the Technical Specification for an Engineering Evaluation of the Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory (BJC 2009b). The evaluators reviewed the Engineering Evaluation Work Plan for Molten Salt Reactor Experiment Residual Salt Removal, Oak Ridge National Laboratory, Oak Ridge, Tennessee (DOE 2008). The Work Plan (DOE 2008) involves installing a salt transfer probe and new drain line into the Fuel Drain Tanks and Fuel Flush Tank and connecting them to the new salt transfer line at the drain tank cell shield. The probe is to be inserted through the tank ball valve and the molten salt to the bottom of the tank. The tank would then be pressurized through the Reactive Gas Removal System to force the salt into the salt canisters. The Evaluation Team reviewed the work plan, interviewed site personnel, reviewed numerous documents on the Molten Salt Reactor (Sects. 7 and 8), and inspected the probes planned to be used for the transfer. Based on several concerns identified during this review, the team recommends not proceeding with the salt transfer via the proposed alternate salt transfer method. The major concerns identified during this evaluation are: (1) Structural integrity of the tanks - The main concern is with the corrosion that occurred during the fluorination phase of the uranium removal process. This may also apply to the salt transfer line for the Fuel Flush Tank. Corrosion Associated with Fluorination in the Oak Ridge National Laboratory Fluoride Volatility Process (Litman 1961) shows that this problem is significant. (2) Continued generation of Fluorine - Although the generation of Fluorine will be at a lower rate than experienced before the uranium removal, it will continue to be generated. This needs to be taken into consideration regardless of what actions are taken with the salt. (3) More than one phase of material

  10. The nonmetal-metal transition in solutions of metals in molten salts

    International Nuclear Information System (INIS)

    Solutions of metals in molten salts present a rich phenomenology: localization of electrons in disordered ionic media, activated electron transport increasing with metal concentration towards a nonmetal-metal (NM-M) transition, and liquid-liquid phase separation. A brief review of progress in the study of these systems is given in this article, with main focus on the NM-M transition. After recalling the known NM-M behaviour of the component elements in the case of expanded fluid alkali metals and mercury and of solid halogens under pressure, the article focuses on liquid metal - molten salt solutions and traces the different NM-M behaviours of the alkalis in their halides and of metals added to polyvalent metal halides. (author). 51 refs, 2 figs

  11. Radiochemical separations in molten salts. Determination of carbon in metals by photon activation

    International Nuclear Information System (INIS)

    A systematic study of radiochemical separations of carbon in molten salts was achieved. It combined forecasting of salt mixtures from literature data and experimental work. The basic three mixtures of the molten salts studied are the following: Pb3O4 - B2O3, NaOH -NaNO3 and H2SO4 - KIO4 with some additions of complexing agents. Selected procedures for the determination of carbon by photon activation were issued for the following samples: Ag, Al, Cr, Fe, Mg, Mo, Ni, Si, Ti, W, Zn, Zr, AlMg, AgZn and ZnMgTe. The detection limit for carbon was 10-8 g and the accuracy from 5 to 10%. Applications to solid state science are given. (author)

  12. Redox condition in molten salts and solute behavior: A first-principles molecular dynamics study

    Science.gov (United States)

    Nam, Hyo On; Morgan, Dane

    2015-10-01

    Molten salts technology is of significant interest for nuclear, solar, and other energy systems. In this work, first-principles molecular dynamics (FPMD) was used to model the solute behavior in eutectic LiCl-KCl and FLiBe (Li2BeF4) melts at 773 K and 973 K, respectively. The thermo-kinetic properties for solute systems such as the redox potential, solute diffusion coefficients and structural information surrounding the solute were predicted from FPMD modeling and the calculated properties are generally in agreement with the experiments. In particular, we formulate an approach to model redox energetics vs. chlorine (or fluorine) potential from first-principles approaches. This study develops approaches for, and demonstrates the capabilities of, FPMD to model solute properties in molten salts.

  13. Disposal of transuranic solid waste using Atomics International Molten Salt Combustion Process

    International Nuclear Information System (INIS)

    The Atomics International Molten Salt Combustion Process for disposal of transuranic solid waste utilizes a molten salt to combust organic materials, to trap inorganic substances including transuranics, and to react chemically with acidic gases formed during combustion. Subsequent processing of the melt in an aqueous system produces three products: (1) ash, which includes the transuranics, (2) salt, which is recycled to the combustor, and (3) solid sodium chloride for disposal. The transuranics are readily leached from the ash and separated via ion exchange techniques. The leached ash is the second solid process product requiring disposal. No liquid wastes are produced in the process. The reductions in weight and volume are 82 and 98 percent respectively, if one considers the products from the process only. (U.S.)

  14. Molten salt extraction of transuranic and reactive fission products from used uranium oxide fuel

    Science.gov (United States)

    Herrmann, Steven Douglas

    2014-05-27

    Used uranium oxide fuel is detoxified by extracting transuranic and reactive fission products into molten salt. By contacting declad and crushed used uranium oxide fuel with a molten halide salt containing a minor fraction of the respective uranium trihalide, transuranic and reactive fission products partition from the fuel to the molten salt phase, while uranium oxide and non-reactive, or noble metal, fission products remain in an insoluble solid phase. The salt is then separated from the fuel via draining and distillation. By this method, the bulk of the decay heat, fission poisoning capacity, and radiotoxicity are removed from the used fuel. The remaining radioactivity from the noble metal fission products in the detoxified fuel is primarily limited to soft beta emitters. The extracted transuranic and reactive fission products are amenable to existing technologies for group uranium/transuranic product recovery and fission product immobilization in engineered waste forms.

  15. Chemical behaviors of actinides and lanthanides in molten salts and liquid metals

    International Nuclear Information System (INIS)

    Separation processes using molten salts or liquid metals are interesting in view of spent fuel reprocessing and partitioning for nuclear transmutation before final radioactive waste disposals. Nevertheless, chemical behaviors of transuranium and lanthanide elements in non-aqueous solvents such as molten salts and liquid metals have been rarely studied. In the present study, thermodynamic properties of La, Ce, Pr, Nd, Gd, Tb, Ho, Er, Tm, Lu, Np, Pu, Am, and Cm in two phase extraction system: molten LiCl-KCl and liquid Bi or Zn were investigated to obtain excess Gibbs free energy experimentally or by using thermodynamic relationships and to examine systematics of 4f and 5f elements in these phases. Thermodynamic stability and specificity of each elements in liquid metals and salts thus obtained can be successfully used to explain systematics of extractability of f-elements in these systems. (Ohno, S.)

  16. Partitioning of actinides and fission products using molten salt electrorefining process

    International Nuclear Information System (INIS)

    Electrorefining is the key step of pyrometallurgical processing for separating actinides from fission products. In this work, the electrorefining process is carried out in a electrorefining cell that contains molten salts (49% LiCl- 51% KCL) floating on a liquid cadmium. The cell is operated under an inert atmosphere at 500 degree C. In this work we describe in detail the construction of the cell and the way of operation

  17. Corrosion of Nickel Exposed to LiF-NaF Molten Salt

    Czech Academy of Sciences Publication Activity Database

    Král, Lubomír; Čermák, Jiří; Matal, O.; Šimo, T.; Nesvadba, L.

    Ostrava : TANGER, spol. s r.o., 2011, paper no. 882. ISBN 978-80-87294-22-2. [Anniversary International Conference on Metallurgy and Materials METAL 2011/20th./. Brno (CZ), 18.05.2011-20.05.2011] R&D Projects: GA MPO 2A-1TP1/067 Institutional research plan: CEZ:AV0Z20410507 Keywords : corrosion * molten salts * nickel * LiF–NaF Subject RIV: JG - Metallurgy

  18. Technology needs assessment: Evaluation of the molten salt oxidation process technology

    International Nuclear Information System (INIS)

    Molten salt oxidation (MSO) is a potential treatment technology for numerous US DOE waste forms. This report is a baseline evaluation of MSO technology (as developed by Energy Technology Engineering Center and Rockwell International) to establish its present and potential readiness to treat DOE wastes, particularly mixed wastes. Much of the information in this report was derived from a peer review meeting in Woodland Hills, California, on November 12--14, 1991. The panel members and other participants provided expertise in treatment technologies, DOE's waste problems, safety and systems analyses, and regulatory and public issues and concerns. The basic concept of MSO is to introduce wastes and air into a bed of molten salt, oxidize organic wastes in the molten salt, use the heat of oxidation to keep the salt molten, retain metals and radionuclides within the salt melt, and remove the salt for disposal or for processing and recycling. Sodium carbonate, or sodium carbonate mixed with other salts (e.g., potassium carbonate or NaCl) is typical of the salts used as the melt. The most common operating temperatures are 900 to 1000 C, and the feed stocks may be gases, solids, organic liquids, aqueous solutions, or slurries. The molten salt acts as catalyst for the oxidation reactions; enhances organic compound destruction; helps retain ash and soot; and helps capture/react with heavy metals, radionuclides, and acid gases. MSO, used as a primary or secondary treatment system, acts as an acid gas scrubber and requires a particulate treatment system to clean the off gas. As part of evaluation process, the Peer Review Panel examined MSO as a treatment technology for specific waste groups. The Peer Review Panel concluded that the MSO process could potentially treat a variety of DOE wastes, particularly the mixed and hazardous wastes. Advantages and disadvantages of MSO over incineration are identified. The formation of an MSO task force is recommended

  19. Disposal of transuranic solid waste using Atomics International's molten salt combustion process. II

    International Nuclear Information System (INIS)

    The Atomics International Molten Salt Combustion Process reduces the weight and volume of combustible transuranic waste by utilizing a molten salt medium to combust organic materials, to trap particulates and fissile material, and to react chemically with any acidic gases produced during combustion. The ''ash'' is retained by the molten salt. To control the amount of noncombustible substances in the melt, a portion of the molten salt is periodically drained from the combustor. There are two options following the combustion step: the salt-ash mixture can be cast into a metal canister for direct storage, which is preferred, or the salt-ash mixture can be processed to separate ash for disposal, to recover the salt for recycle and to recover fissile materials. Either option results in the rapid, complete, and nonpolluting destruction of the combustible waste. Bench-scale (0.2 kg/hr) combustion tests with plutonium-contaminated waste showed that >99.9 percent of the plutonium is retained in the melt during combustion. A similar test with uranium indicated that uranium and plutonium behave identically during combustion. Bench-scale plutonium recovery tests have shown that approx. 98 percent of the plutonium can be recovered from the ash-melt mixture with a single acid leach. Pilot plant combustion tests were conducted with uncontaminated shredded waste consisting of paper, Kimwipes, cardboard, rubber, polyvinyl chloride, and polyethylene at feed rates up to 70 kg/hr. Hydrogen chloride (3 at approx. 7900C to 0.6 g/m3 at 10200C before the venturi scrubber, and 0.01 to 0.04 g/m3, respectively, after the scrubber. Downstream of the HEPA filters, no particulates could be detected

  20. Preparation of Al-La Master Alloy by Thermite Reaction in NaF-NaCl-KCl Molten Salt

    Science.gov (United States)

    Jang, Poknam; Li, Hyonmo; Kim, Wenjae; Wang, Zhaowen; Liu, Fengguo

    2015-05-01

    A NaF-NaCl-KCl ternary system containing La2O3 was investigated for the preparation of Al-La master alloy by the thermite reaction method. The solubility of La2O3 in NaF-NaCl-KCl molten salt was determined by the method of isothermal solution saturation. Inductively coupled plasma-optical emission spectroscopy and x-ray diffraction (XRD) analyses were used to consider the content of La2O3 in molten salt and the supernatant composition of molten salt after dissolution of La2O3, respectively. The results showed that the content of NaF had a positive influence on the solubility of La2O3 in NaF-NaCl-KCl molten salts, and the solubility of La2O3 could reach 8.71 wt.% in molten salts of 50 wt.%NaF-50 wt.% (44 wt.%NaCl + 56 wt.%KCl). The XRD pattern of cooling molten salt indicated the formation of LaOF in molten salt, which was probably obtained by the reaction between NaF and La2O3. The kinetic study showed that the thermite reaction was in accord with a first-order reaction model. The main influence factors on La content in the Al-La master alloy product, including molten salt composition, amount of Al, concentration of La2O3, stirring, reduction time and temperature, were investigated by single-factor experimentation. The content of La in the Al-La master alloy could be reached to 10.1 wt.%.

  1. Study on the phosphate reaction characteristics of lanthanide chlorides in molten salt with operating conditions

    International Nuclear Information System (INIS)

    A minimization of waste salt is one of the most important issues for the optimization of pyroprocessing. The separation of fission products in waste salts and the reuse of purified waste salt are promising strategies for minimizing the waste salt amounts. The phosphate precipitation of lanthanide is currently being considered for eutectic (LiCl–KCl) waste salt purification. In this research, the effects of molten salt temperature (400–550°C) and reaction time (max. 180 min) upon conversion into the phosphate of lanthanides was investigated using 1 and 3 kg of eutectic salt. The conversion efficiency of lanthanides to molten salt-insoluble precipitates and phosphates was increased with an increase in molten salt temperature and operating time until it attained a specific temperature and time. K3PO4 as a precipitant was more favorable than Li3PO4 in terms of reactivity. To obtain over a 99% overall conversion efficiency, about 30 min was required in the case of using K3PO4 at 450°C, but about 120 min in the case of using Li3PO4 at 550°C. The lanthanide precipitates formed by a reaction with phosphate were a mixture of monoclinic structures, usually representing a polyhedron structure, and a tetragonal structure, representing a platelet structure. (author)

  2. CO2 gas decomposition to carbon by electro-reduction in molten salts

    International Nuclear Information System (INIS)

    The electrochemical decomposition of CO2 gas in LiCl–Li2O or CaCl2–CaO molten salt was studied to produce carbon. This process consists of the electrochemical reduction of the oxide, Li2O or CaO, and the thermal reduction of CO2 gas by metallic Li or Ca. Two kinds of ZrO2 solid electrolytes were tested as an oxygen ions conductor and removed oxygen ions from the molten salts to the outside of reactor. After the electrolysis in the both salts, the aggregations of nanometer-scale amorphous carbon and rod-like graphite crystals were observed by transmission electron microscope. When 9.7% CO2–Ar mixed gas was blown into LiCl–Li2O or CaCl2–CaO molten salt, the current efficiency was evaluated to be 89.7% or 78.5%, respectively, by the exhaust gas analysis and the supplied charge. When the solid electrolyte with the better ionic conductivity was used, the current and the carbon production became larger. The rate determining step of this proposal was diffusion of oxygen ions in ZrO2 solid electrolyte

  3. Growth of SiC nanowires on wooden template surface using molten salt media

    International Nuclear Information System (INIS)

    Highlights: • Biomorphic SiC/C ceramics have been produced using a relatively low temperature by molten salt method. • The resulting biomorphic SiC ceramic retains the original structures of wooden template. • The pore size distribution within the porous SiC/C ceramics was determined by automatic mercury porosimetry. - Abstract: This paper examines the growth of SiC nanowires on a wooden template surface through the reaction of wooden template/silicon composites in static argon atmosphere, using molten salt media. The effects of temperature and salt/Si ratio on the growth of wooden template were investigated. Morphology and structure of the biomorphic SiC/C ceramics were characterized by X-ray diffraction (XRD), transmission electron microscopy (TEM), scanning electron microscopy (SEM) and thermogravimetric analysis (TGA). The pore size distribution within the porous SiC/C ceramics was investigated using automatic mercury porosimetry. The results show that the biomorphic cellular morphology of wooden template was remained in the porous SiC ceramic with high precision that consists of β-SiC with traces of α-SiC. SiC in the wooden template exists in the cellular pores in the form of nanowires. The SiC nanowires were formed at about 1250 °C by molten salt reaction between Si and C during the wooden-to-ceramic conversion

  4. A simplified burnup calculation strategy with refueling in static molten salt reactor

    International Nuclear Information System (INIS)

    Molten Salt Reactors, by nature can be refuelled and reprocessed online. Thus, a simulation methodology has to be developed which can consider online refueling and reprocessing aspect of the reactor. To cater such needs a simplified burnup calculation strategy to account for refueling and removal of molten salt fuel at any desired burnup has been identified in static molten salt reactor in batch mode as a first step of way forward. The features of in-house code ITRAN has been explored for such calculations. The code also enables us to estimate the reactivity introduced in the system due to removal of any number of considered nuclides at any burnup. The effect of refueling fresh fuel and removal of burned fuel has been studied in batch mode with in-house code ITRAN. The effect of refueling and burnup on change in reactivity per day has been analyzed. The analysis of removal of 233Pa at a particular burnup has been carried out. The similar analysis has been performed for some other nuclides also. (author)

  5. Growth of SiC nanowires on wooden template surface using molten salt media

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Jun; Zhu, Hongxi, E-mail: wust6168@hotmail.com; Li, Guangqiang; Deng, Chengji; Li, Jun

    2014-11-30

    Highlights: • Biomorphic SiC/C ceramics have been produced using a relatively low temperature by molten salt method. • The resulting biomorphic SiC ceramic retains the original structures of wooden template. • The pore size distribution within the porous SiC/C ceramics was determined by automatic mercury porosimetry. - Abstract: This paper examines the growth of SiC nanowires on a wooden template surface through the reaction of wooden template/silicon composites in static argon atmosphere, using molten salt media. The effects of temperature and salt/Si ratio on the growth of wooden template were investigated. Morphology and structure of the biomorphic SiC/C ceramics were characterized by X-ray diffraction (XRD), transmission electron microscopy (TEM), scanning electron microscopy (SEM) and thermogravimetric analysis (TGA). The pore size distribution within the porous SiC/C ceramics was investigated using automatic mercury porosimetry. The results show that the biomorphic cellular morphology of wooden template was remained in the porous SiC ceramic with high precision that consists of β-SiC with traces of α-SiC. SiC in the wooden template exists in the cellular pores in the form of nanowires. The SiC nanowires were formed at about 1250 °C by molten salt reaction between Si and C during the wooden-to-ceramic conversion.

  6. Enhanced heat transfer performances of molten salt receiver with spirally grooved pipe

    International Nuclear Information System (INIS)

    The enhanced heat transfer performances of solar receiver with spirally grooved pipe were theoretically investigated. The physical model of heat absorption process was proposed using the general heat transfer correlation of molten salt in smooth and spirally grooved pipe. According to the calculation results, the convective heat transfer inside the receiver can remarkably enhance the heat absorption process, and the absorption efficiency increased with the flow velocity and groove height, while the wall temperature dropped. As the groove height increased, the heat losses of convection and radiation dropped with the decrease of wall temperature, and the average absorption efficiency of the heat receiver can be increased. Compared with the heat receiver with smooth pipe, the heat absorption efficiency of heat receiver with spirally grooved pipe e/d = 0.0475 can rise for 0.7%, and the maximum bulk fluid temperature can be increased for 31.1 °C. As a conclusion, spirally grooved pipe can be a very effective way for heat absorption enhancement of solar receiver, and it can also increase the operating temperature of molten salt. - Highlights: • Spirally grooved tube is a very effective way for solar receiver enhancement. • Heat absorption model of receiver is proposed with general heat transfer correlation. • Spirally groove tube increases absorption efficiency and reduces wall temperature. • Operating temperature of molten salt remarkably increases with groove height. • Heat absorption performance is promoted for first and second thermodynamics laws

  7. Thermal Analysis of Surrogate Simulated Molten Salts with Metal Chloride Impurities for Electrorefining Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Toni Y. Gutknecht; Guy L. Fredrickson; Vivek Utgikar

    2012-04-01

    This project is a fundamental study to measure thermal properties (liquidus, solidus, phase transformation, and enthalpy) of molten salt systems of interest to electrorefining operations, which are used in both the fuel cycle research & development mission and the spent fuel treatment mission of the Department of Energy. During electrorefining operations the electrolyte accumulates elements more active than uranium (transuranics, fission products and bond sodium). The accumulation needs to be closely monitored because the thermal properties of the electrolyte will change as the concentration of the impurities increases. During electrorefining (processing techniques used at the Idaho National Laboratory to separate uranium from spent nuclear fuel) it is important for the electrolyte to remain in a homogeneous liquid phase for operational safeguard and criticality reasons. The phase stability of molten salts in an electrorefiner may be adversely affected by the buildup of fission products in the electrolyte. Potential situations that need to be avoided are: (i) build up of fissile elements in the salt approaching the criticality limits specified for the vessel (ii) freezing of the salts due to change in the liquidus temperature and (iii) phase separation (non-homogenous solution) of elements. The stability (and homogeneity) of the phases can potentially be monitored through the thermal characterization of the salts, which can be a function of impurity concentration. This work describes the experimental results of typical salts compositions, consisting of chlorides of strontium, samarium, praseodymium, lanthanum, barium, cerium, cesium, neodymium, sodium and gadolinium (as a surrogate for both uranium and plutonium), used in the processing of used nuclear fuels. Differential scanning calorimetry was used to analyze numerous salt samples providing results on the thermal properties. The property of most interest to pyroprocessing is the liquidus temperature. It was

  8. Concentrating Solar Power - Molten Salt Pump Development, Final Technical Report (Phase 1)

    Energy Technology Data Exchange (ETDEWEB)

    Michael McDowell; Alan Schwartz

    2010-03-31

    The purpose of this project is to develop a long shafted pump to operate at high temperatures for the purpose of producing energy with renewable resources. In Phase I of this three phase project we developed molten salt pump requirements, evaluated existing hardware designs for necessary modifications, developed a preliminary design of the pump concept, and developed refined cost estimates for Phase II and Phase III of the project. The decision has been made not to continue the project into Phases II and III. There is an ever increasing world-wide demand for sources of energy. With only a limited supply of fossil fuels, and with the costs to obtain and produce those fuels increasing, sources of renewable energy must be found. Currently, capturing the sun's energy is expensive compared to heritage fossil fuel energy production. However, there are government requirements on Industry to increase the amount of energy generated from renewable resources. The objective of this project is to design, build and test a long-shafted, molten salt pump. This is the type of pump necessary for a molten salt thermal storage system in a commercial-scale solar trough plant. This project is under the Department of Energy (DOE) Solar Energy Technologies Program, managed by the Office of Energy Efficiency and Renewable Energy. To reduce the levelized cost of energy (LCOE), and to meet the requirements of 'tomorrows' demand, technical innovations are needed. The DOE is committed to reducing the LCOE to 7-10 cents/kWh by 2015, and to 5-7 cents/kWh by 2020. To accomplish these goals, the performance envelope for commercial use of long-shafted molten salt pumps must be expanded. The intent of this project is to verify acceptable operation of pump components in the type of molten salt (thermal storage medium) used in commercial power plants today. Field testing will be necessary to verify the integrity of the pump design, and thus reduce the risk to industry. While the primary

  9. Transport properties of molten-salt reactor fuel mixtures: the case of Na, Li, Be/F and Li, Be, Th/F salts

    International Nuclear Information System (INIS)

    In this paper we have compiled transport properties information, available, on two types of FLiBe based salt mixtures (Na,Li,Be/F and Li,Be,Th/F) that are presently of importance in the design of innovative molten-salt burner reactors. Estimated and/or experimental values measured (particularly, from prior US and Russian studies, as well our recent studies) are given for the following properties: viscosity, thermal conductivity, phase transition behaviour, heat capacity, density and thermal expansion. (author)

  10. Development of Molten-Salt Heat Transfer Fluid Technology for Parabolic Trough Solar Power Plants - Public Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    Grogan, Dylan C. P.

    2013-08-15

    Executive Summary This Final Report for the "Development of Molten-Salt Heat Transfer Fluid (HTF) Technology for Parabolic Trough Solar Power Plants” describes the overall project accomplishments, results and conclusions. Phase 1 analyzed the feasibility, cost and performance of a parabolic trough solar power plant with a molten salt heat transfer fluid (HTF); researched and/or developed feasible component options, detailed cost estimates and workable operating procedures; and developed hourly performance models. As a result, a molten salt plant with 6 hours of storage was shown to reduce Thermal Energy Storage (TES) cost by 43.2%, solar field cost by 14.8%, and levelized cost of energy (LCOE) by 9.8% - 14.5% relative to a similar state-of-the-art baseline plant. The LCOE savings range met the project’s Go/No Go criteria of 10% LCOE reduction. Another primary focus of Phase 1 and 2 was risk mitigation. The large risk areas associated with a molten salt parabolic trough plant were addressed in both Phases, such as; HTF freeze prevention and recovery, collector components and piping connections, and complex component interactions. Phase 2 analyzed in more detail the technical and economic feasibility of a 140 MWe,gross molten-salt CSP plant with 6 hours of TES. Phase 2 accomplishments included developing technical solutions to the above mentioned risk areas, such as freeze protection/recovery, corrosion effects of applicable molten salts, collector design improvements for molten salt, and developing plant operating strategies for maximized plant performance and freeze risk mitigation. Phase 2 accomplishments also included developing and thoroughly analyzing a molten salt, Parabolic Trough power plant performance model, in order to achieve the project cost and performance targets. The plant performance model and an extensive basic Engineering, Procurement, and Construction (EPC) quote were used to calculate a real levelized cost of energy (LCOE) of 11.50

  11. Simplified Reference Electrode for Electrorefining of Spent Nuclear Fuel in High Temperature Molten Salt

    International Nuclear Information System (INIS)

    Pyrochemical processing plays an important role in development of proliferation-resistant nuclear fuel cycles. At the Idaho National Laboratory (INL), a pyrochemical process has been implemented for the treatment of spent fuel from the Experimental Breeder Reactor II (EBR-II) in the last decade. Electrorefining in a high temperature molten salt is considered a signature or central technology in pyroprocessing fuel cycles. Separation of actinides from fission products is being demonstrated by electrorefining the spent fuel in a molten UCl3-LiCl-KCl electrolyte in two engineering scale electrorefiners (ERs). The electrorefining process is current controlled. The reference electrode provides process information through monitoring of the voltage difference between the reference and the anode and cathode electrodes. This information is essential for monitoring the reactions occurring at the electrodes, investigating separation efficiency, controlling the process rate, and determining the process end-point. The original reference electrode has provided good life expectancy and signal stability, but is not easily replaceable. The reference electrode used a vycor-glass ion-permeable membrane containing a high purity silver wire with one end positioned in ∼2 grams of LiCl/KCl salt electrolyte with a low concentration (∼1%) AgCl. It was, however, a complex assembly requiring specialized skill and talent to fabricate. The construction involved multiple small pieces, glass joints, ceramic to glass joints, and ceramic to metal joints all assembled in a high purity inert gas environment. As original electrodes reached end-of-life it was uncertain if the skills and knowledge were readily available to successfully fabricate replacements. Experimental work has been conducted to identify a simpler electrode design while retaining the needed long life and signal stability. This improved design, based on an ion-permeable membrane of mullite has been completed. Use of the silver

  12. Molten salt thermal energy storage systems. Project 8981, final report

    Energy Technology Data Exchange (ETDEWEB)

    Maru, H.C.; Dullea, J.F.; Kardas, A.; Paul, L.

    1978-03-01

    The feasibility of storing thermal energy at temperatures of 450/sup 0/ to 535/sup 0/C (850/sup 0/ to 1000/sup 0/F) in the form of latent heat of fusion has been examined for over 30 inorganic salts and salt mixtures. Alkali carbonate mixtures are attractive as phase-change storage materials in this temperature range because of their relatively high storage capacity and thermal conductivity, moderate cost, low volumetric expansion upon melting, low corrosivity, and good chemical stability. An equimolar mixture of Li/sub 2/CO/sub 3/ and K/sub 2/CO/sub 3/, which melts at 505/sup 0/C with a latent heat of 148 Btu/lb, was chosen for experimental study. The cyclic charge/discharge behavior of laboratory- and engineering-scale systems was determined and compared with predictions based on a mathematical heat-transfer model that was developed during this program. The thermal performance of one engineering-scale unit remained very stable during 1400 hours of cyclic operation. Several means of improving heat conduction through the solid salt were explored. Areas requiring further investigation have been identified.

  13. Metals recovering from waste printed circuit boards (WPCBs) using molten salts

    International Nuclear Information System (INIS)

    Highlights: ► Recovering of valuable metals from WPCBs. ► Low temperature treatment, i.e., 300 °C. ► Copper, and precious metals are recovered, without dissolution or melting. ► Many hazardous gases are dissolved and trapped in the molten salt. ► Under operation without oxygen the flue gas contains large quantities of hydrogen. - Abstract: Recycling of waste electrical and electronic equipments (WEEE) has been taken into consideration in the literature due to the large quantity of concerned wastes and their hazardous contents. The situation is so critical that EU published European Directives imposing collection and recycling with a minimum of material recovery . Moreover, WEEEs contain precious metals, making the recycling of these wastes economically interesting, but also some critical metals and their recycling leads to resource conservation. This paper reports on a new approach for recycling waste printed circuit boards (WPCBs). Molten salts and specifically molten KOH–NaOH eutectic is used to dissolve glasses, oxides and to destruct plastics present in wastes without oxidizing the most valuable metals. This method is efficient for recovering a copper-rich metallic fraction, which is, moreover, cleared of plastics and glasses. In addition, analyses of gaseous emission show that this method is environmentally friendly since most of the process gases, such as carbon monoxide and dioxide and halogens, are trapped in the highly basic molten salt. In other respects, under operation without oxygen, a large quantity of hydrogen is produced and might be used as fuel gas or as synthesis gas, leading to a favourable energy balance for this new process.

  14. Molten Salt Electrodeposition of Silicon in Cu-Si

    Science.gov (United States)

    Sokhanvaran, Samira

    Widespread use of solar energy has not been realized to date because its cost is not competitive with conventional energy sources. The high price of solar grade silicon has been one of the barriers against photovoltaic industry achieving its much anticipated growth. Therefore, developing a method, which is energy efficient and will deliver inexpensive silicon feedstock material is essential. The electrodeposition of Si from a cryolite-based melt was investigated in the present work as a possible solution. This study proposed electrowinning of Si in molten Cu-Si alloy, to decrease the working temperature and increase the efficiency. Solvent refining can be used to recover Si from Cu-Si and also as a second purification method. The physicochemical properties of the potential electrolyte, cryolite-SiO 2 melts, were studied in the first step of this work. The deposition potential of Si on a graphite cathode was measured to determine the working potential and the effect of SiO2 concentration on it. In the next step, the deposition potential of Si from cryolite--SiO2 melt on Cu and Cu-Si cathodes was determined using cyclic voltammetry. Next, the cathodic and the anodic current inefficiencies of the process were measured. Continuous analysis of the evolved gas enabled the instantaneous measurement of the current efficiency and the kinetics of the deposition. Finally, the effectiveness of the process in delivering high purity Si was investigated. Si dendrites were precipitated out of the Cu-Si cathode and recovered to determine the purity of the final product as the final step of this study. The produced Si was separated from the alloy matrix by crushing and acid leaching and the purity was reported. The findings of this research show that the proposed method has the potential to produce high purity silicon with low B content. Further development is required to remove some metallic impurities that are remained in Si.

  15. Titanium Powder Preparation from TiCl4 in the Molten Salt

    OpenAIRE

    Suzuki, R. O.; Deura, T. N.; Ishii, R.; Matsunaga, T.; Harada, T. N.; Wakino, M.; Ono, K.

    2000-01-01

    Two processes are proposed to produce Ti powder directly from TiCl4 gas using molten magnesium as reductant. TiCl4 gas injection into a liquid Mg layer through the molten chloride salts could produce the Ti powder of 1 to 10 μm in diameter. Neither the operation temperature nor the salt composition affected the powder morphology. When TiCl4 was dissolved once in the molten salt as Ti2+, and when Mg successively reduced this Ti2+, the Ti morphology varied from the needle-like to the round shap...

  16. Glovebox design requirements for molten salt oxidation processing of transuranic waste

    Energy Technology Data Exchange (ETDEWEB)

    Ramsey, K.B.; Acosta, S.V. [Los Alamos National Lab., NM (United States); Wernly, K.D. [Molten Salt Oxidation Corp., Bensalem, PA (United States)

    1998-12-31

    This paper presents an overview of potential technologies for stabilization of {sup 238}Pu-contaminated combustible waste. Molten salt oxidation (MSO) provides a method for removing greater than 99.999% of the organic matrix from combustible waste. Implementation of MSO processing at the Los Alamos National Laboratory (LANL) Plutonium Facility will eliminate the combustible matrix from {sup 238}Pu-contaminated waste and consequently reduce the cost of TRU waste disposal operations at LANL. The glovebox design requirements for unit operations including size reduction and MSO processing will be presented.

  17. Molten salt pyrolysis of milled beech wood using an electrostatic precipitator for oil collection

    OpenAIRE

    Nygård, Heidi S.; Espen Olsen

    2015-01-01

    A tubular electrostatic precipitator (ESP) was designed and tested for collection of pyrolysis oil in molten salt pyrolysis of milled beech wood (0.5-2 mm). The voltage-current (V-I) characteristics were studied, showing most stable performance of the ESP when N2 was utilized as inert gas. The pyrolysis experiments were carried out in FLiNaK and (LiNaK)2CO3 over the temperature range of 450-600 ℃. The highest yields of pyrolysis oil were achieved in FLiNaK, with a maximum of 34.2 wt% at 500 ℃...

  18. Molten salt electrolytic reduction of metal oxides with a view to the processing of nuclear materials

    International Nuclear Information System (INIS)

    The winning of metals from their oxides is a subject of huge academic and industrial interest. Molten salt technologies play a key role in this field, as evidenced by the long-established and mature technologies used for the winning of metals such as aluminium, magnesium, lithium and sodium and several others. The objective of this contribution is to review the key features of the FFC Cambridge process, highlight its general advantages and unique versatility and, finally, emphasise its relevance in the reprocessing of spent oxide nuclear fuel in the context of establishing viable nuclear technologies for the future

  19. Nuclear performance of molten salt fusion--fission symbiotic systems for catalyzed DD and DT reactors

    International Nuclear Information System (INIS)

    The nuclear performance of a fusion-fission hybrid reactor having a molten salt composed of Na-Th-F-Be as the blanket fertile material and operating with a catalyzed DD plasma is compared to a similar system utilizing a Li-Th-F-Be salt and operating with a DT plasma. The production of fissile fuel via the 232Th-233U fuel cycle was considered on the basis of its potential nonproliferation aspects. The calculations were performed using one-dimensional discrete ordinates methods to compare neutron balances, fuel producion rates, energy deposition rates, and the radiation damage in the reactor structure

  20. Conceptual Design of Forced Convection Molten Salt Heat Transfer Testing Loop

    Energy Technology Data Exchange (ETDEWEB)

    Manohar S. Sohal; Piyush Sabharwall; Pattrick Calderoni; Alan K. Wertsching; S. Brandon Grover

    2010-09-01

    This report develops a proposal to design and construct a forced convection test loop. A detailed test plan will then be conducted to obtain data on heat transfer, thermodynamic, and corrosion characteristics of the molten salts and fluid-solid interaction. In particular, this report outlines an experimental research and development test plan. The most important initial requirement for heat transfer test of molten salt systems is the establishment of reference coolant materials to use in the experiments. An earlier report produced within the same project highlighted how thermophysical properties of the materials that directly impact the heat transfer behavior are strongly correlated to the composition and impurities concentration of the melt. It is therefore essential to establish laboratory techniques that can measure the melt composition, and to develop purification methods that would allow the production of large quantities of coolant with the desired purity. A companion report describes the options available to reach such objectives. In particular, that report outlines an experimental research and development test plan that would include following steps: •Molten Salts: The candidate molten salts for investigation will be selected. •Materials of Construction: Materials of construction for the test loop, heat exchangers, and fluid-solid corrosion tests in the test loop will also be selected. •Scaling Analysis: Scaling analysis to design the test loop will be performed. •Test Plan: A comprehensive test plan to include all the tests that are being planned in the short and long term time frame will be developed. •Design the Test Loop: The forced convection test loop will be designed including extensive mechanical design, instrument selection, data acquisition system, safety requirements, and related precautionary measures. •Fabricate the Test Loop. •Perform the Tests. •Uncertainty Analysis: As a part of the data collection, uncertainty analysis will

  1. HIGH PURITY ALUMINIUM-LITHIUM MASTER ALLOY BY MOLTEN SALT ELECTROLYSIS

    OpenAIRE

    Watanabe, Y.; Toyoshima, M.; Itoh, K.

    1987-01-01

    The aim of this work is to develop the economical production process of the Al-Li master alloy free from metallic sodium, calcium and potassium. This master alloy can be used for aluminium-lithium alloys for structual materials of aircrafts, automobiles and robots. Moreover the Al-Li master alloy with lithium content of 18-20wt. % is applicable to the blanket of fusion reactors and the active mass of batteries. This Al-Li master alloy can be produced by means of LiCl-KCl molten salt electroly...

  2. Luminescent Properties of Y2O3:Eu3+ Nanocrystals Prepared by Molten Salt Synthesis

    Directory of Open Access Journals (Sweden)

    Lijun Luo

    2013-01-01

    Full Text Available A series of red phosphors Y2O3:Eu3+ were prepared by the molten salt method with different surfactants. Their structures, morphologies, and the photoluminescent properties were investigated at room temperature. The particles size of Y2O3:Eu3+ can be controlled by adjusting the kinds of surfactants. The phosphor Y2O3:Eu3+ prepared with NP-10 [polyoxyethylene (10 nonyl phenyl ether] shows regular morphology and higher crystallinity, and its average particle size is about 200 nm. Bright red light can be observed by naked eyes from the red phosphor under 254 nm excitation.

  3. Glovebox design requirements for molten salt oxidation processing of transuranic waste

    International Nuclear Information System (INIS)

    This paper presents an overview of potential technologies for stabilization of 238Pu-contaminated combustible waste. Molten salt oxidation (MSO) provides a method for removing greater than 99.999% of the organic matrix from combustible waste. Implementation of MSO processing at the Los Alamos National Laboratory (LANL) Plutonium Facility will eliminate the combustible matrix from 238Pu-contaminated waste and consequently reduce the cost of TRU waste disposal operations at LANL. The glovebox design requirements for unit operations including size reduction and MSO processing will be presented

  4. Preparation of niobium nanoparticles by sodiothermic reduction of Nb_2O_5 in molten salts

    Institute of Scientific and Technical Information of China (English)

    2012-01-01

    Niobium nanoparticles with high purity were prepared by a sodiothermic reduction process using Nb2O5 as the raw material, LiCl, NaCl, KCl and CaCl2 as the diluents and sodium as the reducing reagent. The effects of the different molten salt systems, CaCl2 content, reaction time, excessive sodium and reaction temperature on the characteristics of the obtained niobium powder were discussed. The as-prepared niobium nanoparticles under the optimum experimental conditions were obtained by sodiothermic reduction ...

  5. Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Salt Fast Reactor (MSFR)

    International Nuclear Information System (INIS)

    Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor (MSFR) are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated. (authors)

  6. Preparation of Mg-Li alloys by electrolysis in molten salt at low temperature

    Institute of Scientific and Technical Information of China (English)

    Mi Lin Zhang; Yong De Yan; Zhi Yao Hou; Lu An Fan; Zeng Chen; Ding Xiang Tang

    2007-01-01

    A new technology for preparation of low cost Mg-Li alloys was studied. The alloys were prepared by electrolysis in molten were investigated, and optimal electrolysis parameters were obtained. Mg-Li alloys with low lithium content (about 25%) were prepared by the unique method of a higher post-thermal treatment temperature after electrolysis at low temperature. The results showed that the electrolysis can be carried out at low temperature, which resulted in reducing preparation cost due to energy saving.The new technology for the preparation of Mg-Li alloy by electrolysis in molten salt was proved to be feasible.

  7. Stead-state characteristic study of heat exchanger in water-cooled passive heat removal system for molten salt reactor

    International Nuclear Information System (INIS)

    Background: In the water-cooled passive heat removal system for molten salt reactor, the decay heat generated in molten salt can finally be transferred to the heat exchanger placed in water tank by natural circulation. Purpose: Based on the principles of high safety and simplification, there is a need to transfer the decay heat passively without using external power. Methods: The heat exchanger consists of a set of bundles submerged into the water tank with a tube header at each side. Based on the flow process, corresponding numerical model was constructed in the code of C++. Then the total heat exchange coefficient is got and the heat transfer area is calculated. Continually iterate the heat transfer area until the iteration stopping criterion is met, after that the dimensions of water tank are figured out. Results: While the decay power is 100 kW in the initial of the operation, the power of heat exchanger reaches the maximum value of 130 kW due to the low-temperature water in water tank. Then it drops quickly for the decrease of heat exchanger pressure and the rise of water temperature in water tank. When the heat exchanger pressure begins to rise, the heat exchanger power drops slower than before. The heat transfer ability begins to decrease quickly as the temperature difference between inside and outside of heat exchanger tubes lowers. Then it drops gradually as a result of the slowly changed pressure. During early operation, the heat exchanger pressure decreases because the steam generation rate is lower than the steam condensation rate. Then the condition varies as the heat exchanger power declines gradually. When boiling happens inside the water tank, the steam condensation rate raises due to the increasing heat transfer ability which makes the pressure of heat exchanger drops quickly. Afterwards, the heat exchanger pressure changes very slowly as the steam generation rate is approximate to the steam condensation rate. The mass of water in water tank

  8. Effects of operating conditions on molten-salt electrorefining for zirconium recovery from irradiated Zircaloy-4 cladding of pressurized water reactor

    International Nuclear Information System (INIS)

    Highlights: • Computational simulation on electrorefining of irradiated Zircaloy-4 cladding. • Composition of irradiated Zircaloy-4 cladding of pressurized water reactor. • Redox behavior of elements in irradiated Zircaloy cladding during electrorefining. • Effect of electrorefining operating conditions on decontamination factor. - Abstract: To reduce the final waste volume from used nuclear fuel assembly, it is significant to decontaminate irradiated cladding. Electrorefining in high temperature molten salt could be one of volume decontamination processes for the cladding. This study examines the effect of operating conditions on decontamination factor in electrorefining of irradiated Zircaloy-4 cladding of pressurized water reactor. One-dimensional time-dependent electrochemical reaction code, REFIN, was utilized for simulating irradiated cladding electrorefining. Composition of irradiated Zircaloy was estimated based on ORIGEN-2 and other literatures. Co and U were considered in electrorefining simulation with major elements of Zircaloy-4 to represent activation products and actinides penetrating into the cladding respectively. Total 240 cases of electrorefining are simulated including 8 diffusion boundary layer thicknesses, 10 concentrations of contaminated molten salt and 3 termination conditions. Decontamination factors for each case were evaluated and it is revealed that the radioactivity of Co-60 in recovered zirconium on cathode could decrease below the clearance level when initial concentration of chlorides except ZrCl4 is lower than 1 × 10−11 weight fraction if electrorefining is finished before anode potential reaches −1.8 V (vs. Cl2/Cl−)

  9. Corrosion behavior of Hastelloy-N alloys in molten salt fluoride in Ar gas or in air

    International Nuclear Information System (INIS)

    The effects of air on the corrosion of Hastelloy-N alloys in molten salt coolant containing fission product elements were investigated to determine the safety of structural materials in high-temperature reactors cooled with fluoride salt. Corrosion tests of Hastelloy-N in the molten fluoride salt FLiNaK in an alumina crucible and a graphite crucible under argon gas or air were performed at 773–923 K for 100 h. The depth of corrosive attack, as well as the extent of chromium and molybdenum depletion, increased with increasing temperature. The extent of Hastelloy-N corrosion in molten salt under air was significantly greater than under argon gas. The effect of adding the impurity cesium iodide to molten salt containing nuclear waste fuel on the corrosion behavior was negligible. (author)

  10. Kinetic studies on the removal of fission products from molten salt using Zeolite-4A. Contributed Paper RD-15

    International Nuclear Information System (INIS)

    Molten salt electrorefining process is one of the nonaqueous processes, being developed for reprocessing metallic spent fuel. This process uses liquid metals and molten salts and is operated at elevated temperatures. In the electro-refining process, the spent fuel is used as the anode of the electro-refiner and the actinide elements in the spent fuel are electrotransported from the anode through the molten salt electrolyte onto a suitable cathode where they are collected as metals in pure form. After some batches are processed, chlorides of fission products such as alkali, alkaline earth and rare earth metals accumulate in the electrolyte salt. The accumulated FPs in the salt will be removed by adsorption/ion-exchange by using zeolite columns. Hence, kinetic studies on the adsorption of Cs, Ba which are some of the major FP products in LiCI-KCI eutectic, have been carried out

  11. Design and installation of gloveboxes with inert gas atmosphere for molten salt electrolysis and preparation of alloy samples

    International Nuclear Information System (INIS)

    Gloveboxes with argon gas atmosphere were equipped for the investigation of direct electrolysis of actinide mononitride in molten salt and subsequent preparation of alloy samples for experimental use. Design of the gloveboxes was focused on keeping high purity enough to prevent the reaction of nitride and alloy samples and also chloride -molten salt with oxygen and moisture besides the safety for handling actinides such as plutonium and neptunium. As a result, both the oxygen and moisture contents in the gloveboxes could be kept less than 1 ppm, which would be in accord with the above purpose. A molten salt electrorefiner, a cathode processor, an electrolysis testing apparatus, an arc furnace, two furnaces for heat treatment and a differential scanning calorimeter were installed in the gloveboxes. It was confirmed that these apparatus had an excellent ability for the respective experimental use and met with specifications for safety and operation. (author)

  12. Preparation of niobium carbide powder by electrochemical reduction in molten salt

    Energy Technology Data Exchange (ETDEWEB)

    Song, Qiushi [School of Materials Science and Metallurgy, Northeastern University, Shenyang 110819 (China); Xu, Qian, E-mail: qianxu201@mail.neu.edu.cn [School of Materials Science and Metallurgy, Northeastern University, Shenyang 110819 (China); School of Materials Science and Engineering, Shanghai University, Shanghai 200072 (China); Meng, Jingchun; Lou, Taiping; Ning, Zhiqiang [School of Materials Science and Metallurgy, Northeastern University, Shenyang 110819 (China); Qi, Yang [College of Science, Northeastern University, Shenyang 110819 (China); Yu, Kai [School of Materials Science and Metallurgy, Northeastern University, Shenyang 110819 (China)

    2015-10-25

    The niobium carbide powder was prepared via electrochemical reduction of the mixture of Nb{sub 2}O{sub 5} and carbon in molten CaCl{sub 2}–NaCl. The reaction pathway from the sintered precursor to the final product has been investigated. The effect of the working temperature on the reduction of the Nb{sub 2}O{sub 5}/C composite precursor was considered. The role of carbon during the electrochemical reduction of the composite pellet was discussed. The samples were analysed by XRD and SEM. The results indicated that the NbC powder was approximately 200 nm after the reduction. Nb{sub 2}O{sub 5} was gradually reduced to Nb, and NbC was subsequently obtained by the reaction of carbon with Nb metal. In addition, Nb{sub 2}O{sub 5} could spontaneously react with CaO in the melt to form a serious of calcium niobates. The participation of carbon was available for the efficiency of electro-reduction of Nb{sub 2}O{sub 5}. - Graphical abstract: Niobium carbide powder was electrochemically prepared in molten salt, and the reduction pathway was illustrated schematically. - Highlights: • NbC powder was prepared electrochemically in molten salt. • The working temperature was lower than that of carbothermic reduction. • The reduction pathway was discussed compared to direct electro-deoxidation of Nb{sub 2}O{sub 5}.

  13. Molten salt synthesis of sodium lithium titanium oxide anode material for lithium ion batteries

    International Nuclear Information System (INIS)

    Highlights: • Na2Li2Ti6O12 has been successfully synthesized via a molten salt route. • Calcination temperature is an important effect on the component and microstructure of the product. • Pure phase Na2Li2Ti6O12 could be obtained at 700 °C for 2 h. - Abstract: The sodium lithium titanium oxide with composition Na2Li2Ti6O14 has been synthesized by a molten salt synthesis method using sodium chloride and potassium chloride mixture as a flux medium. Synthetic variables on the synthesis, such as sintering temperature, sintering time and the amount of lithium carbonate, were intensively investigated. Powder X-ray diffraction and scanning electron microscopy images of the reaction products indicates that pure phase sodium lithium titanium oxide has been obtained at 700 °C, and impure phase sodium hexatitanate with whiskers produced at higher temperature due to lithium evaporative losses. The results of cyclic voltammetry and discharge–charge tests demonstrate that the synthesized products prepared at various temperatures exhibited electrochemical diversities due to the difference of the components. And the sample obtained at 700 °C revealed highly reversible insertion and extraction of Li+ and displayed a single potential plateau at around 1.3 V. The product obtained at 700 °C for 2 h exhibits good cycling properties and retains the specific capacity of 62 mAh g−1 after 500 cycles

  14. Applicability of molten salt oxidation to the destruction of actinide-contaminated wastes

    International Nuclear Information System (INIS)

    A 1989 ban on incineration in the state of New Mexico caused cessation of actinide-contaminated cheesecloth, paper, and wood incineration within the Plutonium Facility (TA-55) at Los Alamos National Laboratory. Subsequently, plastic wipes were substituted for cheesecloth in the cleaning of glovebox interiors. However, waste minimization is not achieved by these measures since the wipes are discarded as Waste Isolation Pilot Plant certifiable wastes. After the ban was instituted, thermal decomposition of cheesecloth under argon at elevated temperature was examined and found satisfactory although scale of operation and speed were inferior to incineration. In 1991, the ban on incineration was lifted in New Mexico but Alamos has not chosen to pursue renewal of incineration at the Plutonium Facility. This paper reports that Los Alamos is looking from alternatives to incineration and thermal decomposition which are compatible with molten salt processing technology, historically a strength in actinide research at the Laboratory. Also, the technology must significantly reduce the volume of the waste upon treatment, i.e. waste minimization. Molten salt oxidation (MSO) has the promise of such a technology

  15. Electrochemical behavior of the Lanthanide at Cd, Bi electrodes in LiCl-KCl molten salt

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Beomkyu; Lee, Aeri; Kim, Yongil; Park, Byung Gi [Soonchunhyang Univ., Asan (Korea, Republic of)

    2014-05-15

    Electochemical behavior of PrCl{sub 3} and NdCl{sub 3} it the liquid metal electrode studied using cyclic-voltammetry. The electro-reduction of Pr, Nd on electrodes proceeds via the formation of distinct non-homogeneous phases. The measurement for various intermetallic compounds in two-phase coexisting states was carried out in the high temperature. Gibbs free energies of formation Pr-Cd, Pr-Bi, Nd-Cd and Nd-Bi intermetallic compounds was with data of cyclic-voltammetry 4 session. Liquid metal cathode is used to recover actinides from molten salt in an electrowinning among the pyroprocess. Reductive or oxidative extraction between molten salt and liquid metal has been also considered for group partitioning of lanthanide and actinide. Cadmium or bismuth as liquid metal has been extensively studied to investigate the effects of liquid metals on the process. However, there is sparse information about the electrode reaction of lanthanide and actinide at the liquid metal electrodes. The purpose of the present study is to investigate the electrochemical behavior of praseodymium and neodymium among lanthanides(Ln) in molten LiCl-KCl salt at liquid metal electrode such as cadmium and bismuth using cyclic voltammetry and derive the thermochemical properties.

  16. Numerical simulation of induction heating in FLiNaK molten salt high temperature testing loop

    International Nuclear Information System (INIS)

    Background: The graphite pebble test section of FLiNaK molten salt high temperature testing loop was designed to study the heat transfer behavior and corrosion behavior between FLiNaK molten salt and fuel pebbles. In graphite pebble test section, the graphite pebbles are heated using a unique medium frequency inductive heating technique to simulate the heat generation inside the fuel pebbles of Pebble Bed Advanced High Temperature Reactor (PB-AHTR), Purpose: The aim is to study the influence of the induction heating by three kinds of graphite pebbles packing scheme. Methods: Numerical simulation of induction heating using finite element method is implemented and the power loss in graphite pebbles is calculated, and then the three kinds of graphite pebbles packing scheme are compared and analyzed from eddy power loss and power distribution. Results: The results show that the relative axial power distributions of three packing schemes are almost the same and the radial power distribution of the packing scheme with no contact between graphite pebbles in each layer is more uniform than that of other packing schemes. Conclusions: In order to heat graphite pebbles evenly and simulate the heat generation inside fuel pebbles of PB-AHTR better, the packing scheme with no contact between graphite pebbles in each layer is recommended. (authors)

  17. Solidification of Molten Salt Waste by Gel-Route Pre-treatment

    International Nuclear Information System (INIS)

    This study suggested a new method for the solidification of molten salt waste generated from the electro-metallurgical process in the spent fuel treatment. Using binary material system, sodium silicate and phosphoric acid, metal chlorides were converted into metal phosphate in the micro-reaction module formed by SiO2 particles. The volatile element in the reaction module would little vaporized below 1100 .deg. C. After the gel product was mixed with borosilicate glass powder and thermally treated at 1000 .deg. C, Li exists as Li3PO4 separated from glass phase and, Cs and Sr would be incorporated into an amorphous phase from XRD analysis. In case of the addition of ZrCl4 to the binary system, the gel products were transformed into NZP structure considered as an prospective ceramic waste form after heat-treatment above 700 .deg. C. From these results, the gel-route pretreatment can be considered as an effective approach to the solidification of molten salt waste by the confirmed process or waste form and this also would be an alternative method on the ANL method using zeolites in USA by the confirmation of its chemical durability as an future work.

  18. Conceptual Design of a 100 MWe Modular Molten Salt Power Tower Plant

    Energy Technology Data Exchange (ETDEWEB)

    James E. Pacheco; Carter Moursund, Dale Rogers, David Wasyluk

    2011-09-20

    A conceptual design of a 100 MWe modular molten salt solar power tower plant has been developed which can provide capacity factors in the range of 35 to 75%. Compared to single tower plants, the modular design provides a higher degree of flexibility in achieving the desired customer's capacity factor and is obtained simply by adjusting the number of standard modules. Each module consists of a standard size heliostat field and receiver system, hence reengineering and associated unacceptable performance uncertainties due to scaling are eliminated. The modular approach with multiple towers also improves plant availability. Heliostat field components, receivers and towers are shop assembled allowing for high quality and minimal field assembly. A centralized thermal-storage system stores hot salt from the receivers, allowing nearly continuous power production, independent of solar energy collection, and improved parity with the grid. A molten salt steam generator converts the stored thermal energy into steam, which powers a steam turbine generator to produce electricity. This paper describes the conceptual design of the plant, the advantages of modularity, expected performance, pathways to cost reductions, and environmental impact.

  19. Archimede solar energy molten salt parabolic trough demo plant: Improvements and second year of operation

    Science.gov (United States)

    Maccari, Augusto; Donnola, Sandro; Matino, Francesca; Tamano, Shiro

    2016-05-01

    Since July 2013, the first stand-alone Molten Salt Parabolic Trough (MSPT) demo plant, which was built in collaboration with Archimede Solar Energy and Chiyoda Corporation, is in operation, located adjacent to the Archimede Solar Energy (ASE) manufacturing plant in Massa Martana (Italy). During the two year's operating time frame, the management of the demo plant has shown that MSPT technology is a suitable and reliable option. Several O&M procedures and tests have been performed, as Heat Loss and Minimum Flow Test, with remarkable results confirming that this technology is ready to be extended to standard size CSP plant, if the plant design takes into account molten salt peculiarities. Additionally, the plant has been equipped on fall 2014 with a Steam Generator system by Chiyoda Corporation, in order to test even this important MSPT plant subsystem and to extend the solar field active time, overcoming the previous lack of an adequate thermal load. Here, a description of the plant improvements and the overall plant operation figures will be presented.

  20. Tritium permeation and recovery for the helium-cooled molten salt fusion breeder

    International Nuclear Information System (INIS)

    Design concepts are presented to control tritium permeation from a molten salt/helium fusion breeder reactor. This study assumes tritium to be a gas dissolved in molten salt, with TF formation suppressed. Tritium permeates readily through the hot steel tubes of the reactor and steam generator and will leak into the steam system at the rate of about one gram per day in the absence of special permeation barriers, assuming that 1% of the helium coolant flow rate is processed for tritium recovery at 90% efficiency per pass. The proposed permeation barrier for the reactor tubes is a 10 μm layer of tungsten which, in principle, will reduce tritium blanket permeation by a factor of about 300 below the bare-steel rate. A research and development effort is needed to prove feasibility or to develop alternative barriers. A 1 mm aluminum sleeve is proposed to suppress permeation through the steam generator tubes. This gives a calculated reduction factor of more than 500 relative to bare steel, including a factor of 30 due to an assumed oxide layer. The permeation equations are developed in detail for a multi-layer tube wall including a frozen salt layer and with two fluid boundary-layer resistances. Conditions are discussed for which Sievert's or Henry's Law materials become flux limiters. An analytical model is developed to establish the tritium split between wall permeation and reactor-tube flow

  1. Pyridinium molten salts as co-adsorbents in dye-sensitized solar cells

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jui-Cheng; Sun, I-Wen [Department of Chemistry, National Cheng Kung University, Tainan 701 (China); Yang, Cheng-Hsien; Yang, Hao-Hsun; Hsueh, Mao-Lin [Nano-Powder and Thin Film Technology Center, ITRI South, Tainan 709 (China); Ho, Wen-Yueh [Institute of Cosmetic Science, Chia Nan University of Pharmacy and Science, Tainan 717 (China); Chang, Jia-Yaw [Department of Chemical Engineering, National Taiwan University of Science and Technology, Taipei 106 (China)

    2011-01-15

    The influence of using pyridinium molten salts as co-adsorbents to modify the monolayer of a TiO{sub 2} semiconductor on the performance of a dye-sensitized solar cell is studied. The current-voltage characteristics are measured under AM 1.5 (100 mW cm{sup -2}). The pyridinium molten salts significantly enhance the open-circuit photovoltage (V{sub oc}), the short circuit photocurrent density (J{sub sc}) as well as the solar energy conversion efficiency ({eta}). 1-Ethyl-3-carboxypyridinium iodide ([ECP][I]) is applied successfully to prepare an insulating molecular layer with N719, and achieve high energy conversion efficiency as high as 4.49% at 100 mW cm{sup -2} and AM 1.5. The resulting efficiency is 20% higher than that of a non-additive device. This enhancement of conversion efficiency is attributed to the negative shift of the conduction band (CB) edge and the abundant concentration of I{sup -} on the surface of the electrode when using [ECP][I] as the co-adsorbent. (author)

  2. Research on thermal and mechanical behaviour of a Freeze-Valve for molten salt reactor

    International Nuclear Information System (INIS)

    Background: Reactor safety is an important component of developing and designing any nuclear energy systems. The Freeze-Valve is one of the core technologies of the molten salt reactor, which thermal-structural property is directly related to the inherent safety of the molten salt reactor nuclear system. Purpose: The purpose of this paper is to improve the inherent safety of the fourth-generation nuclear energy systems, by researching and optimizing the thermodynamic properties of Freeze-Valve, and exploring reliable safety design. Methods: A 3D Finite Element model to simulate the thermal-structural coupling behaviors was established by the ANSYS software to analyze the performances and the properties of a Freeze-Valve. Results: The analysis result shows that the flat part of the Freeze-Valve plays a very important role in engineering applications. The effects of different factors (heat transfer coefficient, heating power, and insulation size and so on) on the temperature and the stress field of the Freeze-Valve during operation are obtained, which provide some basis for optimization of design and safe operation. The thermal stress of the initial model of the Freeze-Valve is so large that it is easy to cause fatigue failure, owing to the unreasonable initial design (insulation size and heating power, for example). Conclusions: After the relative analysis and structural optimization, the maximum stress of the Freeze-Valve is significantly reduced, and the performance has been greatly improved. (authors)

  3. SAMOFAR - a paradigm shift in reactor safety with the molten salt fast reactor

    International Nuclear Information System (INIS)

    SAMOFAR - Safety Assessment of the Molten Salt Fast Reactor - is a 5M€ project of the European Union research program Horizon 2020. The project consortium consists of 11 participants and the fundamental research part is mainly executed by universities and research laboratories, like CNRS, JRC, ClRTEN, TU Delft and PSI, thereby exploiting each other's unique expertise and infrastructure. The grand objective of SAMOFAR is to prove the innovative safety concepts of the Molten Salt Fast Reactor (MSFR) by advanced experimental and numerical techniques, to deliver a breakthrough in nuclear safety and optimal waste management, and to create a consortium of stakeholders to demonstrate the MSFR beyond SAMOFAR. Furthermore, we will build a software simulator to demonstrate the operational transients, and we will show the mild responses of the MSFR to transients and accident scenarios, using new leading-edge multi-physics simulation tools including uncertainty quantification. All experimental and numerical results will be incorporated into the new reactor design, which will be subjected to a new integral safety assessment method

  4. Corrosion Behavior of Pure Cr, Ni, and Fe Exposed to Molten Salts at High Temperature

    Directory of Open Access Journals (Sweden)

    O. Sotelo-Mazón

    2014-01-01

    Full Text Available Corrosion resistance of pure Fe, Cr, and Ni materials exposed in NaVO3 molten salt at 700°C was evaluated in static air during 100 hours. The corrosion resistance was determined using potentiodynamic polarization, open circuit potential, and lineal polarization resistance. The conventional weight loss method (WLM was also used during 100 hours. The electrochemical results showed that Fe and Cr have a poor corrosion resistance, whereas pure Ni showed the best corrosion performance, which was supported by the passive layer of NiO formed on the metallic surface and the formation of Ni3V2O8 during the corrosion processes, which is a refractory compound with a higher melting point than that of NaVO3, which reduces the corrosivity of the molten salt. Also, the behavior of these materials was associated with the way in which their corresponding oxides were dissolved together with their type of corrosion attack. Through this study, it was confirmed that when materials suffer corrosion by a localized processes such as pitting, the WLM is not reliable, since a certain amount of corrosion products can be kept inside the pits. The corroded samples were analyzed through scanning electron microscopy.

  5. Preparation of niobium carbide powder by electrochemical reduction in molten salt

    International Nuclear Information System (INIS)

    The niobium carbide powder was prepared via electrochemical reduction of the mixture of Nb2O5 and carbon in molten CaCl2–NaCl. The reaction pathway from the sintered precursor to the final product has been investigated. The effect of the working temperature on the reduction of the Nb2O5/C composite precursor was considered. The role of carbon during the electrochemical reduction of the composite pellet was discussed. The samples were analysed by XRD and SEM. The results indicated that the NbC powder was approximately 200 nm after the reduction. Nb2O5 was gradually reduced to Nb, and NbC was subsequently obtained by the reaction of carbon with Nb metal. In addition, Nb2O5 could spontaneously react with CaO in the melt to form a serious of calcium niobates. The participation of carbon was available for the efficiency of electro-reduction of Nb2O5. - Graphical abstract: Niobium carbide powder was electrochemically prepared in molten salt, and the reduction pathway was illustrated schematically. - Highlights: • NbC powder was prepared electrochemically in molten salt. • The working temperature was lower than that of carbothermic reduction. • The reduction pathway was discussed compared to direct electro-deoxidation of Nb2O5

  6. Electrochemical study of nickel from urea-acetamide-LiBr low-temperature molten salt

    International Nuclear Information System (INIS)

    Highlights: • CV results show that the charge transfer process of Ni(II)/Ni in urea-acetamide-LiBr is irreversible. • The reduction process is a single step two-electron transfer process. • Chronoamperometry indicates that the reaction on tungsten electrode involves progressive nucleation. • EDS and XRD analyses confirm that the obtained deposits are pure nickel. -- Abstract: The electrochemical behavior of nickel was studied by cyclic voltammetry and chronoamperometry techniques at 353 K using a tungsten electrode in urea-acetamide-LiBr low-temperature molten salt. The cyclic voltammograms indicate that the reduction of Ni(II) to Ni proceeds via a single-step, two-electron transfer process. Chronoamperometric measurements show that the electrodeposition of nickel on the tungsten electrode involves three-dimensional (3D) progressive nucleation under diffusion-controlled growth at 353 K. Nickel coatings were prepared at different cathodic potentials (−0.70 to −0.85 V) and different temperatures (343–373 K) in urea-acetamide-LiBr molten salt. The deposits were characterized by scanning electron microscope (SEM), energy dispersive spectroscopy (EDS), and X-ray diffraction (XRD). The SEM images reveal that uniform, dense, and compact deposits were obtained at more positive cathodic potentials within the temperature range of 343–363 K. The EDS and XRD analyses confirm that the obtained deposits are pure nickel

  7. New Aluminum-Molten Salt Contactor for Pyrochemical Reprocessing of Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    The demonstration of a new pyrometallurgical reprocessing route requires a technological breakthrough in the design of a new type of device compatible with temperatures up to 1000 deg. C. The authors discuss a pyro-processing device in which actinides are recovered by reductive extraction of actinide fluorides from molten fluoride salt contacted with a liquid aluminum phase. The liquid contactor must separate two liquid phases of the same density in a compact device of simple design with maximum efficiency at each chemical treatment stage, compatible with the thermal constraints and suitable for handling, assembly and disassembly in a hot cell. The liquid-liquid contactor ensures molten Al / molten salt separation using an openwork wall. The principle, patented in 2007, consists in using the surface tension and the interfacial tension properties of aluminum in contact with graphite or boron nitride. The Young-Laplace equation is used to calculate the openwork vessel dimensions. The satisfactory performance of this system was validated at laboratory scale by measuring the aluminum holdup in the openwork vessel with or without molten salt hydrostatic pressure. Mass transfer through the openwork material was also observed and verified by reductive extraction under inactive conditions (NdF3 recovery) and active conditions (UF4 recovery). (authors)

  8. Thermal storage performance of molten salt thermocline system with packed phase change bed

    International Nuclear Information System (INIS)

    Highlights: • Molten salt thermocline storage with packed phase change bed is simulated. • Phase change material can remarkably increase the effective discharging energy. • Thermocline can be divided into three stages including phase change layer. • Melting point of phase change material should be slightly below initial temperature. • The discharging efficiency increases with the phase change material content. - Abstract: Comprehensive transient and two-dimensional numerical model is developed to study energy storage performance of molten salt thermocline thermal storage system with packed phase change bed in solar thermal power. The results show that the packed phase change bed can remarkably increase the effective discharging energy and discharging efficiency. Because of phase change material, the thermocline can be divided into three stages including the high temperature thermocline, low temperature thermocline and phase change layer. As the melting point within the inlet and initial temperature increases, the whole discharging time decreases, while the effective discharging energy remarkably increases, and thus the melting point of phase change material should be within the initial temperature and effective outlet temperature for good heat storage performance. As the phase change material content increases, the effective discharging energy increases with the effective discharging time rising, and the effective discharging efficiency also increases

  9. New rational energy strategy depending on the thorium-molten-salt nuclear energy synergetics

    International Nuclear Information System (INIS)

    In the next century, a - environmental impact of Co2, SOxPV, NOxPV, from fossil fuel systems, b - radiowaste and nuclear proliferation troubles mostly coming from trans-U elements (including several thousand tons Pu) in the U-Pu solid fuel nuclear power stations would become fatal problems of the world. The following should also be taken into consideration to solve the problem: c - resource, d - technological safety and simplicity, e - flexibility in size and siting, and f - integral economy in fuel cycle. A new strategy named ''THORIMSNES'' (Thorium-Molten-Salt Nuclear Energy Synergetics) is proposed. These three technologies composed of Thorium, Molten Salt and Separation of power stations and fissile breeding plants, refusing fission breeders. The typical 155 MWe Small Salt Power Station (Fuji-II) has significant characteristics as follows: 1 - no need of graphite exchange and continuous chemical processing except fission-gas removal, and 2 - fuel self sustainability (no fissile supply) in full life after initial 500 days operation, which means very low excess reactivity (nearly no control rods), few fuel transport, no core meltdown, easy operation and maintenance, few radiowaste and thermal pollution, high thermal efficiency and medium-temperature heat supply, etc. mostly achieving the targets mentioned above. (Auth.)

  10. Molten salt related extensions of the SIMMER-III code and its application for a burner reactor

    International Nuclear Information System (INIS)

    Molten salt reactors (MSRs) can be used as effective burners of plutonium (Pu) and minor actinides (MAs) from light water reactor (LWR) spent fuel. In this paper a study was made to examine the thermal hydraulic behaviour of the conceptual design of the molten salt advanced reactor transmuter (MOSART) [Ignatiev, V., Feynberg, O., Myasnikov, A., Zakirov, R., 2003a. Neutronic properties and possible fuel cycle of a molten salt transmuter. Proceedings of the 2003 ANS/ENS International Winter Meeting (GLOBAL 2003), Hyatt Regency, New Orleans, LA, USA 16-20 November 2003]. The molten salt fuel is a ternary NaF-LiF-BeF2 system fuelled with ca. 1 mol% typical compositions of transuranium-trifluorides (PuF3, etc.) from light water reactor spent fuel. The MOSART reactor core does not contain graphite structure elements to guide the flow, so the neutron spectrum is rather hard in order to improve the burning performance. Without those structure elements in the core, the molten salt in core flows freely and the flow pattern could be potentially complicated and may affect significantly the fuel temperature distribution in the core. Therefore, some optimizations of the salt flow pattern may be needed. Here, the main attention has been paid to the fluid dynamic simulations of the MOSART core with the code SIMMER-III [Kondo, Sa., Morita, K., Tobita, Y., Shirakawa, K., 1992. SIMMER-III: an advanced computer program for LMFBR severe accident analysis. Proceedings of the ANP' 92, Tokyo, Japan; Kondo, Sa., Tobita, Y., Morita, K., Brear, D.J., Kamiyama, K., Yamano, H., Fujita, S., Maschek, W., Fischer, E.A., Kiefhaber, E., Buckel, G., Hesselschwerdt, E., Flad, M., Costa, P., Pigny, S., 1999. Current status and validation of the SIMMER-III LMFR safety analysis code. Proceedings of the ICONE-7, Tokyo, Japan], which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors for the thermo-hydraulic and neutronic models so as to

  11. Facile and scalable synthesis of Ti5Si3 nanoparticles in molten salts for metal-matrix nanocomposites.

    Science.gov (United States)

    Estruga, Marc; Girard, Steven N; Ding, Qi; Chen, Lianyi; Li, Xiaochun; Jin, Song

    2014-02-11

    We report a novel synthesis of Ti5Si3 nanoparticles (NPs) via the magnesio-reduction of TiO2 NPs and SiO2 in eutectic LiCl-KCl molten salts at 700 °C. The Ti5Si3 particle size (∼25 nm) is confined to the nanoscale due to the partial dissolution of Mg and silica in the molten salts and a subsequent heterogeneous reduction on the surface of the TiO2 NPs. PMID:24352506

  12. Review of ORNL's MSR technology and status

    International Nuclear Information System (INIS)

    The current status of molten salt reactor development is discussed with reference to the experience from the Oak Ridge Molten Salt Reactor Experiment. Assessment of the future for this reactor system is reviewed with consideration of both advantages and disadvantages. Application of this concept to ADTT (accelerator driven transmutation technology) needs appears to be feasible by drawing on the MSRE experience. Key chemical considerations remain as: solubility, redox behavior, and chemical activity and their importance to ADTT planning is briefly explained. Priorities in the future development of molten salts for these applications are listed, with the foremost being the acceptance of the 2LiF-BeF2 solvent system. 8 refs, 2 figs

  13. The Chemistry and Thermodynamics of Molten-Salt-Reactor Fluoride Solutions

    International Nuclear Information System (INIS)

    A solvent of lithium and beryllium fluorides (about 2 moles of LiF per mole of BeF2) is used in the fuel salt, the coolant salt, and the flush salt of the Molten Salt Reactor Experiment. As a result of the chemical development work done for this reactor concept, considerable chemical and thermodynamic information has been acquired concerning this solvent and its solutions with actinide, lanthanide, and structural metal fluorides. It is the purpose of this paper to review this information, much of which is not yet generally available. The data were obtained mainly by measurements of heterogeneous equilibria, i.e. by equilibration of melts with gaseous mixtures containing hydrogen, hydrogen fluoride or water and by determinations of solid-liquid phase equilibria. The results of these measurements gave direct information about such important chemical problems as: (1) the corrodibility of structural metals and the reducibility of the structural metal ions, Ni2+, Fe2+, Cr2+; (2) reactions with water vapour to form oxide and hydroxide ions, and the removal of these ions; (3) the precipitation, solubility, and tendency toward solid-solution formation of the oxides of beryllium, uranium, zirconium, thorium and the rare earths; (4) the stability of uranium(IV) toward reduction to the trivalent state and possible subsequent disproportionation; and (5) the solubilities and solid solution formation of rare earth fluorides. Equally important has been the wider usefulness of this information when the methods of thermodynamics are brought to bear. Thus the data obtained could be used to: (1) correlate, revise, and extend existing thermochemical data on fluorides and oxides; (2) determine activity coefficients of the components LiF, BeF2, UF4 , ZrF4, and NiF2 in these molten salt solutions; (3) revise the generally accepted phase diagram for UO2-ZrO2; and (4) estimate solubilities and reactivities of compounds not directly investigated. Thus the chemical development programme for

  14. The mechanics of pressed-pellet separators in molten salt batteries.

    Energy Technology Data Exchange (ETDEWEB)

    Long, Kevin Nicholas; Roberts, Christine Cardinal; Roberts, Scott Alan; Grillet, Anne

    2014-06-01

    We present a phenomenological constitutive model that describes the macroscopic behavior of pressed-pellet materials used in molten salt batteries. Such materials include separators, cathodes, and anodes. The purpose of this model is to describe the inelastic deformation associated with the melting of a key constituent, the electrolyte. At room temperature, all constituents of these materials are solid and do not transport cations so that the battery is inert. As the battery is heated, the electrolyte, a constituent typically present in the separator and cathode, melts and conducts charge by flowing through the solid skeletons of the anode, cathode, and separator. The electrochemical circuit is closed in this hot state of the battery. The focus of this report is on the thermal-mechanical behavior of the separator, which typically exhibits the most deformation of the three pellets during the process of activating a molten salt battery. Separator materials are composed of a compressed mixture of a powdered electrolyte, an inert binder phase, and void space. When the electrolyte melts, macroscopically one observes both a change in volume and shape of the separator that depends on the applied boundary conditions during the melt transition. Although porous flow plays a critical role in the battery mechanics and electrochemistry, the focus of this report is on separator behavior under flow-free conditions in which the total mass of electrolyte is static within the pellet. Specific poromechanics effects such as capillary pressure, pressure-saturation, and electrolyte transport between layers are not considered. Instead, a phenomenological model is presented to describe all such behaviors including the melting transition of the electrolyte, loss of void space, and isochoric plasticity associated with the binder phase rearrangement. The model is appropriate for use finite element analysis under finite deformation and finite temperature change conditions. The model

  15. NaCl-2CsCl molten salt purification technology in dry reprocessing process. Phosphate precipitation experiment

    International Nuclear Information System (INIS)

    Japan Nuclear Cycle Development Institute (JNC) is studying the electro-winning process using NaCl-2CsCl molten salt as a dry reprocessing technology development. It would be necessary to repeatedly purify the used salt after the MOX co-deposition electrolysis and the MA recovery in this process. JNC had selected the phosphate precipitation method as a purification technique of the used salt, and we carried out the phosphate precipitation experiment. In this experiment, Na3PO4 was added into the NaCl-2CsCl molten salt including fission product (FP) ions (Ce3+, Nd3+, Sr2+) of non-radioactive, and we investigated the removing performance of FP ions from the NaCl-2CsCl molten salt by the phosphate precipitation. In conclusion, we confirmed most of FP ions in the molten salt were removed by adding Na3PO4 of 1.5 times the mole amount of the FP content. (author)

  16. Status of benchmark calculations of the neutron characteristics of the cascade molten salt ADS for the nuclear waste incineration

    International Nuclear Information System (INIS)

    The facility for incineration of long-lived minor actinides and some dangerous fission products should be an important feature of the future nuclear power (NP). For many reasons the liquid-fuel reactor driven by accelerator can be considered as the perspective reactor- burner for radioactive waste. The fuel of such reactor is the fluoride molten salt composition with minor actinides (Np, Cm, Am) and some fission products (99Tc, 129I, etc.). Preliminary analysis shows that the values of keff, calculated with different codes and nuclear data differ up to several percents for such fuel compositions. Reliable critical and subcritical benchmark experiments with molten salt fuel compositions with significant quantities of minor actinides are absent. One of the main tasks for the numerical study of this problem is the estimation of nuclear data for such fuel compositions and verification of the different numerical codes used for the calculation of keff, neutron spectra and reaction rates. It is especially important for the resonance region where experimental data are poor or absent. The calculation benchmark of the cascade subcritical molten salt reactor is developed. For the chosen nuclear fuel composition the comparison of the results obtained by three different Monte-Carlo codes (MCNP4A, MCU, and C95) using three different nuclear data libraries are presented. This report concerns the investigation of subcritical molten salt reactor unit main peculiarities carried out at the beginning of ISTC project 1486. (author)

  17. The mechanisms for filling carbon nanotubes with molten salts: carbon nanotubes as energy landscape filters

    International Nuclear Information System (INIS)

    The mechanisms for filling carbon nanotubes with molten salts are investigated using molecular dynamics computer simulation. Inorganic nanotubular structures, whose morphologies can be rationalized in terms of the folding, or the removal of sections from, planes of square nets are found to form. The formation mechanisms are found to follow a 'chain-by-chain' motif in which the structures build systematically from charge neutral M-X-M-Xc chains. The formation mechanisms are rationalized in terms of the ion-ion interactions (intra-chain and inter-chain terms). In addition, the mechanisms of filling are discussed in terms of a 'hopping' between basins on the underlying energy landscape. The role of the carbon nanotube as an energy landscape filter is discussed.

  18. First-principles calculations of the thermodynamic properties of transuranium elements in a molten salt medium

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Seunghyo; Kwak, Dohyun; Lee, Juseung; Kang, Joonhee; Han, Byungchan [Daegu Gyeongbuk Institute of Science and Technology, Daegu (Korea, Republic of)

    2014-03-15

    We utilized first-principles density-functional-theory (DFT) calculations to evaluate the thermodynamic feasibility of a pyroprocessing methodology for reducing the volume of high-level radioactive materials and recycling spent nuclear fuels. The thermodynamic properties of transuranium elements (Pu, Np and Cm) were obtained in electrochemical equilibrium with a LiCl-KCl molten salt as ionic phases and as adsorbates on a W(110) surface. To accomplish the goal, we rigorously calculated the double layer interface structures on an atomic resolution, on the thermodynamically most stable configurations on W(110) surfaces and the chemical activities of the transuranium elements for various coverages of those elements. Our results indicated that the electrodeposition process was very sensitive to the atomic level structures of Cl ions at the double-layer interface. Our studies are easily expandable to general electrochemical applications involving strong redox reactions of transition metals in non-aqueous solutions.

  19. Molten Salt Heat Transport Loop: Materials Corrosion and Heat Transfer Phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Kumar Sridharan; Dr. Mark Anderson; Dr. Michael Corradini; Dr. Todd Allen; Luke Olson; James Ambrosek; Daniel Ludwig

    2008-07-09

    An experimental system for corrosion testing of candidate materials in molten FLiNaK salt at 850 degree C has been designed and constructed. While molten FLiNaK salt was the focus of this study, the system can be utilized for evaluation of materials in other molten salts that may be of interest in the future. Using this system, the corrosion performance of a number of code-certified alloys of interest to NGNP as well as the efficacy of Ni-electroplating have been investigated. The mechanisums underlying corrosion processes have been elucidated using scanning electron microscopy, x-ray diffraction, and x-ray photoelectron spectroscopy of the materials after the corrosion tests, as well as by the post-corrosion analysis of the salts using inductively coupled plasma (ICP) and neutron activation analysis (NAA) techniques.

  20. A new form of MgTa2O6 obtained by the molten salt method

    Indian Academy of Sciences (India)

    Ashok K Ganguli; Shikha Nangia; A Meganathan Thirumala; Pratibha L Gai

    2006-01-01

    Using molten salt route (with NaCl/KCl as the salt) we have been able to synthesize a new form of magnesium tantalate at 850°C. Powder X-ray diffraction data could be indexed on an orthorhombic unit cell with lattice parameters, `' = 15.36(1) Å, ‘’ = 13.38(1) Å and ‘’ = 12.10(1) Å. High resolution transmission electron microscopy and electron diffraction studies confirm the results obtained by X-ray studies. Energy dispersive X-ray spectroscopy helps ascertain the composition of MgTa2O6. The title compound shows a dielectric constant of ∼ 24 with a low dielectric loss of 0.006 at 100 kHz at room temperature. Dielectric constant is nearly unchanged with rise in temperature while the loss shows a very marginal increase (0.007 at 300°C).

  1. Nuclear-grade zirconium prepared by combining combustion synthesis with molten-salt electrorefining technique

    Science.gov (United States)

    Li, Hui; Nersisyan, Hayk H.; Park, Kyung-Tae; Park, Sung-Bin; Kim, Jeong-Guk; Lee, Jeong-Min; Lee, Jong-Hyeon

    2011-06-01

    Zirconium has a low absorption cross-section for neutrons, which makes it an ideal material for use in nuclear reactor applications. However, hafnium typically contained in zirconium causes it to be far less useful for nuclear reactor materials because of its high neutron-absorbing properties. In the present study, a novel effective method has been developed for the production of hafnium-free zirconium. The process includes two main stages: magnesio-thermic reduction of ZrSiO 4 under a combustion mode, to produce zirconium silicide (ZrSi), and recovery of hafnium-free zirconium by molten-salt electrorefining. It was found that, depending on the electrorefining procedure, it is possible to produce zirconium powder with a low hafnium content: 70 ppm, determined by ICP-AES analysis.

  2. Nuclear-grade zirconium prepared by combining combustion synthesis with molten-salt electrorefining technique

    Energy Technology Data Exchange (ETDEWEB)

    Li, Hui [Graduate School of Green Energy Technology, Chungnam National University, Yuseong, Daejeon 305-764 (Korea, Republic of); Nersisyan, Hayk H. [Rapidly Solidified Materials Research Institute, Chungnam National University, Yuseong, Daejeon 305-764 (Korea, Republic of); Park, Kyung-Tae [Graduate School of Green Energy Technology, Chungnam National University, Yuseong, Daejeon 305-764 (Korea, Republic of); Park, Sung-Bin; Kim, Jeong-Guk [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Lee, Jeong-Min [Technical Research Laboratory, Poongsan Corporation, 2222-2 Sandae-ri, Angang-oup, Kyungju, Kyungbuk 780-775 (Korea, Republic of); Lee, Jong-Hyeon, E-mail: jonglee@cnu.ac.kr [Graduate School of Green Energy Technology, Chungnam National University, Yuseong, Daejeon 305-764 (Korea, Republic of)

    2011-06-15

    Zirconium has a low absorption cross-section for neutrons, which makes it an ideal material for use in nuclear reactor applications. However, hafnium typically contained in zirconium causes it to be far less useful for nuclear reactor materials because of its high neutron-absorbing properties. In the present study, a novel effective method has been developed for the production of hafnium-free zirconium. The process includes two main stages: magnesio-thermic reduction of ZrSiO{sub 4} under a combustion mode, to produce zirconium silicide (ZrSi), and recovery of hafnium-free zirconium by molten-salt electrorefining. It was found that, depending on the electrorefining procedure, it is possible to produce zirconium powder with a low hafnium content: 70 ppm, determined by ICP-AES analysis.

  3. Salts separation and removing method from material deposited on molten salt electrolyzing cathode

    International Nuclear Information System (INIS)

    Deposition materials on a cathode obtained by processing highly radioactive drainage discharged from spent fuel reprocessing steps and electrolyzing them in molten salts are incorporated with salts such as LiCl-KCl used as an electrolysis bath. Cadmium is added to the cathode deposition materials comprising lanthanoid and/or actinoid, and melted to form a molten material. The molten material are solidified by cooling to separate a metal portion and salts from the cathode deposition materials. The metal portion is kept at a temperature at which cadmium metal is evaporated to remove cadmium. Subsequently, the metal portion is kept at a temperature at which an intermetallic compound and/or an alloy of cadmium and lanthanoid and/or actinoid is decomposed to remove cadmium. Since salts can be removed efficiently from cathode deposition materials, aimed actinoid metals can be recovered at a high purity. (I.N.)

  4. Numerical Modelling of Induction Heating for a Molten Salts Pyrochemical Process

    International Nuclear Information System (INIS)

    Technological developments in the pyro-chemistry program are required to allow choices for a reprocessing experiment on 100 g of spent nuclear fuel. In this context, a special device must be designed for the solid/gas reaction phases followed by actinide extraction and stripping in molten salt. This paper discusses a modelling approach for designing an induction furnace. Using this numerical approach is a good way to improve thermal performance of the device in terms of magnetic/thermal coupling phenomena. The influence of current frequency is also studied to give another view of the possibilities of an induction furnace. Electromagnetic forces are taken into account in a computational fluid dynamics code derived from a specifically developed exchange library. Induction heating systems are an example of a typical multi-physics problem involving numerically coupled equations. (authors)

  5. Preliminary safety examination on thorium molten-salt nuclear energy synergetics

    International Nuclear Information System (INIS)

    The new global fission industry for the next century should keep a strong public acceptance, which means to ensure an enough rational safety feature not only in the engineering issue but also in all issues of integral fuel-cycle system. In this sense, the safety characteristics of the Thorium Molten-Salt Nuclear Energy Synergetic System (THORIMS-NES) is widely examined relating with the several aspects of environmental (including resources, radio-waste, etc.) social (including anti-nuclear proliferation and terrorism, etc), basic technological, engineering, institutional, and economical aspects. From this examination it seems that this system is verified as one of the most promising measures of North-South problem, greenhouse effect, etc in the world. (author). 11 refs., 3 figs., 5 tabs

  6. Molten-salt Synthesis and Properties of ZnS with Hexagonal Prism Morphology

    Institute of Scientific and Technical Information of China (English)

    LIU, Jin-Song; JI, Guang-Bin; LI, Zi-Quan; CAO, Jie-Ming; ZHENG, Ming-Bo; KE, Xing-Fei

    2007-01-01

    ZnS with hexagonal prism morphology has been synthesized successfully by molten-salt method with ZnS nanoparticles as precursors, and the ZnS nanoparticles were prepared by one-step solid-state reaction of Zn(CH3COO)2·2H2O with Na2S·9H2O at ambient temperature. Crystal structure and morphology of the product were characterized by X-ray diffraction, scanning electron microscopy, transmission electron microscopy and HRTEM. Ultraviolet-visible optical absorption spectrum of the ZnS hexagonal prism shows a distinct red shift from that of bulk ZnS crystals and photoluminescence spectrum exhibits strong emissions at 380 and 500 nm, respectively. Further experiments were designed and the formation mechanism of the ZnS hexagonal prism has been also discussed in brief.

  7. Characteristics of zirconia nanoparticles prepared by molten salts and microwave synthesis

    International Nuclear Information System (INIS)

    Zirconia and yttria stabilized zirconia (3YSZ) nanoparticles were prepared from zirconia and yttria salts using molten salts (MS) and microwave (MW) synthesis. The crystalline ZrO2 and 3YSZ nanoparticles with crystallite size in the range of 3–27 nm were obtained by MS synthesis in NaCl–NaNO3 salts. The zirconia and 3YSZ powders with close characteristics were obtained by combining MW synthesis with calcination of products at 400-800 °C. The crystallite size depends upon synthesis or calcination temperature, and the precursors used. The powders prepared by MS and MW synthesis ensured manufacturing of bulk materials with relative density of 98.6% and 97.2% respectively by using spark plasma sintering at 1300 °C

  8. Direct sensitivity analysis to nuclear data of thorium molten salt reactors at equilibrium using MURE

    International Nuclear Information System (INIS)

    In this paper we propose to use the MURE (MCNP Utility for Reactor Evolution) package to calculate directly nuclear data sensitivities of the breeding gain of a thorium molten salt reactor. The continuous fuel reprocessing is used to control reactivity. This control coupled with the fact that the fuel has reached its equilibrium, induces feedback effects on nuclear data sensitivities. That is why sensitivities are calculated directly by recalculating the fuel equilibrium of a simplified model after modifications of the ACE files for the reactions that are expected to have the strongest contribution to the uncertainty: 233U capture cross section and neutron fission yield. Due to recent improvements in its accuracy, 232Th capture cross-section is not the dominant contributor to the uncertainty. The total uncertainty on the breeding gain is expected to reach 2500 pcm

  9. Molten salt rolling bubble column, reactors utilizing same and related methods

    Science.gov (United States)

    Turner, Terry D.; Benefiel, Bradley C.; Bingham, Dennis N.; Klinger, Kerry M.; Wilding, Bruce M.

    2015-11-17

    Reactors for carrying out a chemical reaction, as well as related components, systems and methods are provided. In accordance with one embodiment, a reactor is provided that includes a furnace and a crucible positioned for heating by the furnace. The crucible may contain a molten salt bath. A downtube is disposed at least partially within the interior crucible along an axis. The downtube includes a conduit having a first end in communication with a carbon source and an outlet at a second end of the conduit for introducing the carbon material into the crucible. At least one opening is formed in the conduit between the first end and the second end to enable circulation of reaction components contained within the crucible through the conduit. An oxidizing material may be introduced through a bottom portion of the crucible in the form of gas bubbles to react with the other materials.

  10. Expedited demonstration of molten salt mixed waste treatment technology. Final report

    International Nuclear Information System (INIS)

    This final report discusses the molten salt mixed waste project in terms of the various subtasks established. Subtask 1: Carbon monoxide emissions; Establish a salt recycle schedule and/or a strategy for off-gas control for MWMF that keeps carbon monoxide emission below 100 ppm on an hourly averaged basis. Subtask 2: Salt melt viscosity; Experiments are conducted to determine salt viscosity as a function of ash composition, ash concentration, temperature, and time. Subtask 3: Determine that the amount of sodium carbonate entrained in the off-gas is minimal, and that any deposited salt can easily be removed form the piping using a soot blower or other means. Subtask 4: The provision of at least one final waste form that meets the waste acceptance criteria of a landfill that will take the waste. This report discusses the progress made in each of these areas

  11. Numerical Modelling of Induction Heating for a Molten Salts Pyrochemical Process

    Energy Technology Data Exchange (ETDEWEB)

    Vu, Xuan-Tuyen; Feraud, Jean-Pierre; Ode, Denis [CEA Marcoule: DTEC/SGCS/LGCI Bat. 57 B17171, 30207 Bagnols/Ceze (France); Du Terrail Couvat, Yves [SIMaP, Grenoble INP, CNRS: ENSEEG, BP 75, 38402 Saint Martin d' Heres Cedex (France)

    2008-07-01

    Technological developments in the pyro-chemistry program are required to allow choices for a reprocessing experiment on 100 g of spent nuclear fuel. In this context, a special device must be designed for the solid/gas reaction phases followed by actinide extraction and stripping in molten salt. This paper discusses a modelling approach for designing an induction furnace. Using this numerical approach is a good way to improve thermal performance of the device in terms of magnetic/thermal coupling phenomena. The influence of current frequency is also studied to give another view of the possibilities of an induction furnace. Electromagnetic forces are taken into account in a computational fluid dynamics code derived from a specifically developed exchange library. Induction heating systems are an example of a typical multi-physics problem involving numerically coupled equations. (authors)

  12. Electrochemical behavior of Cu in the (NaCl-KCl-CuCl)molten salt

    Institute of Scientific and Technical Information of China (English)

    Yungang LI; Jie LI; Kuai ZHANG; Limin LIU

    2011-01-01

    The electrochemical reaction mechanism and electrocrystallization process of Cu on copper electrode in the eutectic NaC1-KC1-CuC1 molten salt were investigated by means of cyclic voltammetry,chronopotentiometry and chronoamperometry technique at 710 ℃.The results show that the electrochemical reaction process of Cu is a quasi-reversible process mix-controlled by Cu+ diffusion rate and electron transport rate; the electrochemical reduction mechanism is Cu++e→Cu; the electrocrystallization process of copper is an instantaneous hemispheroid three-dimensional nucleation process; the Cu+ diffusion coefficient is 4.3×10-4 cm2·s-1 under the experimental conditions.

  13. HYLIFE-II: A molten-salt inertial fusion energy power plant design - final report

    International Nuclear Information System (INIS)

    Enhanced safety and performance improvements have been made to the liquid-wall HYLIFE reactor, yielding the current HYLIFE-II conceptual design. Liquid lithium has been replaced with a neutronically thick array of flowing molten-salt jets (Li2BeF4 or Flibe), which will not burn, has a low tritium solubility and inventory, and protects the chamber walls, giving a robust design with a 30-yr lifetime. The tritium inventory is 0.5 g in the molten salt and 140 g in the metal of the tube walls, where it is less easily released. The 5-MJ driver is a recirculating induction accelerator estimated to cost $570 million (direct costs). Heavy-ion targets yield 350 MJ, six times per second, to produce 940 MW of electrical power for a cost of 6.5 cents/kW·h. Both larger and smaller yields are possible with correspondingly lower and higher pulse rates. When scaled up to 1934 MW (electric), the plant design has a calculated cost of electricity of 4.5 cents/kW·h. The design did not take into account potential improved plant availability and lower operations and maintenance costs compared with conventional power plant experience, resulting from the liquid wall protection. Such improvements would directly lower the electricity cost figures. For example, if the availability can be raised from the conservatively assumed 75% to 85% and the annual cost of component replacement, operations, and maintenance can be reduced from 6% to 3% of direct cost, the cost of electricity would drop to 5.0 and 3.9 cents/kW·h for 1- and 2-GW (electric) cases. 50 refs., 15 figs., 3 tabs

  14. Thermal Properties of LiCl-KCl Molten Salt for Nuclear Waste Separation

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, Kumar [Univ. of Wisconsin, Madison, WI (United States); Allen, Todd [Univ. of Wisconsin, Madison, WI (United States); Anderson, Mark [Univ. of Wisconsin, Madison, WI (United States); Simpson, Mike [Idaho National Lab., (United States)

    2012-11-30

    This project addresses both practical and fundamental scientific issues of direct relevance to operational challenges of the molten LiCl-KCl salt pyrochemical process, while providing avenues for improvements in the process. In order to understand the effects of the continually changing composition of the molten salt bath during the process, the project team will systematically vary the concentrations of rare earth surrogate elements, lanthanum, cerium, praseodymium, and neodymium, which will be added to the molten LiCl-KCl salt. They will also perform a limited number of focused experiments by the dissolution of depleted uranium. All experiments will be performed at 500 deg C. The project consists of the following tasks. Researchers will measure density of the molten salts using an instrument specifically designed for this purpose, and will determine the melting points with a differential scanning calorimeter. Knowledge of these properties is essential for salt mass accounting and taking the necessary steps to prevent melt freezing. The team will use cyclic voltammetry studies to determine redox potentials of the rare earth cations, as well as their diffusion coefficients and activities in the molten LiCl-KCl salt. In addition, the team will perform anodic stripping voltammetry to determine the concentration of the rare earth elements and their solubilities, and to develop the scientific basis for an on-line diagnostic system for in situ monitoring of the cation species concentration (rare earths in this case). Solubility and activity of the cation species are critically important for the prediction of the salt's useful lifetime and disposal.

  15. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    International Nuclear Information System (INIS)

    Molten salt oxidation (MSO) is proposed as a 238Pu waste treatment technology that should be developed for volume reduction and recovery of 238Pu and as an alternative to the transport and permanent disposal of 238Pu waste to the WIPP repository. In MSO technology, molten sodium carbonate salt at 800--900 C in a reaction vessel acts as a reaction media for wastes. The waste material is destroyed when injected into the molten salt, creating harmless carbon dioxide and steam and a small amount of ash in the spent salt. The spent salt can be treated using aqueous separation methods to reuse the salt and to recover 99.9% of the precious 238Pu that was in the waste. Tests of MSO technology have shown that the volume of combustible TRU waste can be reduced by a factor of at least twenty. Using this factor the present inventory of 574 TRU drums of 238Pu contaminated wastes is reduced to 30 drums. Further 238Pu waste costs of $22 million are avoided from not having to repackage 312 of the 574 drums to a drum total of more than 4,600 drums. MSO combined with aqueous processing of salts will recover approximately 1.7 kilograms of precious 238Pu valued at 4 million dollars (at $2,500/gram). Thus, installation and use of MSO technology at LANL will result in significant cost savings compared to present plans to transport and dispose 238Pu TRU waste to the WIPP site. Using a total net present value cost for the MSO project as $4.09 million over a five-year lifetime, the project can pay for itself after either recovery of 1.6 kg of Pu or through volume reduction of 818 drums or a combination of the two. These savings show a positive return on investment

  16. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    Energy Technology Data Exchange (ETDEWEB)

    Wishau, R.

    1998-05-01

    Molten salt oxidation (MSO) is proposed as a {sup 238}Pu waste treatment technology that should be developed for volume reduction and recovery of {sup 238}Pu and as an alternative to the transport and permanent disposal of {sup 238}Pu waste to the WIPP repository. In MSO technology, molten sodium carbonate salt at 800--900 C in a reaction vessel acts as a reaction media for wastes. The waste material is destroyed when injected into the molten salt, creating harmless carbon dioxide and steam and a small amount of ash in the spent salt. The spent salt can be treated using aqueous separation methods to reuse the salt and to recover 99.9% of the precious {sup 238}Pu that was in the waste. Tests of MSO technology have shown that the volume of combustible TRU waste can be reduced by a factor of at least twenty. Using this factor the present inventory of 574 TRU drums of {sup 238}Pu contaminated wastes is reduced to 30 drums. Further {sup 238}Pu waste costs of $22 million are avoided from not having to repackage 312 of the 574 drums to a drum total of more than 4,600 drums. MSO combined with aqueous processing of salts will recover approximately 1.7 kilograms of precious {sup 238}Pu valued at 4 million dollars (at $2,500/gram). Thus, installation and use of MSO technology at LANL will result in significant cost savings compared to present plans to transport and dispose {sup 238}Pu TRU waste to the WIPP site. Using a total net present value cost for the MSO project as $4.09 million over a five-year lifetime, the project can pay for itself after either recovery of 1.6 kg of Pu or through volume reduction of 818 drums or a combination of the two. These savings show a positive return on investment.

  17. Experimental and theoretical studies in Molten Salt Natural Circulation Loop (MSNCL)

    International Nuclear Information System (INIS)

    High Temperature Reactors (HTR) and solar thermal power plants use molten salt as a coolant, as it has low melting point and high boiling point, enabling us to operate the system at low pressure. Molten fluoride salt and molten nitrate salt are proposed as a candidate coolant for High Temperature Reactors (HTR) and solar power plant respectively. BARC is developing a 600 MWth pebble bed high temperature reactor, cooled by natural circulation of fluoride salt and capable of supplying process heat at 1000°C to facilitate hydrogen production by splitting water. Beside this, BARC is also developing a 2MWe solar power tower system using molten nitrate salt. With these requirements, a Molten Salt Natural Circulation Loop (MSNCL) has been designed, fabricated, installed and commissioned in Hall-7, BARC for thermal hydraulic, instrumentation development and material compatibility related studies. Steady state natural circulation experiments with molten nitrate salt (mixture of NaNO3 and KNO3 in 60:40 ratio) have been carried out in the loop at different power level. Various transients viz. startup of natural circulation, step power change, loss of heat sink and heater trip has also been studied in the loop. A well known steady state correlation given by Vijayan et. al. has been compared with experimental data. In-house developed code LeBENC has also been validated against all steady state and transient experimental results. The detailed description of MSNCL, steady state and transient experimental results and validation of in-house developed code LeBENC have been described in this report. (author)

  18. Molten salt synthesis of sodium lithium titanium oxide anode material for lithium ion batteries

    Energy Technology Data Exchange (ETDEWEB)

    Yin, S.Y., E-mail: yshy2004@hotmail.com [College of Environmental and Biological Engineering, Wuhan Technology and Business University, Wuhan 430065 (China); Feng, C.Q. [Hubei Collaborative Innovation Center for Advanced Organic Chemical Materials, Ministry of Education Key Laboratory for Synthesis and Applications of Organic Functional Molecules, Hubei University, Wuhan 430062 (China); Wu, S.J.; Liu, H.L.; Ke, B.Q. [College of Environmental and Biological Engineering, Wuhan Technology and Business University, Wuhan 430065 (China); Zhang, K.L. [College of Chemistry and Molecular Sciences, Wuhan University, Wuhan 430072 (China); Chen, D.H. [College of Environmental and Biological Engineering, Wuhan Technology and Business University, Wuhan 430065 (China); Hubei Key Laboratory for Catalysis and Material Science, College of Chemistry and Material Science, South Central University for Nationalities, Wuhan 430074, Hubei (China)

    2015-09-05

    Highlights: • Na{sub 2}Li{sub 2}Ti{sub 6}O{sub 12} has been successfully synthesized via a molten salt route. • Calcination temperature is an important effect on the component and microstructure of the product. • Pure phase Na{sub 2}Li{sub 2}Ti{sub 6}O{sub 12} could be obtained at 700 °C for 2 h. - Abstract: The sodium lithium titanium oxide with composition Na{sub 2}Li{sub 2}Ti{sub 6}O{sub 14} has been synthesized by a molten salt synthesis method using sodium chloride and potassium chloride mixture as a flux medium. Synthetic variables on the synthesis, such as sintering temperature, sintering time and the amount of lithium carbonate, were intensively investigated. Powder X-ray diffraction and scanning electron microscopy images of the reaction products indicates that pure phase sodium lithium titanium oxide has been obtained at 700 °C, and impure phase sodium hexatitanate with whiskers produced at higher temperature due to lithium evaporative losses. The results of cyclic voltammetry and discharge–charge tests demonstrate that the synthesized products prepared at various temperatures exhibited electrochemical diversities due to the difference of the components. And the sample obtained at 700 °C revealed highly reversible insertion and extraction of Li{sup +} and displayed a single potential plateau at around 1.3 V. The product obtained at 700 °C for 2 h exhibits good cycling properties and retains the specific capacity of 62 mAh g{sup −1} after 500 cycles.

  19. Modular Stellarator Fusion Reactor (MSR) concept

    International Nuclear Information System (INIS)

    A preliminary conceptual study has been made of the Modulator Stellarator Reactor (MSR) as a stedy-state, ignited, DT-fueled, magnetic fusion reactor. The MSR concept combines the physics of classic stellarator confinement with an innovative, modular-coil design. Parametric tradeoff calculations are described, leading to the selection of an interim design point for a 4.8-GWt plant based on Alcator transport scaling and an average beta value of 0.04 in an l = 2 system with a plasma aspect ratio of 11. Neither an economic analysis nor a detailed conceptual engineering design is presented here, as the primary intent of this scoping study is the elucidation of key physics tradeoffs, constraints, and uncertainties for the ultimate power-reactor embodiment

  20. Design of a natural draft air-cooled condenser and its heat transfer characteristics in the passive residual heat removal system for 10 MW molten salt reactor experiment

    International Nuclear Information System (INIS)

    As one of the Generation IV reactors, Molten Salt Reactor (MSR) has its superiorities in satisfying the requirements on safety. In order to improve its inherent safety, a concept of passive residual heat removal system (PRHRS) for the 10 MW Molten Salt Reactor Experiment (MSRE) was put forward, which mainly consisted of a fuel drain tank, a feed water tank and a natural draft air-cooled condenser (NDACC). Besides, several valves and pipes are also included in the PRHRS. A NDACC for the PRHRS was preliminarily designed in this paper, which contained a finned tube bundle and a chimney. The tube bundle was installed at the bottom of the chimney for increasing the velocity of the air across the bundle. The heat transfer characteristics of the NDACC were investigated by developing a model of the PRHRS using C++ code. The effects of the environmental temperature, finned tube number and chimney height on heat removal capacity of the NDACC were analyzed. The results show that it has sufficient heat removal capacity to meet the requirements of the residual heat removal for MSRE. The effects of these three factors are obvious. With the decay heat reducing, the heat dissipation power declines after a short-time rise in the beginning. The operation of the NDACC is completely automatic without the need of any external power, resulting in a high safety and reliability of the reactor, especially once the accident of power lost occurs to the power plant. - Highlights: • A model to study the heat transfer characteristics of the NDACC was developed. • The NDACC had sufficient heat removal capacity to remove the decay heat of MSRE. • NDACC heat dissipation power depends on outside temperature and condenser geometry. • As time grown, the effects of outside temperature and condenser geometry diminish. • The NDACC could automatically adjust its heat removal capacity

  1. Study on optical and thermal performance of a linear Fresnel solar reflector using molten salt as HTF with MCRT and FVM methods

    International Nuclear Information System (INIS)

    Highlights: • A LFR which employs CPC, evacuated tubes and uses molten salt as HTF is designed. • 3D optical and thermal models are developed with MCRT and FVM methods. • The optical and thermal performance, the effects of key parameters are studied. • The instantaneous optical efficiency of 65.0% is achieved at normal incidence. • The collector efficiencies are above 46.0% under all the studied conditions. - Abstract: A novel linear Fresnel reflector which employs the evacuated tube, CPC secondary reflector, and uses molten salt as the heat transfer fluid (HTF) was designed and studied in this paper. A 3D optical model was developed to simulate the radiation transmission within the system with Monte Carlo Ray Tracing (MCRT) method. Based on the model, firstly, the optical performance of the systems using cylindrical and parabolic mirrors was compared. Then the local solar flux distribution on the absorber surface and the optical efficiency were computed. Then the effects of the slope error, time and location, etc. were investigated. Finally, the thermal performance was investigated by coupling the MCRT with the Finite Volume Method (FVM). The optical simulation results indicate that the system with optimized cylindrical mirrors can achieve nearly the same performance as the one with parabolic mirrors. The solar flux distribution on the absorber exhibits a non-uniform characteristic which can be improved by using mirrors with proper slope error. The instantaneous optical efficiency of 65.0% at normal incidence and the annual mean optical efficiency which ranges between 55.2% and 34.8% from the equator to N50° can be achieved. The numerical results indicate that the temperature profiles on the absorber follow the non-uniform solar flux. The collector efficiencies are all above 46.0% under the studied conditions. Both the thermal efficiency and the collector efficiency increase with decreasing salt temperature and with increasing radiation. These results

  2. A furnace and environmental cell for the in situ investigation of molten salt electrolysis using high-energy X-ray diffraction.

    Science.gov (United States)

    Styles, Mark J; Rowles, Matthew R; Madsen, Ian C; McGregor, Katherine; Urban, Andrew J; Snook, Graeme A; Scarlett, Nicola V Y; Riley, Daniel P

    2012-01-01

    This paper describes the design, construction and implementation of a relatively large controlled-atmosphere cell and furnace arrangement. The purpose of this equipment is to facilitate the in situ characterization of materials used in molten salt electrowinning cells, using high-energy X-ray scattering techniques such as synchrotron-based energy-dispersive X-ray diffraction. The applicability of this equipment is demonstrated by quantitative measurements of the phase composition of a model inert anode material, which were taken during an in situ study of an operational Fray-Farthing-Chen Cambridge electrowinning cell, featuring molten CaCl(2) as the electrolyte. The feasibility of adapting the cell design to investigate materials in other high-temperature environments is also discussed. PMID:22186642

  3. An Electrochemical Study of Lanthanide Elements in LiCl-KCl Eutectic Molten Salt to Convert All The Spent Nuclear Fuel into Low and Intermediate Level Waste

    International Nuclear Information System (INIS)

    An additional unit step for the residual actinide recovery, designated as 'Pyro-Reisodex', was proposed to convert all the spent nuclear waste into low and intermediated level water by achieving high decontamination factor for TRH elements. The measurement of basic material properties of lanthanide elements in LiCl-KCl eutectic molten salt is necessary to evaluate the performance of the step. Thus, standard potential, activity coefficient, and diffusion coefficient of lanthanide elements is being tried to determine using conventional electrochemical methods. The cycle voltammetry was measured for LiCl-KCl-SmCl3 mixture and the standard potential, activity coefficient, and diffusion coefficient of this system was determined from the voltammogram data. The calculated data was well-agreed with reference. Based on this results, another techniques for other lanthanide elements will be applied for better understanding of LiCl-KCl-LnCln system

  4. NMR insights on the properties of ZnCl2 molten salt hydrate medium through its interaction with SnCl4 and fructose

    DEFF Research Database (Denmark)

    Qiao, Yan; Pedersen, Christian Marcus; Wang, Yingxiong;

    2014-01-01

    formation during the biomass conversion was evaluated. The ion complex composed by Sn4+ and Zn2+ indicated that there is a synergic catalytic effect between these two Lewis acid ions. 13C NMR spectra of fructose in different ZnCl2 molten salt hydrate concentrations revealed that the concentration of β......The solvent properties of ZnCl2 molten salt medium and its synergic effect with the Lewis acid catalyst, Sn4+, for biomass conversion, were investigated by nuclear magnetic resonance. The tautomeric distribution of fructose in the ZnCl2 molten salt medium was examined, and its effect for humins...

  5. New design for CSP plant with direct-steam solar receiver and molten-salt storage

    Science.gov (United States)

    Ganany, Alon; Hadad, Itay

    2016-05-01

    This paper presents the evolution of BrightSource's Concentrated Solar Power (CSP) technology - from a solar steam generator (SRSG) with no Thermal Energy Storage (TES) to SRSG with TES to Extended-cycle TES. The paper discusses SRSG with TES technology, and the capabilities of this solution are compared with those of an MSR plant.

  6. Engineering Evaluation of Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiment for the Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    This evaluation was performed by Pro2Serve in accordance with the Technical Specification for an Engineering Evaluation of the Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory (BJC 2009b). The evaluators reviewed the Engineering Evaluation Work Plan for Molten Salt Reactor Experiment Residual Salt Removal, Oak Ridge National Laboratory, Oak Ridge, Tennessee (DOE 2008). The Work Plan (DOE 2008) involves installing a salt transfer probe and new drain line into the Fuel Drain Tanks and Fuel Flush Tank and connecting them to the new salt transfer line at the drain tank cell shield. The probe is to be inserted through the tank ball valve and the molten salt to the bottom of the tank. The tank would then be pressurized through the Reactive Gas Removal System to force the salt into the salt canisters. The Evaluation Team reviewed the work plan, interviewed site personnel, reviewed numerous documents on the Molten Salt Reactor (Sects. 7 and 8), and inspected the probes planned to be used for the transfer. Based on several concerns identified during this review, the team recommends not proceeding with the salt transfer via the proposed alternate salt transfer method. The major concerns identified during this evaluation are: (1) Structural integrity of the tanks - The main concern is with the corrosion that occurred during the fluorination phase of the uranium removal process. This may also apply to the salt transfer line for the Fuel Flush Tank. Corrosion Associated with Fluorination in the Oak Ridge National Laboratory Fluoride Volatility Process (Litman 1961) shows that this problem is significant. (2) Continued generation of Fluorine - Although the generation of Fluorine will be at a lower rate than experienced before the uranium removal, it will continue to be generated. This needs to be taken into consideration regardless of what actions are taken with the salt. (3) More than one phase of material

  7. New operating strategies for molten salt in line focusing solar fields - Daily drainage and solar receiver preheating

    Science.gov (United States)

    Eickhoff, Martin; Meyer-Grünefeldt, Mirko; Keller, Lothar

    2016-05-01

    Nowadays molten salt is efficiently used in point concentrating solar thermal power plants. Line focusing systems still have the disadvantage of elevated heat losses at night because of active freeze protection of the solar field piping system. In order to achieve an efficient operation of line focusing solar power plants using molten salt, a new plant design and a novel operating strategy is developed for Linear Fresnel- and Parabolic Trough power plants. Daily vespertine drainage of the solar field piping and daily matutinal refilling of the solar preheated absorber tubes eliminate the need of nocturnal heating of the solar field and reduce nocturnal heat losses to a minimum. The feasibility of this new operating strategy with all its sub-steps has been demonstrated experimentally.

  8. Preparation of SiC/SiO2 core-shell nanowires via molten salt mediated carbothermal reduction route

    Science.gov (United States)

    Zhang, Ju; Yan, Shuai; Jia, Quanli; Huang, Juntong; Lin, Liangxu; Zhang, Shaowei

    2016-06-01

    The growth of silicon carbide (SiC) crystal generally requires a high temperature, especially when low quality industrial wastes are used as the starting raw materials. In this work, SiC/SiO2 core-shell nanowires (NWs) were synthesized from low cost silica fume and sucrose via a molten salt mediated carbothermal reduction (CR) route. The molten salt was found to be effective in promoting the SiC growth and lowering the synthesis temperature. The resultant NWs exhibited a heterostructure composed of a 3C-SiC core of 100 nm in diameter and a 5-10 nm thick amorphous SiO2 shell layer. The photoluminescence spectrum of the achieved SiC NWs displayed a significant blue shift (a dominant luminescence at round 422 nm), which suggested that they were high quality and could be a promising candidate material for future optoelectronic applications.

  9. Surface Morphology and Microstructure of Zinc Deposit From Imidazole with Zinc Chloride Low Temperature Molten Salt Electrolyte in The Presence of Aluminium Chloride

    Directory of Open Access Journals (Sweden)

    Shanmugasigamani Srinivasan, M. Selvam

    2013-07-01

    Full Text Available Low temperature molten salts have variety of applications in organic synthesis, catalytic processing, batteries and electrode position due to their air and water stability. They have wide potential window for their applications in voltage and temperature and hence there is a possibility to deposit metals which could not be deposited from aqueous electrolytes. Our aim and scope of our research was to deposit zinc from low temperature molten salt electrolyte (LTMS containing zinc salt in the presence of aluminium chloride at different current densities and to qualify the nature of deposits. We could identify the effect of current density on the deposit at low temperature molten salt electrolyte by analysing the nature of deposits using different instrumental techniques. Compact, adherent, dense fine grained deposits of zinc with average grain size of 40-150 nm could be obtained from low temperature molten salt electrolyte. (LTMS

  10. The construction, development and application of potential simulation models to the filling of carbon nanotubes by molten salts

    OpenAIRE

    Bishop, C. L.

    2009-01-01

    Inorganic nanotube structures (INTs) can be synthesised through the direct �filling of carbon nanotube templates with molten salts. The resulting structures, usually rationalised in terms of known bulk crystal structures, are shown to be contained within a general set of structures classifi�ed in terms of folded sheets of in�finite squares and hexagons. A flexible model for the carbon nanotube is employed (using a Terso� II potential), a signifi�cant development on previous work in w...

  11. Effects assessment of 10 functioning years on the main components of the molten salt PCS experimental facility of ENEA

    Science.gov (United States)

    Gaggioli, Walter; Di Ascenzi, Primo; Rinaldi, Luca; Tarquini, Pietro; Fabrizi, Fabrizio

    2016-05-01

    In the frame of the Solar Thermodynamic Laboratory, ENEA has improved CSP Parabolic Trough technologies by adopting new advanced solutions for linear tube receivers and by implementing a binary mixture of molten salt (60% NaNO3 and 40% KNO3) [1] as both heat transfer fluid and heat storage medium in solar field and in storage tanks, thus allowing the solar plants to operate at high temperatures up to 550°C. Further improvements have regarded parabolic mirror collectors, piping and process instrumentation. All the innovative components developed by ENEA, together with other standard parts of the plant, have been tested and qualified under actual solar operating conditions on the PCS experimental facility at the ENEA Casaccia Research Center in Rome (Italy). The PCS (Prova Collettori Solari, i.e. Test of Solar Collectors) facility is the main testing loop built by ENEA and it is unique in the world for what concerns the high operating temperature and the fluid used (mixture of molten salt). It consists in one line of parabolic trough collectors (test section of 100 m long life-size solar collectors) using, as heat transfer fluid, the aforesaid binary mixture of molten salt up to 10 bar, at high temperature in the range 270° and 550°C and a flow rate up to 6.5 kg/s. It has been working since early 2004 [2] till now; it consists in a unique closed loop, and it is totally instrumented. In this paper the effects of over ten years qualification tests on the pressurized tank will be presented, together with the characterization of the thermal losses of the piping of the molten salt circuit, and some observations performed on the PCS facility during its first ten years of operation.

  12. The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nevinitsa, V. A., E-mail: Neviniza-VA@nrcki.ru; Dudnikov, A. A.; Blandinskiy, V. Yu.; Balanin, A. L.; Alekseev, P. N. [National Research Centre Kurchatov Institute (Russian Federation); Titarenko, Yu. E.; Batyaev, V. F.; Pavlov, K. V.; Titarenko, A. Yu., E-mail: yuri.titarenko@itep.ru [Institute for Theoretical and Experimental Physics (Russian Federation)

    2015-12-15

    A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing {sup 233}U from {sup 232}Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

  13. An experimental study on the flow and heat transfer of flinak molten salt in small channels for the application to the VHTR intermediate heat exchanger

    International Nuclear Information System (INIS)

    To make the design of the Very High Temperature Reactor (VHTR) complete and plausible, the designs of the Intermediate Heat Transport Loop (IHTL) as well as the Intermediate Heat Exchanger (IHX) are known to be one of the difficult engineering tasks due to its high temperature operating condition (up to 950degC). In this study, the Flinak molten salt, an eutectic mixture of LiF, NaF and KF (46.5:11.5:42.0 mole %) is considered as the heat transporting fluid in the IHTL. To evaluate the flow and heat transfer performance of the Flinak molten salt in small channels of millimeter-range hydraulic diameters, a double-pipe type heat exchanger was constructed using small-diameter tubes for the heat exchange between the Flinak and gas flow. The inner diameters of the inner tube and the outer tube are 1.4 mm and 4.6 mm, respectively, and the length of the tubes is 500 mm. The molten salt flows through the inner tube. The molten salt is prepared in a crucible made of Inconel 600 placed in an electric furnace. The molten salt flow is produced by differential pressure between a twin set of molten salt crucibles without using a mechanical pump. The flow rate of the molten salt is reduced from the weight change of a crucible measured by load cells. Temperatures of the two heat exchanging fluids at various points as well as pressure drop across the test tube are measured to obtain flow and heat transfer characteristics of the molten salt flow. For laminar flow of the Flinak in 1.4 mm inner-diameter circular tube, the measured friction factors were smaller than the 64/Re curve by 50%. Also the measured Nusselt numbers were generally in the range between 3.66 and 4.36, although the data were scattered due to the measurement error in such a high temperature condition. (author)

  14. Separation and recovery of uranium ore by chlorinating, chelate resin and molten salt treatment

    Energy Technology Data Exchange (ETDEWEB)

    Taki, Tomohiro [Japan Nuclear Cycle Development Inst., Kamisaibara, Okayama (Japan). Ningyo Toge Environmental Engineering Center

    2000-12-01

    Three fundamental researches of separation and recovery of uranium from uranium ore are reported in this paper. Three methods used the chloride pyrometallurgy, sodium containing molten salts and chelate resin. When uranium ore is mixed with activated carbon and reacted for one hour under the mixed gas of chlorine and oxygen at 950 C, more than 90% uranium volatilized and vaporization of aluminum, silicone and phosphorus were controlled. The best activated carbon was brown coal because it was able to control the large range of oxygen concentration. By blowing oxygen into the molten sodium hydroxide, the elution rate of uranium attained to about 95% and a few percent of uranium was remained in the residue. On the uranium ore of unconformity-related uranium deposits, a separation method of uranium, molybdenum, nickel and phosphorus from the sulfuric acid elusion solution with U, Ni, As, Mo, Fe and Al was developed. Methylene phosphonic acid type chelate resin (RCSP) adsorbed Mo and U, and then 100 % Mo was eluted by sodium acetate solution and about 100% U by sodium carbonate solution. Ni and As in the passing solution were recovered by imino-diacetic acid type chelate resin and iron hydroxide, respectively. (S.Y.)

  15. Electrochemistry of LiCl-Li2O-H2O Molten Salt Systems

    Energy Technology Data Exchange (ETDEWEB)

    Natalie J. Gese; Batric Pesic

    2013-03-01

    Uranium can be recovered from uranium oxide (UO2) spent fuel through the combination of the oxide reduction and electrorefining processes. During oxide reduction, the spent fuel is introduced to molten LiCl-Li2O salt at 650 degrees C and the UO2 is reduced to uranium metal via two routes: (1) electrochemically, and (2) chemically by lithium metal (Li0) that is produced electrochemically. However, the hygroscopic nature of both LiCl and Li2O leads to the formation of LiOH, contributing hydroxyl anions (OH-), the reduction of which interferes with the Li0 generation required for the chemical reduction of UO2. In order for the oxide reduction process to be an effective method for the treatment of uranium oxide fuel, the role of moisture in the LiCl-Li2O system must be understood. The behavior of moisture in the LiCl-Li2O molten salt system was studied using cyclic voltammetry, chronopotentiometry and chronoamperometry, while reduction to hydrogen was confirmed with gas chromatography.

  16. Stabilization/Solidification of radioactive molten salt waste via gel-route pretreatment.

    Science.gov (United States)

    Park, Hwan-Seo; Kim, In-Tae; Kim, Hwan-Young; Ryu, Seung-Kon; Kim, Joon-Hyung

    2007-02-15

    The volatilization of radionuclides during the stabilization/solidification of radioactive wastes at high temperatures is one of the major problems to be considered in choosing suitable wasteforms, process, material systems, etc. This paper reports a novel method to convert volatile wastes into nonvolatile compounds via a sol-gel process, which is different from the conventional method using metal-alkoxides and organic solvents. The material system was designed with sodium silicate (Si) as a gelling agent, phosphoric acid (P) as a catalyst/stabilizer, aluminum nitrate (Al) as a property promoter, and H20 as a solvent. A novel structural model for the chemical conversion of molten salt waste, named RPRM (Reaction Product in Reaction Module), was established, and the waste could be solidified with glass matrix via a simple procedure. The leached fraction of Cs and Sr by a PCT leaching method was 0.72% and 0.014%, respectively. In conclusion, the RPRM model isto converttargetwastes into stable and manageable products, not to obtain a specific crystalline product for each radionuclide. This paper suggested a new stabilization/solidification method for salt wastes by establishing the gel-forming material system and showing a practical example, not a new synthesis method of stable crystalline phase. This process, named "gel-route stabilization/solidification (GRSS)", will be a prospective alternative with stable chemical process on the immobilization of salt wastes and various mixed radioactive waste for final disposal. PMID:17593740

  17. Separation and recovery of uranium ore by chlorinating, chelate resin and molten salt treatment

    International Nuclear Information System (INIS)

    Three fundamental researches of separation and recovery of uranium from uranium ore are reported in this paper. Three methods used the chloride pyrometallurgy, sodium containing molten salts and chelate resin. When uranium ore is mixed with activated carbon and reacted for one hour under the mixed gas of chlorine and oxygen at 950 C, more than 90% uranium volatilized and vaporization of aluminum, silicone and phosphorus were controlled. The best activated carbon was brown coal because it was able to control the large range of oxygen concentration. By blowing oxygen into the molten sodium hydroxide, the elution rate of uranium attained to about 95% and a few percent of uranium was remained in the residue. On the uranium ore of unconformity-related uranium deposits, a separation method of uranium, molybdenum, nickel and phosphorus from the sulfuric acid elusion solution with U, Ni, As, Mo, Fe and Al was developed. Methylene phosphonic acid type chelate resin (RCSP) adsorbed Mo and U, and then 100 % Mo was eluted by sodium acetate solution and about 100% U by sodium carbonate solution. Ni and As in the passing solution were recovered by imino-diacetic acid type chelate resin and iron hydroxide, respectively. (S.Y.)

  18. A molten Salt Am242M Production Reactor for Space Applications

    Science.gov (United States)

    Emrich, William

    2005-01-01

    The use of Am242m holds great promise for increasing the efficiency nuclear thermal rocket engines. Because Am242m has the highest fission cross section of any known isotope (1000's of barns), its extremely high reactivity may be used to directly heat a propellant gas with fission fragments. Since this isotope does not occur naturally, it must be bred in special production reactors designed for that purpose. The primary advantage to using molten salt reactors for breeding Am242m is that the reactors can be reprocessed continually yielding a constant rate of production of the isotope. Once built and initially fueled, the reactor will continually breed the additional fuel it needs to remain critical. The only feedstock required is a salt of U238. No enriched fuel is required during normal operation and all fissile material, except the Am242m, is maintained in a closed loop. For a reactor operating at 200 MW several kilograms of Am242m may be bred each year.

  19. A final report on the Phase 1 testing of a molten-salt cavity receiver

    Energy Technology Data Exchange (ETDEWEB)

    Chavez, J M [ed.; Smith, D C [Babcock and Wilcox Co., Barberton, OH (United States). Nuclear Equipment Div.

    1992-05-01

    This report describes the design, construction, and testing of a solar central receiver using molten nitrate salt as a heat exchange fluid. Design studies for large commercial plants (30--100 MWe) have shown molten salt to be an excellent fluid for solar thermal plants as it allows for efficient thermal storage. Plant design studies concluded that an advanced receiver test was required to address uncertainties not covered in prior receiver tests. This recommendation led to the current test program managed by Sandia National Laboratories for the US Department of Energy. The 4.5 MWt receiver is installed at Sandia National Laboratories' Central Receiver Test Facility in Albuquerque, New Mexico. The receiver incorporates features of large commercial receiver designs. This report describes the receiver's configuration, heat absorption surface (design and sizing), the structure and supporting systems, and the methods for control. The receiver was solar tested during a six-month period at the Central Receiver Test Facility in Albuquerque, NM. The purpose of the testing was to characterize the operational capabilities of the receiver under a number of solar operating and stand-by conditions. This testing consisted of initial check-out of the systems, followed by steady-state performance, transient receiver operation, receiver operation in clouds, receiver thermal loss testing, receiver start-up operation, and overnight thermal conditioning tests. This report describes the design, fabrication, and results of testing of the receiver.

  20. Molten salt pyrolysis of milled beech wood using an electrostatic precipitator for oil collection

    Directory of Open Access Journals (Sweden)

    Heidi S. Nygård

    2015-07-01

    Full Text Available A tubular electrostatic precipitator (ESP was designed and tested for collection of pyrolysis oil in molten salt pyrolysis of milled beech wood (0.5-2 mm. The voltage-current (V-I characteristics were studied, showing most stable performance of the ESP when N2 was utilized as inert gas. The pyrolysis experiments were carried out in FLiNaK and (LiNaK2CO3 over the temperature range of 450-600 ℃. The highest yields of pyrolysis oil were achieved in FLiNaK, with a maximum of 34.2 wt% at 500 ℃, followed by a decrease with increasing reactor temperature. The temperature had nearly no effect on the oil yield for pyrolysis in (LiNaK2CO3 (19.0-22.5 wt%. Possible hydration reactions and formation of HF gas during FLiNaK pyrolysis were investigated by simulations (HSC Chemistry software and measurements of the outlet gas (FTIR, but no significant amounts of HF were detected.

  1. A descriptive model of the molten salt reactor experiment after shutdown: Review of FY 1995 progress

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D.F.; Del Cul, G.D.; Toth, L.M.

    1996-01-01

    During FY 1995 considerable progress was made toward gaining a better understanding of the chemistry and transport processes that continue to govern the behavior of the Molten Salt Reactor Experiment (MSRE). As measurements in the MSRE proceed, laboratory studies continue, and better analyses are available, our understanding of the state of the MSRE and the best path toward remediation improves. Because of the immediate concern about the deposit in the auxiliary charcoal bed (ACB), laboratory studies in the past year focused on carbon-fluorine chemistry. Secondary efforts were directed toward investigation of gas generation from MSRE salts by both radiolytic and nonradiolytic pathways. In addition to the laboratory studies, field measurements at the MSRE provided the basis for estimating the inventory of uranium and fluorine in the ACB. Analysis of both temperature and radiation measurements provided independent and consistent estimates of about 2.6 kg of uranium deposited in the top of the ACB. Further analysis efforts included a refinement in the estimates of the fuel- salt source term, the deposited decay energy, and the projected rate of radiolytic gas generation. This report also provides the background material necessary to explain new developments and to review areas of particular interest. The detailed history of the MSRE is extensively documented and is cited where appropriate. This work is also intended to update and complement the more recent MSRE assessment reports.

  2. Thermal analysis to support decommissioning of the molten salt reactor experiment

    International Nuclear Information System (INIS)

    As part of the decommissioning process for the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory, several thermal-sciences issues were addressed. Apparently a mixture of UF6 and F2 had diffused into the upper portion of one charcoal column in the MSRE auxiliary charcoal bed (ACB), leading to radiative decay heating and possible chemical reaction sources. A proposed interim corrective action was planned to remove the water from the ACB cell to reduce criticality and reactivity concerns and then fill the ACB cell with an inert material. This report describes design of a thermocouple probe to obtain temperature measurements for mapping the uranium deposit, as well as development of steady-state and transient numerical models for the heat transfer inside the charcoal column. Additional numerical modeling was done to support filling of the ACB cell. Results from this work were used to develop procedures for meeting the goals of the MSRE Remediation Project without exceeding appropriate thermal limits

  3. Customer interface document for the Molten Salt Test Loop (MSTL) system.

    Energy Technology Data Exchange (ETDEWEB)

    Pettit, Kathleen; Kolb, William J.; Gill, David Dennis; Briggs, Ronald D.

    2012-03-01

    The National Solar Thermal Test Facility at Sandia National Laboratories has a unique test capability called the Molten Salt Test Loop (MSTL) system. MSTL is a test capability that allows customers and researchers to test components in flowing, molten nitrate salt. The components tested can range from materials samples, to individual components such as flex hoses, ball joints, and valves, up to full solar collecting systems such as central receiver panels, parabolic troughs, or linear Fresnel systems. MSTL provides realistic conditions similar to a portion of a concentrating solar power facility. The facility currently uses 60/40 nitrate 'solar salt' and can circulate the salt at pressure up to 600psi, temperature to 585 C, and flow rate of 400-600GPM depending on temperature. The purpose of this document is to provide a basis for customers to evaluate the applicability to their testing needs, and to provide an outline of expectations for conducting testing on MSTL. The document can serve as the basis for testing agreements including Work for Others (WFO) and Cooperative Research and Development Agreements (CRADA). While this document provides the basis for these agreements and describes some of the requirements for testing using MSTL and on the site at Sandia, the document is not sufficient by itself as a test agreement. The document, however, does provide customers with a uniform set of information to begin the test planning process.

  4. Reactivity control capability of fuel-salt processing system in a molten-salt breeder reactor

    International Nuclear Information System (INIS)

    An evaluation is made of the reactivity control capability of the fuel processing system (FPS) in a molten-salt breeder reactor. The principal functions required of the FPS are: (a) Isolation of 233Pa from regions of high neutron flux during its decay to 233U, and (b) the removal of fission products from the system. The FPS can very usefully serve also to control the primary system reactivity by appropriately utilizing its function of extracting uranium and reconstituting the fuel contained in the salt. The principles of operation are quite similar to the chemical shim control system currently installed in PWR's whereby the core reactivity, affected by changes in the moderator temperature, fuel burnup and transient Xe, is adjusted by regulating the concentration of boric acid introducted in the moderator as neutron absorber. The present study examines the capability of the FPS to follow transient Xe as in PWR's, and proves that the FPS should effectively serve as a system for adjusting not only long-term changes in reactivity but also short-term transient variations without any accompanying difficulties foreseen in operation. (auth.)

  5. Molten-salt reactors for efficient nuclear fuel utilization without plutonium separation

    International Nuclear Information System (INIS)

    Research and development studies of molten-salt reactors (MSRs) for special purposes have been under way since 1947 and for possible application as possible commercial nuclear electric power generators since 1956. For the latter, the previous emphasis has been on breeding performance and low fissile inventory to help limit the demand on nonrenewable natural resources (uranium) in an expanding nuclear economy; little or no thought has been given to alternative uses of nuclear fuels such as proliferation of nuclear explosives. As a consequence, the conceptual designs that evolved (e.g., the ORNL reference design MSBR) all favored enriched 233U as fuel with an on-site chemical processing facility from which portions of that fuel could be diverted fairly easily. With the current interest in limiting the proliferation potential of nuclear electric power systems, a redirected study of MSRs was undertaken in an effort to identify conceptual systems that would be attractive in this situation. It appears that practical proliferation-resistant MSRs could be designed and built, and the report describes a particularly attractive break-even breeder that includes an on-site chemical reprocessing facility within the reactor primary containment

  6. Engineering development studies for molten-salt breeder reactor processing No. 18

    International Nuclear Information System (INIS)

    A water--mercury system was used to study the effect of geometric variations on mass transfer rates in rectangular contractors similar to those proposed for the molten-salt breeder reactor (MSBR) fuel reprocessing scheme. Since mass transfer rates were not accurately predicted by the Lewis correlation, other correlations were investigated. A correlation which was found to fit the experimental results is given. Mass transfer rates are being measured in a fluoride salt--bismuth contactor. Experimental results indicate that the mass transfer rates in the salt--bismuth system fall between the Lewis correlation and the modified correlation given above. Autoresistance heating tests were continued in the fluorinator mock-up using LiF--BeF2--ThF4 (72-16-12 mole percent) salt. The equipment was returned to operating condition, and five experiments were run. Although correct steady-state operation was not achieved, the results were encouraging. A two-dimensional electrical analog was constructed to study current flow through the electrode sidearm and other critical areas of the test vessel. These studies indicate that no regions of abnormally high current density existed in the first nine runs with the present autoresistance heating equipment. Localized heating had previously been the suspected cause for the failure to achieve proper operation of this equipment. (U.S.)

  7. Conceptual design characteristics of a denatured molten-salt reactor with once-through fueling

    Energy Technology Data Exchange (ETDEWEB)

    Engel, J.R.; Bauman, H.F.; Dearing, J.F.; Grimes, W.R.; McCoy, H.E.; Rhoades, W.A.

    1980-07-01

    A study was made to examine the conceptual feasibility of a molten-salt power reactor fueled with denatured /sup 235/U and operated with a minimum of chemical processing. Because such a reactor would not have a positive breeding gain, reductions in the fuel conversion ratio were allowed in the design to achieve other potentially favorable characteristics for the reactor. A conceptual core design was developed in which the power density was low enough to allow a 30-year life expectancy of the moderator graphite with a fluence limit of 3 x 10/sup 26/ neutrons/m/sup 2/ (E > 50 keV). This reactor could be made critical with about 3450 kg of 20% enriched /sup 235/U and operated for 30 years with routine additions of denatured /sup 235/U and no chemical processing for removal of fission products. A review of the chemical considerations assoicated with the conceptual fuel cycle indicates that no substantial difficulties would be expected if the soluble fission products and higher actinides were allowed to remain in the fuel salt for the life of the plant.

  8. Tungsten coating prepared on molybdenum substrate by electrodeposition from molten salt in air atmosphere

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Fan; Zhang, Yingchun, E-mail: zycustb@163.com; Sun, Ningbo; Leng, Jiaxun

    2015-02-01

    Highlights: • Tungsten coatings were electroplated on molybdenum substrate for the first time. • The electrodeposition was studied in the air atmosphere. • The coating has columnar structure with preferential growth orientation of (1 1 0). • The columnar structure was disappeared after high-temperature annealing. • The coating has an extremely low oxygen content with the value of 0.032 wt%. - Abstract: Compact and smooth tungsten coating on molybdenum substrate was obtained by electrodeposition from Na{sub 2}WO{sub 4}–WO{sub 3} molten salt at 1173 K in atmosphere. Microstructure, morphology and properties were performed on the tungsten coating. The tungsten coating had columnar structure with the preferential growth orientation of (2 0 0). There was about 2 μm thick diffusion layer of tungsten in the molybdenum substrate. The bending test and thermal shock test showed the tungsten coating had good adhesion with the molybdenum substrate. The microhardness of the coating was about 492 HV and the oxygen content of the coating was 0.032 wt%. The high-temperature could enhance the high-temperature oxidation resistance and bond strength of the tungsten coating.

  9. Polarization effects in the molten salt electrorefining of spent nuclear fuel

    International Nuclear Information System (INIS)

    The effects of polarization in the electrorefining of spent nuclear fuel were analytically studied through electrotransport experiments with uranium. When the uranium concentration in molten salt was set at 0.5 wt%, polarization caused the measured cell resistance to increase from 0.16 to 0.33 Ω as the cell voltage was raised from 0.1 to 0.7 V. At 2.0 wt% uranium concentration in salt, on the other hand, the resistance was almost independent of cell voltage. A code named DEVON has been developed to estimate the effects of polarization on the electrorefiner operating condition. The Laplace equation is solved in the bulk salt region by finite element method. In the calculation, the effects of polarization are taken into account by adopting the diffusion layer model on defining the boundary conditions. The result of calculations agreed well with the measured resistance data at two sample concentrations for a diffusion layer thickness of 0.025 cm on the solid cathode. The calculations indicated that significant polarization at the cathode could be avoided by maintaining the uranium concentration in salt above 1 to 2 wt%. When it was held at 5 wt%, which is a typical level for normal operation, polarization proved to exert little influence on the cell resistance, but it was indicated to contribute appreciably toward flattering the current distribution along the cathode surface. (author)

  10. On the proliferation issues of a fusion fission fuel factory using a molten salt

    International Nuclear Information System (INIS)

    The fusion fission fuel factory (FFFF) is a hybrid fusion fission reactor using a neutron source, which is in this case taken similar to the source of the Power Plant Conceptual Study - Water Cooled Lithium Lead (PPCS-A) design, for fissile material production instead of tritium self-sufficiency. As breeding blanket the first wall of the ITER design is attached to a molten salt zone, in which ThF4 and UF4 solute salts are transported by a LiF-BeF2 solvent salt. For this blanket design, the fissile material is assessed in quantity and quality for both the Th-U and the U-Pu fuel cycle. The transport of the initial D-T fusion neutrons and the reaction rates in this breeding blanket are simulated with the Monte Carlo code MCNP4c2. The isotopic evolution of the actinides is calculated with the burn-up code ORIGEN-S. For the Th-U cycle the bred material output remains below 10 g/h with a 232U impurity level of 30 ppm, while for the U-Pu cycle supergrade material is produced at a rate up to 100 g/h.

  11. A preliminary assessment of salt and radionuclide volatilities in the molten salt processor

    International Nuclear Information System (INIS)

    We have applied thermodynamic methods to analyze the volatilities of plutonium (Pu), uranium (U), iodine (I), and salts from molten salt processor baths in bench-scale runs conducted by Rockwell International on oxidation of laboratory mixed waste. Good agreement is obtained between calculated and observed volatilities. The volatilizing species of Pu and U have been identified as gaseous PuO2(OH)2 and UO2(OH)2, both with volatilities that are strongly dependent upon temperature, chemical activity of condensed phase Pu and U, and steam pressure. Hence, for example, orders of magnitude reduction can be achieved in Pu and U volatilities by operating the bath at lower temperatures. Iodine is found to volatilize from the bath as gaseous sodium iodide (NaI) but a small portion of the NaI is converted to gaseous I2 during cooldown of the off-gas. Observations by Rockwell of trace amounts of sodium chloride (NaCl) particulates in the off-gas as a result of NaCl volatilization from the bath are confirmed by our calculations. 24 refs., 1 fig., 16 tabs

  12. Electrodeposition of metallic tungsten coating from binary oxide molten salt on low activation steel substrate

    International Nuclear Information System (INIS)

    Tungsten is considered a promising plasma facing armor material for future fusion devices. An electrodeposited metallic tungsten coating from Na2WO4–WO3 binary oxide molten salt on low activation steel (LAS) substrate was investigated in this paper. Tungsten coatings were deposited under various pulsed currents conditions at 1173 K in atmosphere. Cathodic current density and pulsed duty cycle were investigated for pulsed current electrolysis. The crystal structure and microstructure of tungsten coatings were characterized by X-ray diffractometry, scanning electron microscopy, and energy X-ray dispersive analysis techniques. The results indicated that pulsed current density and duty cycle significantly influence tungsten nucleation and electro-crystallization phenomena. The average grain size of the coating becomes much larger with increasing cathodic current density, which demonstrates that appropriate high cathodic current density can accelerate the growth of grains on the surface of the substrate. The micro-hardness of tungsten coatings increases with the increasing thickness of coatings; the maximum micro-hardness is 482 HV. The prepared tungsten coatings have a smooth surface, a porosity of less than 1%, and an oxygen content of 0.024 wt%

  13. A descriptive model of the molten salt reactor experiment after shutdown: Review of FY 1995 progress

    International Nuclear Information System (INIS)

    During FY 1995 considerable progress was made toward gaining a better understanding of the chemistry and transport processes that continue to govern the behavior of the Molten Salt Reactor Experiment (MSRE). As measurements in the MSRE proceed, laboratory studies continue, and better analyses are available, our understanding of the state of the MSRE and the best path toward remediation improves. Because of the immediate concern about the deposit in the auxiliary charcoal bed (ACB), laboratory studies in the past year focused on carbon-fluorine chemistry. Secondary efforts were directed toward investigation of gas generation from MSRE salts by both radiolytic and nonradiolytic pathways. In addition to the laboratory studies, field measurements at the MSRE provided the basis for estimating the inventory of uranium and fluorine in the ACB. Analysis of both temperature and radiation measurements provided independent and consistent estimates of about 2.6 kg of uranium deposited in the top of the ACB. Further analysis efforts included a refinement in the estimates of the fuel- salt source term, the deposited decay energy, and the projected rate of radiolytic gas generation. This report also provides the background material necessary to explain new developments and to review areas of particular interest. The detailed history of the MSRE is extensively documented and is cited where appropriate. This work is also intended to update and complement the more recent MSRE assessment reports

  14. Direct Conversion of Greenhouse Gas CO2 into Graphene via Molten Salts Electrolysis.

    Science.gov (United States)

    Hu, Liwen; Song, Yang; Jiao, Shuqiang; Liu, Yingjun; Ge, Jianbang; Jiao, Handong; Zhu, Jun; Wang, Junxiang; Zhu, Hongmin; Fray, Derek J

    2016-03-21

    Producing graphene through the electrochemical reduction of CO2 remains a great challenge, which requires precise control of the reaction kinetics, such as diffusivities of multiple ions, solubility of various gases, and the nucleation/growth of carbon on a surface. Here, graphene was successfully created from the greenhouse gas CO2 using molten salts. The results showed that CO2 could be effectively fixed by oxygen ions in CaCl2-NaCl-CaO melts to form carbonate ions, and subsequently electrochemically split into graphene on a stainless steel cathode; O2 gas was produced at the RuO2-TiO2 inert anode. The formation of graphene in this manner can be ascribed to the catalysis of active Fe, Ni, and Cu atoms at the surface of the cathode and the microexplosion effect through evolution of CO in between graphite layers. This finding may lead to a new generation of proceedures for the synthesis of high value-added products from CO2, which may also contribute to the establishment of a low-carbon and sustainable world. PMID:26871684

  15. A study of metallic coatings obtained by electrolysis of molten salts

    International Nuclear Information System (INIS)

    An appropriate technique has been developed for obtaining compact metallic coatings from electrolysis of molten salts. Through the use of this method, it has been possible to produce pure metal deposits which, until now, has been extremely difficult to do. The apparatus used and the main steps of the process such as dehydration of the solvant, degassing of the equipment, and starting of the electrolytic process, are first described. This is followed by a discussion of the deposits of the metals beryllium, uranium, tantalum and tungsten obtained from electrolysis of molten fluorides at temperatures between 600 and 8000C. The metal coatings so obtained are homogeneous and show continuity, their thicknesses varying from a few microns to a millimeter or more. They have been studied by measurements. As potential applications of this new technique, one can mention the growth of diffusion barriers and the production of cathodes for thermoionic emission. The method can also be used for electroforming. An intermetallic diffusion between the deposit and the substrate has been observed in some cases. The advantage of the technique of melt electrolysis in obtaining metal coatings of enhanced thicknesses is illustrated by taking the beryllium-nickel system as an example. It is shown that the thickness obtained is proportional to the square root of growth time and is about 6 to 8 times larger than that obtained by conventional techniques

  16. Estimating steady state and transient characteristics of molten salt natural circulation loop using CFD

    International Nuclear Information System (INIS)

    The steady state and transient characteristics of a molten salt natural circulation loop (NCL) are obtained by 3D CFD simulations. The working fluid is a mixture of NaNO3 and KNO3 in 60:40 ratio. Simulation is performed using PHOENICS CFD software. The computational domain is discretized by a body fitted grid generated using in-built mesh generator. The CFD model includes primary side. Primary side fluid is subjected to heat addition in heater section, heat loss to ambient (in piping connecting heater and cooler) and to secondary side (in cooler section). Reynolds Averaged Navier Stokes equations are solved along with the standard k-ε turbulence model. Validation of the model is done by comparing the computed steady state Reynolds number with that predicted by various correlations proposed previously. Transient simulations were carried out to study the flow initiations transients for different heater powers and different configurations. Similarly the ''power raising'' transient is computed and compared with in-house experimental data. It is found that, using detailed information obtained from 3D transient CFD simulations, it is possible to understand the physics of oscillatory flow patterns obtained in the loop under certain conditions.

  17. TiB2 coating formed on nickel substrates by electroplating in molten salt of fluoride

    Institute of Scientific and Technical Information of China (English)

    LONG Jin-ming; GUO Zhong-cheng; HAN Xia-yun

    2004-01-01

    The TiB2 coatings deposited over nickel substrate by electroplating was investigated, which is in molten salt of a fluoride mixture involving KF, NaF, K2 TiF6 and KBF4. Effects of temperature, cathodic current density (Jc) and duration on the coating's formation were examined. The composition, morphology and structure of the coatings were characterized by scanning electron microscopy (SEM), energy dispersive X-ray detector (EDS) and X-ray diffraction (XRD). The results show that the coatings, with black, smooth and uniform appearance, are composed of predominating TiB2 and small amounts of nickel titanium oxide (Ni0.75 Ti0.125 O). The coatings show a nodular morphology and the grain size is dependent on the Jc and ranges about 1 - 10 μm. There is a linear relationship between the coating's thickness and the time of electrolysis within certain duration range. The reduction of the potassium can take place simultaneously with the electrochemical synthesis of TiB2 as the Jc is in excess of certain level. The hardness of the TiB2 coatings is likely to be deteriorated due to the presence of potassium and Ni0.75Ti0.125 O in the coatings.

  18. Trial Destruction Test of Spent Cationic Resins in a Molten Salt Oxidation Reactor System

    International Nuclear Information System (INIS)

    The spent ion-exchange resins have to be disposed of and as such, spent ion-exchange resins are a significant fraction of the combustible organic waste from the nuclear industries. One effective treatment option is incinerating the spent resins to yield ash and gas. However, there are difficulties associated with this approach. One of the criticisms of a high-temperature incinerator is that radioactive and hazardous metals are not retained in the incinerator. In addition, incineration of the cationic exchange resins, which have the sulphurcontaining functional groups of sulfonic acid (-SO3-H+), has revealed significant problems associated with sulfur dioxide (SO2), a primary air pollutant, which must be kept under control. There is therefore the developing need for an alternative destruction process. Molten salt oxidation, or MSO for short, is a promising alternative technology. Molten carbonate filled in a MSO reactor is capable of trapping sulfur during organic destruction. In addition, the relatively low-operation temperature of the MSO reactor reduces the volatility of the radionuclides, compared to the other available high temperature technologies for organics destruction, such as inductively coupled plasma, incineration, plasma arc and microwave heating. Trial destruction tests of spent cationic exchange resins doped with radioactive metal surrogates were performed in this study. Two typical operating parameters, temperature and oxidizing air rate, which significantly affect the organics destruction, were tested to establish the optimum ranges for those parameters

  19. Molten Salt-Carbon Nanotube Thermal Energy Storage for Concentrating Solar Power Systems Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Michael Schuller; Frank Little; Darren Malik; Matt Betts; Qian Shao; Jun Luo; Wan Zhong; Sandhya Shankar; Ashwin Padmanaban

    2012-03-30

    We demonstrated that adding nanoparticles to a molten salt would increase its utility as a thermal energy storage medium for a concentrating solar power system. Specifically, we demonstrated that we could increase the specific heat of nitrate and carbonate salts containing 1% or less of alumina nanoparticles. We fabricated the composite materials using both evaporative and air drying methods. We tested several thermophysical properties of the composite materials, including the specific heat, thermal conductivity, latent heat, and melting point. We also assessed the stability of the composite material with repeated thermal cycling and the effects of adding the nanoparticles on the corrosion of stainless steel by the composite salt. Our results indicate that stable, repeatable 25-50% improvements in specific heat are possible for these materials. We found that using these composite salts as the thermal energy storage material for a concentrating solar thermal power system can reduce the levelized cost of electricity by 10-20%. We conclude that these materials are worth further development and inclusion in future concentrating solar power systems.

  20. Observation of Fluorescence for some lanthanides in LiCl-KCl molten salt media at high temperature

    International Nuclear Information System (INIS)

    To our knowledge, the fluorescence studies of lanthanides in LiCl-KCl eutectic molten salt at a high temperature are not reported yet. The fluorescence of the lanthanide-ions was generally decreased when the temperature was increased. Moreover, the fluorescence of the lanthanides was strong when the sample was solidified to or from the melt. The temperature, where the fluorescence was decreased, was identified to be different depending on the species of the lanthanides and the substrates was considered to possibly be from quenching of the fluorescence due to either the collisions of melted samples induced by high temperature media or the re-absorption of fluorescence by the samples. Several comparison experiments were performed to explain and understand this phenomenon and improve the fluorescence. In this way, an on-line monitoring of chemical species and the concentration for lanthanides elements in molten salt media of pyrochemical process can be accomplished, and it can replace the currently used trouble for determination of the concentration in molten salt such as a destructive ICP-AES method during process

  1. Comparative study of dielectric properties of MgNb2O6 prepared by molten salt and ceramic method

    Indian Academy of Sciences (India)

    Vishnu Shanker; Ashok K Ganguli

    2003-12-01

    Magnesium niobate (MgNb2O6) powder was synthesized by the conventional ceramic route as well as by the molten salt route using a eutectic mixture of NaCl–KCl as the salt and Mg(NO3)$_2\\cdot$6H2O and TiO2 as the starting materials. Pure phase of MgNb2O6 could be obtained by the molten salt method at 1100°C. However, in ceramic method the pure phase of MgNb2O6 was obtained by heating at 1025°C for 20 h. On sintering at 1100°C the dielectric constant and dielectric loss of MgNb2O6 obtained by the molten salt method was found to be 19.5 and 0.004 at 100 kHz at room temperature. Lower values were obtained for these oxides prepared by the ceramic route, 16.6 and 0.000518, respectively. In both cases the dielectric constant was quite stable with frequency.

  2. The chemistry of molten salt mixtures: application to the reductive extraction of lanthanides and actinides by a liquid metal

    International Nuclear Information System (INIS)

    The design of a process of An/Ln separation by liquid - liquid extraction can be used for on-line purification of the molten salt in a molten salt nuclear reactor (Generation IV) as well as reprocessing various spent fuels. In order to establish the chemical properties of An and Ln in molten salt mediums, E - pO2 - diagrams were established for the relevant chemical elements. With the purpose of checking the possibilities of separating the An from Ln, the real activity coefficients in liquid metals were measured. An experimental protocol was developed and validated on the Gd/Ga system. It was then transferred to radioactive environment to measure the activity coefficient of Pu in Ga. The results made it possible to estimate the effectiveness of the Pu extraction and its separation from Gd and Ce. The selectivity was shown to decrease with the temperature and Al and Ga showed a good selectivity between Pu and the Ce in fluoride medium. (author)

  3. 液态熔盐堆运行安全特性初步研究%Preliminary Study on Safety Characteristics of Molten Salt Reactor

    Institute of Scientific and Technical Information of China (English)

    魏泉; 梅龙伟; 战志超; 郭威; 陈金根; 蔡翔舟

    2014-01-01

    Compared with solid fuel reactors ,there are differences in physics for liquid fuel reactor .As for molten salt reactor (MSR) ,due to fuel flow in primary loop ,the delayed neutron precursors and fission product partly decay out of core , resulting in reactivity loss as well as heat generation in the primary loop .In this paper ,the critical dynamics and safety characteristics of MSR were investigated using Cinsf 1D code .Con‐sidering the loss of delayed neutrons under different fuel flow speeds at zero‐power ,the corresponding control rod positions were calculated under pump start and stop condi‐tions .Keeping reactor power at 2 MW ,the temperature and power were computed for the primary loop system . Finally , the pump stop accident was simulated from rated power 2 MW . After pump stop , the reactor power increases slightly due to the reduction of delayed neutron loss at initial time and then it decreases to approach the decay heat power level quickly .The temperature in core increases slowly and reaches to a balance within safety range .It can be concluded that MSR has intrinsic safety .%液态燃料反应堆与固态燃料反应堆相比,原理上有较大不同。液态熔盐堆中由于燃料流动带走缓发中子先驱核在堆外衰变导致堆芯反应性降低,且裂变产物在堆外回路中衰变也会引起一回路发热。本文使用熔盐堆中子动力学程序Cinsf1D探讨2 M W熔盐堆的临界动力学特性和安全特性,研究零功率临界下不同熔盐流速启泵和停泵导致的缓发中子先驱核流失所需改变的控制棒棒位。同时还计算了2 MW恒定功率情况下稳态运行及降低流速时一回路温度分布,并模拟了2 MW额定功率下停泵事件。停泵后由于缓发中子损失减少反应堆功率先缓慢增加,然后迅速降低到接近余热水平。停泵后堆芯温度缓慢增加后稳定在安全值以内,说明熔盐堆具有本征安全性。

  4. Thorium cycle and molten salt reactors: field parameters and field constraints investigations toward 'thorium molten salt reactor' definition; Cycle thorium et reacteurs a sel fondu: exploration du champ des parametres et des contraintes definissant le 'Thorium Molten Salt Reactor'

    Energy Technology Data Exchange (ETDEWEB)

    Mathieu, L

    2005-09-15

    Producing nuclear energy in order to reduce the anthropic CO{sub 2} emission requires major technological advances. Nuclear plants of 4. generation have to respond to several constraints, as safety improvements, fuel breeding and radioactive waste minimization. For this purpose, it seems promising to use Thorium Cycle in Molten Salt Reactors. Studies on this domain have already been carried out. However, the final concept suffered from serious issues and was discontinued. A new reflection on this topic is being led in order to find acceptable solutions, and to design the Thorium Molten Salt Reactor concept. A nuclear reactor is simulated by the coupling of a neutron transport code with a materials evolution code. This allows us to reproduce the reactor behavior and its evolution all along its operation. Thanks to this method, we have studied a large number of reactor configurations. We have evaluated their efficiency through a group of constraints they have to satisfy. This work leads us to a better understanding of many physical phenomena controlling the reactor behavior. As a consequence, several efficient configurations have been discovered, allowing the emergence of new points of view in the research of Molten Salt Reactors. (author)

  5. Challenges Related to the Use of Liquid Metal and Molten Salt Coolants in Advanced Reactors. Report of the collaborative project COOL of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was launched in 2000, based on a resolution by the IAEA General Conference (GC(44)/RES/21). INPRO aims at helping to ensure that nuclear energy is available in the twenty-first century in a sustainable manner, and seeks to bring together all interested Member States, both technology holders and technology users, to jointly consider actions to achieve desired innovations. INPRO is taking care of the specific needs of developing countries. One of the aims of INPRO is to develop options for enhanced sustainability through promotion of technical and institutional innovations in nuclear energy technology through collaborative projects among IAEA Member States. Collaboration among INPRO members is fostered on selected innovative nuclear technologies to bridge technology gaps. Collaborative projects have been selected so that they complement other national and international R and D activities. The INPRO Collaborative Project COOL on Investigation of Technological Challenges Related to the Removal of Heat by Liquid Metal and Molten Salt Coolants from Reactor Cores Operating at High Temperatures investigated the technological challenges of cooling reactor cores that operate at high temperatures in advanced fast reactors, high temperature reactors and accelerator driven systems by using liquid metals and molten salts as coolants. The project was initiated in 2008 and was led by India; experts from Brazil, China, Germany, India, Italy and the Republic of Korea participated and provided chapters of this report. The INPRO Collaborative Project COOL addressed the following fields of research regarding liquid metal and molten salt coolants: (i) survey of thermophysical properties; (ii) experimental investigations and computational fluid dynamics studies on thermohydraulics, specifically pressure drop and heat transfer under different operating conditions; (iii) monitoring and control of coolant

  6. Electrochemical decomposition of SiO2 pellets to form silicon in molten salts

    International Nuclear Information System (INIS)

    Research highlights: → Increasing temperature or decreasing SiO2 particle size increased the reduction rate. → Addition of NaCl to the electrolyte decreased the reduction rate. → Complete reduction of SiO2 pellets was achieved after 16 h of electrolysis. → The brown color of silicon may be the result of nanometer-scale crystallite size. → Higher purity of Si may be obtained by advancement of cell component materials. - Abstract: Direct electrochemical reduction of porous SiO2 pellets in molten CaCl2 salt and CaCl2-NaCl salt mixture was investigated by applying 2.8 V potential. The study focused on the effects of temperature, particle size of SiO2 powder starting material and the behavior of cathode contacting materials during electrochemical reduction process. The starting materials and the electrolysis products were characterized by X-ray diffraction analysis and scanning electron microscopy mainly. The studies showed that smaller particle sizes and higher temperatures had slightly positive effects in increasing the reduction rate within the ranges covered in this study. The results were interpreted from variations of current and accumulative electrical charge that passed through the cell as a function of duration of electrochemical reduction under different conditions. Microstructures and compositions of the reduced pellets were used to infer that electrochemical reduction of SiO2 in molten salts may become a method to produce silicon that could be used in solar energy utilization. Furthermore, X-ray diffraction analysis results indicated that the silicon produced at the cathode reacts with contacting materials; nickel, and iron in stainless steel to form Ni-Si and Fe-Si compounds due to very reactive nature of silicon especially at high temperatures.

  7. Electrochemical decomposition of SiO{sub 2} pellets to form silicon in molten salts

    Energy Technology Data Exchange (ETDEWEB)

    Erguel, Emre [Middle East Technical University, Engineering Faculty, Department of Metallurgical and Materials Engineering, Inoenue Bulvari, 06531 Ankara (Turkey); Karakaya, Ishak, E-mail: kkaya@metu.edu.tr [Middle East Technical University, Engineering Faculty, Department of Metallurgical and Materials Engineering, Inoenue Bulvari, 06531 Ankara (Turkey); Erdogan, Metehan [Middle East Technical University, Engineering Faculty, Department of Metallurgical and Materials Engineering, Inoenue Bulvari, 06531 Ankara (Turkey)

    2011-01-21

    Research highlights: > Increasing temperature or decreasing SiO{sub 2} particle size increased the reduction rate. > Addition of NaCl to the electrolyte decreased the reduction rate. > Complete reduction of SiO{sub 2} pellets was achieved after 16 h of electrolysis. > The brown color of silicon may be the result of nanometer-scale crystallite size. > Higher purity of Si may be obtained by advancement of cell component materials. - Abstract: Direct electrochemical reduction of porous SiO{sub 2} pellets in molten CaCl{sub 2} salt and CaCl{sub 2}-NaCl salt mixture was investigated by applying 2.8 V potential. The study focused on the effects of temperature, particle size of SiO{sub 2} powder starting material and the behavior of cathode contacting materials during electrochemical reduction process. The starting materials and the electrolysis products were characterized by X-ray diffraction analysis and scanning electron microscopy mainly. The studies showed that smaller particle sizes and higher temperatures had slightly positive effects in increasing the reduction rate within the ranges covered in this study. The results were interpreted from variations of current and accumulative electrical charge that passed through the cell as a function of duration of electrochemical reduction under different conditions. Microstructures and compositions of the reduced pellets were used to infer that electrochemical reduction of SiO{sub 2} in molten salts may become a method to produce silicon that could be used in solar energy utilization. Furthermore, X-ray diffraction analysis results indicated that the silicon produced at the cathode reacts with contacting materials; nickel, and iron in stainless steel to form Ni-Si and Fe-Si compounds due to very reactive nature of silicon especially at high temperatures.

  8. Molten salt coal gasification process development unit. Phase 1. Volume 1. PDU operations. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Kohl, A.L.

    1980-05-01

    This report summarizes the results of a test program conducted on the Molten Salt Coal Gasification Process, which included the design, construction, and operation of a Process Development Unit. In this process, coal is gasified by contacting it with air in a turbulent pool of molten sodium carbonate. Sulfur and ash are retained in the melt, and a small stream is continuously removed from the gasifier for regeneration of sodium carbonate, removal of sulfur, and disposal of the ash. The process can handle a wide variety of feed materials, including highly caking coals, and produces a gas relatively free from tars and other impurities. The gasification step is carried out at approximately 1800/sup 0/F. The PDU was designed to process 1 ton per hour of coal at pressures up to 20 atm. It is a completely integrated facility including systems for feeding solids to the gasifier, regenerating sodium carbonate for reuse, and removing sulfur and ash in forms suitable for disposal. Five extended test runs were made. The observed product gas composition was quite close to that predicted on the basis of earlier small-scale tests and thermodynamic considerations. All plant systems were operated in an integrated manner during one of the runs. The principal problem encountered during the five test runs was maintaining a continuous flow of melt from the gasifier to the quench tank. Test data and discussions regarding plant equipment and process performance are presented. The program also included a commercial plant study which showed the process to be attractive for use in a combined-cycle, electric power plant. The report is presented in two volumes, Volume 1, PDU Operations, and Volume 2, Commercial Plant Study.

  9. Electromigration in molten salts and application to isotopic separation of alkaline and alkaline-earth elements

    International Nuclear Information System (INIS)

    The separation of the isotopes of the alkaline-earth elements has been studied using counter-current electromigration in molten bromides. The conditions under which the cathode operates as a bromine electrode for the highest possible currents have been examined. For the separation of calcium, it has been necessary to use a stable CaBr2 - (CaBr2 + KBr) 'chain'. In the case of barium and strontium, it was possible to employ the pure bromides. Enrichment factors of the order of 10 for 48Ca and of the order of 1.5 for the rare isotopes of barium and strontium have been obtained. In the case of magnesium the method is slightly more difficult to apply because of material loss due to the relatively high vapour pressure of the salt requiring the use of electrolyte chains, MgBr2 - CeBr3. A study has been made that has led to a larger-scale application of the method. These are essentially the inhibition of reversible operation of the cathode by traces of water, limiting the intensity which can be tolerated; evacuation of the heat produced by the Joule effect, in the absence of which the separation efficiency is reduced by thermal gradients; corrosion of the materials by molten salts at high temperature. Several cells capable of treating a few kilograms of substance have been put into operation; none of these has lasted long enough to produce a satisfactory enrichment. The method is thus limited actually to yields of the order of a few grams. (author)

  10. Development of system analysis code for pyrochemical process using molten salt electrorefining

    International Nuclear Information System (INIS)

    This report describes accomplishment of development of a cathode processor calculation code to simulate the mass and heat transfer phenomena with the distillation process and development of an analytical model for cooling behavior of the pyrochemical process cell on personal computers. The pyrochemical process using molten salt electrorefining would introduce new technologies for new fuels of particle oxide, particle nitride and metallic fuels. The cathode processor calculation code with distillation process was developed. A code validation calculation has been conducted on the basic of the benchmark problem for natural convection in a square cavity. Results by using the present code agreed well for the velocity-temperature fields, the maximum velocity and its location with the benchmark solution published in a paper. The functions have been added to advance the reality in simulation and to increase the efficiency in utilization. The test run has been conducted using the code with the above modification for an axisymmetric enclosed vessel simulating a cathode processor, and the capability of the distillation process simulation with the code has been confirmed. An analytical model for cooling behavior of the pyrochemical process cell was developed. The analytical model was selected by comparing benchmark analysis with detailed analysis on engineering workstation. Flow and temperature distributions were confirmed by the result of steady state analysis. In the result of transient cooling analysis, an initial transient peak of temperature occurred at balanced heat condition in the steady-state analysis. Final gas temperature distribution was dependent on gas circulation flow in transient condition. Then there were different final gas temperature distributions on the basis of the result of steady-state analysis. This phenomenon has a potential for it's own metastable condition. Therefore it was necessary to design gas cooling flow pattern without cooling gas circulation

  11. Synthesis of LaMO3 (M = Fe, Co, Ni) using nitrate or nitrite molten salts

    International Nuclear Information System (INIS)

    Research highlights: → The molten salt, synthesis temperature, the oxidising properties, the basicity and oxidation states of the melt play an important role in LaMO3 synthesis. → The system needs a careful optimisation of these parameters to obtain pure orthorhombic LaFeO3, hexagonal LaCoO3 and LaNiO3. Purely hexagonal LaNiO3 nanocrystals were successfully synthesized. → SEM analyses were taken to investigate how the different parameters affect the microstructure of the samples. → The specific surface area and crystallite size of LaMO3 show much difference in different system. - Abstract: Perovskite compound LaMO3 (M = Fe, Co, Ni) nanocrystals were successfully synthesized in molten nitrates or nitrites from a mixture of lanthanum nitrate and an M-containing nitrate for 2 h. The effect of the various process parameters on the phase purity, crystallite size, specific surface area and morphology of the synthesized nanocrystals were systematically studied by XRD, scanning electron microscopy (SEM), simultaneous TG/DSC and BET measurements. The results showed that salt medium, annealing temperatures (mainly 450-850 oC), oxidising properties and basicity of the melt played an important role in the synthesis of LaMO3. The addition of Na2O2 facilitated the reaction between La2O3 and NiO or Co3O4, leading to the formation of LaNiO3 and LaCoO3 at a much lower temperature of 450 oC. Pure hexagonal LaNiO3 nanocrystals were obtained in molten NaNO3-KNO3 eutectic with Na2O2 at 550-750 oC.

  12. Lessons learnt during the design, construction and start-up phase of a molten salt testing facility

    International Nuclear Information System (INIS)

    In 2010, CIEMAT (Centro de investigaciones energéticas medioambientales y tecnológicas) signed a turn-key contract to have an experimental plant for thermal storage using molten salts at its PSA (Plataforma Solar de Almeria) facilities. This plant was designed to evaluate components, instrumentation and operation strategies and to give support to the industry in the qualification and evaluation of components. During the design, construction and start-up phases of this plant, many different aspects regarding design, construction and commissioning have been learnt and these will contribute to the improvement of other plants. Among other tips explained in the paper, we recommend the use of venting valves to eliminate the water present in the system after the pressure test or released by the salts during the first melting. The selection of instrumentation with no electronic components near a heat source, thus preventing them from overheating, is also advisable. The heat exchanger design and dimensioning should take into account not only the thermal losses to the atmosphere and through pipes and supports, but any possible reduction in the heat exchange surface that could have detrimental consequences in the thermal performance. Special attention must be paid when dimensioning and installing the EHT and insulation because both components are decisive in the avoidance of plug formation. Its correct installation in valves and supports and the proper positioning of the temperature control sensors, i.e. where no other heat source can distort the readings, are crucial. Recommendations and strategies for the operation and shutdown of this experimental plant are being gathered for a future paper. -- Highlights: • Description of the experimental molten salt storage system built at CIEMAT. • Design technical considerations for an experimental molten salt storage plant. • Hints for piping and heat exchangers design and thermal losses calculation. • Recommendations for

  13. On the use of a molten salt fast reactor to apply an idealized transmutation scenario for the nuclear phase out

    DEFF Research Database (Denmark)

    Merk, Bruno; Rohde, Ulrich; Glivici-Cotruţă, Varvara;

    2014-01-01

    In the view of transmutation of transuranium (TRU) elements, molten salt fast reactors (MSFRs) offer certain advantages compared to solid fuelled reactor types like sodium cooled fast reactors (SFRs). In the first part these advantages are discussed in comparison with the SFR technology, and the...... developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus onthe determination of the maximal transmutation efficiency and on an as much as possible reduced leftover of transuranium elements at the end of the reactor life. This minimal leftover...

  14. Thermal performance prediction and sensitivity analysis for future deployment of molten salt cavity receiver solar power plants in Algeria

    International Nuclear Information System (INIS)

    Highlights: • Performance of power plant with molten salt cavity receiver is assessed. • A method has been used to optimize the plant solar multiple, capacity factor and LEC. • Comparison of the simulated results to those of PS20 has shown good agreement. • Higher fossil fuel fraction reduces the LEC and increases the capacity factor. • Highland and Sahara regions are suitable for CRS plants deployment. - Abstract: Of all Concentrating Solar Power (CSP) technologies available today, the molten salt solar power plant appears to be the most important option for providing a major share of the clean and renewable electricity needed in the future. In the present paper, a technical and economic analysis for the implementation of a probable molten salt cavity receiver thermal power plant in Algeria has been carried out. In order to do so, we have investigated the effect of solar field size, storage capacity factor, solar radiation intensity, hybridization and power plant capacity on the thermal efficiency and electricity cost of the selected plant. The system advisor model has been used to perform the technical performance and the economic assessment for different locations (coastal, highland and Sahara regions) in Algeria. Taking into account various factors, a method has been applied to optimize the solar multiple and the capacity factor of the plant, to get a trade-off between the incremental investment costs of the heliostat field and the thermal energy storage. The analysis has shown that the use of higher fossil fuel fraction significantly reduces the levelized electricity cost (LEC) and sensibly increases the capacity factor (CF). The present study indicates that hybrid molten salt solar tower power technology is very promising. The CF and the LEC have been found to be respectively of the order of 71% and 0.35 $/kWe. For solar-only power plants, these parameters are respectively about 27% and 0.63 $/kWe. Therefore, hybrid central receiver systems are

  15. Electrodeposited tin coating as negative electrode material for lithium-ion battery in room temperature molten salt

    OpenAIRE

    Fung, YS; Zhu, DR

    2002-01-01

    A new room temperature molten salt (RTMS) [1-methyl-3-ethylimidazolium/AlCl3/SnCl2 (3:2:0.5)] was developed for depositing tin on a copper electrode. Different tin crystallites were deposited at different temperatures, giving widely different performances of the assembled lithium cell [Sn (Cu)/LiCl buffered MEICl-AlCl3 RTMS/lithium]. Tin film deposited at 50°C or higher gave a more desirable crystal structure and an improved performance than films obtained at lower temperatures. Both cyclic v...

  16. Characterization of the effects of continuous salt processing on the performance of molten salt fusion breeder blankets

    International Nuclear Information System (INIS)

    Several continuous salt processing options are available for use in molten salt fusion breeder blanket designs. The effects of processing on blanket performance have been assessed for three levels of processing and various equilibrium uranium concentrations in the salt. A one-dimensional model of the blanket was used in the neutronics analysis which incorporated transport calculations with time-dependent isotope generation and depletion calculations. The level of salt processing was found to have little effect on the behavior of the blanket during reactor operation; however, significant effects were observed during the decay period after reactor shutdown

  17. Synthesis of LiSm_(0.01)Mn_(1.99)O_4 by molten salt technique

    Institute of Scientific and Technical Information of China (English)

    M; Helan; L; John; Berchmans

    2010-01-01

    Samarium substituted lithium manganese oxide powders were successfully prepared by molten-salt synthesis(MSS) using eutectic mixture of LiCl,SmCl3.6H2O and MnO2 salt at 700 °C.The synthesis was carried out in open atmosphere.The crystalline powders were characterized for their phase identification using X-ray diffraction(XRD) analysis.The physico-chemical properties of the samarium substituted lithium manganese oxide powders were investigated by thermal analysis(TGA/DTA),FT-IR spectroscopy,EDAX,electron par...

  18. FISSILE FUEL BREEDING IN THE ARIES-ST FUSION REACTOR BY USING MOLTEN SALT WITH UF4

    OpenAIRE

    ÜBEYLİ, Mustafa

    2010-01-01

    ABSTRACTFissile fuel breeding in the ARIES-ST of the 1000 MWel power plant is investigated by using molten salt containing UF4 Calculations are done with the aid of one-dimensional code of SCALE4.3. In this hybrid model, a substantial amount of fissile fuel can be produced with a fissile fuel breeding ratio of 239 Pu = 0.115 per incident neutron at start-up conditions, that corresponds to 3558 kg 239 Pu/year by a full fusion power of 2740 MW. Tritium breeding ratio is found as 1.14 so that tr...

  19. An electrochemical study of the ruthenium (III) and (IV) hexachlorometallates in a basic room temperature chloroaluminate molten salt

    International Nuclear Information System (INIS)

    This paper reports that the ruthenium (IV) complex, RuCl62-, exhibits two successive voltammetric reduction waves with peak potentials of 0.34 and - 1.09 V, respectively, in the 49.0/51.0 m/o aluminum chloride-1-methyl-3-ethylimidazolium chloride room temperature molten salt vs. the Al3+/Al couple in the 66.7/33.3 m/o melt at 40 degrees C. The first wave corresponds to the reversible uncomplicated reduction of RuCl62- to RuCl63-, and the second appears to arise from the multielectron reduction of the latter complex to produce more than one species. The formal potential of the RuCl62-/3- redox system is 0.389 V, and the average Stokes-Einstein products of RuCl62- and RuCl63- are 1.8 x 10-10 and 1.3 x 10-10 g cm s-2 K-1, respectively, in this melt. The reversible half-wave potentials for a variety of 4d- and 5d-hexachlorometallate redox couples that were measured in this molten salt exhibit the same orderly trends with increasing atomic number found in CH2Cl2. This result was used in conjunction with the half-wave potential of the RuCl62-/3- couple to predict the potential of the (Tc)Cl62-/3- redox system in the melt

  20. Synthesis and piezoelectric properties of KxNa1-xNbO3 ceramic by molten salt method

    International Nuclear Information System (INIS)

    KxNa1-xNbO3 ceramic powder with perovskite structure was synthesized in molten salt with a Na2CO3/K2CO3 molar ratio of 1:1, under different salt-to-oxide weight ratios of 1:10, 1:5, 1:3, 1:2.5 and 1:2 in the temperatures range of 650-900 oC. It is found that the synthesizing temperature and salt-to-oxide ratios had significant effects on the morphology of KxNa1-xNbO3 powder. The X-ray diffraction analysis indicated that a pure perovskite structure of KxNa1-xNbO3 powder could be synthesized at 650 oC. The microstructure observation revealed that the crystal morphology of KxNa1-xNbO3 powder changed from spheroid to cube, and then became irregular after further increasing temperature. The grain size of the synthesized powder increased by an increment of the molten salt content. The KxNa1-xNbO3 ceramics were prepared at x = 0.345 by adding 1.0 mol% ZnO as sintering aid, and the optimized dielectric and piezoelectric properties are obtained as following: d33 = 120 pC/N, Tc = 406 oC, Qm = 126 and kp = 0.302.

  1. Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel

    International Nuclear Information System (INIS)

    The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [∼1000 Mw(e)] with passive safety systems to provide the potential for improved economics

  2. Effect of preparation temperature and cycling voltage range on molten salt method prepared SnO2

    International Nuclear Information System (INIS)

    Highlights: • SnO2 nanoparticles via molten-salt technique at 280 °C. • Studied the effect of particle size and its electrochemical properties. • Best electrochemical performances are obtained with an average particle size of 12 nm. -- Abstract: We prepared nano-sized tin (IV) oxide (SnO2) via molten-salt technique: heating a mixture of tin tetrachloride, lithium nitrate and lithium chloride at 280 °C in air. The powders are characterized by X-ray diffraction and transmission scanning microscopy techniques. The XRD studies showed a structure similar to tetragonal structure. The cyclic voltammetry studies showed characteristic cathodic peak potentials of reduction of Sn4+ to Sn metal in the first cathodic scan, and alloying–de–alloying reaction of Sn at ∼0.25 and ∼0.5 V vs. Li for successive cathodic and anodic scans cycled in the voltage range, 0.005–1.0 V. Galvanostatic cycling studies show that reversible capacities (MSM SnO2 prepared at 280 °C) of 640, 720, 890 mAh g−1 in the voltage range, 0.005–1.0 V, 0.005–1.3 V and 0.005–1.5 V, respectively at a current rate of 100 mA g−1. We also discussed the effect of particle size and its electrochemical properties in the voltage range, 0.005–1.0 V

  3. Recovery of ZrO{sub 2} by leaching from LiF-BeF{sub 2}-ZrO{sub 2} molten salt in distilled water

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Mun Sik; Yoo, Jae Hyung; Park, Hyun Soo; Kang, Young Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kwon, Soo Han [Chungbuk National University, Cheongju (Korea, Republic of)

    2000-12-01

    LiF-BeF{sub 2}-ZrF{sub 4} (63-30-7 mol%) molten salt was dissolved up to 0.02 g in 1 ml of distilled water at room temperature. ZrO{sub 2} oxide made from ZrF{sub 4} through pyrohydrolysis was recovered by leaching in distilled water with LiF-BeF{sub 2Z}rF{sub 4} molten salt including it at room temperature. The crystalline sharpness of recovered ZrO{sub 2} oxide was not damaged. (author)

  4. Dielectric relaxation and underlying dynamics of electrolyte solutions and solvent-molten salt mixtures using terahertz time-domain transmission spectroscopy

    Science.gov (United States)

    Asaki, Melanie Lynette Thongs

    Terahertz (THz) transmission spectroscopy is used to obtain the frequency dependent complex dielectric constants of water, methanol, and propylene carbonate, and solutions of lithium salts in these solvents, as well as mixtures of acetonitrile and a room-temperature molten salt. The behavior of the pure solvents is modeled with either two (water and acetonitrile) or three (methanol and propylene carbonate) Debye relaxations. For solutions of lithium salts, the effects of ionic solvation on the relaxation behavior of the solvents is discussed in terms of modifications to the values of the Debye parameters of the pure solvents. In this way we obtain estimates for numbers of irrotationally bound solvent molecules, the numbers of bonds broken or formed, and the effects of ions on the higher frequency relaxations. The same information was obtained for molten salt-acetonitrile systems. In addition, it was determined that at low molten salt concentrations, the mixtures behave like electrolyte solutions of a crystalline salt dissolved in a solvent. At higher molten salt concentrations, the behavior is that of a mixture of two liquids.

  5. Magnetic properties of La.sub.1-x./sub.Sr.sub.x./sub.MnO.sub.3./sub. nanoparticles prepared in a molten salt

    Czech Academy of Sciences Publication Activity Database

    Kačenka, Michal; Kaman, Ondřej; Jirák, Zdeněk; Maryško, Miroslav; Žvátora, Pavel; Vratislav, S.; Lukeš, I.

    2014-01-01

    Roč. 115, č. 17 (2014), "17B525-1"-"17B525-3". ISSN 0021-8979 R&D Projects: GA ČR(CZ) GAP108/11/0807 Institutional support: RVO:68378271 Keywords : magnetic properties * molten salt * neutron diffraction Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 2.183, year: 2014

  6. Zr electrorefining process for the treatment of cladding hull waste in LiCl-KCl molten salts

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang Hwa; Lee, You Lee; Jeon, Min Ku; Kang, Kweon Ho; Choi, Yong Taek; Park, Geun Il [Korea Atomic Energy Research Institute - KAERI, 989-111 Daeduk-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2013-07-01

    Zr electrorefining for the treatment of Zircaloy-4 cladding hull waste is demonstrated in LiCl-KCl-ZrCl{sub 4} molten salts. Although a Zr oxide layer thicker than 5 μm strongly inhibits the Zr dissolution process, pre-treatment processes increases the dissolution kinetics. For 10 g-scale experiments, the purities of the recovered Zr were 99.54 wt.% and 99.74 wt.% for fresh and oxidized cladding tubes, respectively, with no electrical contact issue. The optimal condition for Zr electrorefining has been found to improve the morphological feature of the recovered Zr, which reduces the salt incorporation by examining the effect of the process parameters such as the ZrCl{sub 4} concentration and the applied potential.

  7. Analysis of a helical coil once-through molten salt steam generator: Experimental results and heat transfer evaluation

    Science.gov (United States)

    Seubert, B.; Rojas, E.; Rivas, E.; Gaggioli, W.; Rinaldi, L.; Fluri, T.

    2016-05-01

    A molten salt helical coil steam generator is an alternative to kettle- or drum-type evaporators which are currently used in commercial-scale solar thermal power plants. A 300 kW prototype was tested during the OPTS project at ENEA. The experimental results presented in this paper have been used to validate a detailed heat transfer analysis of the whole system. The heat transfer analysis deals with the study of both the overall heat transfer coefficient and the shell-side heat transfer coefficient. Due to the specific features of this type of system, no correlations were available in the literature. A new numerical model to predict the performance of large-scale systems is also presented.

  8. Program management plan for the Molten Salt Reactor Experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    The primary mission of the Molten Salt Reactor Experiment (MSRE) Remediation Project is to effectively implement the risk-reduction strategies and technical plans to stabilize and prevent further migration of uranium within the MSRE facility, remove the uranium and fuel salts from the system, and dispose of the fuel and flush salts by storage in appropriate depositories to bring the facility to a surveillance and maintenance condition before decontamination and decommissioning. This Project Management Plan (PMP) for the MSRE Remediation Project details project purpose; technical objectives, milestones, and cost objectives; work plan; work breakdown structure (WBS); schedule; management organization and responsibilities; project management performance measurement planning, and control; conduct of operations; configuration management; environmental, safety, and health compliance; quality assurance; operational readiness reviews; and training

  9. Quality assurance plan for the molten salt reactor experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    This Quality Assurance Plan (QAP) identifies and describes the systems utilized by Molten Salt Reactor Experiment (MSRE) Remediation Project personnel to implement the requirements and associated applicable guidance contained in the Quality Program Description, Y/QD-15 Rev. 2 (Martin Marietta Energy Systems, Inc., 1995) and Environmental Management and Enrichment Facilities Work Smart Standards. This QAP defines the quality assurance (QA) requirements applicable to all activities and operations in and directly pertinent to the MSRE Remediation Project. This QAP will be periodically reviewed, revised, and approved as necessary. This QAP identifies and describes the QA activities and procedures implemented by the various Oak Ridge National Laboratory support organizations and personnel to provide confidence that these activities meet the requirements of this project. Specific support organization (Division) quality requirements, including the degree of implementation of each, are contained in the appendixes of this plan

  10. THERMAL DESTRUCTION OF HIGHLY CHLORINATED MIXED WASTES WITHOUT GENERATING CORROSIVE OFF-GASES USING MOLTEN SALT OXIDATION (1,2)

    International Nuclear Information System (INIS)

    A pilot-scale MSO (Molten Salt Oxidation) system was used to process 45-gallons of a halogenated mixed waste that is difficult to treat with other thermal systems. The mixed waste was a halogenated solvent that consisted mostly of methylchloroform. The 80 weight percent of waste consisting of highly corrosive chlorine was captured in the first process vessel as sodium chloride. The sodium chloride leached chrome from that process vessel and the solidified salt exhibited the toxicity characteristic for chrome as measured by TCLP (Toxicity Characteristic Leaching Procedure) testing. The operating ranges for parameters such as salt bed temperature, off-gas temperature, and feed rate that enable sustained operation were identified. At feed rates below the sustainable limit, both processing capacity and maintenance requirements increased with feed rate. Design and operational modifications to increase the sustainable feed rate limit and reduce maintenance requirements reduced both salt carryover and volumetric gas flows

  11. Molten salt oxidation of chloro-organic compounds: Experimental results for product gas compositions and final forms studies

    Energy Technology Data Exchange (ETDEWEB)

    Rudolph, J.C.; Haas, P.A.; Bell, J.T.; Crosley, S.M.; Calhoun, C.L. Jr.; Gorin, A.H.; Nulf, L.E.

    1995-04-01

    Molten salt oxidation (MSO) has been selected as a promising technology for treatment of some US Department of Energy (DOE) mixed wastes. Mixed wastes are defined as those wastes that contain both radioactive components, which are regulated by the Atomic Energy Act of 1954, and hazardous waste components, which are regulated under the Resource Conservation and Recovery Act (RCRA). Oak Ridge National Laboratory (ORNL) has installed and operated a bench-scale MSO apparatus to obtain experimental information needed before the design and construction of an MSO pilot plant. The primary objective of the experiments performed was to show that dioxin and furan emissions from a molten salt oxidation (MSO) unit were below the proposed regulatory limit of 0.1 ng/m{sup 3} as 2,3,7,8-tetrachlorodibenzo-para-dioxin equivalents or toxic equivalence quotient. The feed stream was to contain 2,4-dichlorophenol, a suspected precursor to the formation of dioxin and furans. The tests were to be done over a range of salt compositions and flow rates expected in a pilot- or full-scale MSO unit. Two other objectives were to demonstrate destruction and removal efficiencies (DREs) greater than US Environmental Protection Agency requirements and to show that levels of products of incomplete combustion (PICs) are the same as, or lower than, those observed in incinerators for two common waste constituents [carbon tetrachloride (CCl{sub 4}) and CH{sub 3}CCl{sub 3}]. A final objective was to perform some initial studies of final waste forms using sulfur polymer cement (SPC). This report presents the results from the operation of the bench-scale MSO system.

  12. Novel band gap-tunable K–Na co-doped graphitic carbon nitride prepared by molten salt method

    International Nuclear Information System (INIS)

    Graphical abstract: K and Na ions co-doped into g-C3N4 crystal lattice can tune the position of CB and VB potentials, influence the structural and optical properties, and thus improve the photocatalytic degradation and mineralization ability. - Highlights: • K, Na co-doped g-C3N4 was prepared in KCl/NaCl molten salt system. • The structural and optical properties of g-C3N4 were greatly influenced by co-doping. • The position of VB and CB can be tuned by controlling the weight ratio of eutectic salts to melamine. • Co-doped g-C3N4 showed outstanding photodegradation ability, mineralization ability, and catalytic stability. - Abstract: Novel band gap-tunable K–Na co-doped graphitic carbon nitride was prepared by molten salt method using melamine, KCl, and NaCl as precursor. X-ray diffraction (XRD), N2 adsorption, Scanning electron microscope (SEM), UV–vis spectroscopy, Photoluminescence (PL), and X-ray photoelectron spectroscopy (XPS) were used to characterize the prepared catalysts. The CB and VB potentials of graphitic carbon nitride could be tuned from −1.09 and +1.55 eV to −0.29 and +2.25 eV by controlling the weight ratio of eutectic salts to melamine. Besides, ions doping inhibited the crystal growth of graphitic carbon nitride, enhanced the surface area, and increased the separation rate of photogenerated electrons and holes. The visible-light-driven Rhodamine B (RhB) photodegradation and mineralization performances were significantly improved after K–Na co-doping

  13. Waste Stream Generated and Waste Disposal Plans for Molten Salt Reactor Experiment at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Haghighi, M. H.; Szozda, R. M.; Jugan, M. R.

    2002-02-26

    The Molten Salt Reactor Experiment (MSRE) site is located in Tennessee, on the U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR), south of the Oak Ridge National Laboratory (ORNL) main plant across Haw Ridge in Melton Valley. The MSRE was run by ORNL to demonstrate the desirable features of the molten-salt concept in a practical reactor that could be operated safely and reliably. It introduced the idea of a homogeneous reactor using fuel salt media and graphite moderation for power and breeder reactors. The MSRE reactor and associated components are located in cells beneath the floor in the high-bay area of Building 7503 (Figure 1). The reactor was operated from June 1965 to December 1969. When the reactor was shut down, fuel salt was drained from the reactor circuit to two drain tanks. A ''clean'' salt was then circulated through the reactor as a decontamination measure and drained to a third drain tank. When operations ceased, the fuel and flush salts were allowed t o cool and solidify in the drain tanks. At shutdown, the MSRE facility complex was placed in a surveillance and maintenance program. As a result of the S&M program, it was discovered in 1994 that gaseous uranium (233U/232U) hexafluoride (UF6) had moved throughout the MSRE process systems. The UF6 was generated when radiolysis of the fluorine salts caused the individual constituents to dissociate to their component atoms, including free fluorine.Some of the free fluorine combined with uranium fluorides (UF4) in the salt to form UF6. UF6 is gaseous at slightly above ambient temperatures; thus, periodic heating of the fuel salts (which was intended to remedy the radiolysis problems) and simple diffusion had allowed the UF6 to move out of the salt and into the process systems of MSRE.

  14. Novel band gap-tunable K–Na co-doped graphitic carbon nitride prepared by molten salt method

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Jiannan [Institute of Eco-environmental Sciences, Liaoning Shihua University, Fushun 113001 (China); School of Environmental and Biological Engineering, Liaoning Shihua University, Fushun 113001 (China); Ma, Lin [School of Petrochemical Engineering, Liaoning Shihua University, Fushun 113001 (China); Wang, Haoying; Zhao, Yanfeng [School of Environmental and Biological Engineering, Liaoning Shihua University, Fushun 113001 (China); Zhang, Jian [School of Petrochemical Engineering, Liaoning Shihua University, Fushun 113001 (China); Hu, Shaozheng, E-mail: hushaozhenglnpu@163.com [Institute of Eco-environmental Sciences, Liaoning Shihua University, Fushun 113001 (China)

    2015-03-30

    Graphical abstract: K and Na ions co-doped into g-C{sub 3}N{sub 4} crystal lattice can tune the position of CB and VB potentials, influence the structural and optical properties, and thus improve the photocatalytic degradation and mineralization ability. - Highlights: • K, Na co-doped g-C{sub 3}N{sub 4} was prepared in KCl/NaCl molten salt system. • The structural and optical properties of g-C{sub 3}N{sub 4} were greatly influenced by co-doping. • The position of VB and CB can be tuned by controlling the weight ratio of eutectic salts to melamine. • Co-doped g-C{sub 3}N{sub 4} showed outstanding photodegradation ability, mineralization ability, and catalytic stability. - Abstract: Novel band gap-tunable K–Na co-doped graphitic carbon nitride was prepared by molten salt method using melamine, KCl, and NaCl as precursor. X-ray diffraction (XRD), N{sub 2} adsorption, Scanning electron microscope (SEM), UV–vis spectroscopy, Photoluminescence (PL), and X-ray photoelectron spectroscopy (XPS) were used to characterize the prepared catalysts. The CB and VB potentials of graphitic carbon nitride could be tuned from −1.09 and +1.55 eV to −0.29 and +2.25 eV by controlling the weight ratio of eutectic salts to melamine. Besides, ions doping inhibited the crystal growth of graphitic carbon nitride, enhanced the surface area, and increased the separation rate of photogenerated electrons and holes. The visible-light-driven Rhodamine B (RhB) photodegradation and mineralization performances were significantly improved after K–Na co-doping.

  15. Preparation and characterization of La9.33Si6O26 powders by molten salt method for solid electrolyte application

    International Nuclear Information System (INIS)

    Research highlights: → Pure La9.33Si6O26 powder was successfully synthesized via the molten salt method. → Pure LSO ceramics via molten salt process exhibited better electrical properties. → The mass ratio of reactants to NaCl greatly affected the crystal structure of LSO powders. → The synthesis temperature greatly affected the crystal structure of LSO powders. → The involvement of Na in LSO lattice deteriorated the conductance of electrolytes. - Abstract: Lanthanum silicate La9.33Si6O26 (LSO) powders with more uniform particle and less agglomeration were obtained at a much lower synthesis temperature by the molten salt method than by the solid-state method. LSO ceramic electrolytes were prepared with these powders and characterized as well. The optimal molten salt synthesis conditions are mass ratio of reactants to NaCl of 1:3 and synthesis temperature of 900 oC. XRD results showed that when the mass ratio of reactants to NaCl was no more than 1:3, pure LSO phase powder was obtained at 900 oC. XRD and XRF results showed that when synthesis temperature was higher than 900 oC, a solid solution type LSO powder with Na replacement for La formed at a fixed mass ratio of reactants to NaCl of 1:3. The involvement of Na in LSO lattice might lead to the lattice contraction in powders and deteriorate the conductance of ceramic electrolytes. The ceramic electrolytes prepared from the pure LSO powder via molten salt process exhibited better electrical properties than those from the powder via solid-state method.

  16. Intergranular tellurium cracking of nickel-based alloys in molten Li, Be, Th, U/F salt mixture

    International Nuclear Information System (INIS)

    In Russia, R and D on Molten Salt Reactor (MSR) are concentrated now on fast/intermediate spectrum concepts which were recognized as long term alternative to solid fueled fast reactors due to their attractive features: strong negative feedback coefficients, easy in-service inspection, and simplified fuel cycle. For high-temperature MSR corrosion of the metallic container alloy in primary circuit is the primary concern. Key problem receiving current attention include surface fissures in Ni-based alloys probably arising from fission product tellurium attack. This paper summarizes results of corrosion tests conducted recently to study effect of oxidation state in selected fuel salt on tellurium attack and to develop means of controlling tellurium cracking in the special Ni-based alloys recently developed for molten salt actinide recycler and tranforming (MOSART) system. Tellurium corrosion of Ni-based alloys was tested at temperatures up to 750 °C in stressed and unloaded conditions in molten LiF–BeF2 salt mixture fueled by about 20 mol% of ThF4 and 2 mol% of UF4 at different [U(IV)]/[U(III)] ratios: 0.7, 4, 20, 100 and 500. Following Ni-based alloys (in mass%): HN80M-VI (Mo—12, Cr—7.6, Nb—1.5), HN80MTY (Mo—13, Cr—6.8, Al—1.1, Ti—0.9), HN80MTW (Mo—9.4, Cr—7.0, Ti—1.7, W—5.5) and EM-721 (W—25.2, Cr—5.7, Ti—0.17) were used for the study in the corrosion facility

  17. Intergranular tellurium cracking of nickel-based alloys in molten Li, Be, Th, U/F salt mixture

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, Victor, E-mail: ignatiev@vver.kiae.ru; Surenkov, Alexander; Gnidoy, Ivan; Kulakov, Alexander; Uglov, Vadim; Vasiliev, Alexander; Presniakov, Mikhail

    2013-09-15

    In Russia, R and D on Molten Salt Reactor (MSR) are concentrated now on fast/intermediate spectrum concepts which were recognized as long term alternative to solid fueled fast reactors due to their attractive features: strong negative feedback coefficients, easy in-service inspection, and simplified fuel cycle. For high-temperature MSR corrosion of the metallic container alloy in primary circuit is the primary concern. Key problem receiving current attention include surface fissures in Ni-based alloys probably arising from fission product tellurium attack. This paper summarizes results of corrosion tests conducted recently to study effect of oxidation state in selected fuel salt on tellurium attack and to develop means of controlling tellurium cracking in the special Ni-based alloys recently developed for molten salt actinide recycler and tranforming (MOSART) system. Tellurium corrosion of Ni-based alloys was tested at temperatures up to 750 °C in stressed and unloaded conditions in molten LiF–BeF{sub 2} salt mixture fueled by about 20 mol% of ThF{sub 4} and 2 mol% of UF{sub 4} at different [U(IV)]/[U(III)] ratios: 0.7, 4, 20, 100 and 500. Following Ni-based alloys (in mass%): HN80M-VI (Mo—12, Cr—7.6, Nb—1.5), HN80MTY (Mo—13, Cr—6.8, Al—1.1, Ti—0.9), HN80MTW (Mo—9.4, Cr—7.0, Ti—1.7, W—5.5) and EM-721 (W—25.2, Cr—5.7, Ti—0.17) were used for the study in the corrosion facility.

  18. Development of Pyro-separation Technology Based on Molten Salt Electrolysis

    International Nuclear Information System (INIS)

    The focus of this study was to develop recovery technologies in the pyroprocessing. The unit processes of the project can be classified into two groups; electro-refining process to recover uranium and long-lived nuclides, and cathode processing to produce a metal ingot both from a salt-contained metal and from Cd-contained metal. This project has been carried out for the third phase period of the long-term nuclear R and D program, and focused on the development of key technologies of the pyroprocessing such as electrorefining, draw down and cathode processing. Mock-up system of 1 kg-U/batch was built for performance tests which were conducted to ensure the adequacy of the research and development of the pyroprocessing technology. The experiments were carried out through bench-scale inactive tests except for uranium. In particular, the sticking problem was inevitable in the US's Mark-V and PEER electrorefiner. As a result of this study, a graphite cathode was developed, which exhibited self-scraping behavior and did not need scraping step. The design of an electrorefiner could be simplified, and the throughput was enhanced due to an increased cathode area. A long-term R and D plan was established to develop pyroprocessing technology. In the near term, the results of the current project will be utilized in the next phase of the R and D plan ('07 - '10), and long-term wise, is expected to contribute to recovering fuel materials for transmutation in a Gen-IV reactor

  19. Development of Pyro-separation Technology Based on Molten Salt Electrolysis

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Joon Bo; Kim, E. H.; Yoo, J. H. (and others)

    2007-06-15

    The focus of this study was to develop recovery technologies in the pyroprocessing. The unit processes of the project can be classified into two groups; electro-refining process to recover uranium and long-lived nuclides, and cathode processing to produce a metal ingot both from a salt-contained metal and from Cd-contained metal. This project has been carried out for the third phase period of the long-term nuclear R and D program, and focused on the development of key technologies of the pyroprocessing such as electrorefining, draw down and cathode processing. Mock-up system of 1 kg-U/batch was built for performance tests which were conducted to ensure the adequacy of the research and development of the pyroprocessing technology. The experiments were carried out through bench-scale inactive tests except for uranium. In particular, the sticking problem was inevitable in the US's Mark-V and PEER electrorefiner. As a result of this study, a graphite cathode was developed, which exhibited self-scraping behavior and did not need scraping step. The design of an electrorefiner could be simplified, and the throughput was enhanced due to an increased cathode area. A long-term R and D plan was established to develop pyroprocessing technology. In the near term, the results of the current project will be utilized in the next phase of the R and D plan ('07 - '10), and long-term wise, is expected to contribute to recovering fuel materials for transmutation in a Gen-IV reactor.

  20. Discriminators for the Accelerator-Based Conversion (ABC) concept using a subcritical molten salt system

    International Nuclear Information System (INIS)

    Discriminators are described that quantify enhancements added to plutonium destruction and/or nuclear waste transmutation systems through use of an accelerator/fluid fuel combination. This combination produces a robust and flexible nuclear system capable of the destruction of all major long-lived actinides (including plutonium) and fission products. The discriminators discussed in this report are (1) impact of subcritical operation on safety, (2) impact of subcritical and fluid fuel operation on plutonium burnout scenarios, and (3) neutron economy enhancements brought about by subcritical operation. Neutron economy enhancements are quantified through assessment of long-term dose reduction resulting from transmutation of key fission products along with relaxation of processing frequencies afforded by subcritical operation

  1. Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for {sup 233}U fuel production: 1. Concept of a stationary Tokamak as a neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Azizov, E. A.; Gladush, G. G., E-mail: gladush@triniti.ru; Dokuka, V. N.; Khayrutdinov, R. R. [State Research Center of the Russian Federation, Troitsk Institute for Innovation and Fusion Research (Russian Federation)

    2015-12-15

    On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of {sup 233}U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.

  2. Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for 233U fuel production: 1. Concept of a stationary Tokamak as a neutron source

    Science.gov (United States)

    Azizov, E. A.; Gladush, G. G.; Dokuka, V. N.; Khayrutdinov, R. R.

    2015-12-01

    On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of 233U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.

  3. Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for 233U fuel production: 1. Concept of a stationary Tokamak as a neutron source

    International Nuclear Information System (INIS)

    On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of 233U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved

  4. Progress in the Application of Molten Salts for New Energy Production%熔盐在新能源领域的应用

    Institute of Scientific and Technical Information of China (English)

    杨武龙; 姜洪涛; 吴靥汝; 樊皓; 华炜; 计建炳

    2012-01-01

    The applications of molten salts in electrolyte cell (including fuel cell, lithium battery and thermal battery), solar thermal power generation and biomass pyrolysis are reviewed. The working characteristics, conversion efficiency and service life of fuel cells with molten salts are evaluated. It is also pointed out that the development of fuel cells will focus on molten carbonate fuel cells. Compared with the cathode material of lithium battery, prepared by high temperature solid phase method, the material by molten salt method is improved on crystalline structure and performance. And in solar thermal power generation field, high temperature molten salt as a low cost and high efficiency heat transfer and storage medium has been also applied, Furthermore, recent research progress in pyrolysis of biomass to bio-oil or hydrogen-rich gas in molten salt media is summarized. By examining molten salt composition and reaction temperature, and referring pyrolysis mechanism and kinetics, it is concluded that pyrolysis of biomass in molten salts is practicable. Finally, in order to enhance the development of molten salts application forward, problems to be solved are listed.%综述了熔盐在电池(燃料电池、锂电池和热电池)、太阳能热发电等新能源领域的应用及其在生物质热裂解领域的研究动态,评价了熔盐为电解质的燃料电池的工作特点、转换效率和使用寿命,指出了熔融碳酸盐电池是燃料电池的发展方向.相对于高温固相法制备的锂电池正极材料,熔盐法制备的正极材料的晶体结构和性能得到改善.在太阳能热发电领域,高温熔盐作为一种低成本、高效率的传热蓄热介质得到了一定的应用.介绍了熔盐热裂解生物质制取生物油或富氢气体的过程,对比了熔盐组成、裂解条件对产物产率的影响,分析了热裂解机理和动力学,指出熔盐裂解生物质技术的可行性.最后对熔盐在新能源领域的应用存在

  5. Superfund record of decision (EPA Region 4): Oak Ridge Reservation (USDOE) Molten Salt Reactor Experiment (MSRE) facility, Oak Ridge, TN, July 7, 1998

    International Nuclear Information System (INIS)

    This documents presents a description of the selected remedy, which includes removing flush salt and fuel salt from their respective storage containers in the Molten Salt Reactor Experiment facility, removing uranium from the salts, treating the uranium to form an oxide for safer storage, placing the uranium oxide into storage, containerizing the fuel and flush salts without uranium, and temporarily storing this salt at the Oak Ridge National Laboratory until final disposition of the salt

  6. Electrochemical Behavior of Europium(III)-Europium(II) in LiF-NaF-KF Molten Salt

    International Nuclear Information System (INIS)

    The transformation of Eu(III) to Eu(II) was confirmed in a fluoride eutectic, LiF-NaF-KF (46.5-11.5-42.0 mol%, FLiNaK) molten salt during a treatment of high temperature as high as 1023 K. The coexistence of Eu(III)-Eu(II) was characterized by X-ray photoelectron spectroscopy (XPS) and voltammetry method, and their concentrations were measured. The electrochemical behavior of Eu(III) and Eu(II) in the fluoride salt was investigated. The mechanism of the electrode reaction was determined using cyclic voltammetry (CV) and square wave voltammetry (SWV). The results indicated a one-electron exchange process, corresponding to the reduction of Eu(III) to Eu(II) and the oxidation of Eu(II) to Eu(III). This process is reversible and diffusion-controlled. The diffusion coefficients (D) of Eu(III) and Eu(II) were determined using the conventional CV by changing the scanning rate and a modified method by changing the area of the working electrode successively. The values obtained by these two different methods were consistent. The temperature dependence of diffusion coefficient was investigated, and the activation energies of diffusion process were calculated to be 38.9 ± 4.6 kJ mol-1 for Eu(III) and 34.7 ± 1.6 kJ mol-1 for Eu(II), respectively

  7. Molten salt solvent synthesis of La2Mo2O9 nano-wires by controlling the subsequent calcinations process

    International Nuclear Information System (INIS)

    Large amounts of La2Mo2O9 nano-wires have been produced using molten-salt synthesis method. The powder X-ray diffraction, field emission scanning electron microscopy, and transmission electron microscopy are used to investigate structure and morphological features of the obtained products. The formed nano-wires have an average diameter of about 100 nm and a length in the range from ten to several tens of micrometers. The analyses of the high resolution transmission electron microscopy and the selected area electron diffraction results show that the nano-wires are single crystalline and grow along the [0 0 1] direction. A growth mechanism of La2Mo2O9 nano-wires is also proposed in this report. It implies that the temperature, chloride ions and cation lattice in β-La2Mo2O9 might be related to the particles morphologies transition. - Highlights: • La2Mo2O9nano-wires are successfully synthesizedand first reported in this paper. • The influences of possible factors on nano-wires synthesis are discussed. • The possible mechanism was discussed

  8. Health and safety plan for the Molten Salt Reactor Experiment remediation project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    Burman, S.N.; Uziel, M.S.

    1995-12-01

    The Lockheed Martin Energy Systems, Inc., (Energy Systems) policy is to provide a safe and healthful workplace for all employees and subcontractors. The accomplishment of the policy requires that operations at the Molten Salt Reactor Experiment (MSRE) facility at the Department of Energy (DOE) Oak Ridge National Laboratory (ORNL) are guided by an overall plan and consistent proactive approach to safety and health (S and H) issues. The policy and procedures in this plan apply to all MSRE operations. The provisions of this plan are to be carried out whenever activities are initiated at the MSRE that could be a threat to human health or the environment. This plan implements a policy and establishes criteria for the development of procedures for day-to-day operations to prevent or minimize any adverse impact to the environment and personnel safety and health and to meet standards that define acceptable management of hazardous and radioactive materials and wastes. The plan is written to utilize past experience and the best management practices to minimize hazards to human health or the environment from events such as fires, explosions, falls, mechanical hazards, or any unplanned release of hazardous or radioactive materials to the air.

  9. Isotopic ratios and effective power determined by gamma-ray spectroscopy vs mass spectroscopy for molten salt extraction residues

    International Nuclear Information System (INIS)

    Impure plutonium metal is routinely processed by molten salt extraction (MSE) to reduce the amount of americium in the metal product. Throughput at various facilities where similar processes are performed has made it essential to evaluate uncertainties and possible discrepancies in the analyses of these difficult MSE materials. In an effort to evaluate the plutonium isotopic ratios and americium concentrations obtained from gamma-ray spectral data analyzed by the computer code GRPAUT, measurements were made on ten MSE salts as received and after pulverization and blending. These results were then compared to the specific powers obtained from isotopic ratios determined by mass spectrometry on these same ten samples. Americium values ranged from a few thousand parts-per-million of total plutonium to greater than 50,000 ppM. Our results indicate a small discrepancy between specific powers as determined by GRPAUT on ''as received'' vs pulverized and blended MSE salts. The specific powers obtained via GRPAUT on the pulverized salts agree somewhat better with specific powers obtained from the mass spectroscopy data. This work may indicate that a small discrepancy exists in the specific powers by using GRPAUT on heterogeneous, high americium samples. 5 refs., 6 tabs

  10. Improvement of the Neutronic Performance of the PACER Fusion Concept Using Thorium Molten Salt with Reactor Grade Plutonium

    Science.gov (United States)

    Acır, Adem

    2013-02-01

    In this study, the improvement of neutronic performance of a dual purpose modified PACER concept has been investigated. Flibe as the main constituent are fixed as 92% coolant. ThF4 is mixed with increased mole-fractions of RG-PuF4 starting by 0 mol % up to 1 mol %. TBR variations for all the investigated salts with respect to the RG-PuF4 contents are computed. Tritium self-sufficiency is provided with the ThF4 when the adding RG-PuF4 content is higher than 0.75%. The energy multiplication of the blanket is increased as 70% with adding RG-PuF4 contents to ThF4. High quality fissile isotope 233U are produced with increasing RG-PuF4. DPA and helium production increases with increased RG-PuF4 content in molten salt. Radiation damage with dpa <1.7 and He <3.3 ppm after a plant operation period of 30 years will be well below the damage limit values.

  11. Al/Pb lightweight grids prepared by molten salt electroless plating for application in lead-acid batteries

    Science.gov (United States)

    Hong, Bo; Jiang, Liangxing; Hao, Ketao; Liu, Fangyang; Yu, Xiaoying; Xue, Haitao; Li, Jie; Liu, Yexiang

    2014-06-01

    In this paper, a lightweight Pb plated Al (Al/Pb) grid was prepared by molten salt electroless plating. The SEM and bonding strength test show that the lead coating is deposited with a smooth surface and firm combination. CV test shows that the electrochemical properties of Al/Pb electrodes are stable. 2.0 V single-cell flooded lead-acid batteries with Al/Pb grids as negative collectors are assembled and the performances including 20 h capacity, rate capacity, cycle life, internal resistance are investigated. The results show that the cycle life of Al/Pb-grid cells is about 475 cycles and can meet the requirement of lead-acid batteries. Al/Pb grids are conducive to the refinement of PbSO4 grain, and thereby reduce the internal resistance of battery and advance the utilization of active mass. Moreover, weight of Al/Pb grid is only 55.4% of the conventional-grid. In this way, mass specific capacity of Al/Pb-grid negatives is 17.8% higher and the utilization of active mass is 6.5% higher than conventional-grid negatives.

  12. Distillation of LiCl from the LiCl-Li2O molten salt of the electrolytic reduction process

    International Nuclear Information System (INIS)

    Electrolytic reduction of the uranium oxide in LiCl-Li2O molten salt for the treatment of spent nuclear fuel requires the separation of the residual salt from the reduced metal product, which contains about 20 wt% salt. In order to separate the residual salt and reuse it in the electrolytic reduction, a vacuum distillation process was developed. Lab-scale distillation equipment was designed and installed in an argon atmosphere glove box. The equipment consisted of an evaporator in which the reduced metal product was contained and exposed to a high temperature and reduced pressure; a receiver; and a vertically oriented condenser that operated at a temperature below the melting point of lithium chloride. We performed experiments with LiCl-Li2O salt to evaluate the evaporation rate of LiCl salt and varied the operating temperature to discern its effect on the behavior of salt evaporation. Complete removal of the LiCl salt from the evaporator was accomplished by reducing the internal pressure to <100 mTorr and heating to 900 deg C. We achieved evaporation efficiency as high as 100 %. (author)

  13. Actinide-Lanthanide separation by an electrolytic method in molten salt media: feasibility assessment of a renewed liquid cathode

    International Nuclear Information System (INIS)

    This study is part of a research program concerning the assessment of pyrochemical methods for the nuclear waste processing. The An-Ln partitioning could be achieved by an electrolytic selective extraction in molten salt media. It has been decided to focus on liquid reactive cathode which better suits to a group actinides co-recycling. The aim of the study is to propose, define and initiate the development of an electrolytic pyro-process dedicated to the quantitative and selective recovery of the actinides. Quantitativeness is related to technology, whereas selectivity is governed by chemistry. The first step consisted in selecting the adequate operating conditions, which enables a sufficient An-Ln separation. The first step consisted, by means of thermodynamic calculi and electrochemical investigations, in selecting a promising combination between molten electrolyte and cathodic material, regarding the process constraints. To improve the recovery yield, it is necessary to develop a disruptive technology: here comes the concept of a dynamic electrodeposition carried out onto liquid metallic drops. The next step consisted in designing and manufacturing a lab-scale device which enables dropping flow studies. Since interfacial phenomena are of primary meaning in such a concept, it has been decided to focus on high temperature liquid-liquid interfacial measurements. (author)

  14. Health and safety plan for the Molten Salt Reactor Experiment remediation project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    The Lockheed Martin Energy Systems, Inc., (Energy Systems) policy is to provide a safe and healthful workplace for all employees and subcontractors. The accomplishment of the policy requires that operations at the Molten Salt Reactor Experiment (MSRE) facility at the Department of Energy (DOE) Oak Ridge National Laboratory (ORNL) are guided by an overall plan and consistent proactive approach to safety and health (S and H) issues. The policy and procedures in this plan apply to all MSRE operations. The provisions of this plan are to be carried out whenever activities are initiated at the MSRE that could be a threat to human health or the environment. This plan implements a policy and establishes criteria for the development of procedures for day-to-day operations to prevent or minimize any adverse impact to the environment and personnel safety and health and to meet standards that define acceptable management of hazardous and radioactive materials and wastes. The plan is written to utilize past experience and the best management practices to minimize hazards to human health or the environment from events such as fires, explosions, falls, mechanical hazards, or any unplanned release of hazardous or radioactive materials to the air

  15. A way to limit the corrosion in the Molten Salt Reactor concept: the salt redox potential control

    International Nuclear Information System (INIS)

    The possibility of controlling the salt redox potential thanks to a redox buffer in the Molten Salt Fast Reactor was investigated, the goal was to limit the oxidation of the reactor structural material. Tests were performed in LiF-CaF2 at 850 °C on two different redox couples to fix the salt potential, Eu(III)/Eu(II) and U(IV)/U(III), where the first one was used as inactive system to validate the methodology to be applied on the uranium system. A metallic reducing agent (Gd plate for Eu, and U plate for U system) was inserted in the salt, leading to a spontaneous reaction: Eu(III) and U(IV) were then reduced. Eu(III) was fully converted into Eu(II) with metallic Gd, validating the approach. On the U system, the U(IV)/U(III) ratio has to be set between 10 and 100 to limit the core material oxidation: addition of metallic U decreased the concentration ratio from the infinite to 1, showing the feasibility of the salt redox potential control with the U system

  16. Reprocessing of spent nitride fuel by chemical dissolution in molten salt: Results on plutonium nitride containing inert matrix materials

    International Nuclear Information System (INIS)

    Solid solutions of actinide mono-nitrides have been proposed as a candidate fuel of the accelerator-driven system (ADS) for transmutation of minor actinides (MA). The pyrochemical process has several advantages over the wet process such as PUREX in the case of treating spent nitride fuel with large decay heat and fast neutron emission, and recovering highly enriched 15N. In the present study, the chemical dissolution of PuN, (PuxZr1-x)N and PuN+TiN, with CdCl2 in LiCl-KCl eutectic melt were investigated to confirm the possibility of the chemical dissolution process of spent nitride fuel. The plutonium nitrides, PuN, (PuxZr1-x)N and PuN+TiN, were dissolved by the reaction with CdCl2 in LiCl- KCl eutectic melt at 823 K and most of Pu was recovered into the molten salt. On the other hand, most of ZrN and TiN were not dissolved and remained as undissolved residues. (authors)

  17. Multicomponent diffusion in molten salt LiF-BeF2: Dynamical correlations and Maxwell–Stefan diffusivities

    International Nuclear Information System (INIS)

    Applying Green–Kubo formalism and equilibrium molecular dynamics (MD) simulations, we have studied the dynamic correlation, Onsager coeeficients and Maxwell–Stefan (MS) Diffusivities of molten salt LiF-BeF2, which is used as coolant in high temperature reactor. All the diffusive flux correlations show back-scattering or cage dynamics which becomes pronouced at higher temperature. Although the MS diffusivities are expected to depend very lightly on the composition due to decoupling of thermodynamic factor, the diffusivity ĐLi-F and ĐBe-F decreases sharply for higher concentration of LiF and BeF2 respectively. Interestingly, all three MS diffusivities have highest magnitude for eutectic mixture at 1000K (except ĐBe-F at lower LiF mole fraction) which is desirable from coolant point of view. Although the diffusivity for positive-positive ion pair is negative it is not in violation of the second law of thermodynamics as it satisfies the non-negative entropic constraints

  18. Calculating the effective delayed neutron fraction in the Molten Salt Fast Reactor: Analytical, deterministic and Monte Carlo approaches

    International Nuclear Information System (INIS)

    Highlights: • Calculation of effective delayed neutron fraction in circulating-fuel reactors. • Extension of the Monte Carlo SERPENT-2 code for delayed neutron precursor tracking. • Forward and adjoint multi-group diffusion eigenvalue problems in OpenFOAM. • Analytical approach for βeff calculation in simple geometries and flow conditions. • Good agreement among the three proposed approaches in the MSFR test-case. - Abstract: This paper deals with the calculation of the effective delayed neutron fraction (βeff) in circulating-fuel nuclear reactors. The Molten Salt Fast Reactor is adopted as test case for the comparison of the analytical, deterministic and Monte Carlo methods presented. The Monte Carlo code SERPENT-2 has been extended to allow for delayed neutron precursors drift, according to the fuel velocity field. The forward and adjoint eigenvalue multi-group diffusion problems are implemented and solved adopting the multi-physics tool-kit OpenFOAM, by taking into account the convective and turbulent diffusive terms in the precursors balance. These two approaches show good agreement in the whole range of the MSFR operating conditions. An analytical formula for the circulating-to-static conditions βeff correction factor is also derived under simple hypotheses, which explicitly takes into account the spatial dependence of the neutron importance. Its accuracy is assessed against Monte Carlo and deterministic results. The effects of in-core recirculation vortex and turbulent diffusion are finally analysed and discussed

  19. Development of pyrochemical process in molten salts applied to the radioactive wastes from nuclear power plants and nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Toshiba started to develop a series of methods of radioactive waste treatment in 1980's. Our original methods developed to date are; an abrasive blast decontamination method, a water jet method, a chemical decontamination method, an electrolytic decontamination method in aqueous solution etc. However, these decontamination methods have one common problem such as difficulties in decontamination of waste materials of complicated shapes and another common problem to cause large volume of secondary wastes such as used reagents and resin. Pyrochemical treatment in molten salts is a promising process for decontamination of radioactive waste in complicated shapes without producing a large amount of secondary wastes because the electric current is distributed effectively entire surface of the waste in high conductive molten salts and the decontaminated species by anodic dissolution are recovered on the cathode. We have applied this process to the radioactive waste such as Magnox end crops, metallic waste contaminated with uranium, wasted chemical traps for UF6 and channel boxes zircaloy wasted from boiling water reactors (BWR). This paper reviews pyrochemical processes in molten salts applied to the radioactive waste treatment. (author)

  20. Theoretical investigation on local structure and transport properties of NaFsbnd AlF3 molten salts under electric field environment

    Science.gov (United States)

    Lv, Xiaojun; Xu, Zhenming; Li, Jie; Chen, Jiangan; Liu, Qingsheng

    2016-08-01

    The effect of electric field and molecular ratio CR (NaF/AlF3) on basic structure and transport properties of NaFsbnd AlF3 molten salts were investigated by molecular dynamics simulations with the Buckingham potential model. The [AlF6]3- groups are the dominant specie in NaFsbnd AlF3 molten salts at CR ≥ 2.6, and followed by the [AlF5]2- groups, while CR ≤ 2.4, [AlF5]2- groups are the protagonists up to 40%. In NaFsbnd AlF3 system, with the increase of CR, the proportion of Fb decreases slightly and the percentage of Ff increases dramatically. The Alsbnd F bonds have ionic characters as well as partial covalently characters due to the hybridization of F-2p and Al-3s, 3p orbitals. The order of ion diffusion ability follows as Na+ > F- > Al3+. Adding more NaF can break some F bridges of structure networks and decrease the polymerization degree of NaFsbnd AlF3 molten salts, the viscosity reduces and ionic conductivity increases as a consequence. The calculated results of ionic conductivity are in agreement with the experimental results. Electric field has no significant impact on the local structure characters, while transport properties are not. The change of CR (NaF/AlF3) can significantly affect these characters of both the structure and transport.