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Sample records for based msr molten-salt

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

    International Nuclear Information System (INIS)

    Brossard, Ph.; Garzenne, C.; Mouney, H.

    2002-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Huntley, W.R.; Silverman, M.D.

    1976-11-01

    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-BeF/sub 2/-ThF/sub 4/-UF/sub 4/ fuel salt at velocities up to 6 m/s (20 fps) and at salt temperatures from 566 to 705/sup 0/C (1050 to 1300/sup 0/F). 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.

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

    International Nuclear Information System (INIS)

    Huntley, W.R.; Silverman, M.D.

    1976-11-01

    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-BeF 2 -ThF 4 -UF 4 fuel salt at velocities up to 6 m/s (20 fps) and at salt temperatures from 566 to 705 0 C (1050 to 1300 0 F). 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

  4. Thorium Molten Salt Nuclear Energy Synergetic System (THORIMS-NES)

    International Nuclear Information System (INIS)

    Yoshioka, Ritsuo; Mitachi, Koshi

    2013-01-01

    The authors have been promoting nuclear energy technology based on thorium molten salt as Thorium Molten Salt Nuclear Energy Synergetic System (THORIMS-NES). This system is a combination of fission power reactor of Molten Salt Reactor (MSR), and Accelerator Molten Salt Breeder (AMSB) for production of fissile 233 U with connecting chemical processing facility. In this paper, concept of THORIMS-NES, advantages of thorium and molten salt recent MSR design results such as FUJI-U3 using 233 U fuel, FUJI-Pu, large sized super-FUJI, pilot plant miniFUJI, AMSB, and chemical processing facility are described. (author)

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

    International Nuclear Information System (INIS)

    Jeong, Yongjin; Choi, Sooyoung; Lee, Deokjung

    2014-01-01

    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

  6. A general overview of generation IV molten salt reactor (MSR) and the use of thorium as fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yamaguchi, Carlos H.; Stefani, Giovanni L.; Santos, Thiago A., E-mail: carlos.yamaguchi@usp.br, E-mail: giovanni.stefani@ipen.br, E-mail: thiago.santos@ufabc.edu.br [Universidade de Sao Paulo (USP), SP (Brazil). Instituto de Fisica; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais Aplicadas

    2017-07-01

    The molten salt reactors (MSRs) make use of fluoride salt as primary cooler, at low pressure. Although considered a generation IV reactor, your concept isn't new, since in the 1960 years the Oak Ridge National Laboratory created a little prototype of 8MWt. Over the 20{sup th} century, other countries, like UK, Japan, Russia, China and France also did research in the area, especially with the use of thorium as fuel. This goes with the fact that Brazil possess the biggest reserve of thorium in the world. In the center of nuclear engineering at IPEN is being created a study group connected to thorium reactors, which purpose is to investigate reactors using thorium to produce {sup 233}U and tailing burn, thus making the MSR using thorium as fuel, an object of study. This present work searches to do a general summary about the researches of MSR's, having as focus the utilization of thorium with the goal being to show it's efficiency and utilization is doable. (author)

  7. 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

  8. 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.

  9. Molten salt reactor related research in Switzerland

    International Nuclear Information System (INIS)

    Krepel, Jiri; Hombourger, Boris; Fiorina, Carlo

    2015-01-01

    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)

  10. Preparation and characterization of molten salt based nanothermic fluids with enhanced thermal properties for solar thermal applications

    International Nuclear Information System (INIS)

    Madathil, Pramod Kandoth; Balagi, Nagaraj; Saha, Priyanka; Bharali, Jitalaxmi; Rao, Peddy V.C.; Choudary, Nettem V.; Ramesh, Kanaparthi

    2016-01-01

    Highlights: • Prepared and characterized inorganic ternary molten salt based nanothermic fluids. • MoS 2 and CuO nanoparticles incorporated ternary molten salts have been prepared. • Thermal properties enhanced by the addition of MoS 2 and CuO nanoparticles. • The amount of nanoparticles has been optimized. - Abstract: In the current energy scenario, solar energy is attracting considerable attention as a renewable energy source with ample research and commercial opportunities. The novel and efficient technologies in the solar energy are directed to develop methods for solar energy capture, storage and utilization. High temperature thermal energy storage systems can deal with a wide range of temperatures and therefore they are highly recommended for concentrated solar power (CSP) applications. In the present study, a systematic investigation has been carried out to identify the suitable inorganic nanoparticles and their addition in the molten salt has been optimized. In order to enhance the thermo-physical properties such as thermal conductivity and specific heat capacity of molten salt based HTFs, we report the utilization of MoS 2 and CuO nanoparticles. The enhancement in the above mentioned thermo-physical properties has been demonstrated for optimized compositions and the morphologies of nanoparticle-incorporated molten salts have been studied by scanning electron microscopy (SEM). Nanoparticle addition to molten salts is an efficient method to prepare thermally stable molten salt based heat transfer fluids which can be used in CSP plants. It is also observed that the sedimentation of nanoparticles in molten salt is negligible compared to that in organic heat transfer fluids.

  11. Molten salt based nanofluids based on solar salt and alumina nanoparticles: An industrial approach

    Science.gov (United States)

    Muñoz-Sánchez, Belén; Nieto-Maestre, Javier; Guerreiro, Luis; Julia, José Enrique; Collares-Pereira, Manuel; García-Romero, Ana

    2017-06-01

    Thermal Energy Storage (TES) and its associated dispatchability is extremely important in Concentrated Solar Power (CSP) plants since it represents the main advantage of CSP technology in relation to other renewable energy sources like photovoltaic (PV). Molten salts are used in CSP plants as a TES material because of their high operational temperature and stability of up to 600°C. Their main problems are their relative poor thermal properties and energy storage density. A simple cost-effective way to improve the thermal properties of molten salts is to dope them with nanoparticles, thus obtaining the so-called salt-based nanofluids. Additionally, the use of molten salt based nanofluids as TES materials and Heat Transfer Fluid (HTF) has been attracting great interest in recent years. The addition of tiny amounts of nanoparticles to the base salt can improve its specific heat as shown by different authors1-3. The application of these nano-enhanced materials can lead to important savings on the investment costs in new TES systems for CSP plants. However, there is still a long way to go in order to achieve a commercial product. In this sense, the improvement of the stability of the nanofluids is a key factor. The stability of nanofluids will depend on the nature and size of the nanoparticles, the base salt and the interactions between them. In this work, Solar Salt (SS) commonly used in CSP plants (60% NaNO3 + 40% KNO3 wt.) was doped with alumina nanoparticles (ANPs) at a solid mass concentration of 1% wt. at laboratory scale. The tendency of nanoparticles to agglomeration and sedimentation is tested in the molten state by analyzing their size and concentration through the time. The specific heat of the nanofluid at 396 °C (molten state) is measured at different times (30 min, 1 h, 5 h). Further research is needed to understand the mechanisms of agglomeration. A good understanding of the interactions between the nanoparticle surface and the ionic media would provide

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

    International Nuclear Information System (INIS)

    Medina C, D.; Hernandez A, P.; Letechipia de L, C.; Vega C, H. R.; Sajo B, L.

    2015-09-01

    This paper presents the design of a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a 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 252 Cf source, also there are two irradiation channels and the other six contain a molten salt ( 7 LiF - BeF 2 - ThF 4 - UF 4 ) as fuel. For the design the k 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 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)

  13. A basic study on fluoride-based molten salt electrolysis technology

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Il Soon [Seoul National University, Seoul (Korea); Kim, Kwang Bum [Yonsei University, Seoul (Korea); Park, Byung Gi [Seoul National University, Seoul (Korea)

    2001-04-01

    The objective of this project is to study on the physicochemical properties of fluoride molten salt, to develop numerical model for simulation of molten salt electrolysis, and to establish experimental technique of fluoride molten salt. Physicochemical data of fluoride molten salt are investigated and summarized. The numerical model, designated as REFIN is developed with diffusion-layer theory and electrochemical reaction kinetics. REFIN is benchmarked with published experimental data. REFIN has a capability to simulate multicomponent electrochemical system at transient conditions. Experimental device is developed to measure electrochemical properties of structural material for fluoride molten salt. Ni electrode is measured with cyclic voltammogram in the conditions of 600 .deg. C LiF-BeF{sub 2} and 700 .deg. C LiF-BeF{sub 2}. 74 refs., 23 figs., 57 tabs. (Author)

  14. A calculational procedure for neutronic and depletion analysis of Molten-Salt reactors based on SCALE6/TRITON

    International Nuclear Information System (INIS)

    Sheu, R.J.; Chang, J.S.; Liu, Y.-W. H.

    2011-01-01

    Molten-Salt Reactors (MSRs) represent one of the selected categories in the GEN-IV program. This type of reactor is distinguished by the use of liquid fuel circulating in and out of the core, which makes it possible for online refueling and salt processing. However, this operation characteristic also complicates the modeling and simulation of reactor core behaviour using conventional neutronic codes. The TRITON sequence in the SCALE6 code system has been designed to provide the combined capabilities of problem-dependent cross-section processing, rigorous treatment of neutron transport, and coupled with the ORIGEN-S depletion calculations. In order to accommodate the simulation of dynamic refueling and processing scheme, an in-house program REFRESH together with a run script are developed for carrying out a series of stepwise TRITON calculations, that makes the work of analyzing the neutronic properties and performance of a MSR core design easier. As a demonstration and cross check, we have applied this method to reexamine the conceptual design of Molten Salt Actinide Recycler & Transmuter (MOSART). This paper summarizes the development of the method and preliminary results of its application on MOSART. (author)

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

    International Nuclear Information System (INIS)

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

    1994-01-01

    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 3 -containing molten salts

  16. The analysis of the initiating events in thorium-based molten salt reactor

    International Nuclear Information System (INIS)

    Zuo Jiaxu; Song Wei; Jing Jianping; Zhang Chunming

    2014-01-01

    The initiation events analysis and evaluation were the beginning of nuclear safety analysis and probabilistic safety analysis, and it was the key points of the nuclear safety analysis. Currently, the initiation events analysis method and experiences both focused on water reactor, but no methods and theories for thorium-based molten salt reactor (TMSR). With TMSR's research and development in China, the initiation events analysis and evaluation was increasingly important. The research could be developed from the PWR analysis theories and methods. Based on the TMSR's design, the theories and methods of its initiation events analysis could be researched and developed. The initiation events lists and analysis methods of the two or three generation PWR, high-temperature gascooled reactor and sodium-cooled fast reactor were summarized. Based on the TMSR's design, its initiation events would be discussed and developed by the logical analysis. The analysis of TMSR's initiation events was preliminary studied and described. The research was important to clarify the events analysis rules, and useful to TMSR's designs and nuclear safety analysis. (authors)

  17. Molten salt reactor concept

    International Nuclear Information System (INIS)

    Sood, D.D.

    1980-01-01

    Molten salt reactor is an advanced breeder concept which is suited for the utilization of thorium for nuclear power production. This reactor is based on the use of solutions of uranium or plutonium fluorides in LiF-BeF 2 -ThF 4 as fuel. Unlike the conventional reactors, no external coolant is used in the reactor core and the fuel salt itself is circulated through heat exchangers to transfer the fission produced heat to a secondary salt (NaF-NaBF 4 ) for steam generation. A part of the fuel stream is continuously processed to isolate 233 Pa, so that it can decay to fissile 233 U without getting converted to 234 Pa, and for the removal of neutron absorbing fission products. This on-line processing scheme makes this reactor concept to achieve a breeding ratio of 1.07 which is the highest for any thermal breeder reactor. Experimental studies at the Bhabha Atomic Research Centre, Bombay, have established the use of plutonium as fuel for this reactor. This molten salt reactor concept is described and the work conducted at the Bhabha Atomic Research Centre is summarised. (auth.)

  18. Absorption and desorption of SO2 in aqueous solutions of diamine-based molten salts.

    Science.gov (United States)

    Lim, Seung Rok; Hwang, Junhyeok; Kim, Chang Soo; Park, Ho Seok; Cheong, Minserk; Kim, Hoon Sik; Lee, Hyunjoo

    2015-05-30

    SO2 absorption and desorption behaviors were investigated in aqueous solutions of diamine-derived molten salts with a tertiary amine group on the cation and a chloride anion, including butyl-(2-dimethylaminoethyl)-dimethylammonium chloride ([BTMEDA]Cl, pKb=8.2), 1-butyl-1,4-dimethylpiperazinium chloride ([BDMP]Cl, pKb=9.8), and 1-butyl-4-aza-1-azoniabicyclo[2,2,2]octane chloride ([BDABCO]Cl, pKb=11.1). The SO2 absorption and desorption performance of the molten salt were greatly affected by the basicity of the molten salt. Spectroscopic, X-ray crystallographic, and computational results for the interactions of SO2 with molten salts suggest that two types of SO2-containg species could be generated depending on the basicity of the unquaternized amino group: a dicationic species comprising two different anions, HSO3(-) and Cl(-), and a monocationic species bearing Cl(-) interacting with neutral H2SO3. Copyright © 2015 Elsevier B.V. All rights reserved.

  19. Thorium and Molten Salt Reactors: Essential Questions for Classroom Discussions

    Science.gov (United States)

    DiLisi, Gregory A.; Hirsch, Allison; Murray, Meredith; Rarick, Richard

    2018-04-01

    A little-known type of nuclear reactor called the "molten salt reactor" (MSR), in which nuclear fuel is dissolved in a liquid carrier salt, was proposed in the 1940s and developed at the Oak Ridge National Laboratory in the 1960s. Recently, the MSR has generated renewed interest as a remedy for the drawbacks associated with conventional uranium-fueled light-water reactors (LWRs) in use today. Particular attention has been given to the "thorium molten salt reactor" (TMSR), an MSR engineered specifically to use thorium as its fuel. The purpose of this article is to encourage the TPT community to incorporate discussions of MSRs and the thorium fuel cycle into courses such as "Physics and Society" or "Frontiers of Physics." With this in mind, we piloted a pedagogical approach with 27 teachers in which we described the underlying physics of the TMSR and posed five essential questions for classroom discussions. We assumed teachers had some preexisting knowledge of nuclear reactions, but such prior knowledge was not necessary for inclusion in the classroom discussions. Overall, our material was perceived as a real-world example of physics, fit into a standards-based curriculum, and filled a need in the teaching community for providing unbiased references of alternative energy technologies.

  20. R and D of On-line Reprocessing Technology for Molten-Salt Reactor Systems

    International Nuclear Information System (INIS)

    Uhlir, Jan; Tulackova, Radka; Chuchvalcova Bimova, Karolina

    2006-01-01

    The Molten Salt Reactor (MSR) represents one of promising future nuclear reactor concept included in the Generation IV reactors family. The reactor can be operated as the thorium breeder or as the actinide transmuter. However, the future deployment of Molten-Salt Reactors will be significantly dependent on the successful mastering of advanced reprocessing technologies dedicated to their fuel cycle. Here the on-line reprocessing technology connected with the fuel circuit of MSR is of special importance because the reactor cannot be operated for a long run without the fuel salt clean-up. Generally, main MSR reprocessing technologies are pyrochemical, majority of them are fluoride technologies. The proposed flow-sheets of MSR on-line reprocessing are based on a combination of molten-salt / liquid metal extraction and electro-separation processes, which can be added to the gas extraction process already verified during the MSRE project in ORNL. The crucial separation method proposed for partitioning of actinides from fission products is based on successive Anodic dissolution and Cathodic deposition processes in molten fluoride media. (authors)

  1. Advanced Instrumentation for Molten Salt Flow Measurements at NEXT

    Science.gov (United States)

    Tuyishimire, Olive

    2017-09-01

    The Nuclear Energy eXperiment Testing (NEXT) Lab at Abilene Christian University is building a Molten Salt Loop to help advance the technology of molten salt reactors (MSR). NEXT Lab's aim is to be part of the solution for the world's top challenges by providing safe, clean, and inexpensive energy, clean water and medical Isotopes. Measuring the flow rate of the molten salt in the loop is essential to the operation of a MSR. Unfortunately, there is no flow meter that can operate in the high temperature and corrosive environment of a molten salt. The ultrasonic transit time method is proposed as one way to measure the flow rate of high temperature fluids. Ultrasonic flow meter uses transducers that send and receive acoustic waves and convert them into electrical signals. Initial work presented here focuses on the setup of ultrasonic transducers. This presentation is the characterization of the pipe-fluid system with water as a baseline for future work.

  2. Studies on components for a molten salt reactor

    International Nuclear Information System (INIS)

    Nejedly, M.; Matal, O.

    2003-01-01

    The aim is contribute to a design of selected components of molten salt reactors with fuel in the molten fluoride salt matrix. Molten salt reactors (MSRs) permit the utilization of plutonium and minor actinides as new nuclear fuel from a traditional nuclear power station with production of electric energy. Results of preliminary feasibility studies of an intermediate heat exchanger, a small power molten salt pump and a modular conception of a steam generator for a demonstration unit of the MSR (30 MW) are summarized. (author)

  3. Characterization and Thermal Properties of Nitrate Based Molten Salt for Heat Recovery System

    Science.gov (United States)

    Faizal Tukimon, Mohd; Muhammad, Wan Nur Azrina Wan; Nor Annuar Mohamad, Md; Yusof, Farazila

    2017-10-01

    Molten salt can acts like a storage medium or heat transfer fluid in heat recovery system. Heat transfer fluid is a fluid that has the capability to deliver heat this one side to another while heat recovery system is a system that transfers heat to produce energy. This studies shows about determining the new formulation of different molten nitrate/nitrite salts consisting of LiNO3, KNO2, KNO3 and NaNO2 that give a low temperature of melting point and high average specific heat capacity. Mixed alkaline molten nitrate/nitrite salt can act as a heat transfer fluid due to their advantageous in terms of its properties that feasible in heat recovery system such as high specific heat capacity, low vapour pressure, low cost and wide range of temperature in its application. The mixing of these primary substances will form a new line of quaternary nitrate salt (LiNO3 - KNO2 - KNO3 - NaNO2). The quaternary mixture was heated inside the box furnace at 150°C for four hours and rose up the temperature to 400°C for eight hours to homogenize the mixture. Through heating process, the elements of nitrate/nitrite base were mixed completely. The temperature was then reduced to 115°C for several hours before removing the mixture from the furnace. The melting point of each sample were testified by using thermal gravimetric analysis, TGA/DTA and experiment of determining the specific heat capacity were conducted by using Differential Scanning Calorimeter, DSC. From the result, it is found that the melting point Sample 1 with percentage of weightage (25.4wt% of LiNO3, 33.8wt% of KNO2, 20.7wt% of KNO3 and 20.1wt% of NaNO2) is 94.4°C whereas the average specific heat capacity was 1.0484/g°C while for Sample 3 with percentages of weightage (30.0wt% of LiNO3, 50.2wt% of KNO2, 3.1wt% of KNO3 and 16.7wt% of NaNO2), the melting point is 86.1°C with average specific heat capacity of 0.7274 J/g°C. In the nut shell, the quaternary mixture salts had been a good mixture with good thermal

  4. 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.

  5. Recommendations for a restart of Molten Salt Reactor development

    International Nuclear Information System (INIS)

    Moir, R. W.

    2007-01-01

    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 2 32U 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

  6. MODERATOR TO FUEL RATIO AND URANIUM FRACTION ANALYSIS OF SQUARE LATTICE MOLTEN SALT TRANSATOMIC POWER

    OpenAIRE

    Dion Bagus Nugraha B; Andang Widi Harto; Sihana Sihana

    2017-01-01

    Molten Salt Reactor Transatomic Power (MSR TAP) is a further development of the nuclear reactor Generation IV Reactor Molten Salt Reactor (MSR). MSR TAP generates clean electric power. It has a passive safety, resistance to proliferation, and low cost. MSR TAP can consume the rest of the nuclear fuel/spent nuclear fuel (SNF) of a commercial Light Water Reactor (LWR) fuel or use the main fuel, a salt solution UF4 - LiF - BeF2. MSR TAP uses Zirconium Hydride material for the moderator. This res...

  7. 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

  8. 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

  9. 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

  10. 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.

  11. Molten salt treatment to minimize and optimize waste

    International Nuclear Information System (INIS)

    Gat, U.; Crosley, S.M.; Gay, R.L.

    1993-01-01

    A combination molten salt oxidizer (MSO) and molten salt reactor (MSR) is described for treatment of waste. The MSO is proposed for contained oxidization of organic hazardous waste, for reduction of mass and volume of dilute waste by evaporation of the water. The NTSO residue is to be treated to optimize the waste in terms of its composition, chemical form, mixture, concentration, encapsulation, shape, size, and configuration. Accumulations and storage are minimized, shipments are sized for low risk. Actinides, fissile material, and long-lived isotopes are separated and completely burned or transmuted in an MSR. The MSR requires no fuel element fabrication, accepts the materials as salts in arbitrarily small quantities enhancing safety, security, and overall acceptability

  12. 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.

  13. Electrolysis of metal oxides in MgCl2 based molten salts with an inert graphite anode.

    Science.gov (United States)

    Yuan, Yating; Li, Wei; Chen, Hualin; Wang, Zhiyong; Jin, Xianbo; Chen, George Z

    2016-08-15

    Electrolysis of solid metal oxides has been demonstrated in MgCl2-NaCl-KCl melt at 700 °C taking the electrolysis of Ta2O5 as an example. Both the cathodic and anodic processes have been investigated using cyclic voltammetry, and potentiostatic and constant voltage electrolysis, with the cathodic products analysed by XRD and SEM and the anodic products by GC. Fast electrolysis of Ta2O5 against a graphite anode has been realized at a cell voltage of 2 V, or a total overpotential of about 400 mV. The energy consumption was about 1 kW h kgTa(-1) with a nearly 100% Ta recovery. The cathodic product was nanometer Ta powder with sizes of about 50 nm. The main anodic product was Cl2 gas, together with about 1 mol% O2 gas and trace amounts of CO. The graphite anode was found to be an excellent inert anode. These results promise an environmentally-friendly and energy efficient method for metal extraction by electrolysis of metal oxides in MgCl2 based molten salts.

  14. Effect of nanoparticles on heat capacity of nanofluids based on molten salts as PCM for thermal energy storage.

    Science.gov (United States)

    Chieruzzi, Manila; Cerritelli, Gian F; Miliozzi, Adio; Kenny, José M

    2013-10-29

    In this study, different nanofluids with phase change behavior were developed by mixing a molten salt base fluid (selected as phase change material) with nanoparticles using the direct-synthesis method. The thermal properties of the nanofluids obtained were investigated. These nanofluids can be used in concentrating solar plants with a reduction of storage material if an improvement in the specific heat is achieved. The base salt mixture was a NaNO3-KNO3 (60:40 ratio) binary salt. The nanoparticles used were silica (SiO2), alumina (Al2O3), titania (TiO2), and a mix of silica-alumina (SiO2-Al2O3). Three weight fractions were evaluated: 0.5, 1.0, and 1.5 wt.%. Each nanofluid was prepared in water solution, sonicated, and evaporated. Measurements on thermophysical properties were performed by differential scanning calorimetry analysis and the dispersion of the nanoparticles was analyzed by scanning electron microscopy (SEM). The results obtained show that the addition of 1.0 wt.% of nanoparticles to the base salt increases the specific heat of 15% to 57% in the solid phase and of 1% to 22% in the liquid phase. In particular, this research shows that the addition of silica-alumina nanoparticles has a significant potential for enhancing the thermal storage characteristics of the NaNO3-KNO3 binary salt. These results deviated from the predictions of the theoretical model used. SEM suggests a greater interaction between these nanoparticles and the salt.

  15. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    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-BeF 2 -ThF 4 -UF 4 ) 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 [fr

  16. Thermal-hydraulics of internally heated molten salts and application to the Molten Salt Fast Reactor

    Science.gov (United States)

    Fiorina, Carlo; Cammi, Antonio; Luzzi, Lelio; Mikityuk, Konstantin; Ninokata, Hisashi; Ricotti, Marco E.

    2014-04-01

    The Molten Salt Reactors (MSR) are an innovative kind of nuclear reactors and are presently considered in the framework of the Generation IV International Forum (GIF-IV) for their promising performances in terms of low resource utilization, waste minimization and enhanced safety. A unique feature of MSRs is that molten fluoride salts play the distinctive role of both fuel (heat source) and coolant. The presence of an internal heat generation perturbs the temperature field and consequences are to be expected on the heat transfer characteristics of the molten salts. In this paper, the problem of heat transfer for internally heated fluids in a straight circular channel is first faced on a theoretical ground. The effect of internal heat generation is demonstrated to be described by a corrective factor applied to traditional correlations for the Nusselt number. It is shown that the corrective factor can be fully characterized by making explicit the dependency on Reynolds and Prandtl numbers. On this basis, a preliminary correlation is proposed for the case of molten fluoride salts by interpolating the results provided by an analytic approach previously developed at the Politecnico di Milano. The experimental facility and the related measuring procedure for testing the proposed correlation are then presented. Finally, the developed correlation is used to carry out a parametric investigation on the effect of internal heat generation on the main out-of-core components of the Molten Salt Fast Reactor (MSFR), the reference circulating-fuel MSR design in the GIF-IV. The volumetric power determines higher temperatures at the channel wall, but the effect is significant only in case of large diameters and/or low velocities.

  17. 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.

  18. 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)

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

    International Nuclear Information System (INIS)

    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

  20. 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)

    Nuttin, Alexis

    2002-01-01

    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 232 Th/ 233 U 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)

  1. The case for the thorium molten salt reactor

    Science.gov (United States)

    Greaves, E. D.; Furukawa, K.; Sajo-Bohus, L.; Barros, H.

    2012-02-01

    Shortcomings of current PWR and BWR, solid uranium-fuel, nuclear power reactors are summarized. It is shown how the Molten Salt Reactor (MSR) created and operated at Oak Ridge National Laboratory (ORNL), USA (1960s-1970s) and developed as FUJI reactor by Furukawa and collaborators (1980s-1990s), addresses all of these shortcomings. Relevant properties of the MSR regarding to simplicity, its impact on capital and operating costs, safety, waste product production, waste reprocessing, power efficiency and non proliferation properties are reviewed. The Thorium MSR within the THORIMS-NES fuel cycle system is described concluding that the superior properties of the MSR make this the technology of choice to provide the required future energy in the South American region.

  2. Molten salt extraction process for the recovery of valued transition metals from land-based and deep-sea minerals

    Science.gov (United States)

    Maroni, Victor A.; von Winbush, Samuel

    1988-01-01

    A process for extracting transition metals and particularly cobalt and manganese together with iron, copper and nickel from low grade ores (including ocean-floor nodules) by converting the metal oxides or other compositions to chlorides in a molten salt, and subsequently using a combination of selective distillation at temperatures below about 500.degree. C., electrolysis at a voltage not more negative than about -1.5 volt versus Ag/AgCl, and precipitation to separate the desired manganese and cobalt salts from other metals and provide cobalt and manganese in metallic forms or compositions from which these metals may be more easily recovered.

  3. Dynamics and control of molten-salt breeder reactor

    Directory of Open Access Journals (Sweden)

    Vikram Singh

    2017-08-01

    Full Text Available Preliminary results of the dynamic analysis of a two-fluid molten-salt breeder reactor (MSBR system are presented. Based on an earlier work on the preliminary dynamic model of the concept, the model presented here is nonlinear and has been revised to accurately reflect the design exemplified in ORNL-4528. A brief overview of the model followed by results from simulations performed to validate the model is presented. Simulations illustrate stable behavior of the reactor dynamics and temperature feedback effects to reactivity excursions. Stable and smooth changes at various nodal temperatures are also observed. Control strategies for molten-salt reactor operation are discussed, followed by an illustration of the open-loop load-following capability of the molten-salt breeder reactor system. It is observed that the molten-salt breeder reactor system exhibits “self-regulating” behavior, minimizing the need for external controller action for load-following maneuvers.

  4. Molten-salt reactor information system

    International Nuclear Information System (INIS)

    Haubenreich, P.N.; Cardwell, D.W.; Engel, J.R.

    1975-06-01

    The Molten-Salt Reactor Information System (MSRIS) is a computer-based file of abstracts of documents dealing with the technology of molten-salt reactors. The file is stored in the IBM-360 system at ORNL, and may be searched through the use of established interactive computer programs from remote terminals connected to the computer via telephone lines. The system currently contains 373 entries and is subject to updating and expansion as additional information is developed. The nature and general content of the data file, a general approach for obtaining information from it, and the manner in which material is added to the file are described. Appendixes provide the list of keywords currently in use, the subject categories under which information is filed, and simplified procedures for searching the file from remote terminals. (U.S.)

  5. Validation of the TRACE code for the system dynamic simulations of the molten salt reactor experiment and the preliminary study on the dual fluid molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    He, Xun

    2016-06-14

    Molten Salt Reactor (MSR), which was confirmed as one of the six Generation IV reactor types by the GIF (Generation IV International Forum in 2008), recently draws a lot of attention all around the world. Due to the application of liquid fuels the MSR can be regarded as the most special one among those six GEN-IV reactor types in a sense. A unique advantage of using liquid nuclear fuel lies in that the core melting accident can be thoroughly eliminated. Besides, a molten salt reactor can have several fuel options, for instance, the fuel can be based on {sup 235}U, {sup 232}Th-{sup 233}U, {sup 238}U-{sup 239}Pu cycle or even the spent nuclear fuel (SNF), so the reactor can be operated as a breeder or as an actinides burner both with fast, thermal or epi-thermal neutron spectrum and hence, it has excellent features of the fuel sustainability and for the non-proliferation. Furthermore, the lower operating pressure not only means a lower risk of the explosion as well as the radioactive leakage but also implies that the reactor vessel and its components can be lightweight, thus lowering the cost of equipments. So far there is no commercial MSR being operated. However, the MSR concept and its technical validation dates back to the 1960s to 1970s, when the scientists and engineers from ORNL (Oak Ridge National Laboratory) in the United States managed to build and run the world's first civilian molten salt reactor called MSRE (Molten Salt Reactor Experiment). The MSRE was an experimental liquid-fueled reactor with 10 MW thermal output using {sup 4}LiF-BeF{sub 2}-ZrF{sub 4}-UF{sub 4} as the fuel also as the coolant itself. The MSRE is usually taken as a very important reference case for many current researches to validate their codes and simulations. Without exception it works also as a benchmark for this thesis. The current thesis actually consists of two main parts. The first part is about the validation of the current code for the old MSRE concept, while the second

  6. 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.

  7. Toward Cost-Effective Manufacturing of Silicon Solar Cells: Electrodeposition of High-Quality Si Films in a CaCl2-based Molten Salt.

    Science.gov (United States)

    Yang, Xiao; Ji, Li; Zou, Xingli; Lim, Taeho; Zhao, Ji; Yu, Edward T; Bard, Allen J

    2017-11-20

    Electrodeposition of Si films from a Si-containing electrolyte is a cost-effective approach for the manufacturing of solar cells. Proposals relying on fluoride-based molten salts have suffered from low product quality due to difficulties in impurity control. Here we demonstrate the successful electrodeposition of high-quality Si films from a CaCl 2 -based molten salt. Soluble Si IV -O anions generated from solid SiO 2 are electrodeposited onto a graphite substrate to form a dense film of crystalline Si. Impurities in the deposited Si film are controlled at low concentrations (both B and P are less than 1 ppm). In the photoelectrochemical measurements, the film shows p-type semiconductor character and large photocurrent. A p-n junction fabricated from the deposited Si film exhibits clear photovoltaic effects. This study represents the first step to the ultimate goal of developing a cost-effective manufacturing process for Si solar cells based on electrodeposition. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  8. High temperature in-situ synchrotron-based XRD study on the crystal structure evolution of C/C composite impregnated by FLiNaK molten salt.

    Science.gov (United States)

    Feng, Shanglei; Yang, Yingguo; Li, Li; Zhang, Dongsheng; Yang, Xinmei; Xia, Huihao; Yan, Long; Tsang, Derek K L; Huai, Ping; Zhou, Xingtai

    2017-09-06

    An in-situ real-time synchrotron-based grazing incidence X-ray diffraction was systematically used to investigate the crystal structural evolution of carbon fiber reinforced carbon matrix (C/C) composite impregnated with FLiNaK molten salt during the heat-treatment process. It was found that the crystallographic thermal expansion and contraction rate of interlayer spacing d 002 in C/C composite with FLiNaK salt impregnation is smaller than that in the virgin sample, indicating the suppression on interlayer spacing from FLiNaK salt impregnated. Meanwhile the crystallite size L C002 of C/C composite with FLiNaK salt impregnation is larger than the virgin one after whole heat treatment process, indicating that FLiNaK salt impregnation could facilitate the crystallization of C/C composite after heat treatment process. This improved crystallization in C/C composite with FLiNaK salt impregnation suggests the synthetic action of the salt squeeze effect on crooked carbon layer and the release of internal residual stress after heating-cooling process. Thus, the present study not only contribute to reveal the interaction mechanism between C/C composite and FLiNaK salt in high temperature environment, but also promote the design of safer and more reliable C/C composite materials for the next generation molten salt reactor.

  9. thermic oil and molten salt

    African Journals Online (AJOL)

    Boukelia T.E, Mecibah M.S and Laouafi A

    1 mai 2016 ... [27] Zavoico, AB. Solar Power Tower Design Basis Document. Tech. rep, Sandia National. Laboratories, SAND2001-2100, 2001. How to cite this article: Boukelia T.E, Mecibah M.S and Laouafi A. Performance simulation of parabolic trough solar collector using two fluids (thermic oil and molten salt).

  10. Molten salts as possible fuel fluids for TRU fuelled systems: ISTC no. 1606 approach

    International Nuclear Information System (INIS)

    Ignatiev, V.; Zakirov, R.; Grebenkine, K.

    2001-01-01

    The principle attraction of the molten salt reactor (MSR) technology is the use of fuel/fertile material flexibility (easy of fuel preparation and processing) for gaining additional profits as compared with solid materials. This approach presents important departures from traditional philosophy, applied in current nuclear power plants, and to some extent contradicts the straightforward interpretation of the defence-in-depth principal. Nevertheless we understand there may be potential to use MSR technology to support back end fuel cycle technologies in future commercial environment. The paper aims at reviewing results of the work performed in Russia, relevant to the problems of MSR technology development. Also this contribution aims at evaluation of remaining uncertainties for molten salt burner concept implementation. Fuel properties and behaviour, container materials, and clean-up of fuels with emphasis on experiments will be of priority. Recommendations are made regarding the types of experimental studies needed on a way to implement molten salt technology to the back-end of the fuel cycle. To better understand the potential and limitations of the molten salts as a fuel for reactor of incinerator type, Russian Institutes have submitted to the ISTC the Task no. 1606 Experimental Study of Molten Salt Technology for Safe and Low Waste Treatment of Plutonium and Minor Actinides in Accelerator Driven and Critical Systems. The project goals, technical approach and expected specific results are discussed. (author)

  11. Molten salt reactor as asymptotic safety nuclear system

    International Nuclear Information System (INIS)

    Novikov, V.M.; Ignatyev, V.V.

    1989-01-01

    Safety is becoming the main and priority problem of the nuclear power development. An increase of the active safety measures could hardly be considered as the proper way to achieve the asymptotically high level of nuclear safety. It seem that the more realistic way to achieve such a goal is to minimize risk factors and to maximize the use of inherent and passive safety properties. The passive inherent safety features of the liquid fuel molten salt reactor (MSR) technology are making it attractive for future energy generation. The achievement of the asymptotic safety in MSR is being connected with the minimization of such risk factors as a reactivity excess, radioactivity stored, decay heat, non nuclear energy stored in core. In this paper safety peculiarities of the different MSR concepts are discussed

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

    International Nuclear Information System (INIS)

    Lelek, Vladimir; Vocka, Radim

    2000-01-01

    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)

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

    International Nuclear Information System (INIS)

    Boussier, H.; Heuer, D.

    2010-01-01

    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 Reactor Fast (MSFR).

  14. {[1,4-DHPyrazine][C(CN)(3)](2)} as a New Nano Molten Salt Catalyst for the Synthesis of Novel Piperazine Based bis(4-hydroxy-2H-chromen-2-one) Derivatives

    NARCIS (Netherlands)

    Baghery, Saeed; Zolfigol, Mohammad Ali; Schirhagl, Romana; Hasani, Masoumeh

    In this article a convenient method for the synthesis of novel piperazine based bis(4-hydroxy-2H-chromen-2-one) derivatives using pyrazine-1,4-diium tricyanomethanide {[1,4-DHPyrazine][C(CN)(3)](2)} as a new nanostructured molten salt (NMS) catalyst has been described. These compounds were

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

    International Nuclear Information System (INIS)

    Wang Haiwei; Mei Longwei; Cai Xiangzhou; Chen Jingen; Guo Wei; Jiang Dazhen

    2013-01-01

    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 233 Th can convert to 233 Pa, which then undergoes beta decay to become 233 U. 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

  16. Application of lithium in molten-salt reduction processes

    International Nuclear Information System (INIS)

    Gourishankar, K. V.

    1998-01-01

    Metallothermic reductions have been extensively studied in the field of extractive metallurgy. At Argonne National Laboratory (ANL), we have developed a molten-salt based reduction process using lithium. This process was originally developed to reduce actinide oxides present in spent nuclear fuel. Preliminary thermodynamic considerations indicate that this process has the potential to be adapted for the extraction of other metals. The reduction is carried out at 650 C in a molten-salt (LiCl) medium. Lithium oxide (Li 2 O), produced during the reduction of the actinide oxides, dissolves in the molten salt. At the end of the reduction step, the lithium is regenerated from the salt by an electrowinning process. The lithium and the salt from the electrowinning are then reused for reduction of the next batch of oxide fuel. The process cycle has been successfully demonstrated on an engineering scale in a specially designed pyroprocessing facility. This paper discusses the applicability of lithium in molten-salt reduction processes with specific reference to our process. Results are presented from our work on actinide oxides to highlight the role of lithium and its effect on process variables in these molten-salt based reduction processes

  17. Improvements and validation of the transient analysis code MOREL for molten salt reactors

    International Nuclear Information System (INIS)

    Zhuang Kun; Zheng Youqi; Cao Liangzhi; Hu Tianliang; Wu Hongchun

    2017-01-01

    The liquid fuel salt used in the molten salt reactors (MSRs) serves as the fuel and coolant simultaneously. On the one hand, the delayed neutron precursors circulate in the whole primary loop and part of them decay outside the core. On the other hand, the fission heat is carried off directly by the fuel flow. These two features require new analysis method with the coupling of fluid flow, heat transfer and neutronics. In this paper, the recent update of MOREL code is presented. The update includes: (1) the improved quasi-static method for the kinetics equation with convection term is developed. (2) The multi-channel thermal hydraulic model is developed based on the geometric feature of MSR. (3) The Variational Nodal Method is used to solve the neutron diffusion equation instead of the original analytic basis functions expansion nodal method. The update brings significant improvement on the efficiency of MOREL code. And, the capability of MOREL code is extended for the real core simulation with feedback. The numerical results and experiment data gained from molten salt reactor experiment (MSRE) are used to verify and validate the updated MOREL code. The results agree well with the experimental data, which prove the new development of MOREL code is correct and effective. (author)

  18. Organic waste processing using molten salt oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, M. G., LLNL

    1998-03-01

    Molten Salt Oxidation (MSO) is a thermal means of oxidizing (destroying) the 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. The U. S. Department of Energy`s Office of Environmental Management (DOE/EM) is currently funding research that will identify alternatives to incineration for the treatment of organic-based mixed wastes. (Mixed wastes are defined as waste streams which have both hazardous and radioactive properties.) One such project is Lawrence Livermore National Laboratory`s Expedited Technology Demonstration of Molten Salt Oxidation (MSO). The goal of this project is to conduct an integrated demonstration of MSO, including off-gas and spent salt treatment, and the preparation of robust solid final forms. Livermore National Laboratory (LLNL) has constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are presently being performed under carefully controlled (experimental) conditions. The system consists of a MSO process vessel with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. In this paper we describe the integrated system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is to identify the most suitable waste streams and waste types for MSO treatment.

  19. Development of flexible support for molten salt reactor

    International Nuclear Information System (INIS)

    Xie, Mingqiang

    2014-01-01

    Supporting member design for equipment and pipes is the requisite factor to realize the concept. It's a challenge to design a reliable supporting structure in molten salt reactor (MSR) due to the extraordinary working temperature (max 750 deg. C). High temperature may cause large expansion and reduce the mechanical strength of material, The support is required both enough strength and flexibility. In this paper, an all-dimensional support was designed, the validation work was carried out on a high temperature test loop. The results indicate that the support has a good performance, it reduce the thermal stress effectively and support the equipment and pipes stably for one year. The support design has a significance referential meaning for MSR construction (authors)

  20. Steady State Analysis of Small Molten Salt Reactor : Effect of Fuel Salt Flow on Reactor Characteristics

    OpenAIRE

    YAMAMOTO, Takahisa; MITACHI, Koshi; SUZUKI, Takashi

    2005-01-01

    The Molten Salt Reactor (MSR) is a thermal neutron reactor with graphite moderation and operates on the thorium-uranium fuel cycle. The feature of the MSR is that fuel salt flows inside the reactor during the nuclear fission reaction. In the previous study, the authors developed numerical model with which to simulate the effects of fuel salt flow on the reactor characteristics. In this study, we apply the model to the steady-state analysis of a small MSR system and estimate the effects of fue...

  1. Nonmetal-metal transition in metal–molten-salt solutions

    NARCIS (Netherlands)

    Silvestrelli, P.-L.; Alavi, A.; Parrinello, M.; Frenkel, D.

    1996-01-01

    The method of ab initio molecular dynamics, based on finite-temperature density-functional theory, is used to study the nonmetal-metal transition in two different metal–molten-salt solutions, Kx(KCl)1-x and Nax(NaBr)1-x. As the excess metal concentration is increased the electronic density becomes

  2. Steady State Analysis of Small Molten Salt Reactor

    Science.gov (United States)

    Yamamoto, Takahisa; Mitachi, Koshi; Suzuki, Takashi

    The Molten Salt Reactor (MSR) is a thermal neutron reactor with graphite moderation and operates on the thorium-uranium fuel cycle. The feature of the MSR is that fuel salt flows inside the reactor during the nuclear fission reaction. In the previous study, the authors developed numerical model with which to simulate the effects of fuel salt flow on the reactor characteristics. In this study, we apply the model to the steady-state analysis of a small MSR system and estimate the effects of fuel flow. The model consists of two-group neutron diffusion equations for fast and thermal neutron fluxes, transport equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and the graphite moderator. The following results are obtained: (1) in the rated operation condition, the peaks of the neutron fluxes slightly move toward the bottom from the center of the reactor and the delayed neutron precursors are significantly carried by the fuel salt flow, and (2) the extension of residence time in the external-loop system and the rise of the fuel inflow temperature show weak negative reactivity effects, which decrease the neutron multiplication factor of the small MSR system.

  3. Molten salt safety study. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    The considerations concerning safety in using molten salt (40% potassium nitrate, 60% sodium nitrate) in a solar central receiver plant are addressed. The considerations are of a general nature and do not cover any details of equipment or plant operation. The study includes salt chemical reaction, experiments with molten salt, dry storage and handling constraints, and includes data from the National Fire Protection Association. The contents of this report were evaluated by two utility companies and they concluded that no major safety problems exist in using a molten salt solar system.

  4. Molten salt reactors: A new beginning for an old idea

    International Nuclear Information System (INIS)

    LeBlanc, David

    2010-01-01

    Molten salt reactors have seen a marked resurgence of interest over the past decade, highlighted by their inclusion as one of six Generation IV reactor types. The most active development period however was between the mid 1950s and early 1970s at Oak Ridge National Laboratories (ORNL) and any new re-examination of this concept must bear in mind the far different priorities then in place. High breeding ratios and short doubling times were paramount and this guided the evolution of the Molten Salt Breeder Reactor (MSBR) program. As the inherent advantages of the molten salt concept have become apparent to an increasing number of researchers worldwide it is important to not simply look to continue where ORNL left off but to return to basics in order to offer the best design using updated goals and abilities. A major potential change to the traditional Single Fluid, MSBR design and a subject of this presentation is a return to the mode of operation that ORNL proposed for the majority of its MSR program. That being the Two Fluid design in which separate salts are used for fissile 233 UF 4 and fertile ThF 4 . Oak Ridge abandoned this promising route due to what was known as the 'plumbing problem'. It will be shown that a simple yet crucial modification to core geometry can solve this problem and enable the many advantages of the Two Fluid design. In addition, another very promising route laid out by ORNL was simplified Single Fluid converter reactors that could obtain far superior lifetime uranium utilization than LWR or CANDU without the need for any fuel processing beyond simple chemistry control. Updates and potential improvements to this very attractive concept will also be explored.

  5. Molten salts processes and generic simulation

    International Nuclear Information System (INIS)

    Ogawa, Toru; Minato, Kazuo

    2001-01-01

    Development of dry separation process (pyrochemical process) using molten salts for the application of spent-nuclear fuel reprocessing requires a rather complete fundamental database as well as process simulation technique with wide applicability. The present report concerns recent progress and problems in this field taking behaviors of co-electrodeposition of UO 2 and PuO 2 in molten salts as an example, and using analytical simulation of local equilibrium combined with generic diffusion. (S. Ohno)

  6. Molten salts processes and generic simulation

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Toru; Minato, Kazuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    Development of dry separation process (pyrochemical process) using molten salts for the application of spent-nuclear fuel reprocessing requires a rather complete fundamental database as well as process simulation technique with wide applicability. The present report concerns recent progress and problems in this field taking behaviors of co-electrodeposition of UO{sub 2} and PuO{sub 2} in molten salts as an example, and using analytical simulation of local equilibrium combined with generic diffusion. (S. Ohno)

  7. Electrochemical ion separation in molten salts

    Energy Technology Data Exchange (ETDEWEB)

    Spoerke, Erik David; Ihlefeld, Jon; Waldrip, Karen; Wheeler, Jill S.; Brown-Shaklee, Harlan James; Small, Leo J.; Wheeler, David R.

    2017-12-19

    A purification method that uses ion-selective ceramics to electrochemically filter waste products from a molten salt. The electrochemical method uses ion-conducting ceramics that are selective for the molten salt cations desired in the final purified melt, and selective against any contaminant ions. The method can be integrated into a slightly modified version of the electrochemical framework currently used in pyroprocessing of nuclear wastes.

  8. Development of a safety analysis code for molten salt reactors

    International Nuclear Information System (INIS)

    Zhang Dalin; Qiu Suizheng; Su Guanghui

    2009-01-01

    The molten salt reactor (MSR) well suited to fulfill the criteria defined by the Generation IV International Forum (GIF) is presently revisited all around the world because of different attractive features of current renewed relevance. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this work, in particular, the attention is focused on the safety characteristic analysis of the MSRs, in which a point kinetic model considering the flow effects of the fuel salt is established for the MSRs and calculated by developing a microcomputer code coupling with a simplified heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the molten salt actinide recycler and transmuter system (MOSART) by simulating three types of basic transient conditions including the unprotected loss of flow, unprotected overcooling accident and unprotected transient overpower. Some reasonable results are obtained for the MOSART, which show that the MOSART conceptual design is an inherently stable reactor design. The present study provides some valuable information for the research and design of the new generation MSRs.

  9. Fuel cycle cost analysis on molten-salt reactors

    International Nuclear Information System (INIS)

    Shimazu, Yoichiro

    1976-01-01

    An evaluation is made of the fuel cycle costs for molten-salt reactors (MSR's), developed at Oak Ridge National Laboratory. Eight combinations of conditions affecting fuel cycle costs are compared, covering 233 U-Th, 235 U-Th and 239 Pu-Th fuels, with and without on-site continuous fuel reprocessing. The resulting fuel cycle costs range from 0.61 to 1.18 mill/kWh. A discussion is also given on the practicability of these fuel cycles. The calculations indicate that somewhat lower fuel cycle costs can be expected from reactor operation in converter mode on 235 U make-up with fuel reprocessed in batches every 10 years to avoid fission product precipitation, than from operation as 233 U-Th breeder with continuous reprocessing. (auth.)

  10. On-line reprocessing of a molten salt reactor: a simulation tool

    International Nuclear Information System (INIS)

    Simon, Nicole; Gastaldi, Olivier; Penit, Thomas; Cohin, Olivier; Campion, Pierre-Yves

    2008-01-01

    The molten salt reactor (MSR) is one of the concepts studied in the frame of GEN IV road-map. Due to the specific features of its liquid fuel, the reprocessing unit may be directly connected to the reactor. A modelling of this unit is presented. The final objective is to create a flexible computer reprocessing code which can use data from neutron calculations and can be coupled to a neutron code. Such a code allows the description of the whole behaviour of MSR, including, in a coupled manner, both the design of the core and the optimised reprocessing scheme effects. (authors)

  11. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-08-31

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

  12. Molten salts and nuclear energy production

    Science.gov (United States)

    Le Brun, Christian

    2007-01-01

    Molten salts (fluorides or chlorides) were considered near the beginning of research into nuclear energy production. This was initially due to their advantageous physical and chemical properties: good heat transfer capacity, radiation insensitivity, high boiling point, wide range solubility for actinides. In addition it was realised that molten salts could be used in numerous situations: high temperature heat transfer, core coolants with solid fuels, liquid fuel in a molten salt reactor, solvents for spent nuclear solid fuel in the case of pyro-reprocessing and coolant and tritium production in the case of fusion. Molten salt reactors, one of the six innovative concepts chosen by the Generation IV international forum, are particularly interesting for use as either waste incinerators or thorium cycle systems. As the neutron balance in the thorium cycle is very tight, the possibility to perform online extraction of some fission product poisons from the salt is very attractive. In this article the most important questions that must be addressed to demonstrate the feasibility of molten salt reactor will be reviewed.

  13. Establishment of a room temperature molten salt capability to measure fundamental thermodynamic properties of actinide elements

    International Nuclear Information System (INIS)

    Smith, W.H.; Costa, D.A.

    1998-01-01

    This is the final report of a six-month, Laboratory Directed Research and Development (LDRD) project at Los Alamos National Laboratory (LANL). The goal of this work was to establish a capability for the measurement of fundamental thermodynamic properties of actinide elements in room temperature molten salts. This capability will be used to study in detail the actinide chloro- and oxo- coordination chemistries that dominate in the chloride-based molten salt media. Uranium will be the first actinide element under investigation

  14. Thorium Molten-Salt Nuclear Energy Synergetics

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Lecocq, A.; Kato, Yoshio; Mitachi, Kohshi.

    1990-01-01

    In the next century, the 'fission breeder' concept will not be practical to solve the global energy problems, including environmental and North-South problems. As a new measure, a simple rational Th molten salt breeding fuel cycle system, named 'Thorium Molten-Salt Nuclear Energy Synergetics (THORIMS-NES)', which composed of simple power stations and fissile producers, is proposed. This is effective to establish the essential improvement in issues of resources, safety, power-size flexibility, anti-nuclear proliferation and terrorism, radiowaste, economy, etc. securing the simple operation, maintenance, chemical processing, and rational breeding fuel cycle. As examples, 155 MWe fuel self-sustaining power station 'FUJI-II', 7 MWe pilot-plant 'miniFUJI-II', 1 GeV-300 mA proton Accelerator Molten-Salt Breeder 'AMSB', and their combined fuel cycle system are explained. (author)

  15. Metal Oxide Solubility and Molten Salt Corrosion.

    Science.gov (United States)

    1982-03-29

    soluble oxides that relations like eq. (3) are significant. The oxides of several metal oxides have been found to be amphoteric , i.e., their solution...METAL OXIDE SOLUBILITY AND MOLTEN SALT CORROSION.(U) MAR 82 K H STERN UNCLASSI E DL R L-4772NL EL .2. MICROCOPY RESOLUTION TEST CHART NATIONAL BURALU...METAL OXIDE SOLUBILITY AND MOLTEN SALT Interim report on a continuing CORROSION NRL problem. S. PERFORMING a4. REPORT NUMlER 7. AuTtwORr) S. CONTRACT OR

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

    Science.gov (United States)

    Sooby, E.; Baty, A.; Beneš, O.; McIntyre, P.; Pogue, N.; Salanne, M.; Sattarov, A.

    2013-09-01

    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. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated. A special thanks is due to Prof. Paul Madden for introducing the ADSMS group to the concept of using the molten salt as the spallation target, rather than a conventional heavy metal spallation target. This feature helps to optimize this core as a Pu/TRU burner.

  17. Design of a helium-cooled molten salt fusion breeder

    Science.gov (United States)

    Moir, R. W.; Lee, J. D.; Fulton, F. J.; Huegel, F.; Neef, W. S., Jr.; Sherwood, A. E.; Berwald, D. H.; Whitley, R. H.; Wong, C. P. C.; Devan, J. H.

    1985-02-01

    A new conceptual blanket design for a fusion reactor is discussed which produces fissile material for fission power plants. Fission is suppressed by using beryllium, rather than uranium, to multiply neutrons and also by minimizing the fissile inventory. The molten-salt breeding media (LiF + BeF2 + TghF4) is circulated through the blanket and on to the online processing system where (233)U and tritium are continuously removed. Helium cools the blanket including the steel pipes containing the molten salt. Austenitic steel was chosen because of its ease of fabrication, adequate radiation-damage lifetime, and low corrosion rate by molten salt. Safety is enhanced because the afterheat is low and the blanket materials do not react with air or water. The fusion breeder based on a pre-MARS mirror is estimated to cost $4.9B or 2.35 time an LWR of the same power. The estimated present value cost of the (233)U produced is $40/g if utility financed or $16/g if government financed.

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

    International Nuclear Information System (INIS)

    Wang, Jiangmeng; Cao, Xinrong

    2015-01-01

    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

  19. The molten salt reactor option for beneficial use of fissile material from dismantled weapons

    International Nuclear Information System (INIS)

    Gat, U.; Engel, J.R.

    1991-01-01

    The Molten Salt Reactor (MSR) option for burning fissile fuel from dismantled weapons is examined and is found very suitable for the beneficial use of this fuel. MSRs can utilize any fissile fuel in continuous operation with no special modifications, as demonstrated in the Molten Salt Reactor Experiment. Thus, MSRs are flexible while maintaining their economy. Furthermore, MSRs require only a minimum of special fuel preparation. They can tolerate denaturing and dilution of their fuel. The size of fuel shipments can be determined to optimize safety and security-all of which supports nonproliferation and resists diversion. In addition, MSRs have inherent safety features that make them acceptable and attractive. They can burn fissile material completely or can convert it to other fuels. MSRs also have the potential for burning the actinides and delivering the waste in an optimal form, thus contributing to the solution of one of the major remaining problems in the deployment of nuclear power

  20. Molten salt combustion of radioactive wastes

    International Nuclear Information System (INIS)

    Grantham, L.F.; McKenzie, D.E.; Richards, W.L.; Oldenkamp, R.D.

    1976-01-01

    The Atomics International Molten Salt Combustion Process reduces the weight and volume of combustible β-γ contaminated transuranic waste by utilizing air in a molten salt medium to combust organic materials, to trap particulates, and to react chemically with any acidic gases produced during combustion. Typically, incomplete combustion products such as hydrocarbons and carbon monoxide are below detection limits (i.e., 3 ) is directly related to the sodium chloride vapor pressure of the melt; >80% of the particulate is sodium chloride. Essentially all metal oxides (combustion ash) are retained in the melt, e.g., >99.9% of the plutonium, >99.6% of the europium, and >99.9% of the ruthenium are retained in the melt. Both bench-scale radioactive and pilot scale (50 kg/hr) nonradioactive combustion tests have been completed with essentially the same results. Design of three combustors for industrial applications are underway

  1. Mixing of zeolite powders and molten salt

    International Nuclear Information System (INIS)

    Pereira, C.; Zyryanov, V.N.; Lewis, M.A.; Ackerman, J.P.

    1996-01-01

    Transuranics and fission products in a molten salt can be incorporated into zeolite A by an ion exchange process and by a batch mixing or blending process. The zeolite is then mixed with glass and consolidated into a monolithic waste form for geologic disposal. Both processes require mixing of zeolite powders with molten salt at elevated temperatures (>700 K). Complete occlusion of salt and a uniform distribution of chloride and fission products are desired for incorporation of the powders into the final waste form. The relative effectiveness of the blending process was studied over a series of temperature, time, and composition profiles. The major criteria for determining the effectiveness of the mixing operations were the level and uniformity of residual free salt in the mixtures. High operating temperatures (>775 K) improved salt occlusion. Reducing the chloride levels in the mixture to below 80% of the full salt capacity of the zeolite significantly reduced the free salt level in the final product

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

    Energy Technology Data Exchange (ETDEWEB)

    Hur, J. M.; Cho, S. H.; Lim, J. H.; Seo, C. S.; Park, S. W

    2006-02-15

    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.

  3. 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...

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

    International Nuclear Information System (INIS)

    Zhang Dalin; Qiu Suizheng; Su Guanghui; Liu Changliang

    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 neutronics 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 neutronics slightly under the steady condition. (authors)

  5. 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.

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

    International Nuclear Information System (INIS)

    Ignatiev, V.; Feinberg, O.; Konakov, S.; Subbotine, S.; Surenkov, A.; Zakirov, R.

    2001-01-01

    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)

  7. 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.

  8. Sensitivity and Uncertainty Study for Thermal Molten Salt Reactors

    Science.gov (United States)

    Bidaud, Adrien; Ivanona, Tatiana; Mastrangelo, Victor; Kodeli, Ivo

    2006-04-01

    The Thermal Molten Salt Reactor (TMSR) using the thorium cycle can achieve the GEN IV objectives of economy, safety, non-proliferation and durability. Its low production of higher actinides, coupled with its breeding capabilities - even with a thermal spectrum - are very valuable characteristics for an innovative reactor. Furthermore, the thorium cycle is more flexible than the uranium cycle since only a small fissile inventory (Molten Salt Breeder Reactor (MSBR). The MSBR concept was developed at Oak Ridge National Laboratory (ORNL) in the 1970's based on the Molten Salt Reactor Experiment (MSRE). The main goals of our current studies are to achieve a reactor concept that enables breeding, improved safety and having chemical reprocessing needs reduced and simplified as much as reasonably possible. The neutronic properties of the new TMSR concept are presented in this paper. As the temperature coefficient is close to zero, we will see that the moderation ratio cannot be chosen to simultaneously achieve a high breeding ratio, long graphite lifetime and low uranium inventory. It is clear that any safety margin taken due to uncertainty in the nuclear data will significantly reduce the capability of this concept, thus a sensitivity analysis is vital to propose measurements which would allow to reduce at present high uncertainties in the design parameters of this reactor. Two methodologies, one based on OECD/NEA deterministic codes and one on IPPE (Obninsk) stochastic code, are compared for keff sensitivity analysis. The uncertainty analysis of keff using covariance matrices available in evaluated files has been performed. Furthermore, a comparison of temperature coefficient sensitivity profiles is presented for the most important reactions. These results are used to review the nuclear data needs for the TMSR thorium fuelled reactor.

  9. Deployment of quasi-digital sensor for high temperature molten salt level measurement in pyroprocessing plants

    Science.gov (United States)

    Sanga, Ramesh; Agarwal, Sourabh; Sivaramakrishna, M.; Rao, G. Prabhakara

    2018-04-01

    Development of a liquid molten salt level sensor device that can detect the level of liquid molten salt in the process vessels of pyrochemical reprocessing of spent metallic fuels is detailed. It is proposed to apply a resistive-type pulsating sensor-based level measurement approach. There are no commercially available sensors due to limitations of high temperature, radiation, and physical dimensions. A compact, simple, rugged, low power, and high precise pulsating sensor-based level probe and simple instrumentation for the molten salt liquid level sensor to work in the extreme conditions has been indigenously developed, with high precision and accuracy. The working principle, design concept, and results have been discussed. This level probe is mainly composed of the variable resistor made up of ceramic rods. This resistor constitutes the part of resistance-capacitance-type Logic Gate Oscillator (LGO). A change in the molten salt level inside the tank causes a small change in the resistance which in turn changes the pulse frequency of the LGO. Thus the frequency, the output of the instrument that is displayed on the LCD of an embedded system, is a function of molten salt level. In the present design, the range of level measurement is about 10 mm. The sensitivity in position measurement up to 10 mm is ˜2.5 kHz/mm.

  10. The source term and waste optimization of molten salt reactors with processing

    International Nuclear Information System (INIS)

    Gat, U.; Dodds, H.L.

    1993-01-01

    The source term of a molten salt reactor (MSR) with fuel processing is reduced by the ratio of processing time to refueling time as compared to solid fuel reactors. The reduction, which can be one to two orders of magnitude, is due to removal of the long-lived fission products. The waste from MSRs can be optimized with respect to its chemical composition, concentration, mixture, shape, and size. The actinides and long-lived isotopes can be separated out and returned to the reactor for transmutation. These features make MSRs more acceptable and simpler in operation and handling

  11. Electrochemical Deposition of High Purity Silicon from Molten Salts

    Science.gov (United States)

    Haarberg, Geir Martin

    Several approaches were tried in order to develop an electrochemical route for producing high purity silicon from molten salts. SiO2, K2SiF6 and metallurgical silicon were used as the source of silicon. Molten electrolytes based on chloride (CaCl2-NaCl) and fluoride (LiF-KF) at temperatures from 550 - 900 oC were used. Transient electrochemical techniques were used to study the electrochemical behaviour of dissolved silicon species. Electrolysis experiments were carried out to deposit silicon.

  12. Electrochemical Deposition of High Purity Silicon in Molten Salts

    Science.gov (United States)

    Haarberg, Geir Martin

    Several approaches were tried in order to develop an electrochemical route for producing high purity silicon from molten salts. SiO2, K2SiF6 and metallurgical silicon were used as the source of silicon. Molten electrolytes based on chloride (CaCl2-NaCl) and fluoride (LiF-KF) at temperatures from 550 - 900 °C were used. Transient electrochemical techniques were used to study the electrochemical behaviour of dissolved silicon species. Electrolysis experiments were carried out to deposit silicon.

  13. Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mueller, Don [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-12-01

    In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.

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

    International Nuclear Information System (INIS)

    Seneda, Jose A.; Lainetti, Paulo E.O.

    2013-01-01

    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 233 U. 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 UO 2 because it produces fewer amounts of actinides, reducing the radiotoxicity of long life nuclear waste. ThO 2 has higher corrosion resistance than UO 2 , 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

  15. Steady RANS methodology for calculating pressure drop in an in-line molten salt compact crossflow heat exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Carasik, Lane B.; Shaver, Dillon R.; Haefner, Jonah B.; Hassan, Yassin A.

    2017-11-01

    The development of molten salt cooled reactors (MSR) and fluoride-salt cooled high temperature reactors (FHR) requires the use of advanced design tools for the primary heat exchanger design. Due to geometric and flow characteristics, compact (pitch to diameter ratios equal to or less than 1.25) heat exchangers with a crossflow flow arrangement can become desirable for these reactors. Unfortunately, the available experimental data is limited for compact tube bundles or banks in crossflow. Computational Fluid Dynamics can be used to alleviate the lack of experimental data in these tube banks. Previous computational efforts have been primarily focused on large S/D ratios (larger than 1.4) using unsteady Reynolds averaged Navier-Stokes and Large Eddy Simulation frameworks. These approaches are useful, but have large computational requirements that make comprehensive design studies impractical. A CFD study was conducted with steady RANS in an effort to provide a starting point for future design work. The study was performed for an in-line tube bank geometry with FLiBe (LiF-BeF2), a frequently selected molten salt, as the working fluid. Based on the estimated pressure drops, the pressure and velocity distributions in the domain, an appropriate meshing strategy was determined and presented. Periodic boundaries in the spanwise direction transverse flow were determined to be an appropriate boundary condition for reduced computational domains. The domain size was investigated and a minimum of 2-flow channels for a domain is recommended to ensure the behavior is accounted for. Lastly, the standard low Re κ-ε (Lien) turbulence model was determined to be the most appropriate for steady RANS of this case at the time of writing.

  16. Accelerator molten-salt breeding and thorium fuel cycle

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Nakahara, Yasuaki; Kato, Yoshio; Ohno, Hideo; Mitachi, Kohshi.

    1990-01-01

    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 233 U 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.)

  17. 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.

  18. Experimental studies of actinides in molten salts

    International Nuclear Information System (INIS)

    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

  19. Applications of molten salts in plutonium processing

    International Nuclear Information System (INIS)

    Bowersox, D.F.; Christensen, D.C.; Williams, J.D.

    1987-01-01

    Plutonium is efficiently recovered from scrap at Los Alamos by a series of chemical reactions and separations conducted at temperatures ranging from 700 to 900 0 C. These processes usually employ a molten salt or salt eutectic as a heat sink and/or reaction medium. Salts for these operations were selected early in the development cycle. The selection criteria are being reevaluated. In this article we describe the processes now in use at Los Alamos and our studies of alternate salts and eutectics

  20. Molten salt reactors and molten salt carriers for industrial heat supply

    International Nuclear Information System (INIS)

    Novikov, V.M.

    1987-01-01

    Structure of industrial heat needs has been analysed. It is shown that main part of heat consumers - 80% demands temperature level up to 600 0 C. Temperature rise from 600 0 C up to 950 0 C could add only about 3% of heat consumers. It is shown that molten salt reactors with out-let temperature 700 0 C meet these conditions optimally

  1. Progress in modeling solidification in molten salt coolants

    Science.gov (United States)

    Tano, Mauricio; Rubiolo, Pablo; Doche, Olivier

    2017-10-01

    Molten salts have been proposed as heat carrier media in the nuclear and concentrating solar power plants. Due to their high melting temperature, solidification of the salts is expected to occur during routine and accidental scenarios. Furthermore, passive safety systems based on the solidification of these salts are being studied. The following article presents new developments in the modeling of eutectic molten salts by means of a multiphase, multicomponent, phase-field model. Besides, an application of this methodology for the eutectic solidification process of the ternary system LiF-KF-NaF is presented. The model predictions are compared with a newly developed semi-analytical solution for directional eutectic solidification at stable growth rate. A good qualitative agreement is obtained between the two approaches. The results obtained with the phase-field model are then used for calculating the homogenized properties of the solid phase distribution. These properties can then be included in a mixture macroscale model, more suitable for industrial applications.

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

    Science.gov (United States)

    Aji, Indarta Kuncoro; Waris, A.

    2014-09-01

    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 UF4 composition. The 235U 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 UF4 with 235U enrichment of at least 93%, 90%, and 90%, respectively. In contrast, the 50 MWth miniFUJI reactor can be critical for 1.96% of UF4 with 235U enrichment of at smallest amount 95%. The neutron spectra are almost similar for each power output.

  3. Analysis of molten salt thermal-hydraulics using computational fluid dynamics

    International Nuclear Information System (INIS)

    Yamaji, B.; Csom, G.; Aszodi, A.

    2003-01-01

    To give a good solution for the problem of high level radioactive waste partitioning and transmutation is expected to be a pro missing option. Application of this technology also could extend the possibilities of nuclear energy. Large number of liquid-fuelled reactor concepts or accelerator driven subcritical systems was proposed as transmutors. Several of these consider fluoride based molten salts as the liquid fuel and coolant medium. The thermal-hydraulic behaviour of these systems is expected to be fundamentally different than the behaviour of widely used water-cooled reactors with solid fuel. Considering large flow domains three-dimensional thermal-hydraulic analysis is the method seeming to be applicable. Since the fuel is the coolant medium as well, one can expect a strong coupling between neutronics and thermal-hydraulics too. In the present paper the application of Computational Fluid Dynamics for three-dimensional thermal-hydraulics simulations of molten salt reactor concepts is introduced. In our past and recent works several calculations were carried out to investigate the capabilities of Computational Fluid Dynamics through the analysis of different molten salt reactor concepts. Homogenous single region molten salt reactor concept is studied and optimised. Another single region reactor concept is introduced also. This concept has internal heat exchanges in the flow domain and the molten salt is circulated by natural convection. The analysis of the MSRE experiment is also a part of our work since it may form a good background from the validation point of view. In the paper the results of the Computational Fluid Dynamics calculations with these concepts are presented. In the further work our objective is to investigate the thermal-hydraulics of the multi-region molten salt reactor (Authors)

  4. 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.

  5. Detection and removal of molten salts from molten aluminum alloys

    Energy Technology Data Exchange (ETDEWEB)

    K. Butcher; D. Smith; C. L. Lin; L. Aubrey

    1999-08-02

    Molten salts are one source of inclusions and defects in aluminum ingots and cast shapes. A selective adsorption media was used to remove these inclusions and a device for detection of molten salts was tested. This set of experiments is described and the results are presented and analyzed.

  6. Advancing Molten Salts and Fuels at Sandia National Laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, Salvador B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-26

    SNL has a combination of experimental facilities, nuclear engineering, nuclear security, severe nuclear accidents, and nuclear safeguards expertise that can enable significant progress towards molten salts and fuels for Molten Salt Reactors (MSRs). The following areas and opportunities are discussed in more detail in this white paper.

  7. Advances in molten salt electrochemistry towards future energy systems

    International Nuclear Information System (INIS)

    Ito, Yasuhiko

    2005-01-01

    This review article describes some selected novel molten salt electrochemical processes which have been created/developed by the author and his coworkers, with emphasis on the applications towards future energy systems. After showing a perspective of the applications of molten salt electrochemistry from the viewpoints of energy and environment, several selected topics are described in detail, which include nitride fuel cycle in a nuclear field, hydrogen energy system coupled with ammonia economy, thermally regenerative fuel cell systems, novel Si production process for solar cell and novel molten salt electrochemical processes for various energy and environment related functional materials including nitrides, rare earth-transition metal alloys, fine particles obtained by plasma-induced electrolysis, and carbon film. And finally, the author stresses again, the importance and potential of molten salt electrochemistry, and encourages young students, scientists and researchers to march in a procession hand in hand towards a bright future of molten salts. (author)

  8. Demonstrating Cost Effective Thermal Energy Storage in Molten Salts: DLR' TESIS Test Facility

    OpenAIRE

    Odenthal, Christian; Klasing, Freerk; Bauer, Thomas

    2017-01-01

    Molten salt thermal energy storage is proven technology with large cost reduction potential --> DLR has built two test facilities TESIS:store and TESIS:com, which help understanding storage behavior, salt chemistry and testing components for faster market application Example based on exergy has shown that thermocline storage with filler can achieve --> high exergetic efficiency and --> significant reduction of salt inventory (investment cost)

  9. DESAIN KONSEP TANGKI PENAMPUNG BAHAN BAKAR PASSIVE COMPACT MOLTEN SALT REACTOR

    Directory of Open Access Journals (Sweden)

    A. Hadiwinata

    2015-04-01

    Full Text Available Passive Compact Molten Salt Reactor (PCMSR merupakan pengembangan dari reaktor MSR. Desain reaktor PCMSR membutuhkan tempat khusus penampung sementara bahan bakar pada saat terjadi insiden, misalnya kecelakaan yang menyebabkan peningkatan suhu bahan bakar. Tangki penampung bahan bakar tersusun dari 3 bagian yang saling terhubung yaitu bagian penampung cairan bahan bakar, cerobong (chimney, dan penukar kalor. Dalam penelitian ini, tangki dimodelkan secara lump dan dilakukan variasi daya awal reaktor dan ketinggian cerobong. Syarat batas model ditetapkan suhu bahan bakar maksimum 1400 °C, yang didasarkan pada titik didih larutan garam LiF-BeF2-ThF4-UF4. Analisis dilakukan dengan cara menghitung rugi tekanan total dan transfer kalor untuk variasi daya awal antara 1800-3000 MWth dan ketinggian cerobong antara 1-10 m. Hasil penelitian menunjukan semakin besar daya reaktor, maka tinggi tangki penampung bahan bakar dan tinggi alat penukar kalor yang dibutuhkan akan semakin besar, tejadi kenaikan suhu fluida pendingin dan suhu udara pendingin, dan menyebabkan kenaikan laju aliran masa fluida pendingin, sedangkan laju aliran masa udara menurun. Peningkatan ketinggian cerobong menyebabkan ketinggian tangki penampung bahan bakar dan ketinggian alat penukar kalor semakin menurun, penurunan suhu fluida pendingin, tetapi suhu udara meningkat, dan menyebabkan peningkatan laju aliran masa fluida pendingin, tetapi laju aliran masa udara akan semakin menurun. Kata kunci: PCMSR, cerobong, alat penukar kalor, variasi daya.   The Passsive Compact Molten Salat Reactor (PCMSR reactor is developed from MSR reactor. The PCMSR reactor design requires special place to temporarily storage for reactor fuel when incident occurs, such as when there is an accident which caused the temperature of the fuel increases. The tank consist of three interconnected parts, the reservoir liquid fuel, chimney, and the heat exchanger. In this research, the tank system is modeled based on

  10. A analysis of molten salt separation system for nuclear wastes transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, In Soon; Park, Byung Gi [Seoul National University, Seoul (Korea, Republic of); Kim, Kwang Bum; Kwon, Ou Sung [Yonsei University, Seoul (Korea, Republic of)

    1997-07-01

    Typical molten salt separation is ANL-IFR pyroprocessing and ORNL-MSRE pyroprocessing. IFR pyroprocessing is based on Chloride chemistry and electrorefining. MSRE pyroprocessing is base on fluoride chemistry and reductive extraction. Major technologies of molten salt separation are electrorefining, electrowining, reductive extraction, and oxide reduction. Common characteristics of this technologies is to utilize reduction-oxidation phenomena in molten salt. Electrorefining process is modeled on the basis of diffusion layer theory and Butler-Volmor relation. This model is numerically solved by LSODA package. To acquire the technology of electrorefining process, 3-electrode electrochemical cell is developed where electrolyte is 500 degree C LiCl-KCl eutectic molten salt, working electrodes are Ni and Au, and reference electrode is Ag/AgCl. We have investigated the stable potential range using cyclic voltammogram of Ni electrode. We have measured steady state polarization curve of Ni electrode. Then corrosion potential of Ni electrode is -0.38V{sub Ag/AgCl} and corrosion current is 1.23 x 10{sup -4} A/cm{sup 2}. 12 refs., 6 tabs., 24 figs. (author)

  11. New primary energy source by thorium molten-salt reactor technology

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Kato, Yoshio; Furuhashi, Akira; Numata, Hiroo; Mitachi, Koushi; Yoshioka, Ritsuo; Sato, Yuzuru; Arakawa, Kazuto

    2005-01-01

    Among the next 30 years, we have to implement a practical measure in the global energy/environmental problems, solving the followings: (1) replacing the fossil fuels without CO 2 emission, (2) no severe accidents, (3) no concern on military, (4) minimizing wastes, (5) economical, (6) few R and D investment and (7) rapid/huge global application supplying about half of the total primary energy till 50 years later. For this purpose the following system was proposed: THORIMS-NES [Thorium Molten-Salt Nuclear Energy Synergetic System], which is composed of (A) simple fission Molten-Salt power stations (FUJI), and (B) fissile-producing Accelerator Molten-Salt Breeder (AMSB). It has been internationally prepared a practical Developmental Program for its huge-size industrialization of Th breeding fuel cycle to produce a new rational primary energy. Here it is explained the social meaning, the conceptual system design and technological bases, especially, including the molten fluoride salt technology, which was developed as the triple-functional medium for nuclear-engineering, heat-transfer and chemical engineering. The complex function of this system is fully achieved by the simplified facility using a single phase molten-salt only. (author)

  12. Molten salt actinide recycler & transforming system and related fuel cycles

    International Nuclear Information System (INIS)

    Ignatiev, V.V.; Feynberg, O.S.

    2015-01-01

    The achievement of a safe, reliable, low waste, flexible, self-sustainable and cheap nuclear power system is the priority task. A study is under progress to examine the feasibility of MOlten Salt Actinide Recycler & Transforming (MOSART) system without and with U-Th support fuelled with different compositions of transuranic elements trifluorides from spent LWR fuel. New design options with homogeneous core and fuel salt with high enough solubility for transuranic elements trifluorides are being examined because of new goals. Consideration is aiming to optimise core neutronic and thermal hydraulic performance, fuel salt / container material, fuel clean up and safety related parameters for MOSART system. Experimental data base created was used for further development of MOSART flowsheet as applied to consumption of transuranic elements trifluorides while extracting their energy. The flexibility of single fluid MOSART concept fuel cycle is underlined, e.g. possibility of its operation in self-sustainable mode using different loadings and feedings. (author)

  13. 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.

  14. 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).

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

    International Nuclear Information System (INIS)

    Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei

    2015-01-01

    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. Boric ester-type molten salt via dehydrocoupling reaction.

    Science.gov (United States)

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

    2014-11-14

    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.

  17. 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.

  18. 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.

  19. Molten salt destruction process for mixed wastes

    International Nuclear Information System (INIS)

    Upadhye, R.S.; Wilder, J.G.; Karisen, C.E.

    1993-01-01

    We are developing an advanced two-stage process for the treatment of mixed wastes, which contain both hazardous and radioactive components (1) The wastes, together with an oxidant gas, such as air, are injected into a bed of molten salt comprising a mixture of sodium-, potassium-, and lithium-carbonates, with a melting point of about 580 degrees C. The organic constituents of the mixed waste are destroyed through the combined effect of pyrolysis and oxidation. Heteroatoms, such as chlorine, in the mixed waste form stable salts, such as sodium chloride, and are retained in the melt. The radioactive actinides in the mixed waste are also retained in the melt because of the combined action of wetting and partial dissolution. The original process, developed by Rockwell International, consists of a one-stage unit, operated at 900-1000 degrees C. The advanced two-stage process has two stages, one for pyrolysis and one for oxidation. The pyrolysis stage is designed to operate at 700 degrees C. The oxidation stage can be operated at a higher temperature, if necessary

  20. Molten salt destruction process for mixed wastes

    International Nuclear Information System (INIS)

    Upadhye, R.S.; Wilder, J.G.; Karlsen, C.E.

    1993-04-01

    We are developing an advanced two-stage process for the treatment of mixed wastes, which contain both hazardous and radioactive components. The wastes, together with an oxidant gas, such as air, are injected into a bed of molten salt comprising a mixture of sodium-, potassium-, and lithium-carbonates, with a melting point of about 580 degree C. The organic constituents of the mixed waste are destroyed through the combined effect of pyrolysis and oxidation. Heteroatoms. such as chlorine, in the mixed waste form stable salts, such as sodium chloride, and are retained in the melt. The radioactive actinides in the mixed waste are also retained in the melt because of the combined action of wetting and partial dissolution. The original process, consists of a one-stage unit, operated at 900--1000 degree C. The advanced two-stage process has two stages, one for pyrolysis and one for oxidation. The pyrolysis stage is designed to operate at 700 degree C. The oxidation stage can be operated at a higher temperature, if necessary

  1. Novel waste printed circuit board recycling process with molten salt

    Science.gov (United States)

    Riedewald, Frank; Sousa-Gallagher, Maria

    2015-01-01

    The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450–470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl–KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. • The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept. • This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L. • The treated PCBs can be removed via leg B while the process is on-going. PMID:26150977

  2. Thermodynamic characterization of salt components for the Molten Salt Reactor Fuel - 15573

    International Nuclear Information System (INIS)

    Capelli, E.; Konings, R.J.M.; Benes, A.

    2015-01-01

    Molten fluoride salts are considered as primary candidates for nuclear fuel in the Molten Salt Reactor (MSR), one of the 6 generation IV nuclear reactor designs. In order to determine the safety limits and to access the properties of the potential fuel mixtures, thermodynamic studies are very important. This study is a combination of experimental work and thermodynamic modelling and focusses on the fluoride systems with alkaline and alkaline earth fluorides as matrix and ThF 4 , UF 4 and PuF 3 as fertile and fissile materials. The purification of the single components was considered as essential first step for the study of more complex systems and ternary phase diagrams were described using Differential Scanning Calorimetry (DSC) and drop calorimetry, which are used to measure phase transitions, enthalpy of mixing and heat capacity. In addition to the calorimetric techniques, Knudsen Effusion Mass Spectrometry (KEMS) and X-ray Diffraction (XRD) were used to collect data on vapour pressure and crystal structure of fluorides. The results are then coupled with thermodynamic modelling using the Calphad method for the assessment of the phase diagrams. A thermodynamic database describing the most important systems for MSR application has been developed and it has been used to optimize the fuel composition in view of the relevant properties such as melting temperature. A reliable database of thermodynamic properties of fluoride salts has been generated. It includes the key systems for the MSR fuel and it is very useful to predict the properties of the fuel

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

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J

    2008-10-23

    . Preliminary design studies 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.

  4. The Equilibrium Between Titanium Ions and Titanium Metal in NaCl-KCl Equimolar Molten Salt

    Science.gov (United States)

    Wang, Qiuyu; Song, Jianxun; Hu, Guojing; Zhu, Xiaobo; Hou, Jungang; Jiao, Shuqiang; Zhu, Hongmin

    2013-08-01

    The equilibrium between metallic titanium and titanium ions, 3Ti2+ ⇌ 2Ti3+ + Ti, in NaCl-KCl equimolar molten salt was reevaluated. At a fixed temperature and an initial concentration of titanium chloride, the equilibrium was achieved by adding an excess amount of sponge titanium in assistant with bubbling of argon into the molten salt. The significance of this work is that the accurate concentrations of titanium ions have been obtained based on a reliable approach for taking samples. Furthermore, the equilibrium constant {{K}}_{{C}} = (x_{{{{Ti}}^{{ 3 { + }}} }}^{{eql}} )3 /(x_{{{{Ti}}^{{ 2 { + }}} }}^{{eql}} )2 was calculated through the best-fitting method under the consideration of the TiOCl dissolution. Indeed, the final results have disclosed that the stable value of KC could be achieved based on all modifications.

  5. Computational Analysis of Nanoparticles-Molten Salt Thermal Energy Storage for Concentrated Solar Power Systems

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Vinod [Univ. of Texas, El Paso, TX (United States)

    2017-05-05

    High fidelity computational models of thermocline-based thermal energy storage (TES) were developed. The research goal was to advance the understanding of a single tank nanofludized molten salt based thermocline TES system under various concentration and sizes of the particles suspension. Our objectives were to utilize sensible-heat that operates with least irreversibility by using nanoscale physics. This was achieved by performing computational analysis of several storage designs, analyzing storage efficiency and estimating cost effectiveness for the TES systems under a concentrating solar power (CSP) scheme using molten salt as the storage medium. Since TES is one of the most costly but important components of a CSP plant, an efficient TES system has potential to make the electricity generated from solar technologies cost competitive with conventional sources of electricity.

  6. Molten Salt: Concept Definition and Capital Cost Estimate

    Energy Technology Data Exchange (ETDEWEB)

    Stoddard, Larry [Black & Veatch, Kansas City, MO (United States); Andrew, Daniel [Black & Veatch, Kansas City, MO (United States); Adams, Shannon [Black & Veatch, Kansas City, MO (United States); Galluzzo, Geoff [Black & Veatch, Kansas City, MO (United States)

    2016-06-30

    The Department of Energy’s (DOE’s) Office of Renewable Power (ORP) has been tasked to provide effective program management and strategic direction for all of the DOE’s Energy Efficiency & Renewable Energy’s (EERE’s) renewable power programs. The ORP’s efforts to accomplish this mission are aligned with national energy policies, DOE strategic planning, EERE’s strategic planning, Congressional appropriation, and stakeholder advice. ORP is supported by three renewable energy offices, of which one is the Solar Energy Technology Office (SETO) whose SunShot Initiative has a mission to accelerate research, development and large scale deployment of solar technologies in the United States. SETO has a goal of reducing the cost of Concentrating Solar Power (CSP) by 75 percent of 2010 costs by 2020 to reach parity with base-load energy rates, and to reduce costs 30 percent further by 2030. The SunShot Initiative is promoting the implementation of high temperature CSP with thermal energy storage allowing generation during high demand hours. The SunShot Initiative has funded significant research and development work on component testing, with attention to high temperature molten salts, heliostats, receiver designs, and high efficiency high temperature supercritical CO2 (sCO2) cycles. DOE retained Black & Veatch to support SETO’s SunShot Initiative for CSP solar power tower technology in the following areas: 1. Concept definition, including costs and schedule, of a flexible test facility to be used to test and prove components in part to support financing. 2. Concept definition, including costs and schedule, of an integrated high temperature molten salt (MS) facility with thermal energy storage and with a supercritical CO2 cycle generating approximately 10MWe. 3. Concept definition, including costs and schedule, of an integrated high temperature falling particle facility with thermal energy storage and with a supercritical CO2

  7. Separation of calcium isotopes by counter-current electromigration in molten salts (1962)

    International Nuclear Information System (INIS)

    Menes, F.; Dirian, G.; Roth, E.

    1962-01-01

    The method of counter-current electromigration in molten salts has been applied to calcium bromide with an alkali metal bromide added to the cathode compartment. Enrichments on calcium-46 greater than a factor of two were obtained at the anode. The mass effect was found to be about 0.06. An estimation of the cost of energy for a process based on this method has been made. (authors) [fr

  8. 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)

  9. Actinides recovery from molten salt/liquid metal system by electrochemical methods

    Science.gov (United States)

    Iizuka, Masatoshi; Koyama, Tadafumi; Kondo, Naruhito; Fujita, Reiko; Tanaka, Hiroshi

    1997-08-01

    Electrochemical methods were examined for the recovery of actinides from the electrorefiner which is used in pyrometallurgical reprocessing of spent metal fuel for fast reactors. Uranium was successfully collected at the solid steel cathode from both liquid cadmium and molten salt solvents. In electrotransport from liquid cadmium, the behavior of uranium and rare earths was as expected by a computer simulation code based on the diffusion layer model at the interface between the electrolyte and the electrodes. In electroreduction from the molten salt electrolyte, a considerable amount of uranium was reduced at the CdLi anode by direct chemical reduction with lithium, especially at a lower anodic current density. The decrease in collection efficiency of uranium due to the direct chemical reduction would be avoided by maintaining the anode potential higher than the deposition potential of uranium.

  10. Accuracy analysis of the thermal diffusivity measurement of molten salts by stepwise heating method

    International Nuclear Information System (INIS)

    Kato, Yoshio; Furukawa, Kazuo

    1976-11-01

    The stepwise heating method for measuring thermal diffusivity of molten salts is based on the electrical heating of a thin metal plate as a plane heat source in the molten salt. In this method, the following estimations on error are of importance: (1) thickness effect of the metal plate, (2) effective length between the plate and a temperature measuring point and (3) effect of the noise on the temperature rise signal. In this report, a measuring apparatus is proposed and measuring conditions are suggested on the basis of error estimations. The measurements for distilled water and glycerine were made first to test the performance; the results agreed well with standard values. The thermal diffusivities of molten NaNO 3 at 320-380 0 C and of molten Li 2 BeF 4 at 470-700 0 C were measured. (auth.)

  11. Study on Utilization of Super Grade Plutonium in Molten Salt Reactor FUJI-U3 using CITATION Code

    Science.gov (United States)

    Wulandari, Cici; Waris, Abdul; Pramuditya, Syeilendra; Asril, Pramutadi AM; Novitrian

    2017-07-01

    FUJI-U3 type of Molten Salt Reactor (MSR) has a unique design since it consists of three core regions in order to avoid the replacement of graphite as moderator. MSR uses floride as a nuclear fuel salt with the most popular chemical composition is LiF-BeF2-ThF4-233UF4. ThF4 and 233UF4 are the fertile and fissile materials, respectively. On the other hand, LiF and BeF2 working as both fuel and heat transfer medium. In this study, the super grade plutonium will be utilized as substitution of 233U since plutonium is easier to be obtained compared to 233U as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2002 code with JENDL 3.2 as nuclear data library.

  12. Deuterium retention in molten salt electrodeposition tungsten coatings

    International Nuclear Information System (INIS)

    Zhou, Hai-Shan; Xu, Yu-Ping; Sun, Ning-Bo; Zhang, Ying-Chun; Oya, Yasuhisa; Zhao, Ming-Zhong; Mao, Hong-Min; Ding, Fang; Liu, Feng; Luo, Guang-Nan

    2016-01-01

    Highlights: • We investigate D retention in electrodeposition W coatings. • W coatings are exposed to D plasmas in the EAST tokamak. • A cathodic current density dependence on D retention is found. • Electrodeposition W exhibits lower D retention than VPS-W. - Abstract: Molten salt electrodeposition is a promising technology to manufacture the first wall of a fusion reactor. Deuterium (D) retention behavior in molten salt electrodeposition tungsten (W) coatings has been investigated by D-plasma exposure in the EAST tokamak and D-ion implantation in an ion beam facility. Tokamak exposure experiments demonstrate that coatings prepared with lower current density exhibit less D retention and milder surface damage. Deuterium-ion implantation experiments indicate the D retention in the molten salt electrodeposition W is less than that in vacuum plasma spraying W and polycrystalline W.

  13. Fission product removal from molten salt using zeolite

    International Nuclear Information System (INIS)

    Pereira, C.; Babcock, B.D.

    1996-01-01

    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

  14. Boric Ester-Type Molten Salt via Dehydrocoupling Reaction

    OpenAIRE

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

    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 molt...

  15. Renewing Liquid Fueled Molten Salt Reactor Research and Development

    Science.gov (United States)

    Towell, Rusty; NEXT Lab Team

    2016-09-01

    Globally there is a desperate need for affordable, safe, and clean energy on demand. More than anything else, this would raise the living conditions of those in poverty around the world. An advanced reactor that utilizes liquid fuel and molten salts is capable of meeting these needs. Although, this technology was demonstrated in the Molten Salt Reactor Experiment (MSRE) at ORNL in the 60's, little progress has been made since the program was cancelled over 40 years ago. A new research effort has been initiated to advance the technical readiness level of key reactor components. This presentation will explain the motivation and initial steps for this new research initiative.

  16. 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.

  17. Surface tension of light rare earth fluoride molten salts electrolyte system

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Xiaoping [School of Materials & Metallurgy, Northeastern University, Shenyang 110819 (China); Sun, Shuchen, E-mail: sunsc@smm.neu.edu.cn [School of Materials & Metallurgy, Northeastern University, Shenyang 110819 (China); Lu, Shuaidan; Huang, Xiaoxiao; Li, Kuanhe; Tu, Ganfeng [School of Materials & Metallurgy, Northeastern University, Shenyang 110819 (China); Huang, Xiaowei [National Engineering Research Center for Rare Earth Materials, General Research Institute for Non-Ferrous Metals, and Grirem Advanced Materials Co., Ltd., Beijing 100088 (China); Huang, Shaodong [School of Materials & Metallurgy, Northeastern University, Shenyang 110819 (China); Inner Mongolia Baotou Steel Rare-earth (Group) Hi-tech. Co., Ltd. Inner Mongolia, Baotou 014030 (China)

    2016-07-20

    Highlights: • The surface tensions of typical RE fluoride melts were systematically studied. • The temperature and composition dependences of the surface tensions were discussed. • Surface tension factor of RE{sub 2}O{sub 3} was studied and the solubility was the key point. • Results of this study are novel and well supplement data base of RE fluoride melts. - Abstract: The surface tensions of REF{sub 3}-LiF (RE = La and Nd) molten salts (REF{sub 3} mol% = 20-40%) were measured using the slide method over the temperature range1223–1423 K which was specifically chosen for meeting the actual electrolysis temperature (about 1323 K). The temperature and composition dependences of the surface tensions were investigated. The results showed that the surface tension decreased linearly with the increasing of temperature and increased nonlinearly with the increasing mole fraction of REF{sub 3}. At 1323 K (near the practical electrolysis temperature), the surface tension of REF{sub 3}-LiF-RE{sub 2}O{sub 3} molten salts firstly increased with the increasing RE{sub 2}O{sub 3} and reached the maximum value near the solubility of RE{sub 2}O{sub 3} (2–3 g) in REF{sub 3}-LiF molten salts, and then decreased with the continuously adding RE{sub 2}O{sub 3}. The accurate knowledge of the surface tension factor of RE{sub 2}O{sub 3} can be useful for well understanding the solubilizing mechanisms of RE{sub 2}O{sub 3} in REF{sub 3}-LiF molten salts system. Results of this study well supplemented the experimental data of physiochemical property on surface tension of rare earth fluoride electrolytes.

  18. Assessment Of Usability Of Molten Salt Mixtures In Metallurgy Of Aluminum Alloys And Recycling Of Composite Materials Based On The Matrix Of Al Alloys

    Directory of Open Access Journals (Sweden)

    Jackowski J.

    2015-09-01

    Full Text Available Effectiveness of the slags used in metallurgy of aluminum alloys and in recycling of composite materials containing these alloys depends on their surface properties at the phase boundaries they are in contact with. An index of surface properties of molten mixtures of slag-forming salts has been formulated. Its calculated values are compared with measured results of surface tension (liquid – atmosphere and interfacial tension (liquid – liquid in the considered systems. It was found that the index can be helpful for purposes of proper choice of the mixtures of slag-forming salts used both in Al alloys metallurgy and in recycling of composite materials based on the matrix of Al alloys.

  19. 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.

  20. Stability of Sodium Electrodeposited From a Series of Room Temperature Chloroaluminate Molten Salts

    National Research Council Canada - National Science Library

    Gray, Gary

    1996-01-01

    .... This work&involved the synthesis of room temperature molten salts and the examination of the electrochemical and transport properties of these salts with the goal of developing a room temperature molten salt...

  1. Thermodynamics of soluble fission products cesium and iodine in the Molten Salt Reactor

    Science.gov (United States)

    Capelli, E.; Beneš, O.; Konings, R. J. M.

    2018-04-01

    The present study describes the full thermodynamic assessment of the Li,Cs,Th//F,I system. The existing database for the relevant fluoride salts considered as fuel for the Molten Salt Reactor (MSR) has been extended with two key fission products, cesium and iodine. A complete evaluation of all the common-ion binary and ternary sub-systems of the LiF-ThF4-CsF-LiI-ThI4-CsI system has been performed and the optimized parameters are presented in this work. New equilibrium data have been measured using Differential Scanning Calorimetry and were used to assess the reciprocal ternary systems and confirm the extrapolated phase diagrams. The developed database significantly contributes to the understanding of the behaviour of cesium and iodine in the MSR, which strongly depends on their concentration and chemical form. Cesium bonded with fluorine is well retained in the fuel mixture while in the form of CsI the solubility of these elements is very limited. Finally, the influence of CsI and CsF on the physico-chemical properties of the fuel mixture was calculated as function of composition.

  2. [Bio-oil production from biomass pyrolysis in molten salt].

    Science.gov (United States)

    Ji, Dengxiang; Cai, Tengyue; Ai, Ning; Yu, Fengwen; Jiang, Hongtao; Ji, Jianbing

    2011-03-01

    In order to investigate the effects of pyrolysis conditions on bio-oil production from biomass in molten salt, experiments of biomass pyrolysis were carried out in a self-designed reactor in which the molten salt ZnCl2-KCl (with mole ratio 7/6) was selected as heat carrier, catalyst and dispersion agent. The effects of metal salt added into ZnCl2-KCl and biomass material on biomass pyrolysis were discussed, and the main compositions of bio-oil were determined by GC-MS. Metal salt added into molten salt could affect pyrolysis production yields remarkably. Lanthanon salt could enhance bio-oil yield and decrease water content in bio-oil, when mole fraction of 5.0% LaCl3 was added, bio-oil yield could reach up to 32.0%, and water content of bio-oil could reduce to 61.5%. The bio-oil and char yields were higher when rice straw was pyrolysed, while gas yield was higher when rice husk was used. Metal salts showed great selectivity on compositions of bio-oil. LiCl and FeCl2 promoted biomass to pyrolyse into smaller molecular weight compounds. CrCl3, CaCl2 and LaCl3 could restrain second pyrolysis of bio-oil. The research provided a scientific reference for production of bio-oil from biomass pyrolysis in molten salt.

  3. Treatment of plutonium process residues by molten salt oxidation

    International Nuclear Information System (INIS)

    Stimmel, J.; Wishau, R.; Ramsey, K.B.; Montoya, A.; Brock, J.; Heslop, M.

    1999-01-01

    Molten Salt Oxidation (MSO) is a thermal process that can remove more than 99.999% of the organic matrix from combustible 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 2 SO 4 , Na 3 PO 4 and NaAsO 2 or Na 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 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

  4. 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.

  5. Conceptual design of Indian molten salt breeder reactor

    Indian Academy of Sciences (India)

    reactor was found to be stable and self-regulating. 4.2 Molten-salt reactor experiment (MSRE) [6]. As ARE had .... Continuous monitoring of the atmo- sphere was carried out by Health Physics Group of ... (4) Breeding ratio should be such that a self-sustaining operation of the reactor is achieved after accounting for losses in ...

  6. Ion diffusion related to structure in molten salts

    International Nuclear Information System (INIS)

    Tosi, M.P.

    1996-08-01

    A model first developed by Zwanzig to derive transport coefficients in cold dense fluids directly from the Green-Kubo time correlation formulae allows one to relate macroscopic diffusion coefficients to the local fluid structure. Applications to various ionic diffusion processes in molten salts are reviewed. Consequences of partial structural quenching are also discussed. (author). 28 refs, 3 tabs

  7. Three-dimensional numerical investigation of a Molten Salt reactor concept with the code CFX-5.5

    International Nuclear Information System (INIS)

    Yamaji, B.; Csom, G.; Aszodi, A.

    2002-01-01

    Partitioning and transmutation of actinides and long-lived fission products is a promising option to extend the possibilities and enhance the environmentally acceptable capabilities of nuclear energy. Also the possible implementation of the thorium cycle is considered as a way to reduce the problem of energy resources in the future. For both objectives different molten salt reactor concepts were proposed mainly based on the Molten Salt Reactor Experiment of the Oak Ridge National Laboratory. Not only critical reactors but also accelerator-driven subcritical systems (ADSs) have advantages worth considering for those aims, especially those ones with liquid fuel, such as molten salts. By using liquid fuel which is the coolant medium, too, a basically different thermalhydraulic behavior is expected than in the case of solid fuel and water coolant. In this work our purpose is to present the possible use of Computational Fluid Dynamics (CFD) technology in molten salt thermal hydraulics. The simulations were performed with the three-dimensional code CFX-5.5.(author)

  8. A study on the reduction of uranium oxide to uranium metal in LiCl molten salt

    International Nuclear Information System (INIS)

    Seo, J. S.; Hur, J. M.; Lee, W. K.; Hong, S. S.; Kang, D. S.; Park, S. W.

    2002-01-01

    Research for the analysis on a metallization process of uranium oxide in LiCl-Li molten salt was carried out. Effect of a concentration of Li 2 O on the metallization process was also studied. The new concept, electrochemical reduction of uranium oxide in LiCl-Li 2 O molten salt was proposed. The concept is based on the integrated process of metallization of UO 2 with simultaneous electrochemical reduction of Li 2 O which is recycled in a closed system. In a LiCl-Li molten salt system, U 3 O 8 whose conversion ratio to U turns out to be 97.1%, showed a better metallization characteristic than UO 2 . It is verified that electrochemically reduced Li is well deposited on the UO 2 powder cathode through a porous magnesia filter in LiCl-Li 2 O molten salt. In that process Li 2 O was from by the reduction process of UO 2 to U. This electrochemical reduction process showed good results to covert UO 2 to U

  9. Influence of the reprocessing on molten salt reactor behaviour

    OpenAIRE

    Merle-Lucotte, E.; Mathieu, L.; Heuer, D.; Billebaud, A.; Brissot, R.; Le Brun, C.; Liatard, E.; Loiseaux, J.M.; Méplan, O.; Nuttin, A.; Wilson, J.

    2005-01-01

    International audience; The Molten Salt Reactor is one of the systems studied as a Generation IV reactor. Its main characteristic is the strong coupling between neutronics and salt reprocessing. Such nuclear reactors use a liquid fuel which is also the coolant. Elements produced during the reactor's operation, like Fission Products (FP) or TransUranians, modify the neutronic balance of the reactor by capturing neutrons. As the fuel is liquid, samples can be extracted and reprocessed to remove...

  10. Fission product ion exchange between zeolite and a molten salt

    Science.gov (United States)

    Gougar, Mary Lou D.

    The electrometallurgical treatment of spent nuclear fuel (SNF) has been developed at Argonne National Laboratory (ANL) and has been demonstrated through processing the sodium-bonded SNF from the Experimental Breeder Reactor-II in Idaho. In this process, components of the SNF, including U and species more chemically active than U, are oxidized into a bath of lithium-potassium chloride (LiCl-KCl) eutectic molten salt. Uranium is removed from the salt solution by electrochemical reduction. The noble metals and inactive fission products from the SNF remain as solids and are melted into a metal waste form after removal from the molten salt bath. The remaining salt solution contains most of the fission products and transuranic elements from the SNF. One technique that has been identified for removing these fission products and extending the usable life of the molten salt is ion exchange with zeolite A. A model has been developed and tested for its ability to describe the ion exchange of fission product species between zeolite A and a molten salt bath used for pyroprocessing of spent nuclear fuel. The model assumes (1) a system at equilibrium, (2) immobilization of species from the process salt solution via both ion exchange and occlusion in the zeolite cage structure, and (3) chemical independence of the process salt species. The first assumption simplifies the description of this physical system by eliminating the complications of including time-dependent variables. An equilibrium state between species concentrations in the two exchange phases is a common basis for ion exchange models found in the literature. Assumption two is non-simplifying with respect to the mathematical expression of the model. Two Langmuir-like fractional terms (one for each mode of immobilization) compose each equation describing each salt species. The third assumption offers great simplification over more traditional ion exchange modeling, in which interaction of solvent species with each other

  11. Lithium-Boron Alloy Anodes for Molten Salt Batteries (II)

    Science.gov (United States)

    1978-05-15

    Eagle - Picher Industries, Inc. Electronics Division, Couples Dept. Attn: D. R. Cottingham J. Dines D. L. Smith J. Wilson P. 0. Box 47 Joplin, MO 64801... Eagle - Picher Industries, Inc. Miami Research Laboratories Attn: P. E. Grayson 200 Ninth Avenue, N.E. Miami, OK 74354 ESB Research Center Attn: Library...777.. -~ -- NSWC/WOL TR 78-63 LITHIUM-BORON ALLOY ANODES FOR MOLTEN SALT BATTERIES (11) BY S.DALLEK, D. W. ERNST, 0 B. F. LARRICK Ott RESEARCH AND

  12. Characteristics of solidified products containing radioactive molten salt waste.

    Science.gov (United States)

    Park, Hwan-Seo; Kim, In-Tae; Cho, Yong-Zun; Eun, Hee-Chul; Kim, Joon-Hyung

    2007-11-01

    The molten salt waste from a pyroprocess to recover uranium and transuranic elements is one of the problematic radioactive wastes to be solidified into a durable wasteform for its final disposal. By using a novel method, named as the GRSS (gel-route stabilization/solidification) method, a molten salt waste was treated to produce a unique wasteform. A borosilicate glass as a chemical binder dissolves the silicate compounds in the gel products to produce one amorphous phase while most of the phosphates are encapsulated by the vitrified phase. Also, Cs in the gel product is preferentially situated in the silicate phase, and it is vitrified into a glassy phase after a heat treatment. The Sr-containing phase is mainly phosphate compounds and encapsulated by the glassy phase. These phenomena could be identified by the static and dynamic leaching test that revealed a high leach resistance of radionuclides. The leach rates were about 10(-3) - 10(-2) g/m2 x day for Cs and 10(-4) - 10(-3) g/m2 x day for Sr, and the leached fractions of them were predicted to be 0.89% and 0.39% at 900 days, respectively. This paper describes the characteristics of a unique wasteform containing a molten salt waste and provides important information on a newly developed immobilization technology for salt wastes, the GRSS method.

  13. Molten salt processing of mixed wastes with offgas condensation

    International Nuclear Information System (INIS)

    Cooper, J.F.; Brummond, W.; Celeste, J.; Farmer, J.; Hoenig, C.; Krikorian, O.H.; Upadhye, R.; Gay, R.L.; Stewart, A.; Yosim, S.

    1991-01-01

    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

  14. Rare Earth Electrochemical Property Measurements and Phase Diagram Development in a Complex Molten Salt Mixture for Molten Salt Recycle

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Jinsuo; Guo, Shaoqiang

    2018-03-30

    Pyroprocessing is a promising alternative for the reprocessing of used nuclear fuel (UNF) that uses electrochemical methods. Compared to the hydrometallurgical reprocessing method, pyroprocessing has many advantages such as reduced volume of radioactive waste, simple waste processing, ability to treat refractory material, and compatibility with fast reactor fuel recycle. The key steps of the process are the electro-refining of the spent metallic fuel in the LiCl-KCl eutectic salt, which can be integrated with an electrolytic reduction step for the reprocessing of spent oxide fuels. During the electro-refining process, actinides and active fission products such rare earth (RE) elements are dissolved into the molten salt from the spent fuel at an anode basket. Then U and Pu are electro-deposited on the cathodes while REs with relatively negative reduction potentials are left in the molten salt bath. However, with the accumulation of lanthanides in the salt, the reduction potentials of REs will approach the values for U and Pu, affecting the recovery efficiency of U and Pu. Hence, RE drawdown is necessary to reduce salt waste after uranium and minor actinides recovery, which can also be performed by electrochemical separations. To separate various REs and optimize the drawdown process, physical properties of REs in LiCl-KCl salt and their concentration dependence are essential. Thus, the primary goal of present research is to provide fundamental data of REs and deduce phase diagrams of LiCl-KCl-RECl3 based complex molten salts. La, Nd and Gd are three representative REs that we are particularly interested in due to the high ratio of La and Nd in UNF, highest standard potential of Gd among all REs, and the existing literature data in dilute solution. Electrochemical measurements are performed to study the thermodynamics and transport properties of LaCl3, GdCl3, NdCl3, and NdCl2 in LiCl-KCl eutectic in the temperature range 723-823 K. Test are conducted in LiCl-KCl melt

  15. 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 <50) from a list of hundreds of thousands of possible options. The operating limits which the heliostat field needs to obey are defined for the ranked surface configurations based on several different design considerations (e.g. mechanical integrity, corrosion limits). The thermal output of the receiver configurations for a given heliostat field is maximized. A combined rank indicating the optimum configurations in descending order of 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

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

    International Nuclear Information System (INIS)

    Mathieu, L.

    2005-09-01

    Producing nuclear energy in order to reduce the anthropic CO 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)

  17. High temperature molten salt reactor in a system of nuclear power supply of technological processes

    International Nuclear Information System (INIS)

    Belousov, I.G.

    1984-01-01

    Features of the high-temperature molten-salt reactor with natural convection of the coolant in the primary power curcuit are considered. An operating process of energy transfer from the core to the energy consumption zone is described. The main energy consumer is the metallurgical production based on the direct reduction of iron oxides by solid carbon. Some other energy consumers are being considered more briefly. A high efficiency usage of nuclear fuel in industrial technological cycles is possible only on fulfilling some definite conditions relating to compatibility limiting power properties of the nuclear heat source and the perfection of the technological cycle of energy consumption

  18. Migrational polarization in high-current density molten salt electrochemical devices

    Energy Technology Data Exchange (ETDEWEB)

    Braunstein, J.; Vallet, C.E.

    1977-01-01

    Electrochemical flux equations based on the thermodynamics of irreversible processes have been derived in terms of experimental transport coefficients for binary molten salt mixtures analogous to those proposed for high temperature batteries and fuel cells. The equations and some numerical solutions indicate steady state composition gradients of significant magnitude. The effects of migrational separation must be considered along with other melt properties in the characterization of electrode behavior, melt composition, operating temperatures and differences of phase stability, wettability and other physicochemical properties at positive and negative electrodes of high current density devices with mixed electrolytes.

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

    International Nuclear Information System (INIS)

    Upadhye, R.S.; Watkins, B.E.; Pruneda, C.O.

    1995-11-01

    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

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

    International Nuclear Information System (INIS)

    Lainetti, Paulo E.O.

    2013-01-01

    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

  1. Transfer characteristics of a lithium chloride–potassium chloride molten salt

    Directory of Open Access Journals (Sweden)

    Eve Mullen

    2017-12-01

    Full Text Available Pyroprocessing is an alternative method of reprocessing spent fuel, usually involving the dissolving spent fuel in a molten salt media. The National Nuclear Laboratory designed, built, and commissioned a molten salt dynamics rig to investigate the transfer characteristics of molten lithium chloride–potassium chloride eutectic salt. The efficacy and flow characteristics of a high-temperature centrifugal pump and argon gas lift were obtained for pumping the molten salt at temperatures up to 500°C. The rig design proved suitable on an industrial scale and transfer methods appropriate for use in future molten salt systems. Corrosion within the rig was managed, and melting techniques were optimized to reduce stresses on the rig. The results obtained improve the understanding of molten salt transport dynamics, materials, and engineering design issues and support the industrialization of molten salts pyroprocessing.

  2. 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.

  3. Transfer characteristics of a lithium chloride–potassium chloride molten salt

    OpenAIRE

    Eve Mullen; Ross Harris; Dave Graham; Chris Rhodes; Zara Hodgson

    2017-01-01

    Pyroprocessing is an alternative method of reprocessing spent fuel, usually involving the dissolving spent fuel in a molten salt media. The National Nuclear Laboratory designed, built, and commissioned a molten salt dynamics rig to investigate the transfer characteristics of molten lithium chloride–potassium chloride eutectic salt. The efficacy and flow characteristics of a high-temperature centrifugal pump and argon gas lift were obtained for pumping the molten salt at temperatures up to 500...

  4. Maintaining molten salt electrolyte concentration in aluminum-producing electrolytic cell

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, Robert J.; Mezner, Michael B.; Bradford, Donald R

    2005-01-04

    A method of maintaining molten salt concentration in a low temperature electrolytic cell used for production of aluminum from alumina dissolved in a molten salt electrolyte contained in a cell free of frozen crust wherein volatile material is vented from the cell and contacted and captured on alumina being added to the cell. The captured volatile material is returned with alumina to cell to maintain the concentration of the molten salt.

  5. Chronopotentiometry of refractory metals, actinides and oxyanions in molten salts: A review

    Science.gov (United States)

    Bansal, Narottam P.

    1992-01-01

    The applications of chronopotentiometry to the study of electrochemical behavior of three technologically important areas of refractory metals, actinides, and oxyanions in molten salts are critically reviewed. Chronopotentiometry is a very versatile diagnostic tool to understand the reaction mechanism of the electrode processes for the electrochemical reduction/oxidation of these electroactive species in molten salt solutions. Well adherent, compact, and uniformly thick coatings of refractory metals may be electrodeposited from their solutions in molten salts.

  6. The molten salt reactor: R and D status and perspectives in Europe

    International Nuclear Information System (INIS)

    Renault, Claude; Delpech, Sylvie; Merle-Lucotte, Elsa; Konings, Rudy; Hron, Miloslav; Ignatiev, Victor

    2010-01-01

    The paper concentrates on molten salt fast reactor (MSFR) concepts which are receiving most attention in the EU context. It shows the main R and D achievements and some remaining issues to be addressed in such essential areas as (a) reactor conceptual design, (b) molten salt properties, (c) fuel salt clean-up scheme and (d) high temperature materials. The status and perspectives of molten salt reactor R and D efforts in Europe are then discussed

  7. Structure and dynamics of molten salts

    International Nuclear Information System (INIS)

    Rovere, M.; Tosi, M.P.

    1986-02-01

    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)

  8. Beryllium research on FFHR molten salt blanket

    International Nuclear Information System (INIS)

    Terai, T.; Tanaka, S.; Sze, D.-K.

    2000-01-01

    Force-free helical reactor, FFHR, is a demo-relevant heliotron-type D-T fusion reactor based on the great amount of R and D results obtained in the LHD project. Since 1993, collaboration works have made great progress in design studies of FFHR with standing on the major advantage of current-less steady operation with no dangerous plasma disruptions. There are two types of reference designs, FFHR-1 and FFHR-2, where molten Flibe (LiF-BeF2) is utilized as tritium breeder and coolant. In this paper, we present the outline of FFHR blanket design and some related R and D topics focusing on Be utilization. Beryllium is used as a neutron multiplier in the design and Be pebbles are placed in the front part of the tritium breeding zone. In a Flibe blanket, HF (TF) generated due to nuclear transmutation will be a problem because of its corrosive property. Though nickel-based alloys are thought to be intact in such a corrosive environment, FFHR blanket design does not adopt the alloys because of their induced radioactivity. The present candidate materials for the structure are low-activated ferritic steel (JLF-1), V-4Cr-4Ti, etc. They are capable to be corroded by HF in the operation condition, and Be is expected to work as a reducing agent in the system as well. Whether Be pebbles placed in a Flibe flow can work well or not is a very important matter. From this point, Be solubility in Flibe, reaction rate of the Redox reaction with TF in the liquid and on the surface of Be pebbles under irradiation, flowing behavior of Flibe through a Be pebble bed, etc. should be investigated. In 1997, in order to establish more practical and new data bases for advanced design works, we started a collaboration work of R and D on blanket engineering, where the Be research above mentioned is included. Preliminary dipping-test of Be sheets and in-situ tritium release experiment from Flibe with Be sheets have got started. (orig.)

  9. Technical review of Molten Salt Oxidation

    International Nuclear Information System (INIS)

    1993-12-01

    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

  10. Fission product behavior in the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Compere, E.L.; Kirslis, S.S.; Bohlmann, E.G.; Blankenship, F.F.; Grimes, W.R.

    1975-10-01

    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 235 U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using 233 U fuel over a period of about 15 months (more than 5100 effective full-power hours)

  11. Concept of an electron accelerator driven molten salt subcritical reactor

    International Nuclear Information System (INIS)

    Brolly, A.; Vertes, P.

    2005-01-01

    Concept and analysis of an electron accelerator driven molten salt subcritical system are presented. The analysis covers the neutron source optimization and burnup history with continuous feeding of TRU into the reactor. Effect of long time operation on TRU consumption and corresponding energy production is considered. It seems that with an electron accelerator of 150 MeV energy and with technically acceptable current it is possible to maintain a subcritical reactor on a reasonable power level while it consumes considerable amount of TRU coming from online chemical processing of spent fuels. (authors)

  12. SYNTHESIS OF ALUMINIUM BORATE WHISKERS THROUGH WET MOLTEN SALT METHOD

    Directory of Open Access Journals (Sweden)

    Xue Zhang

    2017-12-01

    Full Text Available Aluminium borate (Al₄B₂O₉ whiskers were successfully synthesized by the wet molten salt method at 800 oC through control the aluminum/boron atomic ratio and synthesis temperature. The as-received Al₄B₂O₉ whiskers were characterized by scanning electron microscopy (SEM, X-ray diffraction (XRD and thermal analysis. A solution-liquid-solid (SLS mechanism was proposed for the growth mechanism of the whiskers on the basis of the experimental phenomena and the TG-DSC data of the mixed raw materials.

  13. Accelerator-driven molten-salt blankets: Physics issues

    International Nuclear Information System (INIS)

    Houts, M.G.; Beard, C.A.; Buksa, J.J.; Davidson, J.W.; Durkee, J.W.; Perry, R.T.; Poston, D.I.

    1994-01-01

    A number of nuclear physics issues concerning the Los Alamos molten-salt accelerator-driven plutonium converter are discussed. General descriptions of several concepts using internal and external moderation are presented. Burnup and salt processing requirement calculations are presented for four concepts, indicating that both the high power density externally moderated concept and an internally moderated concept achieve total plutonium burnups approaching 90% at salt processing rates of less than 2 m 3 per year. Beginning-of-life reactivity temperature coefficients and system kinetic response are also discussed. Future research should investigate the effect of changing blanket composition on operational and safety characteristics

  14. Polymer Coated Electrodes in Ambient Temperature Molten Salts.

    Science.gov (United States)

    1983-08-01

    PERFORMING ORG. REPORT NUMBER ~7. AUTHOR(*) S. CONTRACT OR GRANT NuMBER(a) P.G. Pickup and Robert A. Osteryoung N00014-79-C-0682 9. PERFORMING ORGANIZATION...glassy carbon ..-r electrodes in various compositions of the am- blent temperature molten salt system aluminum 00 6 ov chloride:n-butylpyridinium...Or TrVT AJf A I I START SECO:REFERENCES N HERE. 1. R. W. Murray, ’Chemically Modified Elect- rodes", Chapter 3 in Electroanalytical Chemistry, Vol. 13

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

    Energy Technology Data Exchange (ETDEWEB)

    Senatore, G.; Parrinello, M.; Tosi, M.P. (Trieste Univ. (Italy). Ist. di Fisica Teorica; Gruppo Nazionale di Struttura dell material del CNR, Trieste (Italy); International Centre for Theoretical Physics, Trieste (Italy))

    1978-12-23

    The theory of liquid structure for fluids of charged hard spheres is applied to an evaluation of the F-centre model for valence electrons in metal-molten salt solutions at high dilution. Minimization of the free energy yields the groundstate radius of the elctron bubble and hence the optical excitation energy in a Franck-Condon transition, the shift and broadening of the transition due to fluctuations in the bubble radius, the volume of mixing, and the activity of the salt in the solution.

  16. Transmutation and inventory analysis in an ATW molten salt system

    Energy Technology Data Exchange (ETDEWEB)

    Sisolak, J.E.; Truebenbach, M.T.; Henderson, D.L. [Univ. of Wisconsin, Madison, WI (United States)

    1995-10-01

    As an extension of earlier work to determine the equilibrium state of an ATW molten salt, power producing, reactor/transmuter, the WAIT code provides a time dependent view of material inventories and reactor parameters. By considering several cases, the authors infer that devices of this type do not reach equilibrium for dozens of years, and that equilibrium design calculations are inapplicable over most of the reactor life. Fissile inventory and k{sub eff} both vary by factors of 1.5 or more between reactor startup and ultimate convergence to equilibrium.

  17. Can the Transport Properties of Molten Salts and Ionic Liquids Be Used To Determine Ion Association?

    Science.gov (United States)

    Harris, Kenneth R

    2016-12-01

    There have long been arguments supporting the concept of ion association in molten salts and ionic liquids, largely based on differences between the conductivity and that predicted from self-diffusion coefficients by the Nernst-Einstein equation for noninteracting ions. It is known from molecular dynamics simulations that even simple models based on charged hard spheres show such a difference due to the (anti)-correlation of ion motions. Formally this is expressed as a difference between the velocity cross-correlation coefficient of the oppositely charged ions and the mean of those for the two like-charged ions. This article examines molten salt and ionic liquid transport property data, comparing simple and model associated salts (ZnCl 2 , PbCl 2 , and TlCl) including weakly dissociated molecular liquids (H 2 O, HCOOH, H 2 SO 4 ). Analysis employing Laity resistance coefficients (r ij ) shows that the common ion-association rationalization is flawed, consistent with recent direct measurements of the degree of ionicity in ionic liquid chlorides and with theoretical studies. However, the protic ionic liquids [PyrOMe][BF 4 ] and [DBUH][CH 3 SO 3 ] have larger than usual NE deviation parameters (>0.5), and large negative like-ion r ii , analogous to those of ZnCl 2 . Structural, spectroscopic, and theoretical studies are suggested to determine whether these are indeed genuine examples of association.

  18. 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.

  19. Direct reduction of uranium dioxide and few other metal oxides to corresponding metals by high temperature molten salt electrolysis

    International Nuclear Information System (INIS)

    Mohandas, K.S.

    2017-01-01

    Molten salt based electro-reduction processes, capable of directly converting solid metal oxides to metals with minimum intermediate steps, are being studied worldwide. Production of metals apart, the process assumes importance in nuclear technology in the context of pyrochemical reprocessing of spent oxide fuels, for it serves as an intermediate step to convert spent oxide fuel to a metal alloy, which in turn can be processed by molten salt electro-refining method to gain the actinides present in it. In the context of future metal fuel fast reactor programme, the electrochemical process was studied for conversion of solid UO 2 to U metal in LiCl-1wt.% Li 2 O melt at 650 °C with platinum anode at the Metal Processing Studies Section, PMPD, IGCAR. A brief overview of the work is presented in the paper

  20. Integrated in situ characterization of molten salt catalyst surface: Evidence of sodium peroxide and OH radical formation

    KAUST Repository

    Takanabe, Kazuhiro

    2017-06-26

    Na-based catalysts (i.e., Na2WO4) were proposed to selectively catalyze OH radical formation from H2O and O2 at high temperatures. This reaction may proceed on molten salt state surfaces due to the lower melting point of the used Na salts compared to the reaction temperature. This study provides direct evidence of the molten salt state of Na2WO4, which can form OH radicals, using in situ techniques including X-ray diffraction (XRD), scanning transmission electron microscopy (STEM), laser induced fluorescence (LIF) spectrometer, and ambient-pressure X-ray photoelectron spectroscopy (AP-XPS). As a result, Na2O2 species, which were hypothesized to be responsible for the formation of OH radicals, has been identified on the outer surfaces at temperatures ≥800°C, and these species are useful for various gas-phase hydrocarbon reactions including the selective transformation of methane to ethane.

  1. Molten salt synthesis of lead lanthanum zirconate titanate ceramic powders

    International Nuclear Information System (INIS)

    Cai Zongying; Xing Xianran; Li Lu; Xu Yeming

    2008-01-01

    Lead lanthanum zirconate titanate (Pb 0.95 La 0.03 )(Zr 0.52 Ti 0.48 )O 3 (PLZT) was synthesized by one step molten salt method with the starting materials of PbC 2 O 4 , La 2 O 3 , ZrO(NO 3 ) 2 .2H 2 O and TiO 2 in the NaCl-KCl eutectic mixtures in the temperature range of 700-1000 deg. C. The single phase of (Pb 0.95 La 0.03 )(Zr 0.52 Ti 0.48 )O 3 powders was prepared at a temperature as low as 850 deg. C for 5 h. The effects of process parameters, such as soaking temperature and time, salt species, and the amount of flux with respect to the starting materials were investigated. The growth process of the PLZT particles in the molten salt undergoes a transition from a diffusion controlled mechanism to an interfacial reaction controlled mechanism at 900 deg. C

  2. Applications of molten salts in reactive metals processing

    Energy Technology Data Exchange (ETDEWEB)

    Mishra, B.; Olson, D.L. [Colorado School of Mines, Golden, CO (United States). Kroll Inst. for Extractive Metallurgy; Averill, W.A. [EG and G Rocky Flats, Inc., Golden, CO (United States). Rocky Flats Plant

    1993-12-31

    Pyrochemical processes using molten salts provide a unique opportunity for the extraction and refining of many reactive and valuable metals either directly from the beneficiated ore or from other process effluent that contain reactive metal compounds. This research program is aimed at developing a process for the production and recovery of reactive and valuable metals, such as zinc, tin, lead, bismuth and silver, in a hybrid reactor combining electrolytic production of the calcium reductant and in-situ utilization of this reductant for pyrochemical reduction of the metal compounds, such as halide or oxides. The process is equally suitable for producing other low melting metals, such as cadmium and antimony. The cell is typically operated below 1000C temperature. Attempts have been made to produce silver, lead, bismuth, tin and cerium by calciothermic reduction in a molten salt media. In a separate effort, calcium has been produced by an electrolytic dissociation of lime in a calcium chloride medium. The most important characteristic of the hybrid technology is its ability to produce metals under ``zero-waste`` conditions.

  3. Establishment of cooperation basis of joint research on the mixed waste molten salt oxidation technology

    International Nuclear Information System (INIS)

    Yang, Hee Chul; Cho, Y. J.; Kim, J. H.; Yoo, J. H.; Yun, H. C.; Lee, D. G.

    2005-08-01

    Molten salt oxidation, MSO for short, is a robust technology that can effectively treat mixed waste (radioactive waste including hazardous metals or organics). It can safely and economically treat the difficult wastes such as not-easily destroyable toxic organic waste, medical waste, chemical warfare and energetic materials such as propellant and explosives, all of which are not easily treated by an incinerator or other currently existing thermal treatment system. Therefore, molten salt oxidation technology should be developed and utilized to treat a lot of niche waste stored in the nuclear and environmental industries. So, if we put the MSO technology to practical use by Korea-Vietnam joint research, we can reduce R and D fund for MSO technology by ourselves and we can expect an export of the outcome of nuclear R and D in Korea. For Establishment of cooperation basis of joint research concerning molten salt oxidation technology between KOREA and VIETNAM, in this research, We invited two Vietnamese researchers and we introduced our experimental scale molten salt oxidation system in order to let them understand molten salt oxidation technology. We also visited Viet man and we consulted about molten salt oxidation process. We held seminar on the mixed waste molten salt oxidation technology, discussed on the joint research on the mixed waste molten salt oxidation technology and finally we wrote MOU for joint research

  4. Establishment of cooperation basis of joint research on the mixed waste molten salt oxidation technology

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Hee Chul; Cho, Y. J.; Kim, J. H.; Yoo, J. H.; Yun, H. C.; Lee, D. G

    2005-08-01

    Molten salt oxidation, MSO for short, is a robust technology that can effectively treat mixed waste (radioactive waste including hazardous metals or organics). It can safely and economically treat the difficult wastes such as not-easily destroyable toxic organic waste, medical waste, chemical warfare and energetic materials such as propellant and explosives, all of which are not easily treated by an incinerator or other currently existing thermal treatment system. Therefore, molten salt oxidation technology should be developed and utilized to treat a lot of niche waste stored in the nuclear and environmental industries. So, if we put the MSO technology to practical use by Korea-Vietnam joint research, we can reduce R and D fund for MSO technology by ourselves and we can expect an export of the outcome of nuclear R and D in Korea. For Establishment of cooperation basis of joint research concerning molten salt oxidation technology between KOREA and VIETNAM, in this research, We invited two Vietnamese researchers and we introduced our experimental scale molten salt oxidation system in order to let them understand molten salt oxidation technology. We also visited Viet man and we consulted about molten salt oxidation process. We held seminar on the mixed waste molten salt oxidation technology, discussed on the joint research on the mixed waste molten salt oxidation technology and finally we wrote MOU for joint research.

  5. Novel concepts in electrochemical solar cells. Second quarterly progress report, August 15, 1979-October 15, 1979. [Molten salt electrolytes

    Energy Technology Data Exchange (ETDEWEB)

    DuBow, J.; Job, R.; Krishnan, R.; Gale, B.

    1979-01-01

    It is considered that the short term stability of n-GaAs PEC's in a ferrocene-based, ambient temperature molten salt electrolyte is reasonably good. However, longer term evaluation is required to determine the extent and significance of corrosion, stability, etc. Extremely few fundamental studies have been made of the semiconductor/molten salt interphase and experiments in this area would be most useful. Indeed, even the design parameters for PECs of any kind have not been quantitatively delineated and present consideration will be given to models for PEC solar cells and limitations caused by ion transport in the electrolyte. The MoSe/sub 2/ and MoS/sub 2/ electrodes appear to have substrate reproducibility and transport limitations that make them unsuitable candidates for efficient PEC's at this time. Similarly, the lack of availability of high quality CuInSe/sub 2/ and CuInS/sub 2/ substrates limits the quantitative experimental evaluation of their utility for PEC applications. We are presently focusing attention on CdSe/CdTe mixtures and CdS as electrodes as well as Si and GaAs in molten salt and polyelectrolyte solutions. The system for solar cell evaluation and network analysis of substrates and cells was mode operational. Preliminary work on economic and theoretical modelling was begun. Progress is reported. (WHK)

  6. ANALISIS TRANSIEN PADA PASSIVE COMPACT MOLTEN SALT REACTOR (PCMSR

    Directory of Open Access Journals (Sweden)

    M. Makrus Imron

    2015-04-01

    Full Text Available Penggunaan bahan bakar cair berupa garam LiF-BeF2-ThF4-UF4 pada Passive Compact Molten Salt Reactor (PCMSR meyebabkan pengendalian daya pada PCMSR dapat dilakukan dengan mengendalikan laju aliran bahan bakar dan pendingin. Sedangkan dari sistem keselamatan, penggunaan bahan bakar cair menjadikan PCMSR memiliki karakter keselamatan melekat (inherent safety yang baik. Pada penelitian ini telah dilakukan analisis transien PCMSR pada tiga kondisi, yaitu: ketika terjadi perubahan laju aliran bahan bakar, ketika terjadi perubahan laju aliran pendingin dan ketika terdapat kegagalan pada sistem pelepasan panas (loss of heat sink. Penelitian dilakukan dengan memodelkan reaktor pada kondisi tunak menggunakan paket program. Standart Reactor Analysis Code (SRAC. Selanjutnya dari keluaran paket program SRAC diperoleh data data yang meliputi fluks netron,konstanta grup, kontanta peluran prekusor netron, fraksi netron kasip untuk perhitungan transien. Penelitian ini menunjukkan bahwa penurunan laju aliran bahan bakar sebesar 50 % dari laju bahan bakar sebelumnya, menyebabkan daya pada PCMSR turun menjadi 78 % dari daya sebelumnya. Dan penurunan laju aliran pendingin sebesar 50 % dari laju pendingin sebelumnya, menyebabkan daya pada PCMSR turun menjadi 63 % dari daya sebelumnya. Sedangkan pada saat terjadi loss of heat sink daya PCMSR menunjukkan penurunan. Kata kunci: PCMSR, transien, daya, laju aliran.   The use of liquid fuels in the form of molten salts LiF-BeF2-ThF4-UF4 in Passive Compact Molten Salt Reactor (PCMSR makes power control at PCMSR can be done by controlling the flow rate of fuel and coolant. In addition, from safety systems aspect, the use of liquid fuels makes PCMSR has good inherent safety characteristics. In this study transient analysis has been carried out on three conditions of PCMSR, namely when the fuel flow rate is changing, when the coolant flow rate is changing and when there is loss of heat sink condition. This research is

  7. REAKTOR INNOVATIVE MOLTEN SALT (IMSR DENGAN SISTEM KESELAMATAN PASIF MENYELURUH

    Directory of Open Access Journals (Sweden)

    Andang Widiharto

    2015-04-01

    Full Text Available Pengembangan Teknologi Reaktor Nuklir pada masa mendatang mengarah pada peningkatan aspek keselamatan, peningkatan pendayagunaan bahan bakar, reduksi limbah radioaktif, ketahanan terhadap proliferasi bahan-bakar nuklir dan peningkatan aspek ekonomi. reaktor Innovative Molten Salt (IMSR adalah reaktor nuklir yang menggunakan bahan bakar cair berupa garam lebur fluoride (7LiF-ThF4-UF4-MaFx. Reaktor IMSR didesain sebagai reaktor pembiak termal, yaitu membiakkan U-233 dari Th-232. Hal ini untuk menjawab permasalahan sustainabilitas ketersedian sumber daya bahan bakar nuklir dan reduksi limbah radioaktif. Dalam aspek keselamatan, desain reaktor IMSR memiliki sifat inherent safe, yaitu koefisien umpan balik daya yang negatif serta memiliki fitur-fitur keselamatan pasif. Fitur-fitur keselamatan pasif terdiri dari sistem shutdown pasif, sistem pendinginan pasif pasca shutdown serta sistem pendinginan pasif untuk produk fisi. Kecelakaan yang berpotensi terjadi pada IMSR, yaitu kecelakaan kehilangan aliran bahan bakar, kecelakaan kehilangan aliran pendingin, kecelakaan kehilangan kemampuan pengambilan kalor serta kecelakaan kerusakan integritas sistem reaktor, dapat ditangani sepenuhnya secara pasif hingga mencapai kondisi shutdown selamat. Kata kunci: keselamatan pasif, inherent safe, IMSR   The next Nuclear Reactor Technology developments are directed to the increasing of the aspects of safety, fuel utility, radioactive waste reduction, proliferation retention and economy. Innovative Molten Salt Reactor (IMSR is a nuclear reactor design that uses fluoride molten salt (7LiF-ThF4-UF4-MaFx. IMSR is designed as a thermal breeder reactor, i.e. to produce U-233 from Th-232. This is the answer of natural nuclear fuel sustainability and radioactive waste problems. In term of safety aspect, IMSR design has inherent safe characteristics, i.e. negative power feedback coefficient, and passive safety features. The passive safety features are passive shutdown

  8. Transient core characteristics of small molten salt reactor coupling problem between heat transfer/flow and nuclear fission reaction

    International Nuclear Information System (INIS)

    Yamamoto, Takahisa; Mitachi, Koshi

    2004-01-01

    This paper performed the transient core analysis of a small Molten Salt Reactor (MSR). The emphasis is that the numerical model employed in this paper takes into account the interaction among fuel salt flow, nuclear reaction and heat transfer. The model consists of two group diffusion equations for fast and thermal neutron fluexs, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The results of transient analysis are that (1) fission reaction (heat generation) rate significantly increases soon after step reactivity insertion, e.g., the peak of fission reaction rate achieves about 2.7 times larger than the rated power 350 MW when the reactivity of 0.15% Δk/k 0 is inserted to the rated state, and (2) the self-control performance of the small MSR effectively works under the step reactivity insertion of 0.56% Δk/k 0 , putting the fission reaction rate back on the rated state. (author)

  9. Coupled optical and thermal detailed simulations for the accurate evaluation and performance improvement of molten salts solar towers

    Science.gov (United States)

    García-Barberena, Javier; Mutuberria, Amaia; Palacin, Luis G.; Sanz, Javier L.; Pereira, Daniel; Bernardos, Ana; Sanchez, Marcelino; Rocha, Alberto R.

    2017-06-01

    The National Renewable Energy Centre of Spain, CENER, and the Technology & Innovation area of ACS Cobra, as a result of their long term expertise in the CSP field, have developed a high-quality and high level of detail optical and thermal simulation software for the accurate evaluation of Molten Salts Solar Towers. The main purpose of this software is to make a step forward in the state-of-the-art of the Solar Towers simulation programs. Generally, these programs deal with the most critical systems of such plants, i.e. the solar field and the receiver, on an independent basis. Therefore, these programs typically neglect relevant aspects in the operation of the plant as heliostat aiming strategies, solar flux shapes onto the receiver, material physical and operational limitations, transient processes as preheating and secure cloud passing operating modes, and more. The modelling approach implemented in the developed program consists on effectively coupling detailed optical simulations of the heliostat field with also detailed and full-transient thermal simulations of the molten salts tube-based external receiver. The optical model is based on an accurate Monte Carlo ray-tracing method which solves the complete solar field by simulating each of the heliostats at once according to their specific layout in the field. In the thermal side, the tube-based cylindrical external receiver of a Molten Salts Solar Tower is modelled assuming one representative tube per panel, and implementing the specific connection layout of the panels as well as the internal receiver pipes. Each tube is longitudinally discretized and the transient energy and mass balances in the temperature dependent molten salts and steel tube models are solved. For this, a one dimensional radial heat transfer model based is used. The thermal model is completed with a detailed control and operation strategy module, able to represent the appropriate operation of the plant. An integration framework has been

  10. A facile molten-salt route to graphene synthesis.

    Science.gov (United States)

    Liu, Xiaofeng; Giordano, Cristina; Antonietti, Markus

    2014-01-15

    Efficient synthetic routes are continuously pursued for graphene in order to implement its applications in different areas. However, direct conversion of simple monomers to graphene through polymerization in a scalable manner remains a major challenge for chemists. Herein, a molten-salt (MS) route for the synthesis of carbon nanostructures and graphene by controlled carbonization of glucose in molten metal chloride is reported. In this process, carbohydrate undergoes polymerization in the presence of strongly interacting ionic species, which leads to nanoporous carbon with amorphous nature and adjustable pore size. At a low precursor concentration, the process converts the sugar molecules (glucose) to rather pure few-layer graphenes. The MS-derived graphenes are strongly hydrophobic and exhibit remarkable selectivity and capacity for absorption of organics. The methodology described may open up a new avenue towards the synthesis and manipulation of carbon materials in liquid media. Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  11. Electrodeposition of alloys or compounds in molten salts and applications

    Directory of Open Access Journals (Sweden)

    Taxil P.

    2003-01-01

    Full Text Available This article deals with the different modes of preparation of alloys or intermetallic compounds using the electrodeposition in molten salts, more particularly molten alkali fluorides. The interest in this process is to obtain new materials for high technology, particularly the compounds of reactive components such as actinides, rare earth and refractory metals. Two ways of preparation are considered: (i electrocoating of the more reactive metal on a cathode made of the noble one and reaction between the two metals in contact, and (ii electrocoating on an inert cathode of the intermetallic compound by coreduction of the ions of each elements. The kinetic is controlled by the reaction at the electrolyte interface. A wide bibliographic survey on the preparation of various compounds (intermetallic compounds, borides, carbides… is given and a special attention is paid to the own experience of the authors in the preparation of these compounds and interpretation of their results.

  12. Molten salt parabolic trough system with synthetic oil preheating

    Science.gov (United States)

    Yuasa, Minoru; Hino, Koichi

    2017-06-01

    Molten salt parabolic trough system (MSPT), which can heat the heat transfer fluid (HTF) to 550 °C has a better performance than a synthetic oil parabolic trough system (SOPT), which can heat the HTF to 400 °C or less. The utilization of HTF at higher temperature in the parabolic trough system is able to realize the design of a smaller size of storage tank and higher heat to electricity conversion efficiency. However, with MSPT there is a great amount of heat loss at night so it is necessary to circulate the HTF at a high temperature of about 290 °C in order to prevent solidification. A new MSPT concept with SOPT preheating (MSSOPT) has been developed to reduce the heat loss at night. In this paper, the MSSOPT system, its performance by steady state analysis and annual performance analysis are introduced.

  13. Filbe molten salt research for tritium breeder applications

    International Nuclear Information System (INIS)

    Anderl, R.A.; Petti, D.A.; Smolik, G.R.

    2004-01-01

    This paper presents an overview of Flibe (2Lif·BeF 2 ) 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)

  14. Modelization of the SECM in molten salts environment

    International Nuclear Information System (INIS)

    Lucas, M.; Slim, C.; Di Caprio, D.; Delpech, S.; Stafiej, J.

    2013-01-01

    We develop a cellular automata simulation of SECM (Scanning Electrochemical Microscopy)experiments to study corrosion in molten salt media for generation IV nuclear reactors. The electrodes used in these experiments are cylindrical glass tips with a coaxial metal wire inside. As the result of simulations we obtain the current approach curves of the electrodes with geometries characterized by several values of the ratios of glass to metal area at the tip. We compare these results with predictions of the known analytic expressions, solutions of partial differential equations for flat uniform geometry of the substrate. We present also the results for other, more complicated substrate surface geometries e. g. regular saw modulated surface, or surface obtained by an Eden model process. We show that with a simple cellular automata model we can reasonably well simulate the results of SECM setup. The stochastic resolution of the diffusion equations is made possible by the parallel code implemented on GPU

  15. Molten salt destruction of rubber and chlorinated solvents

    International Nuclear Information System (INIS)

    Upadhye, R.S.; Wilder, J.G.

    1994-09-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); mckellar, Michael George [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yoon, Su-Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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

  17. On purpose simulation model for molten salt CSP parabolic trough

    Science.gov (United States)

    Caranese, Carlo; Matino, Francesca; Maccari, Augusto

    2017-06-01

    The utilization of computer codes and simulation software is one of the fundamental aspects for the development of any kind of technology and, in particular, in CSP sector for researchers, energy institutions, EPC and others stakeholders. In that extent, several models for the simulation of CSP plant have been developed with different main objectives (dynamic simulation, productivity analysis, techno economic optimization, etc.), each of which has shown its own validity and suitability. Some of those models have been designed to study several plant configurations taking into account different CSP plant technologies (Parabolic trough, Linear Fresnel, Solar Tower or Dish) and different settings for the heat transfer fluid, the thermal storage systems and for the overall plant operating logic. Due to a lack of direct experience of Molten Salt Parabolic Trough (MSPT) commercial plant operation, most of the simulation tools do not foresee a suitable management of the thermal energy storage logic and of the solar field freeze protection system, but follow standard schemes. ASSALT, Ase Software for SALT csp plants, has been developed to improve MSPT plant's simulations, by exploiting the most correct operational strategies in order to provide more accurate technical and economical results. In particular, ASSALT applies MSPT specific control logics for the electric energy production and delivery strategy as well as the operation modes of the Solar Field in off-normal sunshine condition. With this approach, the estimated plant efficiency is increased and the electricity consumptions required for the plant operation and management is drastically reduced. Here we present a first comparative study on a real case 55 MWe Molten Salt Parabolic Trough CSP plant placed in the Tibetan highlands, using ASSALT and SAM (System Advisor Model), which is a commercially available simulation tool.

  18. A study on the corrosion test of equipment material handling hot molten salt

    International Nuclear Information System (INIS)

    Ro, Seung Gy; Jeong, M.S.; Hong, S.S.; Cho, S.H.; Shin, Y.J.; Park, H.S.; Zhang, J.S.

    1999-02-01

    On this technical report, corrosion behavior of austenitic stainless steels of SUS 316L and SUS 304L in molten salt of LiCl-Li 2 O has been investigated in the temperature range of 650 - 850 dg C. Corrosion products of SUS 316L in molten salt consisted of two layers, an outer layer of LiCrO 2 and inner layer of Cr 2 O 3 .The corrosion layer was uniform in molten salt of LiCl, but the intergranular corrosion occurred in addition to the uniform corrosion in mixed molten salt of LiCl-Li 2 O. The corrosion rate increased slowly with the increase of temperature up to 750 dg C, but above 750 dg C rapid increase in corrosion rate observed. SUS 316L stainless steel showed slower corrosion rate and higher activation energy for corrosion than SUS 304L, exhibiting higher corrosion resistance in the molten salt. In heat-resistant alloy, dense protective oxide scale of LiCrO 2 was formed in molten salt of LiCl. Whereas in mixed molten salt of LiCl-Li 2 O, porous non-protective scale of Li(Cr, Ni, Fe)O 2 was formed. (Author). 44 refs., 4 tabs., 16 figs

  19. Molten salt oxidation of organic hazardous waste with high salt content.

    Science.gov (United States)

    Lin, Chengqian; Chi, Yong; Jin, Yuqi; Jiang, Xuguang; Buekens, Alfons; Zhang, Qi; Chen, Jian

    2018-02-01

    Organic hazardous waste often contains some salt, owing to the widespread use of alkali salts during industrial manufacturing processes. These salts cause complications during the treatment of this type of waste. Molten salt oxidation is a flameless, robust thermal process, with inherent capability of destroying the organic constituents of wastes, while retaining the inorganic ingredients in the molten salt. In the present study, molten salt oxidation is employed for treating a typical organic hazardous waste with a high content of alkali salts. The hazardous waste derives from the production of thiotriazinone. Molten salt oxidation experiments have been conducted using a lab-scale molten salt oxidation reactor, and the emissions of CO, NO, SO 2 , HCl and dioxins are studied. Impacts are investigated from the composition of the molten salts, the types of feeding tube, the temperature of molten carbonates and the air factor. Results show that the waste can be oxidised effectively in a molten salt bath. Temperature of molten carbonates plays the most important role. With the temperature rising from 600 °C to 750 °C, the oxidation efficiency increases from 91.1% to 98.3%. Compared with the temperature, air factor has but a minor effect, as well as the composition of the molten salts and the type of feeding tube. The molten carbonates retain chlorine with an efficiency higher than 99.9% and the emissions of dioxins are below 8 pg TEQ g -1 sample. The present study shows that molten salt oxidation is a promising alternative for the disposal of organic hazardous wastes containing a high salt content.

  20. 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 600ÀC. 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.

  1. Actinide removal from molten salts by chemical oxidation and salt distillation

    Energy Technology Data Exchange (ETDEWEB)

    McNeese, J.A.; Garcia, E.; Dole, V.R. [Los Alamos National Laboratory, NM (United States)] [and others

    1995-10-01

    Actinide removal from molten salts can be accomplished by a two step process where the actinide is first oxidized to the oxide using a chemical oxidant such as calcium carbonate or sodium carbonate. After the actinide is precipitated as an oxide the molten salt is distilled away from the actinide oxides leaving a oxide powder heel and an actinide free distilled salt that can be recycled back into the processing stream. This paper discusses the chemistry of the oxidation process and the physical conditions required to accomplish a salt distillation. Possible application of an analogous process sequence for a proposed accelerator driven transmutation molten salt process is also discussed.

  2. Regeneration of WC-Co Nanopowders Via Sodiothermic Reduction in Molten Salts

    Science.gov (United States)

    Wang, Na; Liu, Xue-Mei; Chai, Li-Hua; Wu, Jinyu; Shen, Xuyang

    WC-Co nanoparticles were directly recycled via sodiothermic reduction in NaCl-52mol%CaCl2 molten salts using oxidized hard metal scrap as the raw materials. The as-prepared samples were characterized by X-ray diffraction (XRD) and scanning electron microscopy (SEM). It was found that a series of cemented carbide, such as W, W-Co, WC and WC-Co nanoparticles, were successfully recycled using the molten salt systems. The results show that WC/WC-Co nanoparticles can be recycled form oxidized hard metal scrap via sodiothermic reduction in the NaCl-52mol%CaCl2 molten salts.

  3. 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.

  4. Polyethylene encapsulation of molten salt oxidation mixed low-level radioactive salt residues

    International Nuclear Information System (INIS)

    Lageraaen, P.R.; Kalb, P.D.; Grimmett, D.L.; Gay, R.L.; Newman, C.D.

    1995-01-01

    A limited scope treatability study was conducted for polyethylene encapsulation of salt residues generated by a Molten Salt Oxidation (MSO) technology demonstration at the Energy Technology Engineering Center (ETEC), operated by Rockwell International for the US Department of Energy (DOE). During 1992 and 1993, ETEC performed a demonstration with a prototype MSO unit and treated approximately 50 gallons of mixed waste comprised of radioactively contaminated oils produced by hot cell operations. A sample of the mixed waste contaminated spent salt was used during the BNL polyethylene encapsulation treatability study. A nominal waste loading of 50 wt % was successfully processed and waste form test specimens were made for Toxicity Characteristic Leaching Procedure (TCLP) testing. The encapsulated product was compared with base-line TCLP results for total chromium and was found to be well within allowable EPA guidelines

  5. Incorporating supercritical steam turbines into molten-salt power tower plants :

    Energy Technology Data Exchange (ETDEWEB)

    Pacheco, James Edward; Wolf, Thorsten; Muley, Nishant

    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 600ÀC were evaluated, which resulted in main steam temperatures of 553 and 580ÀC, respectively. Also, the effects of final feedwater temperature (between 260 and 320ÀC) 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 600ÀC and the other 565ÀC. 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 565ÀC. Based on the results of this study, it appears economically and technically feasible to incorporate supercritical steam turbines in molten-salt power tower plants.

  6. 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.

  7. Effect of lithium nitrate and calcium nitrate composition on the thermal properties of quaternary molten salts mixture for heat transfer application

    Science.gov (United States)

    Ahmad, Nuratikah Nadhirah Binti; Yunos, Nursyafiqah Binti; Muhammad, Wan Nur Azrina Binti Wan; Mohamad, Md Nor Anuar Bin; Yusof, Farazila Binti

    2017-10-01

    Mixed molten salt is considered as a promising medium for both heat transfer and energy storage in thermal power because of its many advantages such as low the melting point, large heat capacity, good thermal stability and low cost. In order to determine the thermal properties of the molten salt, the nitrate quaternary mixture of the molten salt are prepared based on the different composition of the lithium nitrate and calcium nitrate. The other salts in the mixture are potassium nitrate and sodium nitrate. Mixture of molten salts were heated in furnace at 150°C for 4 hours and increased the temperature to 400°C for 8 hours for homogenize the salt mixture. Then, decreased the temperature to 115°C for an hour. The melting point and thermal stability of the mixture was determined by using thermogravimetric analysis [TGA] while the heat capacity was determined by using differential scanning calorimetry [DSC]. The lowest melting point using different composition of lithium nitrate, 10wt%NaNO3, 4Owt%KNO3, 20wt%Ca[NO3]2, 30wt%LiNO3 which is 97.1°C while the heat capacity is 6.33 J/g°C. For the experiment using various composition of calcium nitrate, the lowest melting point is from 14wt% NaNO3 + 48wt% KNO3 + 13wt% LiNO3 + 25wt% Ca[NO3]2 which is 111.7°C and the heat capacity is 2.06 J/g °C. From the result, the addition of lithium nitrate in the quaternary molten salts give more effect to the reduction of melting point value but higher heat capacity.

  8. Proceedings of the workshop on molten salts technology and computer simulation

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Hirokazu; Minato, Kazuo (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    Applications of molten salts technology to separation and synthesis of materials have been studied eagerly, which would develop new fields of materials science. Research Group for Actinides Science, Department of Materials Science, Japan Atomic Energy Research Institute (JAERI), together with Reprocessing and Recycle Technology Division, Atomic Energy Society of Japan, organized the Workshop on Molten Salts Technology and Computer Simulation at Tokai Research Establishment, JAERI on July 18, 2001. In the workshop eleven lectures were made and lively discussions were there on the fundamentals and applications of the molten salts technology that covered the structure and basic properties of molten salts, the pyrochemical reprocessing technology and the relevant computer simulation. The 10 of the presented papers are indexed individually. (J.P.N.)

  9. System Requirements Document for the Molten Salt Reactor Experiment 233U conversion system

    International Nuclear Information System (INIS)

    Aigner, R.D.

    2000-01-01

    The purpose of the conversion process is to convert the 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

  10. 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)

  11. Molten salt reactors. Synthesis of studies realized between 1973 and 1983. Metallic materials file

    International Nuclear Information System (INIS)

    1983-03-01

    Metallic materials for molten salt reactors are studied. The corrosion of steels by the eutectic LiF-BeF 2 and molten lead is examined. Fabrication, aging and welding of molybdenum or TZM tubes are also examined. [fr

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

    International Nuclear Information System (INIS)

    Lee, Sung Ho; Kim, In Tae; Park, Sung Bin

    2014-01-01

    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

  13. 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.

  14. Status of the French research in the field of molten salt nuclear reactors

    International Nuclear Information System (INIS)

    Hery, M.; Israel, M.; Fauger, P.; Lecocq, A.

    1977-01-01

    The research program of the CEA in the field of molten salt nuclear reactors has been concerned with MSBR type reactors (Molten Salt Breeder Reactor). The papers written after having performed the theoretical analysis are entitled: core, circuits, chemistry and economy; they include some criticisms and suggestions. The experimental studies consisted in: graphite studies, chemical studies of the salt, metallic materials, the salt loop and the lead loop [fr

  15. Intercalation of FeCl sub 3 into graphite films in molten salts

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Z.D. (Materials Science, Toyohashi Univ. of Tech. (Japan)); Inagaki, M. (Faculty of Engineering, Hokkaido Univ., Sapporo (Japan))

    1991-08-01

    The intercalation of FeCl{sub 3} into graphite films using the FeCl{sub 3}-KCl molten salt system was studied and compared to the intercalation from the vapor phase of FeCl{sub 3}. The intercalation in the molten salt proceeded gradually by changing the stage structure, in contrast to that in vapor phase where initially a stage-1 compound was formed in coexistence with unreacted graphite. The composition of the molten salts (FeCl{sub 3}/KCl ratio) and the addition of chlorine gas were found to affect strongly the intercalation rate and equilibrium stage structure. An exfoliation of the graphite film occurred after intercalation in the molten salts at high temperatures (above 400degC) and produced electrical conductivities much lower than the pristine films. The intercalated films synthesized at 300degC have conductivities about four times higher than the pristine films. From these experimental results, the diffusion rate of FeCl{sub 3} into the graphite galleries is found to be faster in the molten salts than from the vapor phase and to depend on the composition of the molten salts. (orig.).

  16. Use of Molten Salt Fluxes and Cathodic Protection for Preventing the Oxidation of Titanium at Elevated Temperatures

    Science.gov (United States)

    Schwandt, Carsten; Fray, Derek J.

    2014-12-01

    The current study demonstrates that it is possible to protect both solid and liquid titanium and titanium alloys from attack from air by cathodically polarizing the titanium component using an electro-active high-temperature molten salt flux and a moderate polarization potential. The electrolytic cell used comprises a cathode of either solid titanium or liquid titanium alloy, an electrolyte based on molten calcium chloride or fluoride salt, and an anode consisting of an inert oxygen-evolving material such as iridium metal. The new approach renders possible the processing of titanium at elevated temperatures in the presence of oxygen-containing atmospheres.

  17. Molten salt power towers operating at 600–650 °C: Salt selection and cost benefits

    Energy Technology Data Exchange (ETDEWEB)

    Turchi, Craig S. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Vidal, Judith [National Renewable Energy Lab. (NREL), Golden, CO (United States); Bauer, Matthew

    2018-04-01

    This analysis examines the potential benefit of adopting the supercritical carbon dioxide (sCO2) Brayton cycle at 600-650 degrees C compared to the current state-of-the-art power tower operating a steam-Rankine cycle with solar salt at approximately 574 degrees C. The analysis compares a molten-salt power tower configuration using direct storage of solar salt (60:40 wt% sodium nitrate: potassium nitrate) or single-component nitrate salts at 600 degrees C or alternative carbonate- or chloride-based salts at 650 degrees C.

  18. 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.

  19. Electrodeposition nucleation and film formation of cadmium from a low melting molten salt

    Science.gov (United States)

    Agarwal, Rajev

    Consider the simple electrodeposition reaction of a metal: Mz+ + ze- M (at the cathode of the base metal)Whereby metal ions, Mz+, are reduced to metal atoms, M, at the interface of the base metal and electrolyte. The occurrence of such an electrode reaction does not guarantee the nucleation of M or the formation of a compact film of metal M on the base metal. In cases where deposition does occur, metal crystals are formed at fixed sites on the base metal. These metal crystals grow into a polycrystalline film. In this work, this concept of metal deposition is studied. The studies were carried out on cadmium using nickel as a base metal and purified chloroaluminate molten salt electrolyte under purified helium. Cadmium is typically electroplated from aqueous electrolytes. Therefore, the first part of this work consisted of developing an electroplating process for the deposition of this metal from a molten salt electrolyte. This consisted of selection of an electrolyte, cadmium salt, base metal, and experimental proof of cadmium deposition in the selected system. The second part consisted of identification of the cadmium electrode reactions in the chloroaluminate molten salt electrolyte using cyclic voltammetry. When cadmium chloride was added to the chloroaluminate melt, the divalent salt dissolved to form a mixture of divalent and subvalent ions of cadmium in the electrolyte. During cathodic potential scans, these cadmium ions produced two reduction peaks. The first cathodic peak was from reduction of divalent ions to subvalent ions, and the second peak from reduction of subvalent ions to cadmium metal. During anodic potential scans, two oxidation peaks were produced due to reverse reactions of cadmium ions and metal. In addition, a third pair of cadmium peaks was noted to form between the two bulk pairs of cadmium peaks. This third pair of peaks was from chemical dissociation of subvalent ions to divalent ions and cadmium, and coupled electrochemical reactions of

  20. Metallurgical electrochemistry: the interface between materials science and molten salt chemistry

    International Nuclear Information System (INIS)

    Sadoway, D.R.

    1991-01-01

    Even though molten salt electrolysis finds application in the primary extraction of metals (electrowinning), the purification and recycling of metals (electrorefining), and in the formation of metal coatings (electroplating), the technology remains in many respects underexploited. Electrolysis in molten salts as well as other nonaqueous media has enormous potential for materials processing. First, owing to the special attributes of nonaqueous electrolytes electrochemical processing in these media has an important role to play in the generation of advanced materials, i.e., materials with specialized chemistries or tailored microstructures (electrosynthesis). Secondly, as environmental quality standards rise beyond the capabilities of classical metals extraction technologies to comply, molten salt electrolysis may prove to be the only acceptable route from ore to metal. Growing public awareness of pollution from the metals industry could stimulate a renaissance in molten salt electrochemistry. Challenges facing metallurgical electrochemistry as relates to the environment fall into two categories: (1) improving existing electrochemical technology, and (2) developing clean electrochemical technology to displace current nonelectrochemical technology. In both instances success hinges upon the discovery of advanced materials and the ecologically sound extraction of metals, the close coupling between materials science and molten salt chemistry is manifest. (author) 6 refs

  1. Development of high temperature molten salt transport technology for pyrometallurgical reprocessing

    International Nuclear Information System (INIS)

    Hijikata, Takatoshi; Koyama, Tadafumi

    2009-01-01

    Pyrometallurgical reprocessing technology is currently being focused in many countries for closing actinide fuel cycle because of its favorable economic potential and an intrinsic proliferation-resistant feature due to the inherent difficulty of extracting weapons-usable plutonium. The feasibility of pyrometallurgical reprocessing has been demonstrated through many laboratory scale experiments. Hence the development of the engineering technology necessary for pyrometallurgical reprocessing is a key issue for industrial realization. The development of high-temperature transport technologies for molten salt and liquid cadmium is crucial for pyrometallurgical processing; however, there have been very few transport studies on high-temperature fluids. In this study, a salt transport test rig was installed in an argon glove box with the aim of developing technologies for transporting molten salt at approximately 773 K. The gravitation transport of the molten salt at approximately 773 K could be well controlled at a velocity from 0.1 to 1.2 m/s by adjusting the valve. Consequently, the flow in the molten salt can be controlled from laminar flow to turbulent flow. It was demonstrated that; using a centrifugal pump, molten salt at approximately 773 K could be transported at a controlled rate from 2.5 to 8 dm 3 /min against a 1 m head. (author)

  2. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    International Nuclear Information System (INIS)

    Wishau, R.; Ramsey, K.B.; Montoya, A.

    1998-01-01

    This paper presents the technical and economic feasibility of molten salt oxidation technology as a volume reduction and recovery process for 238 Pu contaminated waste. Combustible low-level waste material contaminated with 238 Pu residue is destroyed by oxidation in a 900 C molten salt reaction vessel. The combustible waste is destroyed creating carbon dioxide and steam and a small amount of ash and insoluble 2328 Pu in the spent salt. The valuable 238 Pu is recycled using aqueous recovery techniques. Experimental test results for this technology indicate a plutonium recovery efficiency of 99%. Molten salt oxidation stabilizes the waste converting it to a non-combustible waste. Thus installation and use of molten salt oxidation technology will substantially reduce the volume of 238 Pu contaminated waste. Cost-effectiveness evaluations of molten salt oxidation indicate a significant cost savings when compared to the present plans to package, or re-package, certify and transport these wastes to the Waste Isolation Pilot Plant for permanent disposal. Clear and distinct cost advantages exist for MSO when the monetary value of the recovered 238 Pu is considered

  3. 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

  4. Molten salt thermal energy storage systems: salt selection

    Energy Technology Data Exchange (ETDEWEB)

    Maru, H.C.; Dullea, J.F.; Huang, V.S.

    1976-08-01

    A research program aimed at the development of a molten salt thermal energy storage system commenced in June 1976. This topical report describes Work performed under Task I: Salt Selection is described. A total of 31 inorganic salts and salt mixtures, including 9 alkali and alkaline earth carbonate mixtures, were evaluated for their suitability as heat-of-fusion thermal energy storage materials at temperatures of 850 to 1000/sup 0/F. Thermophysical properties, safety hazards, corrosion, and cost of these salts were compared on a common basis. We concluded that because alkali carbonate mixtures show high thermal conductivity, low volumetric expansion on melting, low corrosivity and good stability, they are attractive as heat-of-fusion storage materials in this temperature range. A 35 wt percent Li/sub 2/CO/sub 3/-65 wt percent K/sub 2/CO/sub 3/ (50 mole percent Li/sub 2/CO/sub 3/-50 mole percent K/sub 2/CO/sub 3/) mixture was selected as a model system for further experimental work. This is a eutectoid mixture having a heat of fusion of 148 Btu/lb (82 cal/g) that forms an equimolar compound, LiKCO/sub 3/. The Li/sub 2/CO/sub 3/-K/sub 2/CO/sub 3/ mixture is intended to serve as a model system to define heat transfer characteristics, potential problems, and to provide ''first-cut'' engineering data required for the prototype system. The cost of a thermal energy storage system containing this mixture cannot be predicted until system characteristics are better defined. However, our comparison of different salts indicated that alkali and alkaline earth chlorides may be more attractive from a salt cost point of view. The long-term corrosion characteristics and the effects of volume change on melting for the chlorides should be investigated to determine their overall suitability as a heat-of-fusion storage medium.

  5. Molten salts as stripping media for radioactive superficial decontamination

    Energy Technology Data Exchange (ETDEWEB)

    De Oliveira Lainetti, Paulo Ernesto [Centro de Quimica e Meio Ambiente - CQMA, Instituto de Pesquisas Energeticas e Nucleares IPEN-CNEN/SP, Av. Prof. Lineu Prestes, 2242 C. Universitaria Sao Paulo - SP, CEP- 5508-000 (Brazil)

    2010-07-01

    developments in these fields were started. The mentioned steel structures, during the operations and after ten or twelve years after the facilities shut down, have presented severe corrosion. In the past, to protect them, several layers of paint were applied. Traditional decontamination methods were tried, such as acid pickling, alkaline washing and ultrasonic baths. Nevertheless, these methods have failed to reach effective decontamination. In this paper are described some aspects and problems in decommissioning of IPEN's nuclear fuel cycle facilities and is presented an innovative method for radioactive superficial decontamination of steel structures, with paint and corrosion products removal, using different molten salt compositions and temperatures as stripping media. (author)

  6. Preliminary analysis of the PreFlexMS molten salt once-through steam generator dynamics and control strategy

    Science.gov (United States)

    Trabucchi, Stefano; Casella, Francesco; Maioli, Tommaso; Elsido, Cristina; Franzini, Davide; Ramond, Mathieu

    2017-06-01

    Concentrated Solar Power plants (CSP) coupled with thermal storage have the potential to guarantee both flexible and continuous energy production, thus being competitive with conventional fossil fuel and hydro power plants, in terms of dispatchability and provision of ancillary services. Hence, the plant equipment and control design have to be focused on flexible operation on one hand, and on plant safety concerning the molten salt freezing on the other hand. The PreFlexMS European project aims to introduce a molten salt Once-Through Steam Generator (OTSG) within a Rankine cycle based power unit, a technology that has greater flexibility potential if compared to steam drum boilers, currently used in CSP plants. The dynamic modelling and simulation from the early design stages is, thus, of paramount importance, to assess the plant dynamic behavior and controllability, and to predict the achievable closed-loop dynamic performance, potentially saving money and time during the detailed design, construction and commissioning phases. The present paper reports the main results of the analysis carried out during the first part of the project, regarding the system analysis and control design. In particular, two different control systems have been studied and tested with the plant dynamic model: a decentralized control strategy based on PI controllers and a Linear Model Predictive Control (LMPC).

  7. 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

  8. 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.

  9. Preparation of Ferrotitanium Alloy from Ilmenite by Electrochemical Reduction in Chloride Molten Salts

    Science.gov (United States)

    Qi, Can-can; Hua, Yi-xin; Chen, Kong-hao; Jie, Ya-fei; Zhou, Zhong-ren; Ru, Juan-jian; Xiong, Li; Gong, Kai

    2016-02-01

    Ferrotitanium alloy is prepared by electrochemical reduction from ilmenite in LiCl-KCl and LiCl-KCl-CaCl2 molten salts, respectively. The products prepared are observed by x-ray diffraction (XRD). It is shown that Fe2Ti can be prepared from ilmenite in LiCl-KCl molten salt at 1073 K with a cell voltage of 3.2 V. Ilmenite can be electrochemically reduced to FeTi in LiCl-KCl-CaCl2 molten salt under the same condition. It is indicated that CaCl2 can promote the reaction and is favors the deoxidization of the FeTiO3.

  10. 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

  11. Comparative study of ATW systems using solid alloy fuel and molten salt fuel

    International Nuclear Information System (INIS)

    Mulyaman, Maman; Silakhuddin

    2002-01-01

    Burning of radioactive wastes can be carried out by transmutation process using a pressurized water reactor, fast breeder reactor and a high energy particle accelerator system. The transmutation using accelerator system which is known as Accelerator Transmutation of Waste (ATW) can be performed in two systems: solid alloy fuel system and molten salt for fuel and cooling system. A comparison of those methods has been assessed. The k e ff values are 0.89 for solid alloy fuel system and 0.92 for molten salt system, with same proton energy of 1.5 GeV

  12. Economic Optimization of a Concentrating Solar Power Plant with Molten-salt Thermocline Storage

    OpenAIRE

    Flueckiger, S. M.; Iverson, B. D.; Garimella, S V

    2014-01-01

    System-level simulation of a molten-salt thermocline tank is undertaken in response to year-long historical weather data and corresponding plant control. Such a simulation is enabled by combining a finite-volume model of the tank that includes a sufficiently faithful representation at low computation cost with a system-level power tower plant model. Annual plant performance of a 100 MWe molten-salt power tower plant is optimized as a function of the thermocline tank size and the plant solar m...

  13. 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.

  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.

  15. Centrifugal separation for miscible solutions: Fundamentals and applications to separation of molten salt nuclear material

    International Nuclear Information System (INIS)

    Ning Li; Camassa, R.; Ecke, R.E.

    1995-01-01

    The authors report on the physical separation of dilute solutions using centrifugal techniques. They use numerical simulations of the diffusion and sedimentation dynamics of centrifugation to model the approach to an equilibrium concentration profile. They verify experimentally the equilibrium profiles for aqueous solutions of different salts under rotation at 25000 rpm corresponding to centrifugal accelerations of about 57,000 g and 75,000 g in two different commercial centrifuges. These measurements provide ratios of sedimentation and diffusion coefficients. The authors show experimental results for the dynamics of separation that confirm the predictions of the theoretical model. They also measure the mass diffusion coefficient for several solutions. Although the relaxation to equilibrium is long, they have determined a method for efficiently extracting enriched components from a ternary mixture based on fast dynamics at early times. These dynamics are modeled in numerical simulations with realistic fluid parameters. Based on these studies the authors show that a multistage centrifugal separation process could provide efficient physical separation of actinides and fission products from a molten-salt solution in proposed transmutation/energy-production systems. The authors consider technical issues in the design of such a separation system

  16. Centrifugal separation for miscible solutions: Fundamentals and applications to separation of molten salt nuclear material

    International Nuclear Information System (INIS)

    Li Ning; Camassa, Roberto; Ecke, Robert E.; Venneri, Francesco

    1995-01-01

    We report on the physical separation of dilute solutions using centrifugal techniques. We use numerical simulations of the diffusion and sedimentation dynamics of centrifugation to model the approach to an equilibrium concentration profile. We verify experimentally the equilibrium profiles for aqueous solutions of different salts under rotation at 25000 rpm corresponding to centrifugal accelerations of about 57,000 g and 75,000 g in two different commercial centrifuges. These measurements provide ratios of sedimentation and diffusion coefficients. We show experimental results for the dynamics of separation that confirm the predictions of the theoretical model. We also measure the mass diffusion coefficient for several solutions. Although the relaxation to equilibrium is long, we have determined a method for efficiently extracting enriched components from a ternary mixture based on fast dynamics at early times. These dynamics are modeled in numerical simulations with realistic fluid parameters. Based on these studies we show that a multistage centrifugal separation process could provide efficient physical separation of actinides and fission products from a molten-salt solution in proposed transmutation/energy-production systems. We consider technical issues in the design of such a separation system

  17. Centrifugal separation for miscible solutions: Fundamentals and applications to separation of molten salt nuclear material

    Energy Technology Data Exchange (ETDEWEB)

    Ning Li; Camassa, R.; Ecke, R.E. [Los Alamos National Laboratory, NM (United States)] [and others

    1995-10-01

    The authors report on the physical separation of dilute solutions using centrifugal techniques. They use numerical simulations of the diffusion and sedimentation dynamics of centrifugation to model the approach to an equilibrium concentration profile. They verify experimentally the equilibrium profiles for aqueous solutions of different salts under rotation at 25000 rpm corresponding to centrifugal accelerations of about 57,000 g and 75,000 g in two different commercial centrifuges. These measurements provide ratios of sedimentation and diffusion coefficients. The authors show experimental results for the dynamics of separation that confirm the predictions of the theoretical model. They also measure the mass diffusion coefficient for several solutions. Although the relaxation to equilibrium is long, they have determined a method for efficiently extracting enriched components from a ternary mixture based on fast dynamics at early times. These dynamics are modeled in numerical simulations with realistic fluid parameters. Based on these studies the authors show that a multistage centrifugal separation process could provide efficient physical separation of actinides and fission products from a molten-salt solution in proposed transmutation/energy-production systems. The authors consider technical issues in the design of such a separation system.

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

    International Nuclear Information System (INIS)

    Uhlir, J.; Priman, V.; Vanicek, J.

    2001-01-01

    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)

  19. First principles study of the thermodynamic and kinetic properties of U in an electrorefining system using molybdenum cathode and LiCl-KCl eutectic molten salt

    International Nuclear Information System (INIS)

    Kwon, Choah; Kang, Joonhee; Kang, Woojong; Kwak, Dohyun; Han, Byungchan

    2016-01-01

    Using first principles density functional theory (DFT) calculations we obtain thermodynamic and kinetic properties of U in an electrorefining process for spent nuclear fuels using a LiCl-KCl eutectic molten salt and Mo as a cathode. The thermodynamic stability of electrodeposited U from the molten salt onto the Mo(110) surface electrode is evaluated by activity coefficients as function of surface coverages of U and Cl. Additionally, ab-initio molecular dynamic simulations combined with the Stokes-Einstein-Sutherland relation enables us to calculate the viscosity of the LiCl-KCl eutectic molten salt. Our results well agree with previously reported experimental data endorsing the credibility. Based on our atomic-level mechanical understanding we propose that an accurate computational model system incorporating the electrochemical conditions of the electrorefining process essential for the purpose of establishing thermodynamic and kinetic database of U, otherwise critical deviations are inevitable. More interestingly, the effect of coadsorption of Cl with U on the Mo(110) surface plays a key role in stabilizing electrodeposited U on the cathode. Our approach can be useful for validating published experimental database and for identifying key factors guiding a rational design of highly efficient electrorefining system for spent nuclear fuels, and thus reducing high-level radioactive nuclear wastes.

  20. 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...

  1. The origin of the conductivity maximum in molten salts. I. Bismuth chloride.

    Science.gov (United States)

    Clay, Adam T; Kuntz, Colin M; Johnson, Keith E; East, Allan L L

    2012-03-28

    A new theory is presented to explain the conductivity maxima of molten salts (versus temperature and pressure). In the new theory, conductivity is due to ions hopping from counterion to counterion, and its temperature dependence can be explained with an ordinary Arrhenius equation in which the frequency prefactor A (for hopping opportunities) and activation energy E(a) (for hopping) are density dependent. The conductivity maximum is due to competing effects: as density decreases, the frequency of opportunities for hopping increases, but the probability that an opportunity is successfully hopped decreases due to rising E(a) caused by the increased hopping distance. The theory is successfully applied to molten bismuth (III) chloride, and supported by density-functional based molecular dynamics simulations which not only reproduce the conductivity maximum, but disprove the long-standing conjecture that this liquid features an equilibrium between BiCl(3) molecules, and BiCl(2)(+) and BiCl(4)(-) ions that shifts to the left with increasing temperature.

  2. Effect of the graphite electrode material on the characteristics of molten salt electrolytically produced carbon nanomaterials

    International Nuclear Information System (INIS)

    Kamali, Ali Reza; Schwandt, Carsten; Fray, Derek J.

    2011-01-01

    The electrochemical erosion of a graphite cathode during the electrolysis of molten lithium chloride salt may be used for the preparation of nano-structured carbon materials. It has been found that the structures and morphologies of these carbon nanomaterials are dependent on those of the graphite cathodes employed. A combination of tubular and spherical carbon nanostructures has been produced from a graphite with a microstructure of predominantly planar micro-sized grains and a minor fraction of more irregular nano-sized grains, whilst only spherical carbon nanostructures have been produced from a graphite with a microstructure of primarily nano-sized grains. Based on the experimental results, a best-fit regression equation is proposed that relates the crystalline domain size of the graphite reactants and the carbon products. The carbon nanomaterials prepared possess a fairly uniform mesoporosity with a sharp peak in pore size distribution at around 4 nm. The results are of crucial importance to the production of carbon nanomaterials by way of the molten salt electrolytic method. - Highlights: → Carbon nanomaterials are synthesised by LiCl electrolysis with graphite electrodes. → The degree of crystallinity of graphite reactant and carbon product are related. → A graphite reactant is identified that enables the preparation of carbon nanotubes. → The carbon products possess uniform mesoporosity with narrow pore size distribution.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, P.C.

    1997-11-01

    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.

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

    International Nuclear Information System (INIS)

    Hsu, P.C.

    1997-01-01

    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

  5. Experimental and numerical thermal-hydraulics investigation of a molten salt reactor concept core

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2017-09-15

    In the paper measurement results of experimental modelling of a molten salt fast reactor concept will be presented and compared with three-dimensional computational fluid dynamics (CFD) simulation results. Purpose of this article is twofold, on one hand to introduce a geometry modification in order to avoid the disadvantages of the original geometry and discuss new measurement results. On the other hand to present an analysis in order to suggest a method of proper numerical modelling of the problem based on the comparison of calculation results and measurement data for the new, modified geometry. The investigated concept has a homogeneous cylindrical core without any internal structures. Previous measurements on the scaled and segmented plexiglas model of the concept core and simulation results have shown that this core geometry could be optimized for better thermal-hydraulics characteristics. In case of the original geometry strong undesired flow separation could develop, that could negatively affect the characteristics of the core from neutronics point of view as well. An internal flow distributor plate was designed and installed with the purpose of optimizing the flow field in the core by enhancing its uniformity. Particle image velocimetry (PIV) measurement results of the modified experimental model will be presented and compared to numerical simulation results with the purpose of CFD model validation.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gay, R.L.; Grantham, L.F.; Fusselman, S.P. [Rockwell International/Rocketdyne Division, Canoga Park, CA (United States)] [and others

    1995-10-01

    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.

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

    International Nuclear Information System (INIS)

    Yustinus, P.; Indra-Gunawan; Wuryanto

    1996-01-01

    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 o C 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 o C 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 A o ; b = 3.8561 - 3.8895 A o and c = 11.6518 - 11.7104 A o , 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 10 5 Amp/m 2

  8. Laser-Induced Breakdown Spectroscopy (LIBS) in a Novel Molten Salt Aerosol System.

    Science.gov (United States)

    Williams, Ammon N; Phongikaroon, Supathorn

    2017-04-01

    In the pyrochemical separation of used nuclear fuel (UNF), fission product, rare earth, and actinide chlorides accumulate in the molten salt electrolyte over time. Measuring this salt composition in near real-time is advantageous for operational efficiency, material accountability, and nuclear safeguards. Laser-induced breakdown spectroscopy (LIBS) has been proposed and demonstrated as a potential analytical approach for molten LiCl-KCl salts. However, all the studies conducted to date have used a static surface approach which can lead to issues with splashing, low repeatability, and poor sample homogeneity. In this initial study, a novel molten salt aerosol approach has been developed and explored to measure the composition of the salt via LIBS. The functionality of the system has been demonstrated as well as a basic optimization of the laser energy and nebulizer gas pressure used. Initial results have shown that this molten salt aerosol-LIBS system has a great potential as an analytical technique for measuring the molten salt electrolyte used in this UNF reprocessing technology.

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

    NARCIS (Netherlands)

    De Falco, Marcello; Gallucci, F.

    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

  10. Molten salt reactors. Synthesis of studies realized between 1973 and 1983. Carbon-materials file

    International Nuclear Information System (INIS)

    1983-03-01

    The study of a molten salt fueled reactor requires a thorough examination of carbon containing materials for moderator, reflectors and structural materials. Are examined: texture, structure, physical and mechanical properties, chemical purity, neutron irradiation, salt-graphite and salt-lead interactions for different types of graphite. [fr

  11. 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 - others: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/

  12. Wall heat transfer coefficient in a molten salt bubble column: testing the experimental setup

    CSIR Research Space (South Africa)

    Skosana, PJ

    2014-10-01

    Full Text Available with molten salt due to leakages on the welds and aggressive corrosion on the column. All the experiments were run at superficial gas velocities of 0.01–0.05 m/s. Three heating tapes, each connected to a corresponding variable AC voltage controller, were used...

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

    International Nuclear Information System (INIS)

    Longmire, V.L.; Hurd, J.R.; Sedlacek, W.E.; Scarborough, A.M.

    1987-01-01

    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. Effect of Ni-Co Ternary Molten Salt Catalysts on Coal Catalytic Pyrolysis Process

    Science.gov (United States)

    Cui, Xin; Qi, Cong; Li, Liang; Li, Yimin; Li, Song

    2017-08-01

    In order to facilitate efficient and clean utilization of coal, a series of Ni-Co ternary molten salt crystals are explored and the catalytic pyrolysis mechanism of Datong coal is investigated. The reaction mechanisms of coal are achieved by thermal gravimetric analyzer (TGA), and a reactive kinetic model is constructed. The microcosmic structure and macerals are observed by scanning electron microscope (SEM). The catalytic effects of ternary molten salt crystals at different stages of pyrolysis are analyzed. The experimental results show that Ni-Co ternary molten salt catalysts have the capability to bring down activation energy required by pyrolytic reactions at its initial phase. Also, the catalysts exert a preferable catalytic action on macromolecular structure decomposition and free radical polycondensation reactions. Furthermore, the high-temperature condensation polymerization is driven to decompose further with a faster reaction rate by the additions of Ni-Co ternary molten salt crystal catalysts. According to pyrolysis kinetic research, the addition of catalysts can effectively decrease the activation energy needed in each phase of pyrolysis reaction.

  15. Destruction of high explosives and wastes containing high explosives using the molten salt destruction process

    International Nuclear Information System (INIS)

    Upadhye, R.S.; Brummond, W.A.; Pruneda, C.O.

    1992-01-01

    This paper reports the Molten Salt Destruction (MSD) Process which has been demonstrated for the destruction of HE and HE-containing wastes. MSD has been used by Rockwell International and by Anti-Pollution Systems to destroy hazardous wastes. MSD converts the organic constituents (including the HE) of the waste into non-hazardous substances such as carbon dioxide, nitrogen and water. In the case of HE-containing mixed wastes, any actinides in the waste are retained in the molten salt, thus converting the mixed wastes into low-level wastes. (Even though the MSD process is applicable to mixed wastes, this paper will emphasize HE-treatment.) The destruction of HE is accomplished by introducing it, together with oxidant gases, into a crucible containing a molten salt, such as sodium carbonate, or a suitable mixture of the carbonates of sodium, potassium, lithium and calcium. The temperature of the molten salt can be between 400 to 900 degrees C. The combustible organic components of the waste react with oxygen to produce carbon dioxide, nitrogen and steam

  16. Variation of diffusivity with the cation radii in molten salts of ...

    Indian Academy of Sciences (India)

    Home; Journals; Journal of Chemical Sciences; Volume 124; Issue 1. Variation of diffusivity with the cation radii in molten salts of superionic conductors containing iodine anion: A molecular dynamics study. Srinivasa R Varanasi S Yashonath. Volume 124 Issue 1 January 2012 pp 159-166 ...

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

    Energy Technology Data Exchange (ETDEWEB)

    Reddy, Ramana G. [Univ. of Alabama, Tuscaloosa, AL (United States)

    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

  18. PERFORMA NEUTRONIK BAHAN BAKAR LiF-BeF2-ThF4-UF4 PADA SMALL MOBILE-MOLTEN SALT REACTOR

    Directory of Open Access Journals (Sweden)

    S. N. Rokhman

    2015-04-01

    Full Text Available Telah dilakukan analisis terhadap performa neutronik bahan bakar garam lebur LiF-BeF2-ThF4-UF4 pada Small Mobile-Molten Salt Reactor (SM-MSR. Penyesuaian konfigurasi teras dan temperatur operasi harus dilakukan untuk penggunaan bahan bakar baru tersebut agar mencapai keff > 1 dan CR (conversion ratio > 1 pada fraksi 0,5% 233U, 20% 232Th, 28% Li, 51,5% Be. Setelah didapat nilai keff ≈ 1 dan CR ≈ 1, dilakukan analisis pengaruh perubahan Th terhadap Be dan Be terhadap Li yang terlihat dalam perubahan parameter keff dan CR. Setelah itu fraksi 233U divariasi antara 0,5–0,46% untuk memperoleh keff > 1 dan CR > 1. Dalam perhitungan koefisien reaktifitas temperatur (αT, temperatur teras dinaikkan sebesar +25K dan +50K., dan untuk koefisien reaktifitas void (αV, densitas bahan bakar dikurangi hingga 90%. Hasil perhitungan menunjukkan bahwa pengurangan Th terhadap Be menyebabkan penurunan nilai CR dan naiknya keff akibat berkurangnya material fertil. Sebaliknya penambahan Be terhadap Li mengakibatkan terjadi kenaikan nilai keff dan menurunkan CR, akibat laju serapan Li lebih besar dari Be. Pada 5 (lima fraksi 233U dalam rentang 0,5–0,49%, hasil perhitungan keff dan CR masing-masing bervariasi dalam rentang 1,00001 - 1,00327 dan 1,00016 - 1,00731. Untuk faktor puncak daya (PPF, hasil perhitungan memberikan nilai dalam rentang 2,4311 -2,4714. Sedangkan untuk parameter keselamatan, koefisien reaktivitas temperatur (αT dan reaktivitas void (αV masingmasing bernilai negatif dalam rentang 4,972×10-5 - 5,909×10-5 dan 2,596×10-2- 2,8287×10-2 ∆k/k/K. Dapat disimpulkan bahwa teras SM-MSR memberikan nilai negatif di kedua koefisien reaktivitas tersebut untuk setiap fraksi,, sehingga memenuhi kriteria keselamatan dan keselamatan melekat. Kata kunci: SM-MSR (small mobile-molten salt reactor, bahan bakar LiF-BeF2-ThF4-UF4, keselamatan melekat, koefisien reaktivitas temperatur, koefisien reaktivitas void   The analysis of neutronic performance has

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

    International Nuclear Information System (INIS)

    Lecarpentier, D.; Garzenne, C.; Vergnes, J.; Mouney, H.; Delpech, M.

    2002-01-01

    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 233 U is about 2.3), MSBR performance is only possible if there are strict constraints set on the in-line reprocessing unit: all the 233 Pa must be removed from the core so that it can decay on the 233 U 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

  20. 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.

  1. Intermediate heat exchanger and steam generator designs for the HYLIFE-II fusion power plant using molten salts

    International Nuclear Information System (INIS)

    Lee, Y.T.; Hoffman, M.A.

    1992-01-01

    The HYLIFE-II fusion power plant employs the molten salt, Flibe, for the liquid jets which form the self-healing 'first wall' of the reactor. The molten salt, sodium fluoroborate then transports the heat from the IHX's to the steam generators. The design and optimization of the IHX's and the steam generators for use with molten salts has been done as part of the HYLIFE-II conceptual design study. The results of this study are described, and reference designs of these large heat exchangers are selected to minimize the cost of electricity while satisfying other important constraints

  2. A Study on the Electrolytic Reduction Mechanism of Uranium Oxide in a LiCl-Li2O Molten Salt

    International Nuclear Information System (INIS)

    Oh, Seung Chul; Hur, Jin Mok; Seo, Chung Seok; Park, Seong Won

    2003-01-01

    This study proposed a new electrolytic reduction technology that is based on the integration of simultaneous uranium oxide metallization and Li 2 O electrowinning. In this electrolytic reduction reaction, electrolytically reduced Li deposits on cathode and simultaneously reacts with uranium oxides to produce uranium metal showing more than 99% conversion. For the verification of process feasibility, the experiments to obtain basic data on the metallization of uranium oxide, investigation of reaction mechanism, the characteristics of closed recycle of Li 2 O and mass transfer were carried out. This evolutionary electrolytic reduction technology would give benefits over the conventional Li-reduction process improving economic viability such as: avoidance of handling of chemically active Li-LiCl molten salt increase of metallization yield, and simplification of process.

  3. A study on the electrolytic reduction of uranium oxide in a LiCl-Li2O molten salt

    International Nuclear Information System (INIS)

    Su, J. S.; Hu, J. M.; Hong, S. S.; Jang, D. S.; Park, S. W.

    2003-01-01

    New electrolytic reduction technology was proposed that is based on the integration of metallization of uranium oxide and Li 2 O electrowinning. In this electrolytic reduction reaction, electrolytically reduced Li deposits on cathode and simultaneously reacts with uranium oxides to produce uranium metal showing more than 99% conversion. For the verification of process feasibility, the experiments to obtain basic data on the metallization of uranium oxide, investigation of reaction mechanism, the characteristics of closed recycle of Li 2 O and mass transfer were carried out. This evolutionary electrolytic reduction technology would give benefits over the conventional Li-reduction process improving economic viability such as: avoidance of handling of chemically active Li-LiCl molten salt, increase of metallization yield, and simplification of process

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

    International Nuclear Information System (INIS)

    Restivo, T.A.G.

    1994-01-01

    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 800 C 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

  5. 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.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

    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-β(double p rime)-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(double p rime)-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

  7. 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

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

    Science.gov (United States)

    Sun, Ningbo; Zhang, Yingchun; Lang, Shaoting; Jiang, Fan; Wang, Lili

    2014-12-01

    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.

  9. 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.

  10. Molten salt synthesis of ZnNb2O6 powder

    International Nuclear Information System (INIS)

    Guo Liangzhai; Dai Jinhui; Tian Jintao; Zhu Zhibin; He Tian

    2007-01-01

    Pure ZnNb 2 O 6 powder was successfully prepared by the molten salt synthesis method using Nb 2 O 5 and ZnO as raw materials and a mixture of NaCl and KCl as the solvent. The phase form and morphology of the prepared powder were characterized by X-ray diffraction and scanning electron microscopy. The effect of reacting temperature on phase formation was investigated. The results indicated that the single phase ZnNb 2 O 6 powder can be obtained by the molten salt synthesis method at 600 deg. C, and the SEM photographs show that the grains of the powder are rod-like particles

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

    International Nuclear Information System (INIS)

    Tosi, M.P.

    1997-04-01

    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

  12. Combined gettering and molten salt process for tritium recovery from lithium

    International Nuclear Information System (INIS)

    Sze, D.K.; Finn, P.A.; Bartlit, J.; Tanaka, S.; Teria, T.; Yamawaki, M.

    1988-02-01

    A new tritium recovery concept from lithium has been developed as part of the US/Japan collaboration on Reversed-Field Pinch Reactor Design Studies. This concept combines the γ-gettering process as the front end to recover tritium from the coolant, and a molten salt recovery process to extract tritium for fuel processing. A secondary lithium is used to regenerate the tritium from the gettering bed and, in the process, increases the tritium concentration by a factor of about 20. That way, the required size of the molten salt process becomes very small. A potential problem is the possible poisoning of the gettering bed by the salt dissolved in lithium. 16 refs., 6 figs

  13. HTGR molten salt sensible energy transmission and storage system design and costs

    International Nuclear Information System (INIS)

    1981-09-01

    This report, which was prepared for Gas-Cooled Reactor Associates by United Engineers and Constructors under Contract No. GCRA/UE and C 81-203, presents the design and cost for a molten salt Sensible Energy Transmission and Storage (SETS) System. Although the reference system for this study is sized to be compatible with an 1170 MW(t) HTGR Nuclear Heat Source, the results and conclusions should be generally applicable to most large scale molten salt energy transmission system applications. A preliminary conceptual design is presented and alternative configurations are discussed. The sensitivity of system costs to variations in important system parameters are also presented. Costs for a reference case conceptual design are reported in constant 1980 dollars and inflated 1995 dollars

  14. Study of tritium removal from fusion reactor blankets of molten salt and lithium--aluminum

    International Nuclear Information System (INIS)

    Talbot, J.B.

    1976-03-01

    The sorption of tritium by molten lithium--bismuth (Li--Bi, approx. 15 at. % lithium) and solid equiatomic lithium--aluminum (Li--Al) was investigated experimentally to evaluate the potential applications of both materials in a controlled thermonuclear reactor. The Li--Bi alloy was proposed to countercurrently extract tritium from a molten salt (Li 2 BeF 4 ) blanket. However, because of the low solubility ( 0 C, the extraction process is not attractive

  15. Structural analysis of molten salt systems by XAFS, XRD and MD

    International Nuclear Information System (INIS)

    Okamoto, Yoshihiro; Minato, Kazuo

    2001-01-01

    Combinational use of X-ray absorption fine structure (XAFS), X-ray diffraction (XRD) and molecular dynamics (MD) simulation has been applied to investigate the structural analysis of high-temperature molten salt systems. The XRD data of molten rare earth trihalides and uranium trichloride were reproduced by the MD with polarizable ionic model. We proposed that the XAFS function is calculated by combinational use of the MD output and the multiscattering XAFS code FEFF. (author)

  16. Dependence of capacitance of metal-molten salt interface on local density profiles near electrode

    International Nuclear Information System (INIS)

    Ballone, P.; Pastore, G.; Tosi, M.P.; Painter, K.R.; Grout, P.J.; March, N.H.

    1983-09-01

    The mean spherical approximation applied to a metal-molten salt interface is generalized to take further account of the local density profiles near the electrode. The temperature dependence of the differential capacitance is shown to arise, in large measure from such local structure. However, the hard wall assumption for the electrode is retained and this must be kept in mind in comparing our model results with experiment. (author)

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

    International Nuclear Information System (INIS)

    Barbero, Jose A.; Wiesztort, Andres; Azcona, Alejandra; Bollini, Edgardo; Forchetti, Alberto; Orce, Alan

    1999-01-01

    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

  18. Concept of the demonstration molten salt unit for the transuranium elements transmutation

    International Nuclear Information System (INIS)

    Alekseev, P.; Dudnikov, A.; Prusakov, V.; Subbotin, S.; Zakirov, R.; Lelek, V.; Peka, I.

    1999-01-01

    In this report it is considered fluorine reprocessing of spent fuel and fluoride molten salt reactor in critical and subcritical modes for plutonium and minor actinides burning. International collaboration for creation of such system is proposed. It is without any doubt that additional neutron source in the core will have positive influence on the transmutation process in the reactor. On the other side there is a lot of problems to realize it technically and to ensure stable work of the whole complex. (Authors)

  19. Effect of Using Thorium Molten Salts on the Neutronic Performance of PACER

    Science.gov (United States)

    Acır, Adem; Übeyli, Mustafa

    2010-04-01

    Utilization of nuclear explosives can produce a significant amount of energy which can be converted into electricity via a nuclear fusion power plant. An important fusion reactor concept using peaceful nuclear explosives is called as PACER which has an underground containment vessel to handle the nuclear explosives safely. In this reactor, Flibe has been considered as a working coolant for both tritium breeding and heat transferring. However, the rich neutron source supplied from the peaceful nuclear explosives can be used also for fissile fuel production. In this study, the effect of using thorium molten salts on the neutronic performance of the PACER was investigated. The computations were performed for various coolants bearing thorium and/or uranium-233 with respect to the molten salt zone thickness in the blanket. Results pointed out that an increase in the fissile content of the salt increased the neutronic performance of the reactor remarkably. In addition, higher energy production was obtained with thorium molten salts compared to the pure mode of the reactor. Moreover, a large quantity of 233U was produced in the blanket in all cases.

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

    International Nuclear Information System (INIS)

    Srivastava, A.K.; Gupta, Anurag; Krishnani, P.D.

    2015-01-01

    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 233 Pa at a particular burnup has been carried out. The similar analysis has been performed for some other nuclides also. (author)

  1. Diagnosis of sources of current inefficiency in industrial molten salt electrolysis cells by Raman spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Sadoway, D.R.

    1988-07-29

    The purpose of this project was to employ Raman spectroscopy in the study of industrial molten salt electrolysis cells. The objective was to improve the understanding of the chemistry and electrochemistry of the relevant melt systems and, in turn, of energy loss mechanisms in the industrial processes. On this basis new ways to improve the energy efficiency of these industrial reactors might be identified. The research plan has several principal elements. First, there was the design and construction of laboratory scale representations of industrial molten salt electrolysis cells that would at the same time serve a spectrocells. Secondly, there was the mastery of the preparation of the molten salt electrolytes, what in industry is called the ''front end.'' Thirdly, there was the adaptation of commercially available Raman instrumentation in order to facilitate the proposed studies. It is the nature of the specimens that so dramatically distinguished this work from conventional Raman studies for which commercial instrumentation is designed: first, the laboratory scale electrolysis cells are large compared to typical spectrocells; and secondly, the cells operate at, what for Raman studies are, extremely high temperatures. 4 refs., 2 figs.

  2. Low temperature synthesis & characterization of lead-free BCZT ceramics using molten salt method

    Science.gov (United States)

    Jai Shree, K.; Chandrakala, E.; Das, Dibakar

    2018-04-01

    Piezoelectric properties are greatly influenced by the synthesis route, microstructure, stoichiometry of the chemical composition, purity of the starting materials. In this study, molten salt method was used to prepare lead-free BCZT ceramics. Molten salt method is one of the simplestmethods to prepare chemically-purified, single phase powders in high yield often at lower temperatures and shorten reaction time. Calcination of the molten salt synthesized powders resulted in asingle-phase perovskite structure at 1000 °C which is ˜ 350 °C less than the conventional solid-sate reaction method. With increasing calcination temperature the average template size was increased (˜ 0.5-2 µm). Formation of well dispersive templates improves the sinterability at lower temperatures. Lead-free BCZT ceramics sintered at 1500 °C for 2 h resulted in homogenous and highly dense microstructure with ˜92% of the theoretical density and a grain size of ˜ 35 µm. This highly dense microstructure could enhance the piezoelectric properties of the system.

  3. 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

  4. 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)

    Ignatiev, V.; Merzlyakov, A.; Afonichkin, V.; Khokhlov, V.; Salyulev, A.

    2003-01-01

    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)

  5. 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

  6. Conceptual design of Indian molten salt breeder reactor

    Indian Academy of Sciences (India)

    2015-08-28

    Aug 28, 2015 ... 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 ...

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

    International Nuclear Information System (INIS)

    Kim Davies; Shelly X Li

    2007-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim Davies; Shelly X Li

    2007-09-01

    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 wire

  9. Application of Proton Conductors to Hydrogen Monitoring for Liquid Metal and Molten Salt Systems

    Science.gov (United States)

    Kondo, Masatoshi; Muroga, Takeo; Katahira, Koji; Oshima, Tomoko

    The chemical control of impurity such as hydrogen and oxygen in coolants is one of the critical issues for the development of liquid metal cooled fast reactors and self-cooled liquid breeder blankets for fusion reactors. Especially, hydrogen (isotopes) level is the key parameter for corrosion and mechanical properties of the in-reactor components. For fission reactors, the monitor of hydrogen level in the melt is important for safety operation. The control of tritium is essential for the tritium breeding performance of the fusion reactors. Therefore, on-line hydrogen sensing is a key technology for these systems. In the present study, conceptual design for the on-line hydrogen sensor to be used in liquid sodium (Na), lead (Pb), lead-bismuth (Pb-Bi), lithium (Li), lead-lithium (Pb-17Li) and molten salt LiF-BeF2 (Flibe) was performed. The cell of hydrogen sensor is made of a solid electrolyte. The solid electrolyte proposed in this study is the CaZrO3-based ceramics, which is well-known as proton conducting ceramics. In this concept, the cell is immersed into the melt which is containing the hydrogen at the activity of PH1 of ambient atmosphere. Then, the cell is filled with Ar-H2 mixture gas at regulated hydrogen activity of PH2. The electromotive force (EMF) is obtained by the proton conduction in the electro chemical system expressed as Pt, Melt(PH1) | Proton conductor | PH2, Pt. The Nernst equation is used for the evaluation of the hydrogen activity from the obtained EMF. The evaluations of expected performance of the sensor in liquid Na, Pb, Pb-Bi, Pb-17Li, Li and Flibe were carried out by means of the measurement test in gas atmosphere at hydrogen activities equivalent to those for the melts in the reactor conditions. In the test, the hydrogen activity in the gas varied from 2.2x10-14 to 1. The sensor exhibited good response, stability and reproducibility.

  10. The Twentieth International Symposium on Molten Salts and Ionic Liquids

    Science.gov (United States)

    2016-11-29

    Characteristics of Capacitors Based on Ionic Liquids: From Dielectric Polymers to Redox-Active Adsorbed Species E. Lust, L. Siinor, H. Kurig, T. Romann, V...Tungsten from Super Hard Alloys in Molten Sodium Hydroxide T. Oishi 633 Red-Ox Reactions in Ionic Liquids and Their Impact on Electrodeposition of

  11. Molten salt steam generator subsystem research experiment. Volume I. Phase 1 - Final report

    Energy Technology Data Exchange (ETDEWEB)

    None

    1984-10-01

    A study was conducted for Phase 1 of a two-phase project whose objectives were to develop a reliable, cost-effective molten salt steam generating subsystem for solar thermal plants, minimize uncertainty in capital, operating, and maintenance costs, and demonstrate the ability of molten salt to generate high-pressure, high-temperature steam. The Phase 1 study involved the conceptual design of molten salt steam generating subsystems for a nominal 100-MWe net stand-alone solar central receiver electric generating plant, and a nominal 100-MWe net hybrid fossil-fueled electric power generating plant that is 50% repowered by a solar central receiver system. As part of Phase 1, a proposal was prepared for Phase 2, which involves the design, construction, testing and evaluation of a Subsystem Research Experiment of sufficient size to ensure successful operation of the full-size subsystem designed in Phase 1. Evaluation of several concepts resulted in the selection of a four-component (preheater, evaporator, superheater, reheater), natural circulation, vertically oriented, shell and tube (straight) heat exchanger arrangement. Thermal hydraulic analysis of the system included full and part load performance, circulation requirements, stability, and critical heat flux analysis. Flow-induced tube vibration, tube buckling, fatigue evaluation of tubesheet junctions, steady-state tubesheet analysis, and a simplified transient analysis were included in the structural analysis of the system. Operating modes and system dynamic response to load changes were identified. Auxiliary equipment, fabrication, erection, and maintenance requirements were also defined. Installed capital costs and a project schedule were prepared for each design.

  12. Application of molten salts in pyrochemical processing of reactive metals

    International Nuclear Information System (INIS)

    Mishra, B.; Olson, D.L.; Averill, W.A.

    1992-01-01

    Various mixes of chloride and fluoride salts are used as the media for conducting pyrochemical processes in the production and purification of reactive metals. These processes generate a significant amount of contaminated waste that has to be treated for recycling or disposal. Molten calcium chloride based salt systems have been used in this work to electrolytically regenerate calcium metal from calcium oxide for the in situ reduction of reactive metal oxides. The recovery of calcium is characterized by the process efficiency to overcome back reactions in the electrowinning cell. A thermodynamic analysis, based on fundamental rate theory, has been performed to understand the process parameters controlling the metal deposition, rate, behavior of the ceramic anode-sheath and influence of the back-reactions. It has been observed that the deposition of calcium is dependent on the ionic diffusion through the sheath. It has also been evidenced that the recovered calcium is completely lost through the back-reactions in the absence of a sheath. A practical scenario has also been presented where the electrowon metal can be used in situ as a reductant to reduce another reactive metal oxide

  13. 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.

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

    International Nuclear Information System (INIS)

    Ramsey, K.B.; Acosta, S.V.; Wernly, K.D.

    1998-01-01

    This paper presents an overview of potential technologies for stabilization of 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 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

  15. Electrochemical Synthesis of Superconductive Boride MgB2 from Molten Salts

    OpenAIRE

    Abe, Hideki; Yoshii, Kenji

    2002-01-01

    Electrolysis has been performed on a fused mixture of magnesium chloride, potassium chloride, and magnesium borate at 600 oC under an Ar atmosphere using a graphite anode and a Pt cathode. Magnetic measurements and X-ray diffractometry have revealed that the resultant deposit on the cathode contains a few tens molar percent of MgB2. A voltanmetric measurement at the initial stage of electrolysis has shown that the threshold cathode voltage for the formation of MgB2 from the molten salts is -1...

  16. Study of tritium removal from fusion reactor blankets of molten salt and lithium--aluminum

    International Nuclear Information System (INIS)

    Talbot, J.B.

    1976-03-01

    The sorption of tritium by molten lithium--bismuth (Li--Bi, approximately 15 at. percent lithium) and solid equiatomic lithium--aluminum (Li--Al) was investigated experimentally to evaluate the potential applications of both materials in a controlled thermonuclear reactor. The Li--Bi alloy was proposed to countercurrently extract tritium from a molten salt (Li 2 BeF 4 ) blanket. However, because of the low solubility (less than 10 ppb) at temperatures ranging from 500 to 700 0 C, the extraction process is not attractive

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

    OpenAIRE

    Rodrigues, Davide; Durán-Klie, Gabriela; Delpech, Sylvie

    2015-01-01

    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...

  18. Conception of electron beam-driven subcritical molten salt ultimate safety reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abalin, S.S.; Alekseev, P.N.; Ignat`ev, V.V. [Kurchatov Institute, Moscow (Russian Federation)] [and others

    1995-10-01

    This paper is a preliminary sketch of a conception to develop the {open_quotes}ultimate safety reactor{close_quotes} using modern reactor and accelerator technologies. This approach would not require a long-range R&D program. The ultimate safety reactor could produce heat and electric energy, expand the production of fuel, or be used for the transmutation of long-lived wastes. The use of the combined double molten salt reactor system allows adequate neutron multiplication to permit using an electron accelerator for the initial neutron flux. The general parameters of such a system are discussed in this paper.

  19. Stress corrosion cracking of Ti-8Al-1 Mo-1V in molten salts

    Science.gov (United States)

    Smyrl, W. H.; Blackburn, M. J.

    1975-01-01

    The stress corrosion cracking (SCC) behavior of Ti-8Al-1 Mo-1V has been studied in several molten salt environments. Extensive data are reported for the alloy in highly pure LiCl-KCl. The influence of the metallurgical heat treatment and texture, and the mechanical microstructure show similarities with aqueous solutions at lower temperature. The fracture path and cracking modes are also similar to that found in other environments. The influence of H2O and H(-) in molten LiCl-KCl lead to the conclusion that hydrogen does not play a major role in crack extension in this environment.

  20. Comparison of slagging pyrolysis and molten salt incinerators for treating TRU waste at the INEL

    International Nuclear Information System (INIS)

    1977-11-01

    For the comparison, it is assumed that the waste product is required to meet the acceptance criteria of the Waste Isolation Pilot Plant, i.e., low leachability. Slagging pyrolysis incinerates combustible waste and melts noncombustible waste; the resulting slag forms a glass of low leachability. In the molten salt incinerator, combustion occurs at low temperatures with no accumulation of explosive gases, but the waste must have been previously sorted into combustibles and noncombustibles and then shredded. The economics, safety, and technical features are compared. Advantages, disadvantages, and areas of technical uncertainty of the two systems are listed. Development costs and schedules for the two types of incinerators are discussed

  1. Development of JURA -A Simulation Code for Electrochemical Processes in Molten Salt-

    OpenAIRE

    小林 嗣幸

    2003-01-01

    A simulation code for electrochemical processes in molten salt named JURA has been developed. This code can simulate the time history of the processes such as the metal pyro-process and the oxide pyro-process by the diffusion layer theory at various temperature and melts. This report describes the specific formulations of the theory for various electrodes such as the solid cathode, the Cd cathode, and the Cd pool anode. Explanation of the input data and sample calculations of the solid cathod...

  2. 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

  3. Thermodynamics investigation of a solar power system integrated oil and molten salt as heat transfer fluids

    International Nuclear Information System (INIS)

    Liu, Qibin; Bai, Zhang; Sun, Jie; Yan, Yuejun; Gao, Zhichao; Jin, Hongguang

    2016-01-01

    Highlights: • A new concentrating solar power system with a dual-solar field is proposed. • The superheated steam with more than 773 K is produced. • The performances of the proposed system are demonstrated. • The economic feasibility of the proposed system is validated. - Abstract: In this paper, a new parabolic trough solar power system that incorporates a dual-solar field with oil and molten salt as heat transfer fluids (HTFs) is proposed to effectively utilize the solar energy. The oil is chosen as a HTF in the low temperature solar field to heat the feeding water, and the high temperature solar field uses molten salt to superheat the steam that the temperature is higher than 773 K. The produced superheated steam enters a steam turbine to generate power. Energy analysis and exergy analysis of the system are implemented to evaluate the feasibility of the proposed system. Under considerations of variations of solar irradiation, the on-design and off-design thermodynamic performances of the system and the characteristics are investigated. The annual average solar-to-electric efficiency and the nominal efficiency under the given condition for the proposed solar thermal power generation system reach to 15.86% and 22.80%, which are higher than the reference system with a single HTF. The exergy losses within the solar heat transfer process of the proposed system are reduced by 7.8% and 45.23% compared with the solar power thermal systems using oil and molten salt as HTFs, respectively. The integrated approach with oil and molten salt as HTFs can make full use of the different physical properties of the HTFs, and optimize the heat transfer process between the HTFs and the water/steam. The exergy loss in the water evaporation and superheated process are reduced, the system efficiency and the economic performance are improved. The research findings provide a new approach for the improvement of the performances of solar thermal power plants.

  4. An overview of radiolysis studies for the molten salt reactor remediation project

    International Nuclear Information System (INIS)

    Icenhour, A.S.; Williams, D.F.; Trowbridge, L.D.; Toth, L.M.; Del Cul, G.D.

    2001-01-01

    A number of radiolysis experiments have been performed in support of the remediation of the Molten Salt Reactor Experiment (MSRE)at the Oak Ridge National Laboratory.Materials studied included simulated MSRE fuel salt,fluorinated charcoal, NH 4 F,2NaFUF 6 ,UO 2 F 2 uranium oxides with a known residual fluoride content,and uranium oxides with a known moisture content.The results from these studies were used as part of the basis for the interim or long-term storage of materials removed from the MSRE. (author)

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

    International Nuclear Information System (INIS)

    Schwandt, Carsten

    2014-01-01

    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

  6. Study of tritium removal from fusion reactor blankets of molten salt and lithium--aluminum

    Energy Technology Data Exchange (ETDEWEB)

    Talbot, J.B

    1976-03-01

    The sorption of tritium by molten lithium--bismuth (Li--Bi, approx. 15 at. % lithium) and solid equiatomic lithium--aluminum (Li--Al) was investigated experimentally to evaluate the potential applications of both materials in a controlled thermonuclear reactor. The Li--Bi alloy was proposed to countercurrently extract tritium from a molten salt (Li/sub 2/BeF/sub 4/) blanket. However, because of the low solubility (< 10 ppB) at temperature ranging from 500 to 700/sup 0/C, the extraction process is not attractive.

  7. Neutronic analysis of two-fluid thorium molten salt reactor

    International Nuclear Information System (INIS)

    Frybort, Jan; Vocka, Radim

    2009-01-01

    The aim of this paper is to evaluate features of the two-fluid MSBR through a parametric study and compare its properties to one-fluid MSBR concepts. The starting point of the analysis is the original ORNL 1000 MWe reactor design, although simplified to some extent. We studied the influence of dimensions of distinct reactor parts - fuel and fertile channels radius, plenum height, design etc. - on fundamental reactor properties: breeding ratio and doubling time, reactor inventory, graphite lifetime, and temperature feedback coefficients. The calculations were carried out using MCNP5 code. Based on obtained results we proposed an improved reactor design. Our results show clear advantages of the concept with two separate fluoride salts if compared to the one fluid concept in breading, doubling time, and temperature feedback coefficients. Limitations of the two-fluid concept - particularly the graphite lifetime - are also pointed out. The reactor design can be a subject of further optimizations, namely from the viewpoint of reactor safety. (author)

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

    International Nuclear Information System (INIS)

    Fa Dan; Yan Changqi; Sun Licheng; Sun Lu; Zhao Hangbin

    2013-01-01

    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

  9. Electrochemical behavior for a reduction of uranium oxide in a LiCl-Li2O molten salt with an integrated cathode assembly

    International Nuclear Information System (INIS)

    Park, Sung Bin; Park, Byung Heung; Seo, Chung Seok; Jung, Ki Jung; Park, Seong Won

    2005-01-01

    Electrolytic reduction of uranium oxide to uranium metal was studied in a LiCl-Li 2 O molten salt system. The reduction mechanism of the uranium oxide to a uranium metal has been studied by means of a cyclic voltammetry. Effects of the layer thickness of the uranium oxide and the thickness of the MgO on the overpotential of the cathode and the anode were investigated by means of a chronopotentiometry. From the cyclic voltamograms, the decomposition potentials of the metal oxides are the determining factors for the mechanism of the reduction of the uranium oxide in a Li Cl-3 wt% Li 2 O molten salt and the two mechanisms of the electrolytic reduction were considered with regards to the applied cathode potential. In the chronopotentiograms, the exchange current and the transfer coefficient based on the Tafel behavior were obtained with regard to the layer thickness of the uranium oxide which is loaded into the porous MgO membrane and the thickness of the porous MgO membrane. The maximum allowable currents for the changes of the layer thickness of the uranium oxide and the thickness of the MgO membrane were also obtained from the limiting potential which is the decomposition potential of LiCl

  10. In Situ Production of Copper Oxide Nanoparticles in a Binary Molten Salt for Concentrated Solar Power Plant Applications

    Directory of Open Access Journals (Sweden)

    Mathieu Lasfargues

    2017-05-01

    Full Text Available Seeding nanoparticles in molten salts has been shown recently as a promising way to improve their thermo-physical properties. The prospect of such technology is of interest to both academic and industrial sectors in order to enhance the specific heat capacity of molten salt. The latter is used in concentrated solar power plants as both heat transfer fluid and sensible storage. This work explores the feasibility of producing and dispersing nanoparticles with a novel one pot synthesis method. Using such a method, CuO nanoparticles were produced in situ via the decomposition of copper sulphate pentahydrate in a KNO3-NaNO3 binary salt. Analyses of the results suggested preferential disposition of atoms around produced nanoparticles in the molten salt. Thermal characterization of the produced nano-salt suspension indicated the dependence of the specific heat enhancement on particle morphology and distribution within the salts.

  11. In Situ Production of Copper Oxide Nanoparticles in a Binary Molten Salt for Concentrated Solar Power Plant Applications.

    Science.gov (United States)

    Lasfargues, Mathieu; Stead, Graham; Amjad, Muhammad; Ding, Yulong; Wen, Dongsheng

    2017-05-19

    Seeding nanoparticles in molten salts has been shown recently as a promising way to improve their thermo-physical properties. The prospect of such technology is of interest to both academic and industrial sectors in order to enhance the specific heat capacity of molten salt. The latter is used in concentrated solar power plants as both heat transfer fluid and sensible storage. This work explores the feasibility of producing and dispersing nanoparticles with a novel one pot synthesis method. Using such a method, CuO nanoparticles were produced in situ via the decomposition of copper sulphate pentahydrate in a KNO₃-NaNO₃ binary salt. Analyses of the results suggested preferential disposition of atoms around produced nanoparticles in the molten salt. Thermal characterization of the produced nano-salt suspension indicated the dependence of the specific heat enhancement on particle morphology and distribution within the salts.

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

    International Nuclear Information System (INIS)

    Sherwood, A.E.

    1984-09-01

    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

  13. Thermal conductivity of molten salt mixtures: Theoretical model supported by equilibrium molecular dynamics simulations

    Science.gov (United States)

    Gheribi, Aïmen E.; Chartrand, Patrice

    2016-02-01

    A theoretical model for the description of thermal conductivity of molten salt mixtures as a function of composition and temperature is presented. The model is derived by considering the classical kinetic theory and requires, for its parametrization, only information on thermal conductivity of pure compounds. In this sense, the model is predictive. For most molten salt mixtures, no experimental data on thermal conductivity are available in the literature. This is a hindrance for many industrial applications (in particular for thermal energy storage technologies) as well as an obvious barrier for the validation of the theoretical model. To alleviate this lack of data, a series of equilibrium molecular dynamics (EMD) simulations has been performed on several molten chloride systems in order to determine their thermal conductivity in the entire range of composition at two different temperatures: 1200 K and 1300 K. The EMD simulations are first principles type, as the potentials used to describe the interactions have been parametrized on the basis of first principle electronic structure calculations. In addition to the molten chlorides system, the model predictions are also compared to a recent similar EMD study on molten fluorides and with the few reliable experimental data available in the literature. The accuracy of the proposed model is within the reported numerical and/or experimental errors.

  14. Assessment of the Neutronic and Fuel Cycle Performance of the Transatomic Power Molten Salt Reactor Design

    Energy Technology Data Exchange (ETDEWEB)

    Robertson, Sean [Transatomic Power Corp., Cambridge, MA (United States); Dewan, Leslie [Transatomic Power Corp., Cambridge, MA (United States); Massie, Mark [Transatomic Power Corp., Cambridge, MA (United States); Davidson, Eva E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    This report presents results from a collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear (GAIN) Nuclear Energy Voucher program. The TAP concept is a molten salt reactor using configurable zirconium hydride moderator rod assemblies to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parameters necessary to simulate the continuously changing physics in this complex system. The implementation of continuous-energy Monte Carlo transport and depletion tools in ChemTriton provide for full-core three-dimensional modeling and simulation. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this concept. Additional analyses of mass feed rates and enrichments, isotopic removals, tritium generation, core power distribution, core vessel helium generation, moderator rod heat deposition, and reactivity coeffcients provide additional information to make informed design decisions. This work demonstrates capabilities of ORNL modeling and simulation tools for neutronic and fuel cycle analysis of molten salt reactor concepts.

  15. Two-Dimensional Neutronic and Fuel Cycle Analysis of the Transatomic Power Molten Salt Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Robertson, Sean [Transatomic Power Corporation, Cambridge, MA (United States); Dewan, Leslie [Transatomic Power Corporation, Cambridge, MA (United States); Massie, Mark [Transatomic Power Corporation, Cambridge, MA (United States)

    2017-01-15

    This status report presents the results from the first phase of the collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear, Nuclear Energy Voucher program. The TAP design is a molten salt reactor using movable moderator rods to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parameters necessary to simulate the continuously changing physics in this complex system. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this design. Additional analyses of time step sizes, mass feed rates and enrichments, and isotopic removals provide additional information to make informed design decisions. This work further demonstrates capabilities of ORNL modeling and simulation tools for analysis of molten salt reactor designs and strongly positions this effort for the upcoming three-dimensional core analysis.

  16. Thermal conductivity of molten salt mixtures: Theoretical model supported by equilibrium molecular dynamics simulations.

    Science.gov (United States)

    Gheribi, Aïmen E; Chartrand, Patrice

    2016-02-28

    A theoretical model for the description of thermal conductivity of molten salt mixtures as a function of composition and temperature is presented. The model is derived by considering the classical kinetic theory and requires, for its parametrization, only information on thermal conductivity of pure compounds. In this sense, the model is predictive. For most molten salt mixtures, no experimental data on thermal conductivity are available in the literature. This is a hindrance for many industrial applications (in particular for thermal energy storage technologies) as well as an obvious barrier for the validation of the theoretical model. To alleviate this lack of data, a series of equilibrium molecular dynamics (EMD) simulations has been performed on several molten chloride systems in order to determine their thermal conductivity in the entire range of composition at two different temperatures: 1200 K and 1300 K. The EMD simulations are first principles type, as the potentials used to describe the interactions have been parametrized on the basis of first principle electronic structure calculations. In addition to the molten chlorides system, the model predictions are also compared to a recent similar EMD study on molten fluorides and with the few reliable experimental data available in the literature. The accuracy of the proposed model is within the reported numerical and/or experimental errors.

  17. Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

    Energy Technology Data Exchange (ETDEWEB)

    Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

    2017-05-15

    Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

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

    International Nuclear Information System (INIS)

    Krepel, J.

    2015-01-01

    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

  19. Design study of advanced nuclear fuel recycle system. Conceptual study of recycle system using molten salt

    International Nuclear Information System (INIS)

    Kakehi, I.; Shirai, N.; Hatano, M.; Kajitani, M.; Yonezawa, S.; Kawai, T.; Kawamura, F.; Tobe, K.; Takahashi, K.

    1996-12-01

    For the purpose of developing the future nuclear fuel recycle system, the design study of the advanced nuclear fuel recycle system is being conducted. This report describes intermediate accomplishments in the conceptual system study of the advanced nuclear fuel recycle system. Fundamental concepts of this system is the recycle system using molten salt which intend to break through the conventional concepts of purex and pellet fuel system. Contents of studies in this period are as follows, 1)feasibility study of the process by Cd-cathode for nitride fuel, 2)application study for the molten salt of low melting point (AlCl3+organic salt), 3)research for decladding (advantage of decladding by heat treatment), 4)behavior of FPs in electrorefining (behavior of iodine and volatile FP chlorides, FPs behavior in chlorination), 5)criticality analysis in electrorefiner, 6)drawing of off-gas flow diagram, 7)drawing of process machinery concept (cathode processor, vibration packing), 8)evaluation for the amounts of the high level radioactive wastes, 9)quality of the recycle fuels (FPs contamination of recycle fuel), 10)conceptual study of in-cell handling system, 11)meaning of the advanced nuclear fuel recycle system. The conceptual system study will be completed in describing concepts of the system and discussing issues for the developments. (author)

  20. Analysis of chromium volatility in the DWTF incinerator and in the molten salt processor

    International Nuclear Information System (INIS)

    Ebbinghaus, B.B.

    1992-01-01

    Thermodynamic methods have been applied to calculate the volatility of chromium both in atmospheres and in milligrams per cubic meter (stp) of offgas for the decontamination and waste treatment facility (DWTF) incinerator and the Rockwell molten salt processor. The known chromium species which have relatively high stabilities under oxidizing conditions and which contain elements found in either the DWTF incinerator or the molten salt processor are CrO 2 (OH) 2 (g), CrO 2 OH(g), CrO 3 (g), CrO 2 Cl 2 (g), and CrOF(g) and CrO 2 F 2 (g). This study demonstrates that these species as well as others such as CrO(OH) 2 (g), CrOOH(g), CrO(OH) 3 (g), CrO 2 Cl(g), CrOC1 2 (g), CrOCl(g), CrOC1 3 (g), CrOC1 4 (g), and CrO 2 F(g) can all be important species which contribute to the overall volatility of chromium in waste oxidation processes

  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. Quantitative Analysis of KF-LiF-ZrF4 Molten Salt by Probe Assisted in-situ LIBS Systems

    International Nuclear Information System (INIS)

    Kim, S.H.; Moon, J.H.; Kim, D.H.; Hwang, I.S.; Lee, J.H.

    2015-01-01

    Full text of publication follows: Pyro-processing draws attention as a recycling process of spent nuclear fuel for future nuclear reactor. In the aspect of process control and safeguards of the pyro-processing, it requires a technology to measure the concentration of molten salt in real-time. The existing technologies measure the concentration by chemical analysis of sampled molten salt in the hot cell but it is disadvantageous in the aspects of cost, safety and time. The LIBS (Laser-Induced Breakdown Spectroscopy) is a form of atomic emission spectroscopy in which a pulsed laser is used as the excitation source. LIBS technology is appropriate to measure sensitive nuclear materials in hot cell because it is capable of measuring specimen quantitatively and qualitatively by exited atom by laser. Spectrum obtained from plasma is largely influenced by laser operation conditions and physical properties of specimens. Also, plasma induction is limited on the surface of specimen, so analysis of composition inside of the molten salt is extremely difficult. Thus, several restrictions should be overcome in order to apply LIBS for the measurement of molten salt (KF-LiF-ZrF 4 ) composition in real-time. In this study probe assisted LIBS system will be introduced with KF-LiF-ZrF 4 to quantitatively measure molten salt composition. Echelle spectrometer was used and the measurable wavelength area was 250-400 nm, the range of UV ray. NIST atomic spectra database measured the wavelength for molten salt composition, and each element was selected high signal intensity and wavelength range that is not overlapped by other elements. (authors)

  3. Accelerator-driven subcritical fission in molten salt core: Closing the nuclear fuel cycle for green nuclear energy

    Science.gov (United States)

    McIntyre, Peter; Assadi, Saeed; Badgley, Karie; Baker, William; Comeaux, Justin; Gerity, James; Kellams, Joshua; McInturff, Al; Pogue, Nathaniel; Phongikaroon, Supathorn; Sattarov, Akhdiyor; Simpson, Michael; Sooby, Elizabeth; Tsvetkov, Pavel

    2013-04-01

    A technology for accelerator-driven subcritical fission in a molten salt core (ADSMS) is being developed as a basis for the destruction of the transuranics in used nuclear fuel. The molten salt fuel is a eutectic mixture of NaCl and the chlorides of the transuranics and fission products. The core is driven by proton beams from a strong-focusing cyclotron stack. This approach uniquely provides an intrinsically safe means to drive a core fueled only with transuranics, thereby eliminating competing breeding terms.

  4. Thorium fuel-cycle development through plutonium incineration by THORIMS-NES (Thorium Molten-Salt nuclear energy synergetics)

    International Nuclear Information System (INIS)

    Furukawa, K.; Furuhashi, A.; Chigrinov, S.E.

    1996-01-01

    Thorium fuel-cycle has benefit on not-only trans-U element reduction but also their incineration. The disadvantage of high gamma activity of fuel, which is useful for improving the resistance to nuclear proliferation and terrorism, can overcome by molten fluorides fuel, and practically by THORIMS-NES, symbiotically coupled with fission Molten-Salt Reactor (FUJI) and fissile-producing Accelerator Molten-Salt Breeder (AMSB). This will have wide excellent advantages in global application, and will be deployed by incinerating Pu and Producing 233 U. Some details of this strategy including time schedule are presented. 14 refs, 2 figs, 4 tabs

  5. 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.

  6. LiCl-LiI molten salt electrolyte with bismuth-lead positive electrode for liquid metal battery

    Science.gov (United States)

    Kim, Junsoo; Shin, Donghyeok; Jung, Youngjae; Hwang, Soo Min; Song, Taeseup; Kim, Youngsik; Paik, Ungyu

    2018-02-01

    Liquid metal batteries (LMBs) are attractive energy storage device for large-scale energy storage system (ESS) due to the simple cell configuration and their high rate capability. The high operation temperature caused by high melting temperature of both the molten salt electrolyte and metal electrodes can induce the critical issues related to the maintenance cost and degradation of electrochemical properties resulting from the thermal corrosion of materials. Here, we report a new chemistry of LiCl-LiI electrolyte and Bi-Pb positive electrode to lower the operation temperature of Li-based LMBs and achieve the long-term stability. The cell (Li|LiCl-LiI|Bi-Pb) is operated at 410 °C by employing the LiCl-LiI (LiCl:LiI = 36:64 mol %) electrolyte and Bi-Pb alloy (Bi:Pb = 55.5:44.5 mol %) positive electrode. The cell shows excellent capacity retention (86.5%) and high Coulombic efficiencies over 99.3% at a high current density of 52 mA cm-2 during 1000th cycles.

  7. Production of Monosugars from Lignocellulosic Biomass in Molten Salt Hydrates: Process Design and Techno-Economic Analysis.

    Science.gov (United States)

    van den Bergh, Johan; Babich, Igor V; O'Connor, Paul; Moulijn, Jacob A

    2017-11-15

    ZnCl 2 hydrate, the main molten salt used in biomass conversion, combined with low concentration HCl is an excellent solvent for the dissolution and hydrolysis of the carbohydrates present in lignocellulosic biomass. The most recalcitrant carbohydrate, cellulose, is dissolved in a residence time less than 1 h under mild conditions without significant degradation. This technology is referred to as BIOeCON-solvent technology. Separation of the sugars from the solution is the main challenge. The earlier conclusion regarding the potential of zeolite beta for selective adsorption has been used as the basis of a scale-up study. The technology of choice is continuous chromatographic separation (e.g., simulated moving bed, SMB). The sugar monomers are separated from the sugar oligomers, allowing the production of monosugars at high yield, using water as an eluent. Results of a pilot plant study are presented showing a stable operation at high selectivity. Several process designs are discussed, and the techno-economic performance of the BIOeCON-solvent technology is demonstrated by comparison with the state-of-the-art technology of NREL (National Renewable Energy Laboratory), which is based on enzymatic conversion of cellulose. It is concluded that the BIOeCON-solvent technology is technically and economically viable and is competitive to the NREL process. Because the BIOeCON-solvent process is in an early stage of development and far from fully optimized, it has the potential to outperform the existing processes.

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

    Energy Technology Data Exchange (ETDEWEB)

    Long, Kevin Nicholas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Roberts, Christine Cardinal [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Roberts, Scott Alan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Grillet, Anne [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    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

  9. Studies of thermal hydraulics and heat transfer in cascade subcritical molten salt reactor

    International Nuclear Information System (INIS)

    Aysen, E.M.; Sedov, A.A.; Subbotin, A.S.

    2005-01-01

    Full text of publication follows: Cascade Subcritical Molten Salt Reactor (CSMSR) consists of three main parts: accelerator-driven proton-bombarded target, central and peripheral zones. External neutrons generated in the result of interaction of protons with the target nuclei are multiplied then in the central zone and leak farther into the peripheral reactor zone, where an efficient burning of Minor Actinides dissolved in a molten salt fluoride composition is produced. The bunch of target and two zones is designed so that preset subcriticality of reactor would not be less than 1% of k eff . A characteristic feature of the reactor is a high density of neutron flux (2.10 15 n/cm 2 s) in the central zone and target and very high volumetric power rate (2000 - 6000 W/cm 3 ) in all the parts of CSMSR. To provide a workability of the core structures under condition of so big level of power rate it is necessary to impose strict limitations on the temperatures and temperature gradients developed in the coolants and constructions. In this reason it has been arranged a calculational-designing study to reveal the problems of heat transfer in the coolant and core structures and to find more appropriate variant of the core and target design, which is a compromise of contradictory requirements: provision of high neutron flux and coolability of the core structures. In this paper the results of studies of thermal hydraulics and heat transfer in the core zones and proton-beam target are presented. Different variants of the target and central zone design as well as application of different kind of coolants in them are discussed and the main problems of heat removal in their structures are analyzed. Multidimensional fields of velocity and temperature got in thermal hydraulics calculations for free flow of fuelled molten salt in cylindrical-cave peripheral CSMSR zone without structures inside are demonstrated. The role of turbulent exchange of momentum and heat for free flow in the

  10. Electrochemical performance of BaSnO3 anode material for lithium-ion battery prepared by molten salt method

    CSIR Research Space (South Africa)

    Nithyadharseni, P

    2016-01-01

    Full Text Available Perovskite-like structure BaSnO(sub3) ceramic oxide has been prepared by low temperature molten salt method using KOH as a flux and Ba(OH)(sub2) and BaCl(sub2) as precursors. The as-prepared compounds were characterized by various techniques...

  11. Production of Monosugars from Lignocellulosic Biomass in Molten Salt Hydrates : Process Design and Techno-Economic Analysis

    NARCIS (Netherlands)

    van den Bergh, Johan; Babich, Igor V.; O'Connor, Paul; Moulijn, J.A.

    2017-01-01

    ZnCl2 hydrate, the main molten salt used in biomass conversion, combined with low concentration HCl is an excellent solvent for the dissolution and hydrolysis of the carbohydrates present in lignocellulosic biomass. The most recalcitrant carbohydrate, cellulose, is dissolved in a residence time less

  12. Critical and accelerator driven sub-critical molten salt reactors for thorium utilisation

    International Nuclear Information System (INIS)

    Degweker, S.B.; Singh, Arun; Rudra, Rashbihari; Ghosh, Biplab

    2013-01-01

    The three stage program for development of nuclear energy in India envisages breeding 233 U in the thorium blankets of fast reactors. Subsequent utilization of the bred fuel should be in a breeding or at least a self sustaining cycle. Molten salt reactors (MSRs) permit better breeding than solid fuelled reactors due to their online fission product removal capability. The fast spectrum MSRs has the advantage that they place much lower demands on the online fission product removal capability. In recent years, there is interest in Accelerator Driven Systems (ADSs) for waste transmutation and power production using Th. Some of these proposals also consider MSRs due to their simpler fuel cycle and higher breeding. The paper presents a brief review of MSRs for utilizing Th and discusses some of the work done by us in recent times. (author)

  13. Electrochemical Carbonitriding of 316L Stainless Steel in Molten Salt System

    Science.gov (United States)

    Ren, Yanjie; Xiao, Bo; Chen, Yaqing; Chen, Jian; Chen, Jianlin

    This paper reports an electrochemical route for carbonitriding 316L stainless steel in molten salts. Carbonitriding process was accomplished in molten alkaline chloride (LiCl/KCl) with the addition of KNO2 at 480∘C using a three-electrode system in which a carbon sheet was the counter electrode. The carbonitriding layer of 316L stainless steel obtained by potentiostatic electrolysis was analyzed by several physical techniques. The results showed that a compact layer with a thickness of about 7μm formed after the treatment. According to X-ray diffraction analysis, chromium nitride and carbide formed on the surface of carbonitriding layer. The microhardness of the carbonitriding layer is HV 336, as compared to HV 265 for the substrate.

  14. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    Science.gov (United States)

    Harto, Andang Widi

    2012-06-01

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  15. Electrorefining of High Carbon Ferromanganese in Molten Salts to Produce Pure Ferromanganese

    Directory of Open Access Journals (Sweden)

    Xiao S. J.

    2017-09-01

    Full Text Available High carbon ferromanganese is used as a starting material to prepare pure ferromanganese by electrorefining in molten salts. High carbon ferromanganese was applied as the anode, molybdenum was the cathode and Ag/AgCl was the reference electrode. The anodic dissolution was investigated by linear polarization in molten NaCl-KCl system. Then potentiostatic electrolysis was carried out to produce pure ferromanganese from high carbon ferromanganese. The cathodic product was determined to be a mixture of manganese and iron by x-ray diffraction (XRD. The content of carbon in the product was analyzed by carbon and sulfur analyzer. The post-electrolysis anode was characterized by scanning electron microscope (SEM. The mechanism of the anode dissolution and the distribution of the main impurity of carbon and silicon after electrolysis were discussed.

  16. Treatment of waste by the Molten Salt Oxidation process at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Crosley, S.M.; Lorenzo, D.K.; Van Cleve, J.E.; Gay, R.L.; Barclay, K.M.; Newcomb, J.C.; Yosim, S.J.

    1993-01-01

    The Molten Salt Oxidation (MSO) process has been under development by the Energy Technology Engineering Center (ETEC) to treat hazardous, radioactive, and mixed waste. Testing of the system was done on a number of wastes to demonstrate the technical feasibility of the process. This testing included simulated intermediate level waste (ILW) from the Oak Ridge National Laboratory. The intermediate level waste stream consisted of a slurry of concentrated aqueous solutions of sodium hydroxide and sodium nitrate, with a small amount of miscellaneous combustible components (PVC, TBP, kerosene, and ion exchange resins). The purpose of these tests was to evaluate the destruction of the organics, evaporation of the water, and conversion of the hazardous salts (hydroxide and nitrate) to non-hazardous sodium carbonate. Results of the tests are discussed and analyzed, and the possibilities of applying the MSO process to different waste streams at ORNL in the future are explored

  17. High-safety and economical small molten-salt fission power stations and their developmental program

    International Nuclear Information System (INIS)

    Furukawa, K.; Mitachi, K.; Minami, K.; Kato, Y.

    1988-01-01

    The nuclear energy industry is not settled yet as one of the sound economical industries. Its establishment should obviously depend on the solution of the following problems: ''natural'' safety (depending on inherent natures), nuclear proliferation resistance - nearly non-production and effective incineration of Pu, Am and Om, universal resource, flexible power-size and excellent economy - wide applicability including Developing Countries. Therefore, some essentially new principles have to introduce in the nuclear energy system design. These are thorium utilization, fluid-fuel concepts, especially molten-fluoride technology, and separation of fissile-breeding and power-generation. This philosophy is named Thorium Molten-Salt Nuclear Energy Synergetics [THOMSNES]. Its practical development program is presented

  18. Denatured Molten Salt Reactors (DMSR): an idea whose time has finally come?

    International Nuclear Information System (INIS)

    LeBlanc, D.

    2010-01-01

    Molten Salt Reactors are one of six next generation designs chosen by the Gen IV program. Traditionally these reactors are thought of as thermal breeder reactors running on the thorium to 233 U cycle and the historical competitor to fast breeder reactors. However, simplified versions running as converter reactors without any fuel processing and consuming low enriched uranium are perhaps a more attractive option. Uranium consumption levels are less than 1/6th that of LWR or a 1/4th of CANDU while at the same time offering clear advantages in safety, capital cost and long lived waste production along with increased proliferation resistance. A review of previous work and potential improvements proposed by the author will be presented. (author)

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

    International Nuclear Information System (INIS)

    Bishop, Clare L; Wilson, Mark

    2009-01-01

    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.

  20. Electrochemical reduction behavior of U3O8 powder in a LiCl molten salt

    International Nuclear Information System (INIS)

    Jeong, Sang Mun; Shin, Ho-Sup; Hong, Sun-Seok; Hur, Jin-Mok; Do, Jae Bum; Lee, Han Soo

    2010-01-01

    The reduction path of the U 3 O 8 powder vol-oxidized at 1200 deg. C has been determined by a series of electrochemical experiments in a 1 wt.% Li 2 O/LiCl molten salt. Various reaction intermediates are observed by during electrolysis of U 3 O 8 . The formation of the metallic uranium is caused from two different reduction paths, a direct reduction of uranium oxide and an electro-lithiothermic reduction. As the uranium oxide is converted to the metallic uranium, the lithium metal is more actively formed in the cathode basket. The reducibility of the rare earth oxides with the U 3 O 8 powder has been tested by constant voltage electrolysis. The results suggest the advanced vol-oxidation could lead to the enhancement in the reducibility of the rare earth fission products.

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

    International Nuclear Information System (INIS)

    Noh, Seunghyo; Kwak, Dohyun; Lee, Juseung; Kang, Joonhee; Han, Byungchan

    2014-01-01

    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.

  2. Solar thermal power & gas turbine hybrid design with molten salt storage tank

    Science.gov (United States)

    Martín, Fernando; Wiesenberg, Ralf; Santana, Domingo

    2017-06-01

    Taking into consideration the need to decelerate the global climatic change, power generation has to shift from burning fossil fuel to renewable energy source in short medium period of time. In this work, we are presenting a new model of a solar-gas natural hybrid power cycle with the main aim of decoupling the solar generation system from the gas turbine system. The objective is to have high solar power contribution compared to conventional ISCC plants [2], producing firm and dispatchable electricity at the same time. The decoupling is motivated by the low solar contribution reached by the ISCC, which is technically limited to maximum of 15%, [4]. In our case, we have implemented a solar tower with molten salts as working fluid. Central receiver systems get higher performance than others systems, like parabolic trough technology [1], due to the higher temperature achieved in the heat transferred fluid HTF, close to 560°C.

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

    International Nuclear Information System (INIS)

    Furukawa, Kazuo

    1991-01-01

    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

  4. 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.

  5. 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.

  6. 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.

  7. Chemico-technological support of transmutation objectives: Solid, molten salt and liquid blanket

    Energy Technology Data Exchange (ETDEWEB)

    Volk, V.I.; Zakharkin, B.S.; Vakhrushin, A.Y. [VNIINM, Moscow (Russian Federation)

    1995-10-01

    Chemical and technological provision for the transmutation process, independantly on the scheme of its conduction, includes: fuel composition separation for fractions of components, subjected to annihilation; their transition into chemical form, in which they are present in the reactor; discharge and return into the form, convenient for chemical reprocessing, providing for the transmutation products separation from the components being transmutated and transferring of short-lived isotopes into the form of their temporary storage. The authors discuss different chemical processes which can be used in these steps to either improve efficiency or minimize additional waste generation and expense associated with decontamination. They consider processes involving molten salts for circulation of wastes thru transmutation steps, and possible advantages in extraction processes.

  8. Estimation of zirconium in various process streams in molten salt electrorefining process

    International Nuclear Information System (INIS)

    Suganthi, S.; Vandarkuzhali, S.; Venkatesh, P.; Prabhakara Reddy, B.; Nagarajan, K.

    2012-01-01

    Molten salt electrorefining process is a non-aqueous pyrochemical process suitable for reprocessing spent metallic fuel. In this process the spent fuel is taken at the anode and the fuel elements are selectively electrotransported to a suitable cathode (either a solid steel cathode or liquid cadmium cathode) using molten LiCl-KCI as electrolyte. We have demonstrated electrorefining of UZr alloy at engineering scale level. 1 Kg U-6%Zr alloy was taken at the anode and pure uranium was recovered at a steel cathode using molten LiCIKCI-5%UCI 3 as electrolyte at 773 K. In this paper we present the method of dissolution, sample preparation and estimation of zirconium in various process streams in the electrorefining experiments carried out in our laboratory

  9. 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)

  10. The effect of different surface treatments on the molten salt hot corrosion of IN-657 superalloy

    International Nuclear Information System (INIS)

    Otero, E.; Utrilla, M.V.

    1996-01-01

    The influence of different surface treatments (laser surface treatment, silicon nitride, and 25 wt.% Al 2 (OH) 4 Si 2 O 5 -75 wt.% Na 2 SiO 3 ), on the hot corrosion behaviour of IN-657 superalloy has been studied. The corrosion tests were performed in an oxidizing atmosphere in contact with a molten mixture of 82% K 2 S 2 O 7 -18% V 2 O 5 . The effects of operation temperature and carbon residues as component of the molten mixture were also analyzed. The best results were obtained for the kaolin and methasilicate mixture coating. Laser surface treatment and the CVD-silicon nitride are not recommended for surface modification of the IN-657 superalloy. Taking into account the results obtained, the authors suggest the average temperature and molten salt compositions where the above coating can work satisfactorily. (orig.)

  11. Porous membrane electrochemical cell for uranium and transuranic recovery from molten salt electrolyte

    Science.gov (United States)

    Willit, James L.

    2007-09-11

    An improved process and device for the recovery of the minor actinides and the transuranic elements (TRU's) from a molten salt electrolyte. The process involves placing the device, an electrically non-conducting barrier between an anode salt and a cathode salt. The porous barrier allows uranium to diffuse between the anode and cathode, yet slows the diffusion of uranium ions so as to cause depletion of uranium ions in the catholyte. This allows for the eventual preferential deposition of transuranics present in spent nuclear fuel such as Np, Pu, Am, Cm. The device also comprises an uranium oxidation anode. The oxidation anode is solid uranium metal in the form of spent nuclear fuel. The spent fuel is placed in a ferric metal anode basket which serves as the electrical lead or contact between the molten electrolyte and the anodic uranium metal.

  12. 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.

  13. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    International Nuclear Information System (INIS)

    Wishau, R.

    1998-05-01

    Molten salt oxidation (MSO) is proposed as a 238 Pu waste treatment technology that should be developed for volume reduction and recovery of 238 Pu and as an alternative to the transport and permanent disposal of 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 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 238 Pu contaminated wastes is reduced to 30 drums. Further 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 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 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

  14. Experimental and theoretical studies in Molten Salt Natural Circulation Loop (MSNCL)

    International Nuclear Information System (INIS)

    Srivastava, A.K.; Borgohain, A.; Jana, S.S.; Bagul, R.K.; Singh, R.R.; Maheshwari, N.K.; Belokar, D.G.; Vijayan, P.K.

    2014-12-01

    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 NaNO 3 and KNO 3 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)

  15. 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.

  16. Thermal Properties of LiCl-KCl Molten Salt for Nuclear Waste Separation

    International Nuclear Information System (INIS)

    Sridharan, Kumar; Allen, Todd; Anderson, Mark; Simpson, Mike

    2012-01-01

    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

  17. Review of ORNL's MSR technology and status

    International Nuclear Information System (INIS)

    Toth, L.M.; Gat, U.; Del Cul, G.D.; Dai, S.; Williams, D.F.

    1996-01-01

    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-BeF 2 solvent system. 8 refs, 2 figs

  18. Properties and heat transfer coefficients of four molten-salt high temperature heat transfer fluid candidates for concentrating solar power plants

    Science.gov (United States)

    Liu, T. L.; Liu, W. R.; Xu, X. H.

    2017-11-01

    Heat transfer fluid is one critical component for transferring and storing heat energy in concentrating solar power systems. Molten-salt mixtures can be used as high temperature heat transfer fluids because of their thermophysical properties. This paper studied the thermophysical properties of Li2CO3-Na2CO3-K2CO3 eutectic salt and three eutectic chloride salts NaCl-KCl-ZnCl2 with different compositions in the range of 450-600°C and 250-800°C, respectively. Properties including specific heat capacity, thermal conductivity, density and viscosity were determined based on imperial correlations and compared at different operating temperatures. The heat transfer coefficients of using different eutectic salts as heat transfer fluids were also calculated and compared in their operating temperature range. It is concluded that all the four eutectic salts can satisfy the requirements of a high-temperature heat transfer fluid.

  19. 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)

    Bae, Judong; Hwang, Il Soon; Park, Byung Gi; Hong, Kwang

    2010-01-01

    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-SmCl 3 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-LnCl n system

  20. A study on the electrolytic reduction of U3O8 to uranium metal in LiCl-Li2O molten salt

    International Nuclear Information System (INIS)

    Seo, J. S.; Heo, J. M.; Hong, S. S.; Kang, D. S.; Park, S. W.

    2002-01-01

    New electrolytic reduction technology was proposed that is based on the intregration of metallization of U 3 O 8 and Li 2 O electrowinning. In this electrolytic reduction reaction, electrolytically reduced Li deposits on cathode and simultaneously reacts with uranium oxide to produce uranium metal showing more than 99% conversion. For the verification of process feasibility, the experiments to obtain basic data on the metallization of uranium oxide, materials for cathode and anode electrode, the characteristics of closed recycle of Li 2 O and mass transfer were carried out. This evolutionary electrolytic reduction technology would give benefits over the conventional Li-reduction process improving economic viability such as: avoidance of handling of chemically active Li-LiCl molten salt, increase of metallization yield, and simplification of process

  1. 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)

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

    2009-01-01

    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

  2. Electrochemistry of Centered Hexanuclear Zirconium Halide Clusters in Ambient-Temperature Chloroaluminate Molten Salts.

    Science.gov (United States)

    Sun, Dong; Hughbanks, Timothy

    1999-03-08

    Ambient temperature AlCl(3)-1-ethyl-3-methylimidazolium chloride (ImCl) molten salts, both basic (40/60 mol % AlCl(3)/ImCl) and acidic (60/40 mol % AlCl(3)/ImCl), were used in an electrochemical investigation of centered hexanuclear zirconium halide clusters. In the basic molten salt, these [(Zr(6)ZCl(12))Cl(6)](n-) (Z = Be, B, C, Mn, Fe) centered clusters exhibit the following electrochemical reactions on a glassy carbon electrode (potentials vs Al/Al(3+)): [(Zr(6)BeCl(12))Cl(6)](n)()(-) + e(-) right harpoon over left harpoon [(Zr(6)BeCl(12))Cl(6)](()(n)()(+1))(-), E(1/2) = -0.613 V (n = 4), E(1/2) = -1.085 V (n = 5); [(Zr(6)BCl(12))Cl(6)](n)()(-) + e(-) right harpoon over left harpoon [(Zr(6)BCl(12))Cl(6)](()(n)()(+1))(-), E(1/2) = -0.365 V (n = 4), E(1/2) = 0.072 V (n = 3); [(Zr(6)CCl(12))Cl(6)](n)()(-) + e(-) right harpoon over left harpoon [(Zr(6)CCl(12))Cl(6)](()(n)()(+1))(-), E(1/2) = 0.230 V (n = 3); [(Zr(6)MnCl(12))Cl(6)](4)(-) + e(-) right harpoon over left harpoon [(Zr(6)MnCl(12))Cl(6)](5)(-), E(1/2) = -0.432 V. In the acidic melt, only electrochemical reactions [(Zr(6)BeCl(12))(AlCl(4))(6)](n)()(-) + e(-) right harpoon over left harpoon [(Zr(6)BeCl(12))(AlCl(4))(6)](()(n)()(+1))(-), E(1/2) = -0.069 V (n = 5), E(1/2) = 0.504 V (n = 4), and [(Zr(6)BCl(12))(AlCl(4))(6)](4)(-) + e(-) right harpoon over left harpoon [(Zr(6)BCl(12))(AlCl(4))(6)](5)(-), E(1/2) = 0.700 V are clearly observed. These data is consistent with less systematic observations of oxidation of these clusters in solution. No unambiguous one-electron electrochemical reduction nor oxidation is observable for the Fe-centered cluster in the ionic liquids. The half-wave potentials of the above reactions of the Be-, B-, C-, and Mn-centered clusters are controlled largely by clusters' charges. This correlation of redox potentials allows a useful direct comparison with data for related hexanuclear niobium clusters in the literature.

  3. 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.

  4. 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.

  5. Morphology-controlled synthesis of CdWO4 nanorods and nanoparticles via a molten salt method

    International Nuclear Information System (INIS)

    Wang Yonggang; Ma Junfeng; Tao Jiantao; Zhu Xiaoyi; Zhou Jun; Zhao Zhongqiang; Xie Lijin; Tian Hua

    2006-01-01

    Cadmium tungstate (CdWO 4 ) nanoparticles and nanorods have been successfully synthesized by a molten salt method at 270 deg. C, and the morphology of the nanocrystals can be controlled by adjusting such reaction conditions as the calcined time and the weight ratio of the salt to the CdWO 4 precursor. The resultant sample is a pure phase of CdWO 4 without any other impurities

  6. Surface Morphology and Microstructure of Zinc Deposit From Imidazole with Zinc Chloride Low Temperature Molten Salt Electrolyte in The Presence of Aluminium Chloride

    OpenAIRE

    Shanmugasigamani Srinivasan, M. Selvam

    2013-01-01

    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 p...

  7. Measuring lanthanide concentrations in molten salt using laser-induced breakdown spectroscopy (LIBS).

    Science.gov (United States)

    Weisberg, Arel; Lakis, Rollin E; Simpson, Michael F; Horowitz, Leo; Craparo, Joseph

    2014-01-01

    The versatility of laser-induced breakdown spectroscopy (LIBS) as an analytical method for high-temperature applications was demonstrated through measurement of the concentrations of the lanthanide elements europium (Eu) and praseodymium (Pr) in molten eutectic lithium chloride-potassium chloride (LiCl-KCl) salts at a temperature of 500 °C. Laser pulses (1064 nm, 7 ns, 120 mJ/pulse) were focused on the top surface of the molten salt samples in a laboratory furnace under an argon atmosphere, and the resulting LIBS signals were collected using a broadband Echelle-type spectrometer. Partial least squares (PLS) regression using leave-one-sample-out cross-validation was used to quantify the concentrations of Eu and Pr in the samples. The root mean square error of prediction (RMSEP) for Eu was 0.13% (absolute) over a concentration range of 0-3.01%, and for Pr was 0.13% (absolute) over a concentration range of 0-1.04%.

  8. 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.

  9. Electrodeposition of metallic tungsten coating from binary oxide molten salt on low activation steel substrate

    International Nuclear Information System (INIS)

    Liu, Y.H.; Zhang, Y.C.; Jiang, F.; Fu, B.J.; Sun, N.B.

    2013-01-01

    Tungsten is considered a promising plasma facing armor material for future fusion devices. An electrodeposited metallic tungsten coating from Na 2 WO 4 –WO 3 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%

  10. 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.

  11. Engineering development studies for molten-salt breeder reactor processing No. 18

    International Nuclear Information System (INIS)

    Hightower, J.R. Jr.

    1975-03-01

    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--BeF 2 --ThF 4 (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.)

  12. Molten salt oxidation of mixed waste: Preliminary bench-scale experiments without radioactivity

    International Nuclear Information System (INIS)

    Haas, P.A.; Rudolph, J.C.; Bell, J.T.

    1994-06-01

    Molten salt oxidation (MSO) is a process in which organic wastes are oxidized by sparging them with air through a bed of molten sodium carbonate (bp 851 degrees C) at ≥ 900 degrees C. This process is readily applicable to the mixed waste because acidic products from Cl, S, P, etc., in the waste, along with most metals and most radionuclides, are retained within the melt as oxides or salts. Rockwell International has studied the application of MSO to various wastes, including some mixed waste. A unit used by Rockwell to study the mixed waste treatment is presently in use at Oak Ridge National Laboratory (ORNL). ORNL's studies to date have concentrated on chemical flowsheet questions. Concerns that were studied included carbon monoxide (CO) emissions, NO x , emissions, and metal retention under a variety of conditions. Initial experiments show that CO emissions increase with increasing NaCl content in the melt, increasing temperature, and increasing airflow. Carbon monoxide content is especially high (> 2000 ppm) with high chlorine content (> 10%). Thermal NO x , emissions are relatively low ( x , The metal contents of the melt and of knockout pot samples of condensed salt show high volatilities of Cs as CsCl. Average condensed salt concentrations were 60% for barium and 100% for strontium and cobalt. The cerium disappeared -- perhaps from deposition on the alumina reactor walls

  13. A feasibility study of scaling-up the electrolytic production of carbon nanotubes in molten salts

    International Nuclear Information System (INIS)

    Dimitrov, Aleksandar T.; Chen, George Z.; Kinloch, Ian A.; Fray, Derek J.

    2002-01-01

    The feasibility of scaling-up the electrolytic production of carbon nanotubes in molten salts has been investigated with the aid of electron microscopy (TEM and SEM). Using molten LiCl as the electrolyte and commercial graphite as both cathode and anode materials, carbon nanomaterials, including nanotubes, were prepared by constant voltage electrolysis. The cell was more than 20 times as large as that used in previous work. The nanotube concentration in the final product increased with cell voltage (including iR drop) from 1 vol.% at 4.0 V to 35 vol.% at 8.4 V. Under desired conditions, the charge and energy consumption for the cathode erosion was 0.28 Ah/g and 4.1 Wh/g, of which 60-70 wt.% were for producing nanomaterials (nanotubes: >30 vol.%). When adding 1 wt.% SnCl 2 to the electrolyte, partial and fully filled nanotubes were obtained with the nanomaterials containing up to 20 wt.% Sn. Preliminary results from applying the product as the electrode in lithium ion batteries are reported

  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. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  15. The electrodeposition and rare earths reduction in the molten salt actinides recovery systems using liquid metal

    International Nuclear Information System (INIS)

    Shim, J-B.; Lee, J-H.; Kwon, S-W.; Ahn, B-G.; Woo, M-S.; Lee, B-J.; Kim, E-H.; Park, H-S.; Yoo, J-H.

    2005-01-01

    A pyrochemical partitioning system uses liquid metals such as cadmium and bismuth in order to recover the actinide metals from a molten salt mixture containing rare earth fission product metals. The liquid metals play roles as a cathode in the electrowinning or an extracting phase in the reductive extraction operation. The product resulting from the above operations is metal-cadmium or-bismuth alloy, which should contain the rare earth element amounts as low as possible for a transmutation purpose. In this study, the electrodeposition behaviours of uranium and lanthanide elements such as La, Ce and Nd were investigated for solid molybdenum and liquid cadmium electrodes in a molten LiCl-KCl eutectic salt. Electrochemical methods used are a cyclic voltammetry (CV) and a chronopotentiometry for monitoring the salt phase and recovering the metals, respectively. The CV graphs for monitoring the oxidizing agent CdCl 2 in the salt phase were obtained. These show a time dependently disappearance of the oxidizing agent corresponding to the formation of UCl 3 by inserting the uranium metal into the salt. Also, a sequential oxidation technique which is added at a controlled amount of the oxidizing agents into the salt phase was applied. It was found that this method is feasible for the selective reduction of the rare earths content in liquid metal alloys. (author)

  16. 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.

  17. 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.

  18. Diffusion mechanism in molten salt baths during the production of carbide coatings via thermal reactive diffusion

    Science.gov (United States)

    Ghadi, Aliakbar; Saghafian, Hassan; Soltanieh, Mansour; Yang, Zhi-gang

    2017-12-01

    The diffusion mechanism of carbide-forming elements from a molten salt bath to a substrate surface was studied in this research, with particular focus on the processes occurring in the molten bath at the time of coating. Metal, oxide, and metal-oxide baths were investigated, and the coating process was performed on H13 steel substrates. Scanning electron microscopy and electron-probe microanalysis were used to study the coated samples and the quenched salt bath. The thickness of the carbide coating layer was 6.5 ± 0.5, 5.2 ± 0.5, or 5.7 ± 0.5 μm depending on whether it was deposited in a metal, oxide, or metal-oxide bath, respectively. The phase distribution of vanadium-rich regions was 63%, 57%, and 74% of the total coating deposited in metal, oxide, and metal-oxide baths, respectively. The results obtained using the metal bath indicated that undissolved suspended metal particles deposited onto the substrate surface. Then, carbon subsequently diffused to the substrate surface and reacted with the metal particles to form the carbides. In the oxide bath, oxide powders dissolved in the bath with or without binding to the oxidative structure (Na2O) of borax; they were then reduced by aluminum and converted into metal particles. We concluded that, in the metal and oxide baths, the deposition of metal particles onto the sample surface is an important step in the formation of the coating.

  19. A novel rechargeable zinc-air battery with molten salt electrolyte

    Science.gov (United States)

    Liu, Shuzhi; Han, Wei; Cui, Baochen; Liu, Xianjun; Zhao, Fulin; Stuart, Jessica; Licht, Stuart

    2017-02-01

    Zinc-air batteries have been proposed for EV applications and large-scale electricity storage such as wind and solar power. Although zinc-air batteries are very promising, there are numerous technological barriers to overcome. We demonstrate for the first time, a new rechargeable zinc-air battery that utilizes a molten Li0.87Na0.63K0.50CO3 eutectic electrolyte with added NaOH. Cyclic voltammetry reveals that a reversible deposition/dissolution of zinc occurs in the molten Li0.87Na0.63K0.50CO3 eutectic. At 550 °C, this zinc-air battery performs with a coulombic efficiency of 96.9% over 110 cycles, having an average charging potential of ∼1.43 V and discharge potential of ∼1.04 V. The zinc-air battery uses cost effective steel and nickel electrodes without the need for any precious metal catalysts. Moreover, the molten salt electrolyte offers advantages over aqueous electrolytes, avoiding the common aqueous alkaline electrolyte issues of hydrogen evolution, Zn dendrite formation, "drying out", and carbonate precipitation.

  20. 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.

  1. Thermal analysis to support decommissioning of the molten salt reactor experiment

    Energy Technology Data Exchange (ETDEWEB)

    Sulfredge, C.D.; Morris, D.G.; Park, J.E.; Williams, P.T.

    1996-06-01

    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 UF{sub 6} and F{sub 2} 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.

  2. A descriptive model of the molten salt reactor experiment after shutdown: Review of FY 1995 progress

    International Nuclear Information System (INIS)

    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

  3. Experimental results from a laboratory-scale molten salt thermocline storage

    Science.gov (United States)

    Seubert, Bernhard; Müller, Ralf; Willert, Daniel; Fluri, Thomas

    2017-06-01

    Single-tank storage presents a valid option for cost reduction in thermal energy storage systems. For low-temperature systems with water as storage medium this concept is widely implemented and tested. For high-temperature systems very limited experimental data are publicly available. To improve this situation a molten salt loop for experimental testing of a single-tank storage prototype was designed and built at Fraunhofer ISE. The storage tank has a volume of 0.4 m3 or a maximum capacity of 72 kWhth. The maximum charging and discharging power is 60 kW, however, a bypass flow control system enables to operate the system also at a very low power. The prototype was designed to withstand temperatures up to 550 °C. A cascaded insulation with embedded heating cables can be used to reduce the effect of heat loss on the storage which is susceptible to edge effects due to its small size. During the first tests the operating temperatures were adapted to the conditions in systems with thermal oil as heat transfer fluid and a smaller temperature difference. A good separation between cold and hot fluid was achieved with temperature gradients of 95 K within 16 cm.

  4. A study of metallic coatings obtained by electrolysis of molten salts

    International Nuclear Information System (INIS)

    Broc, Michel.

    1978-06-01

    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 800 0 C. 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 [fr

  5. The preparation and thermoelectric properties of molten salt electrodeposited boron wafers

    International Nuclear Information System (INIS)

    Kumashiro, Y.; Ozaki, S.; Sato, K.; Kataoka, Y.; Hirata, K.; Yokoyama, T.; Nagatani, S.; Kajiyama, K.

    2004-01-01

    We have prepared electrodeposited boron wafer by molten salts with KBF 4 -KF at 680 deg. C using graphite crucible for anode and silicon wafer and nickel plate for cathodes. Experiments were performed by various molar ratios KBF 4 /KF and current densities. Amorphous p-type boron wafers with purity 87% was deposited on nickel plate for 1 h. Thermal diffusivity by ring-flash method and heat capacity by DSC method produced thermal conductivity showing amorphous behavior in the entire temperature range. The systematical results on thermoelectric properties were obtained for the wafers prepared with KBF 4 -KF (66-34 mol%) under various current densities in the range 1-2 A/cm 2 . The temperature dependencies of electrical conductivity showed thermal activated type with activation energy of 0.5 eV. Thermoelectric power tended to increase with increasing temperature up to high temperatures with high values of (1-10) mV/K. Thermoelectric figure-of-merit was 10 -4 /K at high temperatures. Estimated efficiency of thermoelectric energy conversion would be calculated to be 4-5%

  6. 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.

  7. Separation and recovery of uranium ore by chlorinating, chelate resin and molten salt treatment

    International Nuclear Information System (INIS)

    Taki, Tomohiro

    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.)

  8. Modular Stellarator Fusion Reactor (MSR) concept

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.

    1981-01-01

    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

  9. 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)

  10. 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)

    2013-05-01

    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

  11. Laminar iridium coating produced by pulse current electrodeposition from chloride molten salt

    Science.gov (United States)

    Zhu, Li'an; Bai, Shuxin; Zhang, Hong; Ye, Yicong

    2013-10-01

    Due to the unique physical and chemical properties, Iridium (Ir) is one of the most promising oxidation-resistant coatings for refractory materials above 1800 °C in aerospace field. However, the Ir coatings prepared by traditional methods are composed of columnar grains throughout the coating thickness. The columnar structure of the coating is considered to do harm to its oxidation resistance. The laminar Ir coating is expected to have a better high-temperature oxidation resistance than the columnar Ir coating does. The pulse current electrodeposition, with three independent parameters: average current density (Jm), duty cycle (R) and pulse frequency (f), is considered to be a promising method to fabricate layered Ir coating. In this study, laminar Ir coatings were prepared by pulse current electrodeposition in chloride molten salt. The morphology, roughness and texture of the coatings were determined by scanning electron microscope (SEM), profilometer and X-ray diffraction (XRD), respectively. The results showed that the laminar Ir coatings were composed of a nucleation layer with columnar structure and a growth layer with laminar structure. The top surfaces of the laminar Ir coatings consisted of cauliflower-like aggregates containing many fine grains, which were separated by deep grooves. The laminar Ir coating produced at the deposition condition of 20 mA/cm2 (Jm), 10% (R) and 6 Hz (f) was quite smooth (Ra 1.01 ± 0.09 μm) with extremely high degree of preferred orientation of , and its laminar structure was well developed with clear boundaries and uniform thickness of sub-layers.

  12. 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.

  13. Electromigration in molten salts and application to isotopic separation of alkaline and alkaline-earth elements

    International Nuclear Information System (INIS)

    Menes, F.

    1969-01-01

    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 CaBr 2 - (CaBr 2 + 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 48 Ca 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, MgBr 2 - CeBr 3 . 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) [fr

  14. An Investigation on the Thermophysical Properties of a Binary Molten Salt System Containing Both Aluminum Oxide and Titanium Oxide Nanoparticle Suspensions

    Science.gov (United States)

    Giridhar, Kunal

    Molten salts are showing great potential to replace current heat transfer and thermal energy storage fluids in concentrated solar plants because of their capability to maximize thermal energy storage, greater stability, cost effectiveness and significant thermal properties. However one of the major drawbacks of using molten salt as heat transfer fluid is that they are in solid state at room temperature and they have a high freezing point. Hence, significant resources would be required to maintain it in liquid form. If molten salt freezes while in operation, it would eventually damage piping network due to its volume shrinkage along with rendering the entire plant inoperable. It is long known that addition of nanoparticle suspensions has led to significant changes in thermal properties of fluids. In this investigation, aluminum oxide and titanium oxide nanoparticles of varying concentrations are added to molten salt/solar salt system consisting of 60% sodium nitrate and 40% potassium nitrate. Using differential scanning calorimeter, an attempt will be made to investigate changes in heat capacity of system, depression in freezing point and changes in latent heat of fusion. Scanning electron microscope will be used to take images of samples to study changes in micro-structure of mixture, ensure uniform distribution of nanoparticle in system and verify authenticity of materials used for experimentation. Due to enormous magnitude of CSP plant, actual implementation of molten salt system is on a large scale. With this investigation, even microscopic enhancement in heat capacity and slight lowering of freezing point will lead to greater benefits in terms of efficiency and cost of operation of plant. These results will further the argument for viability of molten salt as a heat transfer fluid and thermal storage system in CSP. One of the objective of this experimentation is to also collect experimental data which can be used for establishing relation between concentration

  15. Molten salts activated by high-energy milling: A useful, low-temperature route for the synthesis of multiferroic compounds

    Energy Technology Data Exchange (ETDEWEB)

    Hernández-Ramírez, Anayantzin; Martínez-Luévanos, Antonia [Facultad de Ciencias Químicas, Universidad Autónoma de Coahuila, V. Carranza s/n, Saltillo, Coahuila 25280 (Mexico); Fuentes, Antonio F. [CINVESTAV Unidad Saltillo, Apdo. Postal 663, Saltillo, Coahuila 25000 (Mexico); Earth and Environmental Science, University of Michigan, 3514 C.C. Little Building, 1100 N. University Avenue, Ann Arbor, MI 48109-1005 (United States); Nelson, Anna-Gay D.; Ewing, Rodney C. [Earth and Environmental Science, University of Michigan, 3514 C.C. Little Building, 1100 N. University Avenue, Ann Arbor, MI 48109-1005 (United States); Montemayor, Sagrario M., E-mail: smmontemayor@gmail.com [Facultad de Ciencias Químicas, Universidad Autónoma de Coahuila, V. Carranza s/n, Saltillo, Coahuila 25280 (Mexico); Earth and Environmental Science, University of Michigan, 3514 C.C. Little Building, 1100 N. University Avenue, Ann Arbor, MI 48109-1005 (United States)

    2014-01-25

    Highlights: • The synthesis route purposed demonstrates the formation of BiFeO{sub 3} at only 500 °C. • The magnetic and ferroelectric properties are comparable to those of bulk BiFeO{sub 3}. • By this route, several phases in Bi{sub 1−x}La{sub x}FeO{sub 3} system are obtained at only 500 °C. • The route developed here could be useful to synthesize other perovskite-type oxides. -- Abstract: There are only a few multiferroic compounds, among which BiFeO{sub 3} is the most important. Research the synthesis of bismuth ferrite, with novel and improved magnetic and electrical properties, has been mainly based on the use of hydrothermal or sol gel methods. However, these methods require either rather extreme conditions or several steps for synthesis. We demonstrate that the use of molten salts, activated by high energy milling, results in pure nanometric BiFeO{sub 3}, LaFeO{sub 3} and intermediate phases in the Bi{sub 1−x}La{sub x}FeO{sub 3} system. The chemical reagents used are Bi(NO{sub 3}){sub 3}⋅5H{sub 2}O, La(NO{sub 3}){sub 3}⋅6H{sub 2}O, Fe(NO{sub 3}){sub 3}⋅9H{sub 2}O and NaOH. A brief milling process of the reagents creates an amorphous precursor and crystalline NaNO{sub 3}. The thermal treatment of the precursors, at 500 °C for two hours, produces a crystalline mixture of Bi{sub 1−x}La{sub x}FeO{sub 3} and NaNO{sub 3}. Simple washing eliminates the NaNO{sub 3}. The characterization of intermediates and final products, through thermal analysis, X-ray diffraction and scanning electronic microscopy, allows the inference of possible mechanism. In addition, vibrating sample magnetometry (VSM) and ferroelectric tests show the typical magnetic and electric polarization loops characteristic of these materials even when formed at the nano-scale.

  16. Molten salt corrosion behavior of structural materials in LiCl-KCl-UCl3 by thermogravimetric study

    Science.gov (United States)

    Rao, Ch Jagadeeswara; Ningshen, S.; Mallika, C.; Mudali, U. Kamachi

    2018-04-01

    The corrosion resistance of structural materials has been recognized as a key issue in the various unit operations such as salt purification, electrorefining, cathode processing and injection casting in the pyrochemical reprocessing of spent metallic nuclear fuels. In the present work, the corrosion behavior of the candidate materials of stainless steel (SS) 410, 2.25Cr-1Mo and 9Cr-1Mo steels was investigated in molten LiCl-KCl-UCl3 salt by thermogravimetric analysis under inert and reactive atmospheres at 500 and 600 °C, for 6 h duration. Insignificant weight gain (less than 1 mg/cm2) in the inert atmosphere and marginal weight gain (maximum 5 mg/cm2) in the reactive atmosphere were observed at both the temperatures. Chromium depletion rates and formation of Cr-rich corrosion products increased with increasing temperature of exposure in both inert and reactive atmospheres as evidenced by SEM and EDS analysis. The corrosion attack by LiCl-KCl-UCl3 molten salt, under reactive atmosphere for 6 h duration was more in the case of SS410 than 9Cr-1Mo steel followed by 2.25Cr-1Mo steel at 500 °C and the corrosion attack at 600 °C followed the order: 9Cr-1Mo steel >2.25Cr-1Mo steel > SS410. Outward diffusion of the minor alloying element, Mo was observed in 9Cr-1Mo and 2.25Cr-1Mo steels at both temperatures under reactive atmosphere. Laser Raman spectral analysis of the molten salt corrosion tested alloys under a reactive atmosphere at 500 and 600 °C for 6 h revealed the formation of unprotected Fe3O4 and α-as well as γ-Fe2O3. The results of the present study facilitate the selection of structural materials for applications in the corrosive molten salt environment at high temperatures.

  17. Front-end and back-end electrochemistry of molten salt in accelerator-driven transmutation systems

    International Nuclear Information System (INIS)

    Williamson, M.A.; Venneri, F.

    1995-01-01

    The objective of this work is to develop preparation and clean-up processes for the fuel and carrier salt in the Los Alamos Accelerator-Driven Transmutation Technology molten salt nuclear system. The front-end or fuel preparation process focuses on the removal of fission products, uranium, and zirconium from spent nuclear fuel by utilizing electrochemical methods (i.e., electrowinning). The same method provides the separation of the so-called noble metal fission products at the back-end of the fuel cycle. Both implementations would have important diversion safeguards. The proposed separation processes and a thermodynamic analysis of the electrochemical separation method are presented

  18. Plutonium (TRU) transmutation and {sup 233}U production by single-fluid type accelerator molten-salt breeder (AMSB)

    Energy Technology Data Exchange (ETDEWEB)

    Furukaw, Kazuo [Tokai Univ., Kanagawa (Japan); Kato, Yoshio [Japan Atom. Ene. Res. Inst., Ibaraki (Japan); Chigrinov, Sergey E. [Academy of Science, Minsk (Belarus)

    1995-10-01

    For practical/industrial disposition of Pu(TRU) by accelerator facility, not only physical soundness and safety but also the following technological rationality should be required: (1) few R&D items including radiation damage, heat removal and material compatibility; (2) few operation/maintenance/processing works: (3) few reproduction of radioactivity; (4) effective energy production in parallel. This will be achieved by the new modification of Th-fertilizing Single-Fluid type Accelerator Molten-Salt Breeder (AMSB), by which a global nuclear energy strategy for next century might be prepared.

  19. Thermal performance prediction and sensitivity analysis for future deployment of molten salt cavity receiver solar power plants in Algeria

    International Nuclear Information System (INIS)

    Boudaoud, S.; Khellaf, A.; Mohammedi, K.; Behar, O.

    2015-01-01

    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 $/kW e . For solar-only power plants, these parameters are respectively about 27% and 0.63 $/kW e . Therefore, hybrid central receiver systems are

  20. Overview of the recovery and processing of 233U from the Oak Ridge molten salt reactor experiment (MSRE) remediation activities

    International Nuclear Information System (INIS)

    Del Cul, G.D.; Icenhour, A.S.; Simmons, D.W.; Trowbridge, L.D.; Williams, D.F.; Toth, L.M.; Dai, S.

    2001-01-01

    The Molten Salt Reactor Experiment (MSRE) was operated at Oak Ridge National Laboratory (ORNL) from 1965 to 1969 to test the concept of a high-temperature, homogeneous, fluid-fueled reactor. The discovery that UF 6 and F 2 migrated from the storage tanks into distant pipes and a charcoal bed resulted in significant activities to remove and recover the 233 U and to decommission the reactor. The recovered fissile uranium will be converted into uranium oxide (U 3 O 8 ), which is a suitable form for long-term storage. This publication reports the research and several new developments that were needed to carry out these unique activities. (author)

  1. 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......-furanose and α-furanose tautomers, which lead to 5-HMF, were significantly decreased with increased salt concentration. Meanwhile the β-pyranose tautomer, which is correlated with humins formation, was increased significantly.......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...

  2. Interaction of molten salts with a semi-anthracite char at 743 K. Influence on the gasification in air

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Serrano, V.; Alfaro-Dominguez, M.; Higes-Rolando, F.J. [Universidad de Extremadura, Badajoz (Spain). Dept. de Quimica Inorganica; Martin-Aranda, M.; Rojas-Cervantes, M.L.; Lopez-Peinado, A.J. [Universidad Nacional de Educacion a Distancia (UNED), Madrid (Spain). Dept. de Quimica Inorganica

    1997-12-31

    The treatments of a semi-anthracite char (AC) with molten salts followed by washing of the intermediate products with distilled water as a rule produce an increase in the surface area of micropores (S{sub mi}) and in the mean equivalent pore diameter (MEPD), which is greater for AC-MgO. The variation of MEPD is only slight for AC-CaO and AC-CoO. The reactivity of the carbon in air at 823 K enhances for a number of samples but especially for AC-CaO and AC-MgO. Washing with HCl generally results in a small increase in S{sub mi} and in a more uniform MEPD. The reactivity of the carbon decreases for most samples including AN and AC. This suggests that mineral components of the starting materials and chemical species present in distilled waterwashed samples (i.e. metal oxides in excess and reaction products of molten salts with the mineral fraction of AC) are catalytically active and that their removal from such samples occurs by washing with HCl. The presence of catalytic species seems to be greater for AC-MgO and AC-CaO than for the other samples. (orig.)

  3. Renewable and high efficient syngas production from carbon dioxide and water through solar energy assisted electrolysis in eutectic molten salts

    KAUST Repository

    Wu, Hongjun

    2017-07-13

    Over-reliance on non-renewable fossil fuel leads to steadily increasing concentration of atmospheric CO2, which has been implicated as a critical factor contributing to global warming. The efficient conversion of CO2 into useful product is highly sought after both in academic and industry. Herein, a novel conversion strategy is proposed to one-step transform CO2/H2O into syngas (CO/H2) in molten salt with electrolysis method. All the energy consumption in this system are contributed from sustainable energy sources: concentrated solar light heats molten salt and solar cell supplies electricity for electrolysis. The eutectic Li0.85Na0.61K0.54CO3/nLiOH molten electrolyte is rationally designed with low melting point (<450 °C). The synthesized syngas contains very desirable content of H2 and CO, with tuneable molar ratios (H2/CO) from 0.6 to 7.8, and with an efficient faradaic efficiency of ∼94.5%. The synthesis of syngas from CO2 with renewable energy at a such low electrolytic temperature not only alleviates heat loss, mitigates system corrosion, and heightens operational safety, but also decreases the generation of methane, thus increases the yield of syngas, which is a remarkable technological breakthrough and this work thus represents a stride in sustainable conversion of CO2 to value-added product.

  4. Synthesis and piezoelectric properties of KxNa1-xNbO3 ceramic by molten salt method

    International Nuclear Information System (INIS)

    Li Yueming; Wang Jinsong; Liao Runhua; Huang Dan; Jiang Xiangping

    2010-01-01

    K x Na 1-x NbO 3 ceramic powder with perovskite structure was synthesized in molten salt with a Na 2 CO 3 /K 2 CO 3 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 o C. It is found that the synthesizing temperature and salt-to-oxide ratios had significant effects on the morphology of K x Na 1-x NbO 3 powder. The X-ray diffraction analysis indicated that a pure perovskite structure of K x Na 1-x NbO 3 powder could be synthesized at 650 o C. The microstructure observation revealed that the crystal morphology of K x Na 1-x NbO 3 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 K x Na 1-x NbO 3 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: d 33 = 120 pC/N, T c = 406 o C, Q m = 126 and k p = 0.302.

  5. Renewable and high efficient syngas production from carbon dioxide and water through solar energy assisted electrolysis in eutectic molten salts

    Science.gov (United States)

    Wu, Hongjun; Liu, Yue; Ji, Deqiang; Li, Zhida; Yi, Guanlin; Yuan, Dandan; Wang, Baohui; Zhang, Zhonghai; Wang, Peng

    2017-09-01

    Over-reliance on non-renewable fossil fuel leads to steadily increasing concentration of atmospheric CO2, which has been implicated as a critical factor contributing to global warming. The efficient conversion of CO2 into useful product is highly sought after both in academic and industry. Herein, a novel conversion strategy is proposed to one-step transform CO2/H2O into syngas (CO/H2) in molten salt with electrolysis method. All the energy consumption in this system are contributed from sustainable energy sources: concentrated solar light heats molten salt and solar cell supplies electricity for electrolysis. The eutectic Li0.85Na0.61K0.54CO3/nLiOH molten electrolyte is rationally designed with low melting point (decreases the generation of methane, thus increases the yield of syngas, which is a remarkable technological breakthrough and this work thus represents a stride in sustainable conversion of CO2 to value-added product.

  6. Dip-coated ZrO2-Y2O3 coatings tested in molten salts for CSP applications

    Science.gov (United States)

    Pérez, Francisco Javier; Encinas-Sánchez, Víctor; Lasanta, María Isabel; de Miguel, María Teresa; García-Martín, Gustavo

    2017-06-01

    In the present work, the behaviour of ZrO2 - Y2O3 coatings in contact with molten salts at 500 °C has been studied. The coatings were prepared by sol-gel and deposited by dip-coating on AISI 304 specimens previously prepared by sanding and polishing. The behaviour in contact with molten salt was studied through static corrosion tests by the immersion of the coated samples in an alkali-nitrate mixture with a composition of 60 wt.% NaNO3/40 wt.% KNO3 (commonly known as Solar Salt). Prior to test, the deposited coatings were characterized using Scanning Electron Microscopy and X-Ray Diffraction, showing a compacted, homogeneous and uniform aspect and t-YSZ as main component. After corrosion tests, the samples were characterized via gravimetric, Scanning Electron Microscopy and X-Ray Diffraction. The results show a good behaviour of the coated samples compared with the bare coupon samples. However after 1000 h of testing m-ZrO2 appears in the composition,. At this preliminary study, results confirm the suitability of ZrO2 - Y2O3 coatings in solar applications after those working hours, although it is necessary to optimize the coating and study its behaviour at longer times.

  7. A nightly conditioning method to reduce parasitic power consumption in molten-salt central-receiver solar-power plants

    Energy Technology Data Exchange (ETDEWEB)

    Pacheco, J.E.

    1995-06-01

    A method to reduce nightly parasitic power consumption in a molten salt central receiver is discussed where salt is drained from the piping and heat tracing is turned off to allow the piping to cool to ambient overnight, then in the morning the pipes are filled while they are cold. Since the piping and areas of the receiver in a molten-nitrate salt central-receiver solar power plant must be electrically heated to maintain their temperatures above the nitrate salt freezing point (430{degrees}F, 221{degrees}C), considerable energy could be used to maintain such temperatures during nightly shut down and bad weather. Experiments and analyses have been conducted to investigate cold filling receiver panels and piping as a way of reducing parasitic electrical power consumption and increasing the availability of the plant. The two major concerns with cold filling are: (1) how far can the molten salt penetrate cold piping before freezing closed and (2) what thermal stresses develop during the associated thermal shock. Experiments and analysis are discussed.

  8. Numerical research on natural convection in molten salt reactor with non-uniformly distributed volumetric heat generation

    International Nuclear Information System (INIS)

    Qian Libo; Qiu Suizheng; Zhang Dalin; Su Guanghui; Tian Wenxi

    2010-01-01

    Molten salt reactor is one of the six Generation IV systems capable of breeding and transmutation of actinides and long-lived fission products, which uses the liquid molten salt as the fuel solvent, coolant and heat generation simultaneously. The present work presents a numerical investigation on natural convection with non-uniform heat generation through which the heat generated by the fluid fuel is removed out of the core region when the reactor is under post-accident condition or zero-power condition. The two-group neutron diffusion equation is applied to calculated neutron flux distribution, which leads to non-uniform heat generation. The SIMPLER algorithm is used to calculate natural convective heat transfer rate with isothermal or adiabatic rigid walls. These two models are coupled through the temperature field and heat sources. The peculiarities of natural convection with non-uniform heat generation are investigated in a range of Ra numbers (10 3 ∼ 10 7 ) for the laminar regime of fluid motion. In addition, the numerical results are also compared with those containing uniform heat generation.

  9. The results of the investigations of Russian Research Center - {open_quotes}Kurchatov Institute{close_quotes} on molten salt applications to problems of nuclear energy systems

    Energy Technology Data Exchange (ETDEWEB)

    Novikov, V.M. [Russian Research Center, Moscow (Russian Federation)

    1995-10-01

    The results of investigations on molten salt (MS) applications to problems of nuclear energy systems that have been conducted in Russian Research {open_quotes}Kurchatov Institute{close_quotes} are presented and discussed. The spectrum of these investigations is rather broad and covers the following items: physical characteristics of molten salt nuclear energy systems (MSNES); nuclear and radiation safety of MSNES; construction materials compatible with MS of different compositions; technological aspects of MS loops; in-reactor loop testing. It is shown that main findings of completed program support the conclusion that there are no physical nor technological obstacles on way of MS application to different nuclear energy systems.

  10. Hot corrosion behavior of nanostructured Gd2O3 doped YSZ thermal barrier coating in presence of Na2SO4 + V2O5 molten salts

    Directory of Open Access Journals (Sweden)

    Yixiong Wang

    2017-08-01

    Full Text Available Nickel-based superalloy DZ125 was first sprayed with a NiCrAlY bond coat and followed with a nanostructured 2 mol% Gd2O3−4.5 mol% Y2O3-ZrO2 (2GdYSZ topcoat using air plasma spraying (APS. Hot corrosion behavior of the as-sprayed thermal barrier coatings (TBCs were investigated in the presence of 50 wt% Na2SO4 + 50 wt% V2O5 as the corrosive molten salt at 900 °C for 100 h. The analysis results indicate that Gd doped YVO4 and m-ZrO2 crystals were formed as corrosion products due to the reaction of the corrosive salts with stabilizers (Y2O3, Gd2O3 of zirconia. Cross-section morphology shows that a thin layer called TGO was formed at the bond coat/topcoat interface. After hot corrosion test, the proportion of m-ZrO2 phase in nanostructured 2GdYSZ coating is lower than that of nano-YSZ coating. The result reveals that nanostructured 2GdYSZ coating exhibits a better hot corrosion resistance than nano-YSZ coating.

  11. Effect of sand and method of mixing on molten salt properties for an open direct absorption solar receiver/storage system

    Science.gov (United States)

    AlQaydi, Muna; Delclos, Thomas; AlMheiri, Saif; Calvet, Nicolas

    2017-06-01

    The concept of CSPonD Demo project is based on a single and open molten salt tank as a thermal solar receiver and storage unit. Therefore, the effect of external environment such as sand and air on the thermophysical properties of nitrate salt (60 wt. % sodium nitrate, 40 wt. % potassium nitrate) has been investigated in this work. Differential Scanning Calorimeter (DSC) was used to determine the melting, solidification temperatures while the thermal stability and mass loss measurements were carried on Thermal Gravimetric Analysis (TGA). Measurements under nitrogen indicate that the adding 2% (w/w) sand has negative impact by increasing the solidification temperature, mass loss percentage and decreasing the stability limit. While the melting temperature was not affected by the sand and by the preparation method. On the other hand, measurement under air showed an increase of the stability limit and decrease of the mass loss percentage. Furthermore, the measurements for the mass loss under air did not reach a stable value, which required further investigation.

  12. Experimental tests about the cooling/freezing of the molten salts in the receiver tubes of a solar power plant with parabolic trough

    Science.gov (United States)

    Gaggioli, Walter; Fabrizi, Fabrizio; Rinaldi, Luca; Di Ascenzi, Primo

    2017-06-01

    In 2003 ENEA realized the PCS experimental Facility at Casaccia Research Centre (Rome, Italy), in order to test in real operating conditions the components of a parabolic trough solar plant, and to evaluate the technical feasibility of using the solar molten salts mixture (60% NaNO3, 40% KNO3, melting point 220÷240°C) in such a type of plant. ENEA also had the need to assess the behaviour of the solar receiver tubes during abnormal operating situations (wrong operation, pump block, power failure, etc.), when a block of the circulation of the molten salts may occur and cause the cooling or, worse, the freezing of the salts mixture inside the pipes. Some experimental tests have been performed, aimed to examine what happens in such a cases. In fact, without quick maneuvers to restart the circulation of the molten salts or to readily empty the receiver tubes, the molten salts contained in them may cool down to temperatures near/below the solidification. In this report are shown the results of the experimental tests carried out on the receiver tubes of the PCS Facility by cooling the process fluid down to temperatures near or below its freezing point. The tests show that the solidification of the salts does not damage the components of the plant but it is manageable and reversible, provided you apply the correct procedures.

  13. 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

  14. Corrosion resistance of ceramic materials in pyrochemical reprocessing atmosphere by using molten salt for spent nuclear oxide fuel. Corrosion research under chlorine gas condition

    International Nuclear Information System (INIS)

    Takeuchi, Masayuki; Hanada, Keiji; Koizumi, Tsutomu; Aose, Shinichi; Kato, Toshihiro

    2002-12-01

    Pyrochemical reprocessing using molten salts (RIAR process) has been recently developed for spent nuclear oxide fuel and discussed in feasibility study. It is required to improve the corrosion resistance of equipments such as electrolyzer because the process is operated in severe corrosion environment. In this study, the corrosion resistance of ceramic materials was discussed through the thermodynamic calculation and corrosion test. The corrosion test was basically carried out in alkali molten salt under chlorine gas condition. And further consideration about the effects of oxygen, carbon and main fission product's chlorides were evaluated in molten salt. The result of thermodynamic calculation shows most of ceramic oxides have good chemical stability on chlorine, oxygen and uranyl chloride, however the standard Gibb's free energies with carbon have negative value. On the other hand, eleven kinds of ceramic materials were examined by corrosion test, then silicon nitride, mullite and cordierite have a good corrosion resistance less than 0.1 mm/y. Cracks were not observed on the materials and flexural strength did not reduce remarkably after 480 hours test in molten salt with Cl 2 -O 2 bubbling. In conclusion, these three ceramic materials are most applicable materials for the pyrochemical reprocessing process with chlorine gas condition. (author)

  15. Structural and Redox Properties of Vanadium Complexes in Molten Salts of Interest for the Catalytic Oxidation of Sulfur Dioxide

    DEFF Research Database (Denmark)

    Boghosian, S.; Chrissanthopoulos, A.; Fehrmann, Rasmus

    2000-01-01

    and vibrational properties of the vanadium complexes formed in the molten salt-gas system V2O5-M2S2O7-M2SO4/SO2-O-2 (M = K or Cs). The spectral features and the exploitation of the relative Raman intensities indicate that the (VO)(2)O(SO4)(4)(+) dimeric complex unit which possesses a V-O-V bridge is formed...... in the V2O5-M2S2O7 molten mixtures. Interaction with SO2 or addition of sulfate leads to cleavage of the V-O-V bridge and formation of the (VO)-O-IV(SO4)(2)(2-) or (VO2)-O-V(SO4)(2)(3-) complex units respectively. The most prominent Raman bands due to the V-V and V-IV complexes formed are assigned...

  16. Quality assurance plan for the molten salt reactor experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1998-02-01

    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

  17. Program management plan for the Molten Salt Reactor Experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1996-09-01

    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

  18. Program management plan for the Molten Salt Reactor Experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    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.

  19. 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.

  20. 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.

  1. A study for an electrolytic reduction of tantalum oxide in a LiCl-Li2O molten salt

    International Nuclear Information System (INIS)

    Park, Sung Bin; Park, Byung Heung; Seo, Chung Seok; Kang, Dae Seung; Kwon, Seon Gil; Park, Seong Won

    2005-01-01

    Korea Atomic Energy Research Institute (KAERI) has developed the Advanced Spent Fuel Conditioning Process (ACP) to be an innovative technology for handling the PWR spent fuel. As part of ACP, the electrolytic reduction process (ER process) is the electrochemical reduction process of uranium oxide to uranium metal in a molten salt. The ER process has advantages in a technical stability, an economic potential and a good proliferation resistance. KAERI has reported on the good experimental results of an electrochemical reduction of the uranium oxide in a 20 kg HM/batch lab-scale. The ER process can be applicable to the reduction of other metal oxides. Metal tantalum powder has attracted attention for a variety of applications. A tantalum capacitor made from superfine and pliable tantalum powders is very small in size and it has a higher-capacitance part, therefore it is useful for microelectronic devices. By the ER process the metal tantalum can be obtained from tantalum pentoxide. In this work, a 40 g Ta 2 O 5 /batch electrochemical reactor was used for the synthesis of the metal tantalum. From the results of the cyclic voltammograms for the Ta 2 O 5 -LiCl-Li 2 O system, the mechanism of the tantalum reduction in a molten LiCl-Li 2 O salt system was investigated. Tantalum pentoxide is chemically reduced to tantalum metal by the lithium metal which is electrochemically deposited into an integrated cathode assembly in the LiCl-Li 2 O molten salt. The experiments for the tantalum reduction were performed with a chronopotentiometry in the reactor cell, the reduced products were analyzed from an analysis of the X-ray diffraction (XRD), scanning electron microscope and energy dispersive X-ray (SEM-EDX). From the results, the electrolytic reduction process is applicable to the synthesis of metal tantalum

  2. Delayed Neutrons Effect on Power Reactor with Variation of Fluid Fuel Velocity at MSR Fuji-12

    Science.gov (United States)

    Kuncoro Aji, Indarta; Pramuditya, Syeilendra; Novitrian; Irwanto, Dwi; Waris, Abdul

    2017-01-01

    As the nuclear reactor operate with liquid fuel, controlling velocity of the fuel flow on the Molten salt reactor very influence on the neutron kinetics in that reactor system. The effect of the pace fuel changes to the populations number of neutrons and power density on vertical direction (1 dimension) from the first until fifth year reactor operating had been analyzed on this research. This research had been conducted on MSR Fuji-12 with a two meters core high, and LiF-BeF2-ThF4-233UF4 as fuel composition respectively 71.78%-16%-11.86%-0.36%. Data of reactivity, neutron flux, and the macroscopic fission cross section obtained from ouput of SRAC (neutronic calculation code has been developed by JAEA, with JENDL-4.0 as data library on the SRAC calculation) was being used for the calculation process of this research. The calculation process of this research had been performed numerically by SOR (successive over relaxation) and finite difference methode, as well as using C programing language. From the calculation, regarding to the value of power density resulting from delayed neutrons, concluded that 20 m/s is the optimum fuel flow velocity in all the years reactor had operated. Where the increases number of power are inversely proportional with the fuel flow speed.

  3. 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.

  4. The effect of molten salt on high temperature behavior of stainless steel and titanium alloy with the presence of water vapor

    Science.gov (United States)

    Baharum, Azila; Othman, Norinsan Kamil; Salleh, Emee Marina

    2018-04-01

    The high temperature oxidation experiment was conducted to study the behavior of titanium alloy Ti6A14V and stainless steel 316 in Na2SO4-50%NaCl + Ar-20%O2 (molten salt) and Na2SO4-50%NaCl + Ar-20%O2 + 12% H2O (molten salt + water vapor) environment at 900°C for 30 hours using horizontal tube furnace. The sample then was investigated using weight change measurement analysis and X-ray diffraction (XRD) analysis to study the weight gained and the phase oxidation that occurred. The weight gained of the titanium alloy was higher in molten salt environment compared to stainless steel due to the rapid growth in the oxide scale but showed almost no change of weight gained upon addition of water vapor. This is due to the alloy was fully oxidized. Stainless steel showed more protection and better effect in molten salt environment compared to mixed environment showed by slower weight gain and lower oxidation rate. Meanwhile, the phase oxidation test of the samples showed that the titanium alloy consist of multi oxide layer of rutile (TiO2) and Al2O3 on the surface of the exposed sample. While stainless steel show the formation of both protective Cr-rich oxide and non-protective Fe-rich oxide layer. This can be concluded that stainless steel is better compared to Ti alloy due to slow growing of chromia oxide. Therefore it is proven that stainless steel has better self-protection upon high temperature exposure.

  5. On the Burning of Plutonium Originating from Light Water Reactor Use in a Fast Molten Salt Reactor—A Neutron Physical Study

    OpenAIRE

    Merk, Bruno; Litskevich, Dzianis

    2015-01-01

    An efficient burning of the plutonium produced during light water reactor (LWR) operation has the potential to significantly improve the sustainability indices of LWR operations. The work offers a comparison of the efficiency of Pu burning in different reactor configurations—a molten salt fast reactor, a LWR with mixed oxide (MOX) fuel, and a sodium cooled fast reactor. The calculations are performed using the HELIOS 2 code. All results are evaluated against the plutonium burning efficiency d...

  6. Research on electrochemical methods for concentration measurement of dissolved ion in molten salt to apply to electrolytic process control. Innovative research adopted in 2002 fiscal year

    International Nuclear Information System (INIS)

    Nagai, Takayuki

    2005-03-01

    The purpose of this research is to establish the online (in-situ) technique for concentration measuring of dissolved ion in the molten salt, and this technique is due to the electrochemical method for the concentration measuring of dissolved ion in solutions like the polarization curve measurement. This research executed the following four items. 1) Examination of possibility for concentration measuring of dissolved ion in molten salt by cyclic voltammetry. 2) Examination of possibility for concentration measuring of dissolved ion in molten salt by various electrochemical methods. 3) Examination of suitable electrochemical method for concentration measuring of dissolved ion. 4) Confirmation of selected electrochemical method for concentration measuring of dissolved ion. It has been understood that the differential pulse voltammetry (DPV) is a promising electrochemical technique for the concentration measuring of dissolved ion in the molten salt as a result of this research. An appropriate measurement condition is as follows, the potential sweep rate is -0.1 V/s, the pulse cycle is 0.1 s, the pulse width is 10 ms, and the pulse voltage is 50 mV. As for the electrodes, the platinum working electrode, the glassy carbon counter electrode, and silver/silver chloride reference electrode are suitable. Moreover, the molar absorptivities of U 3+ , U 4+ , UO 2 + , UO 2 2+ , and the standard redox potentials of couples of U 4+ /U 3+ and UO 2 2+ /UO 2 + were acquired as a basic data of the uranium and the uranyl ion in molten NaCl-2CsCl. (author)

  7. Large-scale synthesis of Pb1-xLa xTiO3 ceramic powders by molten salt method

    International Nuclear Information System (INIS)

    Cai Zongying; Xing Xianran; Yu Ranbo; Liu Guirong; Xing Qifeng

    2006-01-01

    The ferroelectric perovskite type lanthanum doped lead titanate (PLT) ceramic powders were synthesized in one step with the starting materials of PbC 2 O 4 , La 2 O 3 and TiO 2 in NaCl-KCl molten salts in the temperature range of 700-950 deg. C. It was found that molten salt method was a large scale and easy preparation way to produce PLT powders with high dispersity. Tetragonal phase Pb 1-x La x TiO 3 ceramic powders were identified by XRD in the composition range 0 ≤ x ≤ 0.3 and mono-dispersed particles with spheric shape and less than 100 nm size were observed by SEM. The grain sizes of Pb 1-x La x TiO 3 ceramic powders increased with the increase of La content and decreased with calcination temperature. The grain growth progress and the possible reaction mechanism in molten salts and its influencing factors were discussed in this work. The grain growth process was the main influencing factor of the grain size, which depended on the solubility in the flux

  8. Electrochemical transient techniques for determination of uranium and rare-earth metal separation coefficients in molten salts

    Energy Technology Data Exchange (ETDEWEB)

    Kuznetsov, S.A. [Institute of Chemistry, Kola Science Centre, RAS, Apatity, Murmansk Region 184200, Russia (Russian Federation)]. E-mail: kuznet@chemy.kolasc.net.ru; Hayashi, H. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Minato, K. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Gaune-Escard, M. [Ecole Polytechnique, IUSTI-CNRS UMR 6595, University of Provence, F-13453 Marseille, Cedex 13 (France)

    2006-02-25

    The main step in the pyrometallurgical recycling process of spent nuclear fuel is a molten salt electrorefining. The knowledge of separation coefficients of actinides (U, Np, Pu and Am) and rare-earth metals (Y, La, Ce, Nd and Gd) is very important for this step. Usually the separation coefficients are evaluated from the formal standard potentials of metals in melts containing their own ions, i.e. values obtained by potentiometric method. Electrochemical experiments were carried out at 723-823 K in order to estimate separation coefficients in LiCl-KCl eutectic melt containing uranium and lanthanum trichlorides. The electrochemical behaviour of UCl{sub 3} in LiCl-KCl melt was studied by different electrochemical methods. The diffusion coefficients of U(III) were determined by linear sweep voltammetry, chronopotentiometry and chronoamperometry. The standard rate constants of charge transfer for electroreduction of uranium, U(III) + 3e{sup -} {sup {yields}} U, were calculated by the impedance spectroscopy method. The values of constants testify that electroreduction of U(III) to U is mainly controlled by the rate of charge transfer. La(III) discharge on uranium electrode was also investigated. It was shown that for the calculation of uranium and lanthanum separation coefficients it is necessary to determine the voltammetric peak potentials of U(III) and La(III), their concentration in the melt and the kinetic parameters relating to U(III) discharge such as transfer and diffusion coefficients, and standard rate constants of charge transfer.

  9. The Electrochemistry of Li-LiCl-Li2O Molten Salt Systems and the Role of Moisture

    Science.gov (United States)

    Gese, Natalie J.

    Uranium can be recovered from uranium-oxide (UO2) spent fuel through the combination of oxide reduction and electrorefining processes. During oxide reduction, the spent fuel is introduced to molten LiCl-Li 2O salt at 650°C, and the UO2 is reduced to uranium metal via two routes: (1) electrochemically, and (2) chemically by lithium metal (Li°) 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 Li° generation required for the chemical reduction of UO 2. 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-Li 2O system must be understood. The behavior of moisture in the LiCl-Li 2O molten-salt system was studied using cyclic voltammetry, chronopotentiometry, and chronoamperometry while reduction to hydrogen was confirmed with gas chromatography.

  10. 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)

    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

  11. Actinide-Lanthanide separation by an electrolytic method in molten salt media: feasibility assessment of a renewed liquid cathode

    International Nuclear Information System (INIS)

    Huguet, A.

    2009-12-01

    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)

  12. 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.

  13. Nitridation of U and Pu recovered in liquid Cd cathode by molten salt electrorefining of (U,Pu)N

    Energy Technology Data Exchange (ETDEWEB)

    Satoh, Takumi; Iwai, Takashi; Arai, Yasuo [Japan Atomic Energy Agency (Japan)

    2009-06-15

    Solid solutions of actinide mono-nitrides have been proposed as a candidate fuel of the accelerator-driven system (ADS) and Gen.IV-type fast reactors because the thermal conductivity and metal density are higher than those of actinide oxides and also they have high melting temperature. Pyrochemical process has several advantages over conventional wet process in treating of spent nitride fuel. One of the key technologies of the pyrochemical reprocessing of nitride fuel is the formation of the nitrides from actinides in the liquid Cd cathode. The nitridation-distillation combined method was developed and has been adopted to convert the actinides to the nitrides. In this method, the nitridation of actinides and the distillation of Cd occurred simultaneously by heating the actinide-Cd alloys in N{sub 2} gas stream. In the present study, the nitride formation behavior of U and Pu recovered in Cd cathode by molten salt electrorefining of (U,Pu)N was experimentally investigated. In addition, the nitride pellet was prepared form the powder obtained by the nitridation of U and Pu recovered in Cd cathode. (U,Pu)N (PuN = 80 mol %) was used as the starting material in the experiment. Molten salt electrorefining of (U,Pu)N pellet was carried out in the LiCl-KCl eutectic salt with 1.2 wt% PuCl{sub 3} and 0.3 wt% UCl{sub 3} of about 110 g at the constant anodic potential of -0.60 to -0.55 V vs. Ag/AgCl for about 9 hours at 773 K. After the electrorefining, about 42 % of U and Pu in the starting (U,Pu)N pellet was dissolved at the anode and recovered into the liquid Cd cathode. The recovered U-Pu-Cd alloy was heated in an alumina crucible at 973 K for 10 hours under N{sub 2} gas (99.999 %) stream (0.015 L/min). Fine black powder was recovered after heating the U-Pu-Cd alloy. The powder was identified as the single phase solid solution of (U,Pu)N by the XRD analysis. After milling in the agate mortar for 1 hour, the powder was compacted into green pellet under a pressure of about

  14. Characterization of the effects of continuous salt processing on the performance of molten salt fusion breeder blankets

    Science.gov (United States)

    Patterson-Hine, F. A.; Davidson, J. W.; Klein, D. E.; Lee, J. D.

    1985-12-01

    Several continuous salt processing options are available for use in molten salt fusion breeder blanket designs: fluorination only, fluorination plus reductive extraction, and fluorination, plus reductive extraction, plus metal transfer. The effects of processing on blanket performance have been assessed for these 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 method of salt processing was found to have little affect on the level of radioactivity, toxicity, or the thermal behavior of the salt during operation of the reactor. The processing rates necessary to maintain the desired uranium concentrations in the suppressed-fission environment were quite low, which permitted only long-lived species to be removed from the salt. The effects of the processing therefore became apparent only after the radioactivity due to the short-lived species diminished. The effects of the additional processing (reductive extraction and metal transfer) could be seen after approximately 1 year of decay, but were not significant at times closer to shutdown. The reduced radioactivity and corresponding heat deposition were thus of no consequence in accident or maintenance situations. Net fissile production in the Be/MS blanket concept at a fusion power level of 3000 MW at 70% capacity ranged from 5100 kg/year to 5170 kg/year for uranium concentrations of 0.11% and 1.0%233U in thorium, respectively, with fluorination-only processing. The addition of processing by reductive extraction resulted in 5125 kg/year for the 0.11%233U case and 5225 kg/year for the 1.0%233U case.

  15. On the use of a molten salt fast reactor to apply an idealized transmutation scenario for the nuclear phase out.

    Directory of Open Access Journals (Sweden)

    Bruno Merk

    Full Text Available 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 research challenges are analyzed. In the second part cycle studies for the MSFR are given for different configurations--a core with U-238 fertile, a fertile free core, and a core with Th-232 as fertile material. For all cases, the transmutation potential is determined and efficient transmutation performance for the case with thorium as a fertile material as well as for the fertile free case is demonstrated and the individual advantages are discussed. The time evolution of different important isotopes is analyzed. In the third part a strategy for the optimization of the transmutation efficiency is developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus on the 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 is achieved by a two step procedure of a first transmuter operation phase followed by a second deep burning phase. There the U-233, which is bred in the blanket of the core consisting of thorium containing salt, is used as feed. It is demonstrated, that transmutation rates up to more than 90% can be achieved for all transuranium isotopes, while the production of undesired high elements like californium is very limited. Additionally, the adaptations needed for the simulation of a MSFR, and the used tool HELIOS 1.10 is described.

  16. Development of a new process for radioactive decontamination of painted carbon steel structures by molten salt stripping

    International Nuclear Information System (INIS)

    Lainetti, Paulo Ernesto de O.

    2009-01-01

    The main practical difficulty associated to the task of the dismantling and decommissioning of the old Nuclear Fuel Cycle facilities of the IPEN has been the large amount of radioactive waste generated in the dismantling operations. The waste is mainly in the form of contaminated carbon steel structures. In the IPEN, the presence of contamination in the equipment, structures and buildings, although restricted to low and average activity levels, constituted an important concern due, on one hand, to the great volume of radioactive wastes generated during the operations. On the other hand, it should be outstanding that the capacity of stockpiling the radioactive wastes in IPEN found been exhausted. Basically, for the dismantling operations of the units, the main radionuclides of interest, from the radioprotection point of view, are U of natural isotopic composition and the thorium-232. Some attempts were done to reduce the volume of those wastes. Nevertheless, the only decontamination available methods were chemical methods such as pickling/rinsing treatments employing acid solutions (with nitric or citric acids) and alkaline solutions (sodium hydroxide). Different concentrations of such solutions were tested. The results obtained in the employed processes were not satisfactory. Ultrasonic equipment available was also employed in an attempt to increase the efficiency of decontamination. The choice of a coating removal process for radioactive material in the form of carbon steel pieces must have into account, among other factors, that it is not necessary a high quality of finishing, since the main objective is the release of the material as iron scrap. This paper describes the development of a new method for surface decontamination by immersion in molten salt baths. (author)

  17. Diagnosis of sources of current inefficiency in industrial molten salt electrolysis cells by Raman spectroscopy: A topical report on chlorides: Topical report, June 1982-June 1987

    Energy Technology Data Exchange (ETDEWEB)

    Sadoway, D. R.

    1987-06-01

    Molten salt electrolysis, a very energy-intensive process, is used in the extraction of light metals. Aluminum production by the Hall process and magnesium production in the Dow and I.G. Farbenindustrie cells constitute the major commercial applications of metal electrowinning from molten-salt media at present. The energy input into the electrolysis cell is in the form of direct current, and the energy efficiencies in the magnesium or aluminum processes are only in the 30 to 40% range. Major energy reductions are achieved by reducing the cell voltage or by increasing the current efficiency. Goal of the research is to identify the sources of the current losses occurring in molten salt electrolysis. This research worked on the systems of I.G. Farben magnesium chloride and Alcoa smelting aluminum chloride processes. Raman spectra were measured and analyzed for each component or their mixtures of the electrolyte for magnesium and aluminum reduction in chloride melts. Raman measurements were also conducted on the melts of industrial composition for aluminum and magnesium electrolysis. In laboratory-scale cells which imitated industrial practice, Raman spectra were measured in situ during electrolysis in attempts to identify the streamers, coloration of electrolyte, and any subvalent species. They were known to occur only during electrolysis, and they have been reported to be possible current losses. Cyclic voltammetry was conducted to obtain information about the generation of subvalent species which were not detected by Raman measurement. These were thought to be kinetic entities present only during electrolysis. Results of Raman spectroscopy and electrochemistry of magnesium and aluminum reduction from molten chloride bath are presented. The results would be useful to establish the basis for the study of electrolysis of aluminum from molten fluoride media. 119 refs., 66 figs.

  18. Methionine sulfoxide reductase A (MsrA) associated with bipolar I disorder and executive functions in A Han Chinese population.

    Science.gov (United States)

    Ni, Peiyan; Ma, Xiaohong; Lin, Yin; Lao, Guohui; Hao, Xiaoyu; Guan, Lijie; Li, Xuan; Jiang, Zeyu; Liu, Yuping; Ye, Biyu; Liu, Xiang; Wang, Yingcheng; Zhao, Liansheng; Cao, Liping; Li, Tao

    2015-09-15

    The oxidative stress hypothesis proposed to explain bipolar I disorder (BD I) pathogenesis has gained growing attention based on its association with cognitive impairment. The aim of the present study was to explore the association of the methionine sulfoxide reductase A (MsrA) gene with BD I as well as executive functions of BD I patients. A total of 44 tagging single-nucleotide polymorphisms within the MsrA gene were selected to analyze gene association with BD I in 375 BD I patients and 475 controls in a Han Chinese population. The association of MsrA haplotypes with executive functions was analyzed in 157 clinically stable BD I patients and 210 controls. Allele frequencies of the rs4840463 polymorphism were significantly different between BD I patients and controls, and between patients with psychotic symptoms and controls. BD I patients performed more poorly in 11 of the 13 neurocognitive measurements compared with controls. Three MsrA haplotypes showed significant associations with different executive functions. The limited sample size requires a cautious conclusion, and further comprehensive approaches are needed to explore the mechanism of MsrA's effect on BD I. The rs4840463 polymorphism in the MsrA gene may be associated with the increased risk of BD I in a Chinese population. The association of MsrA haplotypes with executive functions indicated that MsrA is associated with executive function defects in BD I patients. Copyright © 2015 Elsevier B.V. All rights reserved.

  19. From molten salts to room temperature ionic liquids: simulation studies on chloroaluminate systems.

    Science.gov (United States)

    Salanne, Mathieu; Siqueira, Leonardo J A; Seitsonen, Ari P; Madden, Paul A; Kirchner, Barbara

    2012-01-01

    An interaction potential including chloride anion polarization effects, constructed from first-principles calculations, is used to examine the structure and transport properties of a series of chloroaluminate melts. A particular emphasis was given to the study of the equimolar mixture of aluminium chloride with 1-ethyl-3-methylimidazolium chloride, which forms a room temperature ionic liquid EMI+ -AlCl4-. The structure yielded by the classical simulations performed within the framework of the polarizable ion model is compared to the results obtained from entirely electronic structure-based simulations: An excellent agreement between the two flavors of molecular dynamics is observed. When changing the organic cation EMI+ by an inorganic cation with a smaller ionic radius (Li+, Na+, K+), the chloroaluminate speciation becomes more complex, with the formation of Al2Cl7-, in small amounts. The calculated transport properties (diffusion coefficients, electrical conductivity and viscosity) of EMI+ -AlCl4- are in good agreement with experimental data.

  20. Production of Nd-Fe Foundry Alloy by Electrolysis in Molten Salts

    Directory of Open Access Journals (Sweden)

    Makaseev Yuriy

    2017-01-01

    Full Text Available Nd-Fe-B-based permanent magnets have the highest magnetic characteristics. To improve their quality, one generally applies the technique of solid-phase alloying by materials with high content of neodymium, dysprosium, terbium, i.e. rare-earth metal (REM-Fe(Co foundry alloys. The paper presents the results of the study of the manufacturing process of Nd-Fe foundry alloys, with composition being close to eutectic, by electrolysis of neodymium oxides and fluorides from chloride and fluoride melts. It is shown that in the oxide electrolysis a sufficiently high current efficiency (over 98 % is obtained. This is a promising method of obtaining of rare earth-containing materials.

  1. Demand driven salt clean-up in a molten salt fast reactor - Defining a priority list.

    Science.gov (United States)

    Merk, B; Litskevich, D; Gregg, R; Mount, A R

    2018-01-01

    The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified.

  2. MSR Founders Narrative and Content Analysis of Scholarly Papers

    DEFF Research Database (Denmark)

    Tackney, Charles T.; Chappell, Stacie F.; Sato, Toyoko

    2017-01-01

    the Interest Group. The founders interviewed were identified through preliminary inquiry and from archival sources. As complement and extension, we concurrently conducted a content analysis of the 15 years of MSR Best Papers and Carolyn Dexter Award MSR nominated papers for Academy internationalization...... of spirituality in management practices that infuse MSR with its enduring organizational vitality....

  3. A novel molten-salt electrochemical cell for investigating?the reduction of uranium dioxide to uranium metal by lithium using in situ synchrotron radiation

    OpenAIRE

    Brown, Leon D.; Abdulaziz, Rema; Jervis, Rhodri; Bharath, Vidal; Mason, Thomas J.; Atwood, Robert C.; Reinhard, Christina; Connor, Leigh D.; Inman, Douglas; Brett, Daniel J. L.; Shearing, Paul R.

    2017-01-01

    A novel electrochemical cell has been designed and built to allow for in situ energy-dispersive X-ray diffraction measurements to be made during reduction of UO2 to U metal in LiCl?KCl at 500?C. The electrochemical cell contains a recessed well at the bottom of the cell into which the working electrode sits, reducing the beam path for the X-rays through the molten-salt and maximizing the signal-to-noise ratio from the sample. Lithium metal was electrodeposited onto the UO2 working electrode b...

  4. Identification and evaluation of alternatives for the disposition of fluoride fuel and flush salts from the molten salt reactor experiment at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1996-01-01

    This document presents an initial identification and evaluation of the alternatives for disposition of the fluoride fuel and flush salts stored in the drain tanks at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). It will serve as a resource for the U.S. Department of Energy contractor preparing the feasibility study for this activity under the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA). This document will also facilitate further discussion on the range of credible alternatives, and the relative merits of alternatives, throughout the time that a final alternative is selected under the CERCLA process

  5. Identification and evaluation of alternatives for the disposition of fluoride fuel and flush salts from the molten salt reactor experiment at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-15

    This document presents an initial identification and evaluation of the alternatives for disposition of the fluoride fuel and flush salts stored in the drain tanks at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). It will serve as a resource for the U.S. Department of Energy contractor preparing the feasibility study for this activity under the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA). This document will also facilitate further discussion on the range of credible alternatives, and the relative merits of alternatives, throughout the time that a final alternative is selected under the CERCLA process.

  6. Molten Salt Reactor Experiment Facility (Building 7503) standards/requirements identification document adherence assessment plan at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1996-02-01

    This is the Phase 2 (adherence) assessment plan for the Building 7503 Molten Salt Reactor Experiment (MSRE) Facility standards/requirements identification document (S/RID). This document outlines the activities to be conducted from FY 1996 through FY 1998 to ensure that the standards and requirements identified in the MSRE S/RID are being implemented properly. This plan is required in accordance with the Department of Energy Implementation Plan for Defense Nuclear Facilities Safety Board Recommendation 90-2, November 9, 1994, Attachment 1A. This plan addresses the major aspects of the adherence assessment and will be consistent with Energy Systems procedure QA-2. 7 ''Surveillances.''

  7. High temperature corrosion performance of FeAl intermetallic alloys in molten salts

    Energy Technology Data Exchange (ETDEWEB)

    Amaya, M.; Espinosa-Medina, M.A.; Porcayo-Calderon, J.; Martinez, L.; Gonzalez-Rodriguez, J.G

    2003-05-25

    The corrosion performance of FeAl base intermetallic alloys fabricated by spray-atomization and deposition during their immersion in molten sodium metavanadate (NaVO{sub 3}), 80% (wt.%) sodium pentoxide (V{sub 2}O{sub 5}) +20% sodium sulfate (Na{sub 2}SO{sub 4}) and pure Na{sub 2}SO{sub 4} in the temperature range of 600-1000 deg. C during 200 h was investigated. The experiments were realized by the weight loss method in the intermetallic alloys of composition FeAl40(at.%), FeAl40+0.1B and FeAl40+0.1B+10Al{sub 2}O{sub 3}. In all cases, the FeAl40+0.1B+10Al{sub 2}O{sub 3} alloy showed the best corrosion resistance in the temperatures interval studied here. This behavior was discussed in terms of the formation of a protective Al{sub 2}O{sub 3} layer and its dissolution by vanadate phases and internal sulfidation in the case of experiments carried out in pure Na{sub 2}SO{sub 4}. The morphology of the external layers and the corrosion products formed during the tests revealed that the corrosion rate of this type alloy depends on the corrosion compounds that are formed and the development of protective alumina scales.

  8. High-temperature molten salt thermal energy storage systems for solar applications

    Science.gov (United States)

    Petri, R. J.; Claar, T. D.; Ong, E.

    1983-01-01

    Experimental results of compatibility screening studies of 100 salt/containment/thermal conductivity enhancement (TCE) combinations for the high temperature solar thermal application range of 704 deg to 871 C (1300 to 1600 F) are presented. Nine candidate containment/HX alloy materials and two TCE materials were tested with six candidate solar thermal alkali and alkaline earth carbonate storage salts (both reagent and technical grade of each). Compatibility tests were conducted with salt encapsulated in approx. 6.0 inch x 1 inch welded containers of test material from 300 to 3000 hours. Compatibility evaluations were end application oriented, considering the potential 30 year lifetime requirement of solar thermal power plant components. Analyses were based on depth and nature of salt side corrosion of materials, containment alloy thermal aging effects, weld integrity in salt environment, air side containment oxidation, and chemical and physical analyses of the salt. A need for more reliable, and in some cases first time determined thermophysical and transport property data was also identified for molten carbonates in the 704 to 871 C temperature range. In particular, accurate melting point (mp) measurements were performed for Li2CO3 and Na2CO3 while melting point, heat of fusion, and specific heat determinations were conducted on 81.3 weight percent Na2CO3-18.7 weight percent K2CO3 and 52.2 weight percent BaCO3-47.8 weight percent Na2CO3 to support future TES system design and ultimate scale up of solar thermal energy storage (TES) subsystems.

  9. Molten salt electrolysis device

    International Nuclear Information System (INIS)

    Ota, Kazuaki; Takasawa, Hiroshi

    1998-01-01

    A rotational shaft is disposed vertically downwardly from an upper portion so as to be immersed in a liquid metal in a vessel, impellers for stirring the liquid metal are disposed on the lower portion of the rotational shaft, and a cylindrical body is disposed in the inside of the vessel so as to surround the impellers. When the rotational shaft is rotated, the impellers suck the liquid metal upwardly from the lower portion of the cylindrical body and flow the metal from the upper portion to cause a downwarding stream on the outside of the cylindrical body. As a result, materials reduced and deposited on the upper surface of a liquid metal cathode along with the stream of the liquid metal are precipitated effectively on the lower portion of the vessel. In addition, since a liquid metal with no deposition of uranium is always frown over from the upper portion of the cylindrical body, growth of uranium on the surface of the cathode of liquid metal is prevented, so that uranium, plutonium and transuranium elements can be recovered stably by electrolytic reduction. (N.H.)

  10. A novel molten-salt electrochemical cell for investigating the reduction of uranium dioxide to uranium metal by lithium using in situ synchrotron radiation.

    Science.gov (United States)

    Brown, Leon D; Abdulaziz, Rema; Jervis, Rhodri; Bharath, Vidal; Mason, Thomas J; Atwood, Robert C; Reinhard, Christina; Connor, Leigh D; Inman, Douglas; Brett, Daniel J L; Shearing, Paul R

    2017-03-01

    A novel electrochemical cell has been designed and built to allow for in situ energy-dispersive X-ray diffraction measurements to be made during reduction of UO 2 to U metal in LiCl-KCl at 500°C. The electrochemical cell contains a recessed well at the bottom of the cell into which the working electrode sits, reducing the beam path for the X-rays through the molten-salt and maximizing the signal-to-noise ratio from the sample. Lithium metal was electrodeposited onto the UO 2 working electrode by exposing the working electrode to more negative potentials than the Li deposition potential of the LiCl-KCl eutectic electrolyte. The Li metal acts as a reducing agent for the chemical reduction of UO 2 to U, which appears to proceed to completion. All phases were fitted using Le Bail refinement. The cell is expected to be widely applicable to many studies involving molten-salt systems.

  11. Preparation of tungsten coatings on graphite by electro-deposition via Na2WO4–WO3 molten salt system

    International Nuclear Information System (INIS)

    Sun, Ning-bo; Zhang, Ying-chun; Jiang, Fan; Lang, Shao-ting; Xia, Min

    2014-01-01

    Highlights: • Tungsten coatings on graphite were firstly obtained by electro-deposition method via Na 2 WO 4 –WO 3 molten salt system. • Uniform and dense tungsten coatings could be easily prepared in each face of the sample, especially the complex components. • The obtained tungsten coatings are with high purity, ultra-low oxygen content (about 0.022 wt%). • Modulate pulse parameters can get tungsten coatings with different thickness and hardness. - Abstract: Tungsten coating on graphite substrate is one of the most promising candidate materials as the ITER plasma facing components. In this paper, tungsten coatings on graphite substrates were fabricated by electro-deposition from Na 2 WO 4 –WO 3 molten salt system at 1173 K in atmosphere. Tungsten coatings with no impurities were successfully deposited on graphite substrates under various pulsed current densities in an hour. By increasing the current density from 60 mA cm −2 to 120 mA cm −2 an increase of the average size of tungsten grains, the thickness and the hardness of tungsten coatings occurs. The average size of tungsten grains can reach 7.13 μm, the thickness of tungsten coating was in the range of 28.8–51 μm, and the hardness of coating was higher than 400 HV. No cracks or voids were observed between tungsten coating and graphite substrate. The oxygen content of tungsten coating is about 0.022 wt%

  12. A Tunable Molten-Salt Route for Scalable Synthesis of Ultrathin Amorphous Carbon Nanosheets as High-Performance Anode Materials for Lithium-Ion Batteries.

    Science.gov (United States)

    Wang, Yixian; Tian, Wei; Wang, Luhai; Zhang, Haoran; Liu, Jialiang; Peng, Tingyue; Pan, Lei; Wang, Xiaobo; Wu, Mingbo

    2018-02-14

    Amorphous carbon is regarded as a promising alternative to commercial graphite as the lithium-ion battery anode due to its capability to reversibly store more lithium ions. However, the structural disorder with a large number of defects can lead to low electrical conductivity of the amorphous carbon, thus limiting its application for high power output. Herein, ultrathin amorphous carbon nanosheets were prepared from petroleum asphalt through tuning the carbonization temperature in a molten-salt medium. The amorphous nanostructure with expanded carbon interlayer spacing can provide substantial active sites for lithium storage, while the two-dimensional (2D) morphology can facilitate fast electrical conductivity. As a result, the electrodes deliver a high reversible capacity, outstanding rate capability, and superior cycling performance (579 and 396 mAh g -1 at 2 and 5 A g -1 after 900 cycles). Furthermore, full cells consisting of the carbon anodes coupled with LiMn 2 O 4 cathodes exhibit high specific capacity (608 mAh g -1 at 50 mA g -1 ) and impressive cycling stability with slow capacity loss (0.16% per cycle at 200 mA g -1 ). The present study not only paves the way for industrial-scale synthesis of advanced carbon materials for lithium-ion batteries but also deepens the fundamental understanding of the intrinsic mechanism of the molten-salt method.

  13. Computational Thermodynamic Modeling of Hot Corrosion of Alloys Haynes 242 and HastelloyTM N for Molten Salt Service in Advanced High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    V. Glazoff, Michael; Charit, Indrajt; Sabharwall, Piyush

    2014-09-17

    An evaluation of thermodynamic aspects of hot corrosion of the superalloys Haynes 242 and HastelloyTM N in the eutectic mixtures of KF and ZrF4 is carried out for development of Advanced High Temperature Reactor (AHTR). This work models the behavior of several superalloys, potential candidates for the AHTR, using computational thermodynamics tool (ThermoCalc), leading to the development of thermodynamic description of the molten salt eutectic mixtures, and on that basis, mechanistic prediction of hot corrosion. The results from these studies indicated that the principal mechanism of hot corrosion was associated with chromium leaching for all of the superalloys described above. However, HastelloyTM N displayed the best hot corrosion performance. This was not surprising given it was developed originally to withstand the harsh conditions of molten salt environment. However, the results obtained in this study provided confidence in the employed methods of computational thermodynamics and could be further used for future alloy design efforts. Finally, several potential solutions to mitigate hot corrosion were proposed for further exploration, including coating development and controlled scaling of intermediate compounds in the KF-ZrF4 system.

  14. Fabrication and oxidation resistance of titanium carbide-coated carbon fibres by reacting titanium hydride with carbon fibres in molten salts

    International Nuclear Information System (INIS)

    Dong, Z.J.; Li, X.K.; Yuan, G.M.; Cong, Y.; Li, N.; Jiang, Z.Y.; Hu, Z.J.

    2009-01-01

    Using carbon fibres and titanium hydride as a reactive carbon source and a metal source, respectively, a protective titanium carbide (TiC) coating was formed on carbon fibres in molten salts, composed of LiCl-KCl-KF, at 750-950 o C. The structure and morphology of the TiC coatings were characterised by X-ray diffraction and scanning electron microscopy, respectively. The oxidation resistance of the TiC-coated carbon fibres was measured by thermogravimetric analysis. The results reveal that control of the coating thickness is very important for improvement of the oxidation resistance of TiC-coated carbon fibres. The oxidative weight loss initiation temperature for the TiC-coated carbon fibres increases significantly when an appropriate coating thickness is used. However, thicker coatings lead to a decrease of the carbon fibres' weight loss initiation temperature due to the formation of cracks in the coating. The TiC coating thickness on carbon fibres can be controlled by adjusting the reaction temperature and time of the molten salt synthesis.

  15. Synthesis of hollandite-type LixMnO2 by Li+ ion-exchange in molten salt and lithium insertion characteristics

    International Nuclear Information System (INIS)

    Kadoma, Yoshihiro; Oshitari, Satoru; Ui, Koichi; Kumagai, Naoaki

    2007-01-01

    The Li + ion-exchange reaction of K + -type α-K 0.14 MnO 1.93 .nH 2 O containing different amounts of water molecules (n = 0-0.15) with a large (2 x 2) tunnel structure has been investigated in a LiNO 3 -LiCl molten salt at 300 deg. C. The Li + ion-exchanged products were examined by chemical analysis, X-ray diffraction, and transmission electron microscopy measurements. The K + ions and the hydrogens of the water molecules in the (2 x 2) tunnels of α-MnO 2 were exchanged by Li + ions in the molten salt, resulting in the Li + -type α-MnO 2 containing different amounts of Li + ions and lithium oxide (Li 2 O) in the (2 x 2) tunnels with maintaining the original hollandite structure. The electrochemical properties and structural variation with initial discharge and charge-discharge cycling of the Li + ion-exchanged α-MnO 2 samples have been investigated as insertion compounds in the search for new cathode materials for rechargeable lithium batteries. The Li + ion-exchanged α-MnO 2 samples provided higher capacities and higher Li + ion diffusivity than the parent K + -type materials on initial discharge and charge-discharge cyclings, probably due to the structural stabilization with the existence of Li 2 O in the (2 x 2) tunnels

  16. Molten salt-mediated formation of g-C3N4-MoS2 for visible-light-driven photocatalytic hydrogen evolution

    Science.gov (United States)

    Li, Ni; Zhou, Jing; Sheng, Ziqiong; Xiao, Wei

    2018-02-01

    Construction of two-dimensional/two-dimensional (2D/2D) hybrid with well-defined composition and microstructure is a general protocol to achieve high-performance catalysts. We herein report preparation of g-C3N4-MoS2 hybrid by pyrolysis of affordable melamine and (NH4)2MoS4 in molten LiCl-NaCl-KCl at 550 °C. Molten salts are confirmed as ideal reaction media for formation of homogeneous hybrid. Characterizations suggest a strong interaction between g-C3N4 and MoS2 in the hybrid, which results in an enhanced visible-light-driven photocatalytic hydrogen generation of the hybrid with an optimal g-C3N4/MoS2 ratio. The present study highlights the merits of molten salt methods on preparation of 2D photocatalysts and provides a rational design of 2D/2D hybrid catalysts for advanced environmental and energy applications.

  17. The magnetic and neutron diffraction studies of La.sub.1-x./sub.Sr.sub.x./sub.MnO.sub.3./sub. nanoparticles prepared via molten salt synthesis

    Czech Academy of Sciences Publication Activity Database

    Kačenka, Michal; Kaman, Ondřej; Jirák, Zdeněk; Maryško, Miroslav; Veverka, Pavel; Veverka, Miroslav; Vratislav, S.

    2015-01-01

    Roč. 221, Jan (2015), s. 364-372 ISSN 0022-4596 R&D Projects: GA ČR(CZ) GAP108/11/0807 Institutional support: RVO:68378271 Keywords : magnetic nanoparticles * manganite * molten salt synthesis * neutron diffraction Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 2.265, year: 2015

  18. Density of molten salt Mixtures of eutectic LiCl-KCl containing UCl{sub 3}, CeCl{sub 3}, or LaCl{sub 3}

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, C.; Simpson, M. F. [Dept. of Metallurgical Engineering, University of Utah, Salt Lake City (United States)

    2017-06-15

    Densities of molten salt mixtures of eutectic LiCl-KCl with UCl{sub 3}, CeCl{sub 3}, or LaCl{sub 3} at various concentrations (up to 13 wt%) were measured using a liquid surface displacement probe. Linear relationships between the mixture density and the concentration of the added salt were observed. For LaCl{sub 3} and CeCl{sub 3}, the measured densities were signifcantly higher than those previously reported from Archimedes’ method. In the case of LiCl-KCl-UCl{sub 3}, the data ft the ideal mixture density model very well. For the other salts, the measured densities exceeded the ideal model prediction by about 2%.

  19. A study of integrated cathode assembly for electrolytic reduction of uranium oxide in LiCl-Li2O molten salt

    International Nuclear Information System (INIS)

    Park, Sung Bin; Seo, Jung Seok; Kang, Dae Seung; Kwon, Sun Kil; Park, Seong Won

    2004-01-01

    Interest of electrolytic reduction of uranium oxide is increasing in treatment of spent metal fuels. Argonne National Laboratory (ANL) has reported the experimental results of electrochemical reduction of uranium oxide fuel in bench-scale apparatus with cyclic voltammetry, and has designed high-capacity reduction (HCR) cells and conducted three kg-scale UO 2 reduction runs. From the cyclic voltammograms, the mechanism of electrolytic reduction of metal oxides is analyzed. The uranium oxide in LiCl-Li 2 O is converted to uranium metal according to the two mechanism; direct and indirect electrolytic reduction. In this study, cyclic voltammograms for LiCl-3wt% Li 2 O system and U 3 O 8 -LiCl-3wt% Li 2 O system using the 325-mesh stainless steel screen in cathode assembly have been obtained. Direct electrolytic reduction of uranium oxide in LiCl-3wt% Li 2 O molten salt has been conducted

  20. Addendum to Engineering Evaluation of Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiment for the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Wilson, Guy

    2009-01-01

    The purpose of this addendum is to graphically publish data which indicate moisture in leakage and corrosion may have occurred during heating of the tanks at the Molten Salt Reactor Experiment (MSRE) for and during hydrofluorination, fluorination and transfer of uranium. Corrosion, especially by hydrofluoric acid, is not expected to occur uniformly over the tank and piping inner surfaces and therefore is not easily measured by nondestructive techniques that can measure only limited areas. The rate of corrosion exponentially escalates with both temperature and moisture. The temperature, pressure, and concentration data in this addendum indicate periods when elevated corrosion rates were likely to have been experienced. This data was not available in time to be considered as part of the evaluation that was the focus of the report. Pressure and temperature data were acquired via the LabView(trademark) Software, while concentration data was acquired from the Fourier Transform InfraRed (FTIR) system

  1. On the Burning of Plutonium Originating from Light Water Reactor Use in a Fast Molten Salt Reactor—A Neutron Physical Study

    Directory of Open Access Journals (Sweden)

    Bruno Merk

    2015-11-01

    Full Text Available An efficient burning of the plutonium produced during light water reactor (LWR operation has the potential to significantly improve the sustainability indices of LWR operations. The work offers a comparison of the efficiency of Pu burning in different reactor configurations—a molten salt fast reactor, a LWR with mixed oxide (MOX fuel, and a sodium cooled fast reactor. The calculations are performed using the HELIOS 2 code. All results are evaluated against the plutonium burning efficiency determined in the Consommation Accrue de Plutonium dans les Réacteurs à Neutrons RApides (CAPRA project. The results are discussed with special view on the increased sustainability of LWR use in the case of successful avoidance of an accumulation of Pu which otherwise would have to be forwarded to a final disposal. A strategic discussion is given about the unavoidable plutonium production, the possibility to burn the plutonium to avoid a burden for the future generations which would have to be controlled.

  2. Results of molten salt panel and component experiments for solar central receivers: Cold fill, freeze/thaw, thermal cycling and shock, and instrumentation tests

    Energy Technology Data Exchange (ETDEWEB)

    Pacheco, J.E.; Ralph, M.E.; Chavez, J.M.; Dunkin, S.R.; Rush, E.E.; Ghanbari, C.M.; Matthews, M.W.

    1995-01-01

    Experiments have been conducted with a molten salt loop at Sandia National Laboratories in Albuquerque, NM to resolve issues associated with the operation of the 10MW{sub e} Solar Two Central Receiver Power Plant located near Barstow, CA. The salt loop contained two receiver panels, components such as flanges and a check valve, vortex shedding and ultrasonic flow meters, and an impedance pressure transducer. Tests were conducted on procedures for filling and thawing a panel, and assessing components and instrumentation in a molten salt environment. Four categories of experiments were conducted: (1) cold filling procedures, (2) freeze/thaw procedures, (3) component tests, and (4) instrumentation tests. Cold-panel and -piping fill experiments are described, in which the panels and piping were preheated to temperatures below the salt freezing point prior to initiating flow, to determine the feasibility of cold filling the receiver and piping. The transient thermal response was measured, and heat transfer coefficients and transient stresses were calculated from the data. Freeze/thaw experiments were conducted with the panels, in which the salt was intentionally allowed to freeze in the receiver tubes, then thawed with heliostat beams. Slow thermal cycling tests were conducted to measure both how well various designs of flanges (e.g., tapered flanges or clamp type flanges) hold a seal under thermal conditions typical of nightly shut down, and the practicality of using these flanges on high maintenance components. In addition, the flanges were thermally shocked to simulate cold starting the system. Instrumentation such as vortex shedding and ultrasonic flow meters were tested alongside each other, and compared with flow measurements from calibration tanks in the flow loop.

  3. Development of electrolytic process in molten salt media for light rare-earth metals production. The metallic cerium electrodeposition; Desenvolvimento de processo de eletrolise em meio de sais fundidos para a producao de metais de terras-raras leves. A obtencao do cerio metalico

    Energy Technology Data Exchange (ETDEWEB)

    Restivo, T.A.G.

    1994-12-31

    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 800{sup C} 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.

  4. [download] (941Data Analysis with the Morse-Smale Complex: The msr Package for R

    Directory of Open Access Journals (Sweden)

    Samuel Gerber

    2012-07-01

    Full Text Available In many areas, scientists deal with increasingly high-dimensional data sets. An important aspect for these scientists is to gain a qualitative understanding of the process or system from which the data is gathered. Often, both input variables and an outcome are observed and the data can be characterized as a sample from a high-dimensional scalar function. This work presents the R package msr for exploratory data analysis of multivariate scalar functions based on the Morse-Smale complex. The Morse-Smale complex provides a topologically meaningful decomposition of the domain. The msr package implements a discrete approximation of the Morse-Smale complex for data sets. In previous work this approximation has been exploited for visualization and partition-based regression, which are both supported in the msr package. The visualization combines the Morse-Smale complex with dimension-reduction techniques for a visual summary representation that serves as a guide for interactive exploration of the high-dimensional function. In a similar fashion, the regression employs a combination of linear models based on the Morse-Smale decomposition of the domain. This regression approach yields topologically accurate estimates and facilitates interpretation of general trends and statistical comparisons between partitions. In this manner, the msr package supports high-dimensional data understanding and exploration through the Morse-Smale complex.

  5. Data Analysis with the Morse-Smale Complex: The msr Package for R

    KAUST Repository

    Gerber, Samuel

    2012-01-01

    In many areas, scientists deal with increasingly high-dimensional data sets. An important aspect for these scientists is to gain a qualitative understanding of the process or system from which the data is gathered. Often, both input variables and an outcome are observed and the data can be characterized as a sample from a high-dimensional scalar function. This work presents the R package msr for exploratory data analysis of multivariate scalar functions based on the Morse-Smale complex. The Morse-Smale complex provides a topologically meaningful decomposition of the domain. The msr package implements a discrete approximation of the Morse-Smale complex for data sets. In previous work this approximation has been exploited for visualization and partition-based regression, which are both supported in the msr package. The visualization combines the Morse-Smale complex with dimension-reduction techniques for a visual summary representation that serves as a guide for interactive exploration of the high-dimensional function. In a similar fashion, the regression employs a combination of linear models based on the Morse-Smale decomposition of the domain. This regression approach yields topologically accurate estimates and facilitates interpretation of general trends and statistical comparisons between partitions. In this manner, the msr package supports high-dimensional data understanding and exploration through the Morse-Smale complex.

  6. Stabilization/solidification of radioactive salt waste by using xSiO2-yAl2O3-zP2O5 (SAP) material at molten salt state.

    Science.gov (United States)

    Park, Hwan-Seo; Kim, In-Tae; Cho, Yong-Zun; Eun, Hee-Chul; Lee, Han-Soo

    2008-12-15

    The molten salt waste from the pyroprocess is one of the problematic wastes to directly apply a conventional process such as vitrification or ceramization. This study suggested a novel method using a reactive material for metal chlorides at a molten temperature of salt waste, and then converting them into manageable product at a high temperature. The inorganic composite, SAP (SiO2-Al2O3-P2O5), synthesized by a conventional sol-gel process has three or four distinctive domains that are bonded sequentially, Si-O-Si-O-A-O-P-O-P. The P-rich phase in the SAP composite is unstable for producing a series of reactive sites when in contact with a molten LiCl salt. After the reaction, metal aluminosilicate, metal aluminophosphate, metal phosphates and gaseous chlorines are generated. From this process, the volatile salt waste is stabilized and it is possible to apply a high temperature process. The reaction products were fabricated successfully by using a borosilicate glass with an arbitrary composition as a chemical binder. There was a low possibility for the valorization of radionuclides up to 1200 degrees C, based on the result of the thermo gravimetric analysis. The Cs and Sr leach rates by the PCT-A method were about 1 x 10(-3) g/(m2 day). For the final disposal of the problematic salt waste, this approach suggested the design concept of an effective stabilizer for metal chlorides and revealed the chemical route to the fabrication of monolithic wasteform by using a composite as an example. Using this method, we could obtain a higher disposal efficiency and lower waste volume, compared with the present immobilization methods.

  7. Electrochemistry of vanadium(II and the electrodeposition of aluminum-vanadium alloys in the aluminum chloride-1-ethyl-3-methylimidazolium chloride molten salt

    Directory of Open Access Journals (Sweden)

    Tsuda T.

    2003-01-01

    Full Text Available The electrochemical behavior of vanadium(II was examined in the 66.7-33.3 mole percent aluminum chloride-1-ethyl-3-methylimidazolium chloride molten salt containing dissolved VCl2 at 353 K. Voltammetry experiments revealed that V(II could be electrochemically oxidized to V(III and V(IV. However at slow scan rates the V(II/V(III electrode reaction is complicated by the rapid precipitation of V(III as VCl3. The reduction of V(II occurs at potentials considerably negative of the Al(III/Al electrode reaction, and Al-V alloys cannot be electrodeposited from this melt. However electrodeposition experiments conducted in VCl2-saturated melt containing the additive, 1-ethyl-3-methylimidazolium tetrafluoroborate, resulted in Al-V alloys. The vanadium content of these alloys increased with increasing cathodic current density or more negative applied potentials. X-ray analysis of Al-V alloys that were electrodeposited on a rotating copper wire substrate indicated that these alloys did not form or contain an intermetallic compound, but were non-equilibrium or metastable solid solutions. The chloride-pitting corrosion properties of these alloys were examined in aqueous NaCl by using potentiodynamic polarization techniques. Alloys containing ~10 a/o vanadium exhibited a pitting potential that was 0.3 V positive of that for pure aluminum.

  8. Prototype Tests for the Recovery and Conversion of UF6 Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

    International Nuclear Information System (INIS)

    Del Cul, G.D.; Icenhour, A.S.; Simmons, D.W.

    2000-01-01

    The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of -11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form (uranium oxide), which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design

  9. Prototype Tests for the Recovery and Conversion of UF6Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

    Energy Technology Data Exchange (ETDEWEB)

    Del Cul, G.D.

    2000-06-07

    The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of {approx}11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form [uranium oxide (U{sub 3}O{sub 8})], which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design.

  10. Prototype Tests for the Recovery and Conversion of UF6 Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

    Energy Technology Data Exchange (ETDEWEB)

    Del Cul, G.D.; Icenhour, A.S.; Simmons, D.W.

    2000-04-01

    The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of -11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form [uranium oxide], which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design.

  11. Strong magnetoelectric coupling in CoFe2O4-BaTiO3 composites prepared by molten-salt synthesis method

    International Nuclear Information System (INIS)

    Nie Junwu; Xu Guoyue; Yang Ying; Cheng Chuanwei

    2009-01-01

    Magnetoelectric nano-composites (1 - x)CoFe 2 O 4 + (x)BaTiO 3 with x varies as 0, 0.5, 0.65 and 1.0 in molar ratio were prepared by molten-salt synthesis method. The structural analysis carried out by X-ray diffraction (XRD) technique has confirmed that both phases are present in all the nano-composites powders and ceramic composites. The TEM images show that the nano-particle crystallite size is about 50-80 nm, which is consistent to the result calculated by XRD. The dielectric constant was studied as a function of frequency for ceramic composites sintered by using those nano-composite powders. The saturation magnetization (Ms) and remnant polarization (Pr) were calculated from the magnetic hysteresis loop and electric hysteresis loop, respectively. And a large ME coefficient of about 17.04 mV cm -1 Oe -1 was observed for 0.5CoFe 2 O 4 + 0.5BaTiO 3 ME composite under the ac superimposed magnetic signal with 20 kHz frequency by using the lock-in technique

  12. Environmental 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)

    Burman, S.N.; Tiner, P.F.; Gosslee, R.C.

    1998-01-01

    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 this 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 environmental protection and 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

  13. Environmental 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.; Tiner, P.F.; Gosslee, R.C.

    1998-01-01

    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 this 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 environmental protection and 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.

  14. Synthesis of Bi{sub 25}FeO{sub 39} by molten salts method and its Mössbauer spectrum

    Energy Technology Data Exchange (ETDEWEB)

    Flores Morales, Shirley Saraí; León Flores, Jesús A.; Pérez Mazariego, José Luis, E-mail: mazariego@ciencias.unam.mx; Marquina Fábrega, Vivianne; Gómez González, Raúl W.

    2017-01-01

    A sample of Bi{sub 25}FeO{sub 39} was synthesized by the molten salts method in 2 h at 750 °C with KCl-KBr as reaction media. The sample was characterized with X-ray powder diffraction and Mössbauer spectroscopy. The Rietveld refinement of the X-ray diffraction confirms that a sillenite type structure with a cell parameter a=10.190(4) Å was achieved, with only small traces of Bi{sub 2}O{sub 3}. In order to get a Mössbauer spectrum with a good signal to noise ratio, the synthesis was made using 95% enriched α–{sup 57}Fe{sub 2}O{sub 3}. To our knowledge, this is the first time that a Mössbauer spectrum for the Bi{sub 25}FeO{sub 39} has been recorded. The nonappearance of a magnetic sextet confirms its paramagnetic behavior. The measured Mössbauer parameters (IS=0.20±0.004 mm/s and QS=0.2±0.07 mm/s) correspond to Fe{sup +3} in very symmetric surroundings, and rule out some interpretations given to the doublet observed in some Mössbauer spectra of BiFeO{sub 3}.

  15. Study of the Electrolytic Reduction of Uranium Oxide in LiCl-Li2O Molten Salts with an Integrated Cathode Assembly

    International Nuclear Information System (INIS)

    Park, Sung Bin; Seo, Chung Seok; Kang, Dae Seung; Kwon, Seon Gil; Park, Seong Won

    2005-01-01

    The electrolytic reduction of uranium oxide in a LiCl-Li 2 O molten salt system has been studied in a 10 g U 3 O 3 /batch-scale experimental apparatus with an integrated cathode assembly at 650 .deg. C. The integrated cathode assembly consists of an electric conductor, the uranium oxide to be reduced and the membrane for loading the uranium oxide. From the cyclic voltammograms for the LiCl-3 wt% Li 3 O system and the U 3 O 3 -LiCl-3 wt% Li 2 O system according to the materials of the membrane in the cathode assembly, the mechanisms of the predominant reduction reactions in the electrolytic reactor cell were to be understood; direct and indirect electrolytic reduction of uranium oxide. Direct and indirect electrolytic reductions have been performed with the integrated cathode assembly. Using the 325-mesh stainless steel screen the uranium oxide failed to be reduced to uranium metal by a direct and indirect electrolytic reduction because of a low current efficiency and with the porous magnesia membrane the uranium oxide was reduced successfully to uranium metal by an indirect electrolytic reduction because of a high current efficiency.

  16. Interplay between structure and transport properties of molten salt mixtures of ZnCl2-NaCl-KCl: A molecular dynamics study.

    Science.gov (United States)

    Manga, Venkateswara Rao; Swinteck, Nichlas; Bringuier, Stefan; Lucas, Pierre; Deymier, Pierre; Muralidharan, Krishna

    2016-03-07

    Molten mixtures of network-forming covalently bonded ZnCl2 and network-modifying ionically bonded NaCl and KCl salts are investigated as high-temperature heat transfer fluids for concentrating solar power plants. Specifically, using molecular dynamics simulations, the interplay between the extent of the network structure, composition, and the transport properties (viscosity, thermal conductivity, and diffusion) of ZnCl2-NaCl-KCl molten salts is characterized. The Stokes-Einstein/Eyring relationship is found to break down in these network-forming liquids at high concentrations of ZnCl2 (>63 mol. %), while the Eyring relationship is seen with increasing KCl concentration. Further, the network modification due to the addition of K ions leads to formation of non-bridging terminal Cl ions, which in turn lead to a positive temperature dependence of thermal conductivity in these melts. This new understanding of transport in these ternary liquids enables the identification of appropriate concentrations of the network formers and network modifiers to design heat transfer fluids with desired transport properties for concentrating solar power plants.

  17. Underpotential and overpotential deposition of Al onto Cu(111 from the AlCl3–EtMeImCl room temperature molten salt

    Directory of Open Access Journals (Sweden)

    V. D. JOVIC

    2006-04-01

    Full Text Available The processes of underpotential (UPD and overpotential (OPD deposition ofAl onto Cu(111, from the room temperature molten salt AlCl3–EtMeImCl of different compositions, has been investigated by the cyclic volatmmetry (CV and potentiostatic pulse techniques. It was shown that the CVs of the UPD are characterized by two sharp peaks, while the potentiostatic cathodic and anodic j–t-transients of this process are characterized by two waves, indicating that the UPD of Al results in the formation of two structures. The first, less dense one,most probably the (sqrt3 x sqrt3 R30 º ordered structure of Al, is formed at a more positive potential of about 200 mV vs. Al, while the second one, a complete momolayer of Al, is formed at about 20 mV vs. Al, just before the reversible potential of Al in these melts (–20mV vs. Al. The OPD of Al was detected at potentials more negative than –30 mV vs. Al, occurring through the progressive 3D nucleation and growth mechanism. Slow surface alloying of Al with Cu was found to occur at a potential close to the reversible potential of Al.

  18. X-ray absorption spectroscopic and magnetic characterization of cobalt-doped zinc oxide nanocrystals prepared by the molten-salt method

    International Nuclear Information System (INIS)

    Han, Ah Reum; Hwang, Seong-Ju; Zhao, Yongnan; Kwon, Young-Uk

    2008-01-01

    The local atomic arrangement and electronic structure of the Co-doped Zn 1-x Co x O nanocrystal have been quantitatively examined along with its magnetic properties. According to our analysis using powder X-ray diffraction, electron microscopy, and Zn K-edge X-ray absorption spectroscopy (XAS), phase-pure wurzite-structured Zn 1-x Co x O nanocrystals have been successfully synthesized via the molten-salt method. The Co K-edge XAS analysis clearly demonstrates that all the Co 2+ ions are substituted for the tetrahedral Zn sites of the Wurzite structure with a coordination number of 3.9 and a bond distance of 1.97 A, ruling out the presence of magnetic impurity phase and Co-metal cluster. Magnetization measurements reveal that the present Zn 1-x Co x O sample does not show any ferromagnetic transition down to 2 K. In this regard, we can conclude that Co-doped zinc oxide is not ferromagnetic but the previously reported ferromagnetism in this phase would be an extrinsic property

  19. Corrosion Behavior of Alloy 625 in PbSO4-Pb3O4-PbCl2-ZnO-10 Wt Pct CdO Molten Salt Medium

    Science.gov (United States)

    Mohammadi Zahrani, E.; Alfantazi, A. M.

    2012-08-01

    Corrosion behavior and degradation mechanisms of alloy 625 under a 47.288 PbSO4-12.776 Pb3O4-6.844PbCl2-23.108ZnO-10CdO (wt pct) molten salt mixture under air atmosphere were studied at 873 K, 973 K, and 1073 K (600 °C, 700 °C, and 800 °C). Electrochemical impedance spectroscopy (EIS), open circuit potential (OCP) measurements, and potentiodynamic polarization techniques were used to evaluate the degradation mechanisms and characterize the corrosion behavior of the alloy. Morphology, chemical composition, and phase structure of the corrosion products and surface layers of the corroded specimens were studied by scanning electron microscopy/energy-dispersive X-ray (SEM/EDX) and X-ray map analyses. Results confirmed that during the exposure of alloy 625 to the molten salt, chromium was mainly dissolved through an active oxidation process as CrO3, Cr2O3, and CrNbO4, while nickel dissolved only as NiO in the system. Formation of a porous and nonprotective oxide layer with low resistance is responsible for the weak protective properties of the barrier layer at high temperatures of 973 K and 1073 K (700 °C and 800 °C). There were two kinds of attack for INCONEL 625, including general surface corrosion and pitting. Pitting corrosion occurred due to the breakdown of the initial oxide layer by molten salt dissolution of the oxide or oxide cracking.

  20. Demand driven salt clean-up in a molten salt fast reactor – Defining a priority list

    Science.gov (United States)

    Litskevich, D.; Gregg, R.; Mount, A. R.

    2018-01-01

    The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified. PMID:29494604

  1. Converting hcp Mg-Al-Zn alloy into bcc Mg-Li-Al-Zn alloy by electrolytic deposition and diffusion of reduced lithium atoms in a molten salt electrolyte LiCl-KCl

    International Nuclear Information System (INIS)

    Lin, M.C.; Tsai, C.Y.; Uan, J.Y.

    2007-01-01

    A body-centered cubic (bcc) Mg-12Li-9Al-1Zn (wt.%) alloy was fabricated in air by electrolysis from LiCl-KCl molten salt at 500 deg. C. Electrolytic deposition of Li atoms on cathode (Mg-Al-Zn alloy) and diffusion of the Li atoms formed the bcc Mg-Li-Al-Zn alloy with 12 wt.% Li and only 0.264 wt.% K. Low K concentration in the bcc Mg alloy strip after the electrolysis process resulted from 47% atomic size misfit between K and Mg atoms and low solubility of K in Mg matrix

  2. SiC Schottky Diode Detectors for Measurement of Actinide Concentrations from Alpha Activities in Molten Salt Electrolyte

    Energy Technology Data Exchange (ETDEWEB)

    Windl, Wolfgang [The Ohio State Univ., Columbus, OH (United States); Blue, Thomas [The Ohio State Univ., Columbus, OH (United States)

    2013-01-28

    In this project, we have designed a 4H-SiC Schottky diode detector device in order to monitor actinide concentrations in extreme environments, such as present in pyroprocessing of spent fuel. For the first time, we have demonstrated high temperature operation of such a device up to 500 °C in successfully detecting alpha particles. We have used Am-241 as an alpha source for our laboratory experiments. Along with the experiments, we have developed a multiscale model to study the phenomena controlling the device behavior and to be able to predict the device performance. Our multiscale model consists of ab initio modeling to understand defect energetics and their effect on electronic structure and carrier mobility in the material. Further, we have developed the basis for a damage evolution model incorporating the outputs from ab initio model in order to predict respective defect concentrations in the device material. Finally, a fully equipped TCAD-based device model has been developed to study the phenomena controlling the device behavior. Using this model, we have proven our concept that the detector is capable of performing alpha detection in a salt bath with the mixtures of actinides present in a pyroprocessing environment.

  3. Separation of rare earths from transition metals by liquid-liquid extraction from a molten salt hydrate to an ionic liquid phase.

    Science.gov (United States)

    Rout, Alok; Binnemans, Koen

    2014-02-28

    The solvent extraction of trivalent rare-earth ions and their separation from divalent transition metal ions using molten salt hydrates as the feed phase and an undiluted fluorine-free ionic liquid as the extracting phase were investigated in detail. The extractant was tricaprylmethylammonium nitrate, [A336][NO3], and the hydrated melt was calcium nitrate tetrahydrate, Ca(NO3)2·4H2O. The extraction behavior of rare-earth ions was studied for solutions of individual elements, as well as for mixtures of rare earths in the hydrated melt. The influence of different extraction parameters was investigated: the initial metal loading in the feed phase, percentage of water in the feed solution, equilibration time, and the type of hydrated melt. The extraction of rare earths from Ca(NO3)2·4H2O was compared with extraction from CaCl2·4H2O by [A336][Cl] (Aliquat 336). The nitrate system was found to be the better one. The extraction and separation of rare earths from the transition metals nickel, cobalt and zinc were also investigated. Remarkably high separation factors of rare-earth ions over transition metal ions were observed for extraction from Ca(NO3)2·4H2O by the [A336][NO3] extracting phase. Furthermore, rare-earth ions could be separated efficiently from transition metal ions, even in melts with very high concentrations of transition metal ions. Rare-earth oxides could be directly dissolved in the Ca(NO3)2·4H2O phase in the presence of small amounts of Al(NO3)3·9H2O or concentrated nitric acid. The efficiency of extraction after dissolving the rare-earth oxides in the hydrated nitrate melt was identical to extraction from solutions with rare-earth nitrates dissolved in the molten phase. The stripping of the rare-earth ions from the loaded ionic liquid phase and the reuse of the recycled ionic liquid were also investigated in detail.

  4. Functional Predominance of msr(D), Which Is More Effective as mef(A)-Associated Than mef(E)-Associated, Over mef(A)/mef(E) in Macrolide Resistance in Streptococcus pyogenes.

    Science.gov (United States)

    Tatsuno, Ichiro; Isaka, Masanori; Masuno, Katsuaki; Hata, Nanako; Matsumoto, Masakado; Hasegawa, Tadao

    2018-02-06

    Although mef(A) and its subclass mef(E) genes have long been considered to play a central role in macrolide efflux-based resistance, we have previously demonstrated that the msr(D) gene located immediately downstream of the mef(A) gene plays a predominant role in Streptococcus pyogenes macrolide resistance. The mef(A) and mef(E) genes are carried by different genetic elements and the resistance associated with mef(A) was reported to be higher than that associated with mef(E); therefore, we further investigated the functional relevance of mef(A)/mef(E) and its associated msr(D). We established additional mef(A)-, mef(E)-, and their associated msr(D)-knockout strains and confirmed the predominance of msr(D) over mef(A)/mef(E). In addition, we performed experiments introducing mef(A), mef(E), and their associated msr(D) genes into mef(A)/mef(E)-msr(D) double-knockout and mef(A)/mef(E) negative strains. Neither mef(A) nor mef(E) genes had effects on erythromycin resistance. However, both associated msr(D) showed significant effects, and the mef(A)-associated msr(D) exhibited more effect than the mef(E)-associated one. These results suggest that an overall functional predominance of msr(D) over mef(A)/mef(E) is conceivable in efflux-based macrolide resistance in at least some S. pyogenes strains. Furthermore, the higher resistance of mef(A) system over mef(E) system could be derived at least in part from functional differences of mef(A)- and mef(E)-associated msr(D).

  5. Design of MSR primary circuit with minimum pressure losses

    Directory of Open Access Journals (Sweden)

    Noga Tomáš

    2017-01-01

    Full Text Available This article describes a design of a MSR primary circuit with minimum pressure losses. It includes a brief description of this type of a reactor and its integral layout, properties, purpose, etc. The objective of this paper is to define problems of pressure losses calculation and to design a proper device for a primary circuit of MSR reactor, including its basic dimensions. Thanks to this, it can become an initial project for a construction of a real piece of work. This is the main contribution of the carried out study. Of course, this article is not a detailed solution, but it points out facts and problems, which future designers may have to face. The further step of our work will be a reconstruction of the current experiment for a two-stage flowing.

  6. Development of Pyrochemical Reprocessing of the Spent Nuclear Fuel and Prospects of Closed Fuel Cycle

    OpenAIRE

    Tulackova, R; Chuchvalcova-Bimova, K; Precek, M; Marecek, M; Uhlir, J

    2007-01-01

    Molten-Salt Reactor (MSR) is a design of an advanced reactor system from the GEN IV family working in thermal or epithermal neutron spectrum and using thorium or transuranium fuel in the form of molten fluorides. It is based on the experience with the development of the molten-salt reactor technology in the Oak-Ridge National Laboratory in the United States. The MSR fuel cycle with integrated reprocessing represents one of the potential ways both for significant decrease of total amount of ra...

  7. Molten salt extraction (MSE) of americium from plutonium metal in CaCl2-KCl-PuCl3 and CaCl2-PuCl3 salt systems

    International Nuclear Information System (INIS)

    Dodson, K.E.

    1992-01-01

    Molten salt extraction (MSE) of americium-241 from reactor-grade plutonium has been developed using plutonium trichloride salt in stationary furnaces. Batch runs with oxidized and oxide-free metal have been conducted at temperature ranges between 750 and 945C, and plutonium trichloride concentrations from one to one hundred mole percent. Salt-to-metal ratios of 0.10, 0.15, and 0 30 were examined. The solvent salt was either eutectic 74 mole percent CaCl 2 endash 26 mole percent KCl or pure CaCl 2 . Evidence of trivalent product americium, and effects of temperature, salt-to-metal ratio, and oxide contamination on the americium extraction efficiency are given. 24 refs, 20 figs, 13 tabs

  8. Quality assurance plan for the Molten Salt Reactor Experiment Remediation Project at the Oak Ridge National Laboratory. Phase 1 -- Interim corrective measures and Phase 2 -- Purge and trap reactive gases

    International Nuclear Information System (INIS)

    1995-11-01

    This Quality Assurance Plan (QAP) identifies and describes the systems utilized by the Molten Salt Reactor Experiment Remediation Project (MSRERP) personnel to implement the requirements and associated applicable guidance contained in the Quality Program Description Y/QD-15 Rev. 2 (Energy Systems 1995f). This QAP defines the quality assurance (QA) requirements applicable to all activities and operations in and directly pertinent to the MSRERP Phase 1--Interim Corrective Measures and Phase 2--Purge and Trap objectives. This QAP will be reviewed, revised, and approved as necessary for Phase 3 and Phase 4 activities. 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

  9. Evaluation of chloride-ion-specific electrodes as in situ chemical sensors for monitoring total chloride concentration in aqueous solutions generated during the recovery of plutonium from molten salts used in plutonium electrorefining operations

    International Nuclear Information System (INIS)

    Smith, W.H.

    1992-10-01

    Two commercially available chloride-ion-specific electrodes (CLISEs), a solid-state type and a membrane type, were evaluated as potential in situ chemical sensors for determining total chloride ion concentration in mixed sodium chloride/potassium chloride/hydrochloric acid solutions generated during the recovery of plutonium from molten salts used in plutonium electrorefining operations. Because the response of the solid-state CLISE was closer than was the response of the membrane-type CLISE to the theoretical response predicted by the Nernst equation, the solid-state CLISE was selected for further evaluation. A detailed investigation of the characteristics of the chloride system and the corresponding CLISE response to concentration changes suggested four methods by which the CLISE could be used either as a direct, in situ sensor or as an indirect sensor through which an analysis could be performed on-line with a sample extracted from the process solution

  10. Synthesis of hollandite-type Li yMn 1- xCo xO 2 (x = 0-0.15) by Li + ion-exchange in molten salt and the electrochemical property for rechargeable lithium battery electrodes

    Science.gov (United States)

    Kumagai, Naoaki; Oshitari, Satoru; Komaba, Shinichi; Kadoma, Yoshihiro

    The Li + ion-exchange reaction of K +-type α-K 0.14MnO 1.93·0.18H 2O and its Co-doped α-K 0.14(Mn 0.85Co 0.15)O 1.96·0.21H 2O with a large (2 × 2) tunnel structure has been investigated in a LiNO 3/LiCl molten salt at 300 °C. The Li + ion-exchanged products were examined by chemical analysis, X-ray diffraction, and scanning and transmission electron microscopic measurements. Almost all the K + ions and the hydrogens of water molecules in the (2 × 2) tunnel of α-MnO 2 and its Co-doped one were exchanged by Li + ions in the molten salt, resulting in Li +-type α-MnO 2 and its Co-doped one containing Li + ions as well as Li 2O (lithium oxide) in the (2 × 2) tunnel with maintaining the original hollandite structure. The electrochemical properties including charge-discharge cycling of the Li + ion-exchanged α-MnO 2 and its Co-doped samples have been investigated as insertion compounds in the search for new cathode materials for rechargeable lithium batteries. The Li + ion-exchanged α-MnO 2 and its Co-doped samples provided higher capacities than the K +-type parent materials on initial discharge and charge-discharge cyclings, probably due to the structural stabilization with the existence of Li 2O in the (2 × 2) tunnels.

  11. MSR Founding Narratives and Content Analysis of Best/Dexter Award Nominee Papers (2001-2015)

    DEFF Research Database (Denmark)

    Tackney, Charles T.; Chappell, Stacie; Sato, Toyoko

    International Award nominated papers. Taken together, our study is an important benchmark of founding and institutionalization for MSR members and potential members, while serving as an indicator of emerging trends in the research domain. By tracing the particular research paths that MSR has taken in light...... of its founding aspirations, we illuminate the distinctive values, tensions, management practices, and skills that give MSR its organizational vitality....

  12. High-resolution wave number spectrum using multi-point measurements in space – the Multi-point Signal Resonator (MSR technique

    Directory of Open Access Journals (Sweden)

    Y. Narita

    2011-02-01

    Full Text Available A new analysis method is presented that provides a high-resolution power spectrum in a broad wave number domain based on multi-point measurements. The analysis technique is referred to as the Multi-point Signal Resonator (MSR and it benefits from Capon's minimum variance method for obtaining the proper power spectral density of the signal as well as the MUSIC algorithm (Multiple Signal Classification for considerably reducing the noise part in the spectrum. The mathematical foundation of the analysis method is presented and it is applied to synthetic data as well as Cluster observations of the interplanetary magnetic field. Using the MSR technique for Cluster data we find a wave in the solar wind propagating parallel to the mean magnetic field with relatively small amplitude, which is not identified by the Capon spectrum. The Cluster data analysis shows the potential of the MSR technique for studying waves and turbulence using multi-point measurements.

  13. High-resolution wave number spectrum using multi-point measurements in space - the Multi-point Signal Resonator (MSR) technique

    Science.gov (United States)

    Narita, Y.; Glassmeier, K.-H.; Motschmann, U.

    2011-02-01

    A new analysis method is presented that provides a high-resolution power spectrum in a broad wave number domain based on multi-point measurements. The analysis technique is referred to as the Multi-point Signal Resonator (MSR) and it benefits from Capon's minimum variance method for obtaining the proper power spectral density of the signal as well as the MUSIC algorithm (Multiple Signal Classification) for considerably reducing the noise part in the spectrum. The mathematical foundation of the analysis method is presented and it is applied to synthetic data as well as Cluster observations of the interplanetary magnetic field. Using the MSR technique for Cluster data we find a wave in the solar wind propagating parallel to the mean magnetic field with relatively small amplitude, which is not identified by the Capon spectrum. The Cluster data analysis shows the potential of the MSR technique for studying waves and turbulence using multi-point measurements.

  14. Parametric systems analysis of the Modular Stellarator Reactor (MSR)

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.

    1982-05-01

    The close coupling in the stellarator/torsatron/heliotron (S/T/H) between coil design (peak field, current density, forces), magnetics topology (transform, shear, well depth), and plasma performance (equilibrium, stability, transport, beta) complicates the reactor assessment more so than for most magnetic confinement systems. In order to provide an additional degree of resolution of this problem for the Modular Stellarator Reactor (MSR), a parametric systems model has been developed and applied. This model reduces key issues associted ith plasma performance, first-wall/blanket/shield (FW/B/S), and coil design to a simple relationship between beta, system geometry, and a number of indicators of overall plant performance. The results of this analysis can then be used to guide more detailed, multidimensional plasma, magnetics, and coil design efforts towards technically and economically viable operating regimes. In general, it is shown that beta values > 0.08 may be needed if the MSR approach is to be substantially competitive with other approaches to magnetic fusion in terms of system power density, mass utilization, and cost for total power output around 4.0 GWt; lower powers will require even higher betas

  15. Lack of a functional methionine sulfoxide reductase (MsrB) increases oxacillin and H₂O₂ stress resistance and enhances pigmentation in Staphylococcus aureus.

    Science.gov (United States)

    Singh, Vineet K

    2014-09-01

    Staphylococcus aureus produces 3 MsrA enzymes (MsrA1, MsrA2, and MsrA3) and 1 MsrB enzyme. The genes encoding MsrA1 and MsrB are the first and second genes of a 4-gene operon in S. aureus. In a previous study, MsrA1-deficient S. aureus cells showed increased sensitivity to oxidative stress conditions in spite of a higher production of MsrB. In this study, an msrB mutant of S. aureus was created by site-directed mutagenesis that left the first gene of this locus, msrA1, intact. Studies with this mutant suggest that a deletion of MsrB increases resistance of S. aureus to H2O2 and oxacillin and that the mutant cells produce a higher level of carotenoids relative to wild-type S. aureus cells.

  16. Aluminothermic Reduction-Molten Salt Electrolysis Using Inert Anode for Oxygen and Al-Base Alloy Extraction from Lunar Soil Simulant

    Science.gov (United States)

    Xie, Kaiyu; Shi, Zhongning; Xu, Junli; Hu, Xianwei; Gao, Bingliang; Wang, Zhaowen

    2017-10-01

    Aluminothermic reduction-electrolysis using an inert anode process is proposed to extract oxygen and metals from Minnesota Lunar Simulant-1 (MLS-1). Effective aluminothermic reduction between dissolved MLS-1 and dissolved metal aluminum was achieved in cryolite salt media. The product phases obtained by aluminothermic reduction at 980°C for 4 h were Al, Si, and Al5FeSi, while the chemical components were 79.71 mass% aluminum, 12.03 mass% silicon, 5.91 mass% iron, and 2.35 mass% titanium. The cryolite salt containing Al2O3 was subsequently electrolyzed with Fe0.58-Ni0.42 inert anode at 960°C for 4 h. Oxygen was evolved at the anode with an anodic current efficiency of 78.28%. The results demonstrate that this two-step process is remarkably feasible for the extraterrestrial extraction of oxygen and metals. This process will help expand the existing in situ resource utilization methods.

  17. Metallocene electrochemistry. 3. Reduction-oxidation study of dicyclopentadienyl compounds of ruthenium and titanium in Lewis acid-base molten salts

    Energy Technology Data Exchange (ETDEWEB)

    Gale, R.J.; Job, R.

    1981-01-01

    Voltammetry at vitreous C electrodes of Ru(eta/sup 5/-C/sub 5/H/sub 5/)/sub 2/ and Ti(eta/sup 5/-C/sub 5/H/sub 5/)/sub 2/Cl/sub 2/ in the 0.8:1 molar ratio AlCl/sub 3/-1 butylpyridinium chloride melt reveals an irreversible oxidation (E/sub p/2/ = = +0.68 V, 100 mV s/sup -1/) and a reversible 1e reduction (E/sub 1/2/ = -0.65 V vs. SCE), respectively. These electrode potentials are comparable to those found for each species in nonaqueous, organic electrolytes. In the acidic melts (excess AlCl/sub 3/), Ru(eta/sup 5/-C/sub 5/H/sub 5/)/sub 2/ is oxidized by multistep pathways sensitive to the melt composition, while the Ti(eta/sup 5/-C/sub 5/H/sub 5/)/sup 2 +/ dication is reduced by a reversible 1e step (E/sub 1/2/ = +0.67 V vs. SCE, 2:1 mole ratio melt) in a pC1/sup -/ dependent process. Electrochemical behavior of each of these metallocenes in neutral 1:1 molar ratio melts is complicated.

  18. The chemistry of molten salt mixtures: application to the reductive extraction of lanthanides and actinides by a liquid metal; Chimie des melanges de sels fondus. Application a l'extraction reductrice d'actinides et de lanthanides par un metal liquide

    Energy Technology Data Exchange (ETDEWEB)

    Finne, J

    2005-10-15

    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 - pO{sub 2} - 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)

  19. Optimizing the space-time-yield and the specific energy consumption of molten salt electrolysis processes for the electrowinning of metals in subgroups 4 and 5 of the periodic table of elements

    International Nuclear Information System (INIS)

    Koeck, W.

    1988-04-01

    Disadvantages of molten salt electrolysis are its low space-time-yield (kg/m 3 h) combined with its high specific energy consumption (kWh/kg). These factors essentially determine how electrolysis is applied on an industrial scale. The electrolysis of tantalum was selected as an example representative for other electrolytic processes; this series of tests allow statements also on the winning of the other elements from subgroups 4 and 5 of the periodic table, and on electrolytic aluminium extraction. Optimal operating conditions for direct current electrolysis were determined in the laboratory by varying the current density and the electrolysis temperature. In order to improve the space-time-yield from an existing electrolytic cell with a given electrolyte composition beyond the optimal range of direct current electrolysis, the process of periodic current reversal is applied. In this process, the polarity is reversed for a short time at constant periodic intervals. If the forward time period and the backward time period are chosen in a suitable way, both the current efficiency and the space-time-yield can be improved without increasing the energy consumption. 59 refs., 48 figs., 8 tabs. (Author)

  20. Infinite Multiplication Factor and Temperature Coefficient of MSR Calculated by HELIOS

    International Nuclear Information System (INIS)

    Nemeth, Z.; Necas, V.; Darilak, P.

    2006-01-01

    The aim of this study was to make a correct model of an elementary MSR cell calculated by HELIOS. Infinite multiplication factor and temperature coefficient of reactivity were calculated for various fuel channel diameters (Authors)