National Library of Energy BETA

Sample records for molten salt reactor

  1. Accelerators for Subcritical Molten-Salt Reactors

    SciTech Connect (OSTI)

    Johnson, Roland

    2011-08-03

    Accelerator parameters for subcritical reactors have usually been based on using solid nuclear fuel much like that used in all operating critical reactors as well as the thorium burning accelerator-driven energy amplifier proposed by Rubbia et al. An attractive alternative reactor design that used molten salt fuel was experimentally studied at ORNL in the 1960s, where a critical molten salt reactor was successfully operated using enriched U235 or U233 tetrafluoride fuels. These experiments give confidence that an accelerator-driven subcritical molten salt reactor will work better than conventional reactors, having better efficiency due to their higher operating temperature, having the inherent safety of subcritical operation, and having constant purging of volatile radioactive elements to eliminate their accumulation and potential accidental release in dangerous amounts. Moreover, the requirements to drive a molten salt reactor can be considerably relaxed compared to a solid fuel reactor, especially regarding accelerator reliability and spallation neutron targetry, to the point that much of the required technology exists today. It is proposed that Project-X be developed into a prototype commercial machine to produce energy for the world by, for example, burning thorium in India and nuclear waste from conventional reactors in the USA.

  2. Liquid fuel molten salt reactors for thorium utilization (Journal Article)

    Office of Scientific and Technical Information (OSTI)

    | SciTech Connect Journal Article: Liquid fuel molten salt reactors for thorium utilization Citation Details In-Document Search This content will become publicly available on April 8, 2017 Title: Liquid fuel molten salt reactors for thorium utilization Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and

  3. Hybrid Molten Salt Reactor (HMSR) System Study

    SciTech Connect (OSTI)

    Woolley, Robert D; Miller, Laurence F

    2014-04-01

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

  4. Fast Spectrum Molten Salt Reactor Options

    SciTech Connect (OSTI)

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

    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.

  5. Fast Thorium Molten Salt Reactors Started with Plutonium

    SciTech Connect (OSTI)

    Merle-Lucotte, E.; Heuer, D.; Le Brun, C.; Brissot, R.; Liatard, E.; Meplan, O.; Nuttin, A.

    2006-07-01

    One of the pending questions concerning Molten Salt Reactors based on the {sup 232}Th/{sup 233}U fuel cycle is the supply of the fissile matter, and as a consequence the deployment possibilities of a fleet of Molten Salt Reactors, since {sup 233}U does not exist on earth and is not yet produced in the current operating reactors. A solution may consist in producing {sup 233}U in special devices containing Thorium, in Pressurized Water or Fast Neutrons Reactors. Two alternatives to produce {sup 233}U are examined here: directly in standard Molten Salt Reactors started with Plutonium as fissile matter and then operated in the Th/{sup 233}U cycle; or in dedicated Molten Salt Reactors started and fed with Plutonium as fissile matter and Thorium as fertile matter. The idea is to design a critical reactor able to burn the Plutonium and the minor actinides presently produced in PWRs, and consequently to convert this Plutonium into {sup 233}U. A particular reactor configuration is used, called 'unique channel' configuration in which there is no moderator in the core, leading to a quasi fast neutron spectrum, allowing Plutonium to be used as fissile matter. The conversion capacities of such Molten Salt Reactors are excellent. For Molten Salt Reactors only started with Plutonium, the assets of the Thorium fuel cycle turn out to be quickly recovered and the reactor's characteristics turn out to be equivalent to Molten Salt Reactors operated with {sup 233}U only. Using a combination of Molten Salt Reactors started or operated with Plutonium and of Molten Salt Reactors started with {sup 233}U, the deployment capabilities of these reactors fully satisfy the condition of sustainability. (authors)

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

    SciTech Connect (OSTI)

    Pauzi, Anas Muhamad; Cioncolini, Andrea; Iacovides, Hector

    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.

  7. Liquid fuel molten salt reactors for thorium utilization

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Gehin, Jess C.; Powers, Jeffrey J.

    2016-04-08

    Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with themore » online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing

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

    SciTech Connect (OSTI)

    Betzler, Benjamin R; Mays, Gary T

    2016-01-01

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

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

    SciTech Connect (OSTI)

    Zitek, Pavel E-mail: klimko@kke.zcu.cz; Valenta, Vaclav E-mail: klimko@kke.zcu.cz; Klimko, Marek E-mail: klimko@kke.zcu.cz

    2014-08-06

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

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

    SciTech Connect (OSTI)

    Notz, K.J.

    1988-01-01

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

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

    SciTech Connect (OSTI)

    Williams, D.F.; Peretz, F.J.

    1996-05-01

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

  12. Hybrid Molten Salt Reactor (HMSR): Method and System to fully fission

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    actinides for electric power production without fuel enrichment, fabrication, or reprocessing | Princeton Plasma Physics Lab Hybrid Molten Salt Reactor (HMSR): Method and System to fully fission actinides for electric power production without fuel enrichment, fabrication, or reprocessing A method for integrating an external source of high-energy neutrons with a conventional moderated high conversion ratio molten salt reactor, thereby creating a self-contained hybrid system which fissions any

  13. Transient Analyses for a Molten Salt Transmutation Reactor Using the Extended SIMMER-III Code

    SciTech Connect (OSTI)

    Wang, Shisheng; Rineiski, Andrei; Maschek, Werner; Ignatiev, Victor

    2006-07-01

    Recent developments extending the capabilities of the SIMMER-III code for the dealing with transient and accidents in Molten Salt Reactors (MSRs) are presented. These extensions refer to the movable precursor modeling within the space-time dependent neutronics framework of SIMMER-III, to the molten salt flow modeling, and to new equations of state for various salts. An important new SIMMER-III feature is that the space-time distribution of the various precursor families with different decay constants can be computed and took into account in neutron/reactivity balance calculations and, if necessary, visualized. The system is coded and tested for a molten salt transmuter. This new feature is also of interest in core disruptive accidents of fast reactors when the core melts and the molten fuel is redistributed. (authors)

  14. Development of fluoride reprocessing technologies devoted to molten-salt reactor systems

    SciTech Connect (OSTI)

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

    2007-07-01

    Main fuel processing and reprocessing technologies proposed for Molten Salt Reactor fuel cycle are pyrochemical or pyrometallurgical, majority of them are fluoride technologies. It is based on the fact that Molten Salt Reactor fuel is in the chemical form of molten fluorides and the reprocessing technology is needed to be an 'on-line' process. The corresponding pyrochemical separation processes proposed for MSR fuel processing and reprocessing are mainly fluoride volatilization processes, molten salt / liquid metal extraction processes, electrochemical separation processes from the molten salt media and gas extraction from the molten salt medium. Techniques based on fluoride volatilization and on electrochemical separation from fluoride molten salt media are under development in the Czech Republic. Whereas the Fluoride Volatility Method is proposed to be the main 'Front-end' technology of the MSR used as the actinide burner (transmuter), the electro-separation methods should be dedicated to the 'on-line' reprocessing of the circulating MSR fuel and should be used as for MSR incinerating transuranium fuel as for MSR working within the {sup 232}Th - {sup 233}U fuel cycle. (authors)

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

    SciTech Connect (OSTI)

    Harto, Andang Widi

    2012-06-06

    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.

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

    SciTech Connect (OSTI)

    Patton, Bruce; Sorensen, Kirk

    2002-07-01

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

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

    SciTech Connect (OSTI)

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

    2006-07-01

    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)

  18. Reactor Controllability of 3-Region-Core Molten Salt Reactor System - A Study on Load Following Capability

    SciTech Connect (OSTI)

    Takahisa Yamamoto; Koshi Mitachi; Masatoshi Nishio

    2006-07-01

    The Molten Salt Reactor (MSR) systems are liquid-fueled reactors that can be used for actinide burning, production of electricity, production of hydrogen, and production of fissile fuels (breeding). Thorium (Th) and uranium-233 ({sup 233}U) are fertile and fissile of the MSR systems, and dissolved in a high-temperature molten fluoride salt (fuel salt) with a very high boiling temperature (up to 1650 K), that is both the reactor nuclear fuel and the coolant. The MSR system is one of the six advanced reactor concepts identified by the Generation IV International Forum (GIF) as a candidate for cooperative development. In the MSR system, fuel salt flows through a fuel duct constructed around a reactor core and fuel channel of a graphite moderator accompanied by fission reaction and heat generation, and flows out to an external-loop system consisted of a heat exchanger and a circulation pump. Due to the motion of fuel salt, delayed neutron precursors that are one of the source of neutron production make to change their position between the fission reaction and neutron emission events and decay even occur in the external loop system. Hence the reactivity and effective delayed neutron precursor fraction of the MSR system are lower than those of solid fuel reactor systems such as Boiling Water Reactors (BWRs) and Pressurised Water Reactor (PWRs). Since all of the presently operating nuclear power reactors utilize solid fuel, little attention had been paid to the MSR analysis of the reactivity loss and reactor characteristics change caused by the fuel salt circulation. Sides et al. and Shimazu et al. developed MSR analytical models based on the point reactor kinetics model to consider the effect of fuel salt flow. Their models represented a reactor as having six zones for fuel salt and three zones for the graphite moderator. Since their models employed the point reactor kinetics model and the rough temperature approximation, their results were not sufficiently accurate to

  19. Preliminary safety calculations to improve the design of Molten Salt Fast Reactor

    SciTech Connect (OSTI)

    Brovchenko, M.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Capellan, N.; Ghetta, V.; Laureau, A.

    2012-07-01

    Molten salt reactors are liquid fuel reactors so that they are flexible in operation but very different in the safety approach from solid fuel reactors. This study bears on the specific concept named Molten Salt Fast Reactor (MSFR). Since this new nuclear technology is in development, safety is an essential point to be considered all along the R and D studies. This paper presents the first step of the safety approach: the systematic description of the MSFR, limited here to the main systems surrounding the core. This systematic description is the basis on which we will be able to devise accidental scenarios. Thanks to the negative reactivity feedback coefficient, most accidental scenarios lead to reactor shut down. Because of the decay heat generated in the fuel salt, it must be cooled. After the description of the tools developed to calculate the residual heat, the different contributions are discussed in this study. The decay heat of fission products in the MSFR is evaluated to be low (3% of nominal power), mainly due to the reprocessing that transfers the fission products to the gas reprocessing unit. As a result, the contribution of the actinides is significant (0.5% of nominal power). The unprotected loss of heat sink transients are studied in this paper. It appears that slow transients are favorable (> 1 min) to minimize the temperature increase of the fuel salt. This work will be the basis of further safety studies as well as an essential parameter for the design of the draining system. (authors)

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

    SciTech Connect (OSTI)

    Aji, Indarta Kuncoro; Waris, Abdul Permana, Sidik

    2015-09-30

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

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

    SciTech Connect (OSTI)

    Uhlir, J.; Straka, M.; Szatmary, L.

    2012-07-01

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

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

    SciTech Connect (OSTI)

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

    2012-06-06

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

  3. Random effects of fissile lumps in molten salt reactors

    SciTech Connect (OSTI)

    Dulla, S.; Ravetto, P.; Prinja, A. K.

    2013-07-01

    The problem of the effect of fissile lumps spatially appearing in a random fashion inside a fluid fuel reactor is addressed. The effect on reactivity is evaluated by means of first-order perturbation theory. The analysis is carried out in diffusion theory with the presence of delayed neutron emissions in one dimensional plane geometry. The estimation of the mean value and standard deviation of the reactivity inserted is performed by Monte Carlo simulations and a deterministic quadrature approach, to compare the methods in terms of computational effort and the accuracy of the results. The results presented show that the effects constitute an important issue in the assessment of these innovative systems. (authors)

  4. Sandia Energy - Molten Salt Test Loop Melted Salt

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Salt Home Renewable Energy Energy News Concentrating Solar Power Solar Molten Salt Test Loop Melted Salt Previous Next Molten Salt Test Loop Melted Salt The Molten Salt Test...

  5. Molten salt electrolyte separator

    DOE Patents [OSTI]

    Kaun, T.D.

    1996-07-09

    The patent describes a molten salt electrolyte/separator for battery and related electrochemical systems including a molten electrolyte composition and an electrically insulating solid salt dispersed therein, to provide improved performance at higher current densities and alternate designs through ease of fabrication. 5 figs.

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

    SciTech Connect (OSTI)

    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

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

    SciTech Connect (OSTI)

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

    2010-03-01

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

  8. Gaseous fission product management for molten salt reactors and vented fuel systems

    SciTech Connect (OSTI)

    Messenger, S. J.; Forsberg, C.; Massie, M.

    2012-07-01

    Fission gas disposal is one of the unresolved difficulties for Molten Salt Reactors (MSRs) and advanced reactors with vented fuel systems. As these systems operate, they produce many radioactive isotopes of xenon and krypton (e.g. {sup 135}Xe t{sub 1/2} = 9.14 hours and {sup 85}Kr t{sub 1/2}= 10.73 years). Removing these gases proves vital to the success of such reactor designs for two reasons. First, the gases act as large neutron sinks which decrease reactivity and must be counterbalanced by increasing fuel loading. Second, for MSRs, inert fission product gases naturally separate quickly from high temperature salts, thus creating high vapor pressure which poses safety concerns. For advanced reactors with solid vented fuel, the gases are allowed to escape into an off-gas system and thus must be managed. Because of time delays in transport of fission product gases in vented fuel systems, some of the shorter-lived radionuclides will decay away thereby reducing the fission gas source term relative to an MSR. To calculate the fission gas source term of a typical molten salt reactor, we modeled a 1000 MWe graphite moderated thorium MSR similar to that detailed in Mathieu et al. [1]. The fuel salt used in these calculations was LiF (78 mole percent) - (HN)F 4 (22 mole percent) with a heavy nuclide composition of 3.86% {sup 233}U and 96.14% {sup 232}Th by mass. Before we can remove the fission product gases produced by this reactor configuration, we must first develop an appropriate storage mechanism. The gases could be stored in pressurized containers but then one must be concerned about bottle failure. Methods to trap noble gases in matrices are expensive and complex. Alternatively, there are direct storage/disposal options: direct injection into the Earth or injecting a grout-based product into the Earth. Advances in drilling technologies, hydro fracture technologies, and methods for the sequestration of carbon dioxide from fossil fuel plants are creating new options

  9. Molten salt rolling bubble column, reactors utilizing same and related methods

    SciTech Connect (OSTI)

    Turner, Terry D.; Benefiel, Bradley C.; Bingham, Dennis N.; Klinger, Kerry M.; Wilding, Bruce M.

    2015-11-17

    Reactors for carrying out a chemical reaction, as well as related components, systems and methods are provided. In accordance with one embodiment, a reactor is provided that includes a furnace and a crucible positioned for heating by the furnace. The crucible may contain a molten salt bath. A downtube is disposed at least partially within the interior crucible along an axis. The downtube includes a conduit having a first end in communication with a carbon source and an outlet at a second end of the conduit for introducing the carbon material into the crucible. At least one opening is formed in the conduit between the first end and the second end to enable circulation of reaction components contained within the crucible through the conduit. An oxidizing material may be introduced through a bottom portion of the crucible in the form of gas bubbles to react with the other materials.

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

    SciTech Connect (OSTI)

    Aji, Indarta Kuncoro; Waris, A.

    2014-09-30

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

  11. Conceptual design characteristics of a denatured molten-salt reactor with once-through fueling

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    Powers, J.J.; Worrall, A.; Gehin, J.C.; Harrison, T.J.; Sunny, E.E.

    2013-07-01

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

  13. The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

    SciTech Connect (OSTI)

    Nevinitsa, V. A. Dudnikov, A. A.; Blandinskiy, V. Yu.; Balanin, A. L.; Alekseev, P. N.; Titarenko, Yu. E.; Batyaev, V. F.; Pavlov, K. V.; Titarenko, A. Yu.

    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.

  14. Waste Stream Generated and Waste Disposal Plans for Molten Salt Reactor Experiment at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Haghighi, M. H.; Szozda, R. M.; Jugan, M. R.

    2002-02-26

    The Molten Salt Reactor Experiment (MSRE) site is located in Tennessee, on the U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR), south of the Oak Ridge National Laboratory (ORNL) main plant across Haw Ridge in Melton Valley. The MSRE was run by ORNL to demonstrate the desirable features of the molten-salt concept in a practical reactor that could be operated safely and reliably. It introduced the idea of a homogeneous reactor using fuel salt media and graphite moderation for power and breeder reactors. The MSRE reactor and associated components are located in cells beneath the floor in the high-bay area of Building 7503 (Figure 1). The reactor was operated from June 1965 to December 1969. When the reactor was shut down, fuel salt was drained from the reactor circuit to two drain tanks. A ''clean'' salt was then circulated through the reactor as a decontamination measure and drained to a third drain tank. When operations ceased, the fuel and flush salts were allowed t o cool and solidify in the drain tanks. At shutdown, the MSRE facility complex was placed in a surveillance and maintenance program. As a result of the S&M program, it was discovered in 1994 that gaseous uranium (233U/232U) hexafluoride (UF6) had moved throughout the MSRE process systems. The UF6 was generated when radiolysis of the fluorine salts caused the individual constituents to dissociate to their component atoms, including free fluorine.Some of the free fluorine combined with uranium fluorides (UF4) in the salt to form UF6. UF6 is gaseous at slightly above ambient temperatures; thus, periodic heating of the fuel salts (which was intended to remedy the radiolysis problems) and simple diffusion had allowed the UF6 to move out of the salt and into the process systems of MSRE.

  15. Sandia Energy - Molten Salt Test Loop Commissioning

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Energy Energy News EC News & Events Concentrating Solar Power Solar Molten Salt Test Loop Commissioning Previous Next Molten Salt Test Loop Commissioning The Molten Salt...

  16. Molten salt destruction of energetic waste materials

    DOE Patents [OSTI]

    Brummond, William A.; Upadhye, Ravindra S.; Pruneda, Cesar O.

    1995-01-01

    A molten salt destruction process is used to treat and destroy energetic waste materials such as high explosives, propellants, and rocket fuels. The energetic material is pre-blended with a solid or fluid diluent in safe proportions to form a fluid fuel mixture. The fuel mixture is rapidly introduced into a high temperature molten salt bath. A stream of molten salt is removed from the vessel and may be recycled as diluent. Additionally, the molten salt stream may be pumped from the reactor, circulated outside the reactor for further processing, and delivered back into the reactor or cooled and circulated to the feed delivery system to further dilute the fuel mixture entering the reactor.

  17. Molten salt destruction of energetic waste materials

    DOE Patents [OSTI]

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

    1995-07-18

    A molten salt destruction process is used to treat and destroy energetic waste materials such as high explosives, propellants, and rocket fuels. The energetic material is pre-blended with a solid or fluid diluent in safe proportions to form a fluid fuel mixture. The fuel mixture is rapidly introduced into a high temperature molten salt bath. A stream of molten salt is removed from the vessel and may be recycled as diluent. Additionally, the molten salt stream may be pumped from the reactor, circulated outside the reactor for further processing, and delivered back into the reactor or cooled and circulated to the feed delivery system to further dilute the fuel mixture entering the reactor. 4 figs.

  18. MOLTEN FLUORIDE NUCLEAR REACTOR FUEL

    DOE Patents [OSTI]

    Barton, C.J.; Grimes, W.R.

    1960-01-01

    Molten-salt reactor fuel compositions consisting of mixtures of fluoride salts are reported. In its broadest form, the composition contains an alkali fluoride such as sodium fluoride, zirconium tetrafluoride, and a uranium fluoride, the latter being the tetrafluoride or trifluoride or a mixture of the two. An outstanding property of these fuel compositions is a high coeffieient of thermal expansion which provides a negative temperature coefficient of reactivity in reactors in which they are used.

  19. High-value use of weapons-plutonium by burning in molten salt accelerator-driven subcritical systems or reactors

    SciTech Connect (OSTI)

    Bowman, C.D.; Venneri, F.

    1993-11-01

    The application of thermal-spectrum molten-salt reactors and accelerator-driven subcritical systems to the destruction of weapons-return plutonium is considered from the perspective of deriving the maximum societal benefit. The enhancement of electric power production from burning the fertile fuel {sup 232}Th with the plutonium is evaluated. Also the enhancement of destruction of the accumulated waste from commercial nuclear reactors is considered using the neutron-rich weapons plutonium. Most cases examined include the concurrent transmutation of the long-lived actinide and fission product waste ({sup 99}Tc, {sup 129}I, {sup 135}Cs, {sup 126}Sn and {sup 79}Se).

  20. Molten salt lithium cells

    DOE Patents [OSTI]

    Raistrick, Ian D.; Poris, Jaime; Huggins, Robert A.

    1982-02-09

    Lithium-based cells are promising for applications such as electric vehicles and load-leveling for power plants since lithium is very electropositive and light weight. One type of lithium-based cell utilizes a molten salt electrolyte and is operated in the temperature range of about 400.degree.-500.degree. C. Such high temperature operation accelerates corrosion problems and a substantial amount of energy is lost through heat transfer. The present invention provides an electrochemical cell (10) which may be operated at temperatures between about 100.degree.-170.degree. C. Cell (10) comprises an electrolyte (16), which preferably includes lithium nitrate, and a lithium or lithium alloy electrode (12).

  1. Molten salt lithium cells

    DOE Patents [OSTI]

    Raistrick, Ian D.; Poris, Jaime; Huggins, Robert A.

    1983-01-01

    Lithium-based cells are promising for applications such as electric vehicles and load-leveling for power plants since lithium is very electropositive and light weight. One type of lithium-based cell utilizes a molten salt electrolyte and is operated in the temperature range of about 400.degree.-500.degree. C. Such high temperature operation accelerates corrosion problems and a substantial amount of energy is lost through heat transfer. The present invention provides an electrochemical cell (10) which may be operated at temperatures between about 100.degree.-170.degree. C. Cell (10) comprises an electrolyte (16), which preferably includes lithium nitrate, and a lithium or lithium alloy electrode (12).

  2. Molten salt lithium cells

    DOE Patents [OSTI]

    Raistrick, I.D.; Poris, J.; Huggins, R.A.

    1980-07-18

    Lithium-based cells are promising for applications such as electric vehicles and load-leveling for power plants since lithium is very electropositive and light weight. One type of lithium-based cell utilizes a molten salt electrolyte and is operated in the temperature range of about 400 to 500/sup 0/C. Such high temperature operation accelerates corrosion problems and a substantial amount of energy is lost through heat transfer. The present invention provides an electrochemical cell which may be operated at temperatures between about 100 to 170/sup 0/C. The cell is comprised of an electrolyte, which preferably includes lithium nitrate, and a lithium or lithium alloy electrode.

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

    SciTech Connect (OSTI)

    Xia, B.; Lu, Y.

    2012-07-01

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

  4. Molten Salt Mixture Properties (KF-ZrF4 and KCl-MgCl2) for Use in RELAP5-3D for High Temperature Reactor Application

    SciTech Connect (OSTI)

    N. A. Anderson; P. Sabharwall

    2012-06-01

    Molten salt coolants are being investigated as primary coolants for a fluoride high-temperature reactor and as secondary coolants for high temperature reactors such as the next generation nuclear plant. This work provides a review of the thermophysical properties of candidate molten salt coolants for use as a secondary heat transfer medium from a high temperature reactor to a processing plant. The molten salts LiF-NaF-KF, KF-ZrF4 and KCl-MgCl2 were considered for use in the secondary coolant loop. The thermophysical properties necessary to add the molten salts KF-ZrF4 and KCl-MgCl2 to RELAP5-3D were gathered for potential modeling purposes. The properties of the molten salt LiF-NaF-KF were already available in RELAP5-3D. The effect that the uncertainty in individual properties had on the Nusselt number was evaluated. This uncertainty in the Nusselt number was shown to be nearly independent of the molten salt temperature.

  5. Molten metal reactors

    DOE Patents [OSTI]

    Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

    2013-11-05

    A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

  6. Batteries using molten salt electrolyte

    DOE Patents [OSTI]

    Guidotti, Ronald A.

    2003-04-08

    An electrolyte system suitable for a molten salt electrolyte battery is described where the electrolyte system is a molten nitrate compound, an organic compound containing dissolved lithium salts, or a 1-ethyl-3-methlyimidazolium salt with a melting temperature between approximately room temperature and approximately 250.degree. C. With a compatible anode and cathode, the electrolyte system is utilized in a battery as a power source suitable for oil/gas borehole applications and in heat sensors.

  7. Evaluation of the Molten Salt Reactor Experiment drain tanks for reuse in salt disposal, Oak Ridge National Laboratory, Oak Ridge, Tennessee

    SciTech Connect (OSTI)

    1998-05-01

    This report was prepared to identify the source documentation used to evaluate the drain tanks in the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). The evaluation considered the original quality of the tanks, their service history, and their intended use during the removal of fluoride salts. It also includes recommendations for a quality verification plan. The estimates of corrosion damage to the salt containing system at the MSRE are low enough to lend optimism that the system will be fit for its intended use, which is disposal of the salt by transferring it to transport containers. The expected corrosion to date is estimated between 10 and 50 mil, or 2 to 10% of the shell wall. The expected corrosion rate when the tanks are used to remove the salt at 110 F is estimated to be .025 to 0.1 mil per hour of exposure to HF and molten salt. To provide additional assurance that the estimates of corrosion damage are accurate, cost effective nondestructive examination (NDE) has been recommended. The NDE procedures are compared with industry standards and give a perspective for the extent of additional measures taken in the recommendation. A methodology for establishing the remaining life has been recommended, and work is progressing towards providing an engineering evaluation based upon thickness and design conditions for the future use of the tanks. These extra measures and the code based analysis will serve to define the risk of salt or radioactive gases leaking during processing and transfer of the salt as acceptable.

  8. Delivery system for molten salt oxidation of solid waste

    DOE Patents [OSTI]

    Brummond, William A. (Livermore, CA); Squire, Dwight V. (Livermore, CA); Robinson, Jeffrey A. (Manteca, CA); House, Palmer A. (Walnut Creek, CA)

    2002-01-01

    The present invention is a delivery system for safety injecting solid waste particles, including mixed wastes, into a molten salt bath for destruction by the process of molten salt oxidation. The delivery system includes a feeder system and an injector that allow the solid waste stream to be accurately metered, evenly dispersed in the oxidant gas, and maintained at a temperature below incineration temperature while entering the molten salt reactor.

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

    SciTech Connect (OSTI)

    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.

  10. ALARA Controls and the Radiological Lessons Learned During the Uranium Fuel Removal Projects at the Molten Salt Reactor Experiment

    SciTech Connect (OSTI)

    Gilliam, B. J.; Chapman, J. A.; Jugan, M. R.

    2002-02-26

    The removal of uranium-233 (233 U) from the auxiliary charcoal bed (ACB) of the Molten Salt Reactor Experiment (MSRE), performed from January through May 2001, created both unique radiological challenges and widely-applicable lessons learned. In addition to the criticality concerns and alpha contamination, 233U has an associated intense gamma photon from the cocontaminant uranium-232 (232U) decaying to thallium-208 (208Tl). Therefore, rigorous contamination controls and significant shielding were implemented. Extensive, timed mock-up training was also imperative to minimize individual and collective personnel exposures. Back-up shielding and containment techniques (that had been previously developed for defense in depth) were used successfully to control significant, changed conditions. Additional controls were placed on tests and on recovery designs to assure a higher level of safety throughout the removal operations. This paper delineates the manner in which each difficulty was solved, while relating the relevance of the results and the methodology to other projects with high dose-rate, highly-contaminated ionizing radiation hazards. Because of the distinctive features of and current interest in molten salt technology, a brief overview is provided. Also presented is the detailed, practical application of radiological controls integrated into, rather than added after, each evolution of the project--thus demonstrating the broad-based benefits of radiological engineering and ALARA reviews. The resolution of the serious contamination-control problems caused by unexpected uranium hexafluoride (UF6) gaseous diffusion is also explicated. Several tables and figures document the preparations, equipment and operations. A comparison of the pre-job dose calculations for the various functions of the uranium deposit removal (UDR) and the post-job dose-rate data are included in the conclusion.

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

    SciTech Connect (OSTI)

    Wilson, Guy

    2009-10-06

    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.

  12. Health and safety plan for the Molten Salt Reactor Experiment remediation project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    SciTech Connect (OSTI)

    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. Preliminary Design For Conventional and Compact Secondary Heat Exchanger in a Molten Salt Reactor

    SciTech Connect (OSTI)

    Piyush Sabharwall; Mike Patterson; Ali Siahpush; Eung Soo Kim

    2012-07-01

    The strategic goal of the Advance Reactors such as AHTR is to broaden the environmental and economic benefits of nuclear energy in the United States by producing power to meet growing energy demands and demonstrating its applicability to market sectors not being served by light water reactors

  14. Design of a californium source-driven measurement system for accountability of material recovered from the Molten Salt Reactor Experiment charcoal bed

    SciTech Connect (OSTI)

    Bentzinger, D.L.; Perez, R.B.; Mattingly, J.K.; Valentine, T.E.; Mihalczo, J.T.

    1998-05-01

    The Molten Salt Reactor Experiment Facility (MSRE) operated from 1965 to 1969. The fuel was a molten salt that flowed through the reactor core which consisted of uranium tetrafluoride with molten lithium and beryllium salt used as the coolant. In 1968 the fuel was switched from {sup 235}U to {sup 233}U. The Molten Salt Reactor Experiment was canceled in 1969 at which time approximately 4800 kg of salt was transferred to the fuel drain tanks. There was about 36.3 kg of uranium, 675 grams of plutonium and various fission products present in the fuel salt. The salt was allowed to solidify in the fuel drain tanks. The salt was heated on a yearly basis to recombine the fluorine gas with the uranium salt mixture. In March 1994, a gas sample was taken from the off gas system that indicated {sup 233}U had migrated from the fuel drain tank system to the off gas system. It was found that approximately 2.6 kg of uranium had migrated to the Auxiliary Charcoal Bed (ACB). The ACB is located in the concrete-lined charcoal bed cell which is below ground level located outside the MSRE building. Therefore, there was a concern for the potential of a nuclear criticality accident, although water would have to leak into the chamber for a criticality accident to occur. Unstable carbon/fluorine compounds were also formed when the fluorine reacted with the charcoal in the charcoal bed. The purpose of the proposed measurement system was to perform an accountability measurement to determine the fissile mass of {sup 233}U in the primary vessel. The contents of the primary containment assembly will then be transferred to three smaller containers for long term storage. Calculations were performed using MCNP-DSP to determine the configuration of the measurement system. The information obtained from the time signatures can then be compared to the measurement data to determine the amount of {sup 233}U present in the primary containment assembly.

  15. Advanced heat exchanger development for molten salts

    SciTech Connect (OSTI)

    Sabharwall, Piyush; Clark, Denis; Glazoff, Michael; Zheng, Guiqiu; Sridharan, Kumar; Anderson, Mark

    2014-12-01

    This study addresses present work concerned with advanced heat exchanger development for molten salt in nuclear and non nuclear thermal systems. The molten salt systems discussed herein use alloys, such as Hastelloy N and 242, which show corrosion resistance to molten salt at nominal operating temperatures up to 700°C. These alloys were diffusion welded, and the corresponding information is presented. Test specimens were prepared for exposing diffusion welds to molten salt environments. Hastelloy N and 242 were found to be weldable by diffusion welding, with ultimate tensile strengths about 90% of base metal values. Both diffusion welds and sheet material in Hastelloy N were corrosion tested in?58 mol% KF and 42 mol% ZrF4 at 650, 700, and 850°C for 200, 500, and 1,000 hours. Corrosion rates found were similar between welded and nonwelded materials, typically <10 mils per year. For materials of construction, nickel and alloys with dense nickel coatings are effectively inert to corrosion in fluorides, but not so in chlorides. Hence, additional testing of selected alloys for resistance to intergranular corrosion is needed, as is a determination of corrosion rate as a function of contaminant type and alloy composition with respect to chromium and carbon to better define the optimal chromium and carbon composition, independent of galvanic or differential solubility effects. Also presented is the division of the nuclear reactor and high temperature components per ASME standards, along with design requirements for a subcritical Rankine power cycle heat exchanger that has to overcome pressure difference of about 17 MPa.

  16. Advanced heat exchanger development for molten salts

    SciTech Connect (OSTI)

    Sabharwall, Piyush; Clark, Denis; Glazoff, Michael; Zheng, Guiqiu; Sridharan, Kumar; Anderson, Mark

    2014-12-01

    This study addresses present work concerned with advanced heat exchanger development for molten salt in nuclear and non nuclear thermal systems. The molten salt systems discussed herein use alloys, such as Hastelloy N and 242, which show corrosion resistance to molten salt at nominal operating temperatures up to 700C. These alloys were diffusion welded, and the corresponding information is presented. Test specimens were prepared for exposing diffusion welds to molten salt environments. Hastelloy N and 242 were found to be weldable by diffusion welding, with ultimate tensile strengths about 90% of base metal values. Both diffusion welds and sheet material in Hastelloy N were corrosion tested in?58 mol% KF and 42 mol% ZrF4 at 650, 700, and 850C for 200, 500, and 1,000 hours. Corrosion rates found were similar between welded and nonwelded materials, typically <10 mils per year. For materials of construction, nickel and alloys with dense nickel coatings are effectively inert to corrosion in fluorides, but not so in chlorides. Hence, additional testing of selected alloys for resistance to intergranular corrosion is needed, as is a determination of corrosion rate as a function of contaminant type and alloy composition with respect to chromium and carbon to better define the optimal chromium and carbon composition, independent of galvanic or differential solubility effects. Also presented is the division of the nuclear reactor and high temperature components per ASME standards, along with design requirements for a subcritical Rankine power cycle heat exchanger that has to overcome pressure difference of about 17 MPa.

  17. Advanced heat exchanger development for molten salts

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Sabharwall, Piyush; Clark, Denis; Glazoff, Michael; Zheng, Guiqiu; Sridharan, Kumar; Anderson, Mark

    2014-12-01

    This study addresses present work concerned with advanced heat exchanger development for molten salt in nuclear and non nuclear thermal systems. The molten salt systems discussed herein use alloys, such as Hastelloy N and 242, which show corrosion resistance to molten salt at nominal operating temperatures up to 700°C. These alloys were diffusion welded, and the corresponding information is presented. Test specimens were prepared for exposing diffusion welds to molten salt environments. Hastelloy N and 242 were found to be weldable by diffusion welding, with ultimate tensile strengths about 90% of base metal values. Both diffusion welds and sheet materialmore » in Hastelloy N were corrosion tested in?58 mol% KF and 42 mol% ZrF4 at 650, 700, and 850°C for 200, 500, and 1,000 hours. Corrosion rates found were similar between welded and nonwelded materials, typically <10 mils per year. For materials of construction, nickel and alloys with dense nickel coatings are effectively inert to corrosion in fluorides, but not so in chlorides. Hence, additional testing of selected alloys for resistance to intergranular corrosion is needed, as is a determination of corrosion rate as a function of contaminant type and alloy composition with respect to chromium and carbon to better define the optimal chromium and carbon composition, independent of galvanic or differential solubility effects. Also presented is the division of the nuclear reactor and high temperature components per ASME standards, along with design requirements for a subcritical Rankine power cycle heat exchanger that has to overcome pressure difference of about 17 MPa.« less

  18. Update on Small Modular Reactors Dynamics System Modeling Tool -- Molten Salt Cooled Architecture

    SciTech Connect (OSTI)

    Hale, Richard Edward; Cetiner, Sacit M.; Fugate, David L.; Qualls, A L.; Borum, Robert C.; Chaleff, Ethan S.; Rogerson, Doug W.; Batteh, John J.; Tiller, Michael M.

    2014-08-01

    The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the third year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled) concepts, including the use of multiple coupled reactors at a single site. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor SMR models, ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface (ICHMI) technical area, and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the Modular Dynamic SIMulation (MoDSIM) tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the program, (2) developing a library of baseline component modules that can be assembled into full plant models using existing geometry and thermal-hydraulic data, (3) defining modeling conventions for interconnecting component models, and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

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

    DOE Patents [OSTI]

    Brummond, W.A.; Upadhye, R.S.

    1996-02-13

    An injector nozzle has been designed for safely injecting energetic waste materials, such as high explosives, propellants, and rocket fuels, into a molten salt reactor in a molten salt destruction process without premature detonation or back burn in the injection system. The energetic waste material is typically diluted to form a fluid fuel mixture that is injected rapidly into the reactor. A carrier gas used in the nozzle serves as a carrier for the fuel mixture, and further dilutes the energetic material and increases its injection velocity into the reactor. The injector nozzle is cooled to keep the fuel mixture below the decomposition temperature to prevent spontaneous detonation of the explosive materials before contact with the high-temperature molten salt bath. 2 figs.

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

    DOE Patents [OSTI]

    Brummond, William A. (Livermore, CA); Upadhye, Ravindra S. (Pleasanton, CA)

    1996-01-01

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

  1. Sandia Energy - Molten Salt Test Loop Pump Installed

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Energy Energy News Concentrating Solar Power Solar Energy Storage Systems Molten Salt Test Loop Pump Installed Previous Next Molten Salt Test Loop Pump Installed The pump was...

  2. Molten salt heat transfer fluids and thermal storage technology...

    Office of Scientific and Technical Information (OSTI)

    Molten salt heat transfer fluids and thermal storage technology. Citation Details In-Document Search Title: Molten salt heat transfer fluids and thermal storage technology. No ...

  3. Development of Molten-Salt Heat Trasfer Fluid Technology for...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Development of Molten-Salt Heat Trasfer Fluid Technology for Parabolic Trough Solar Power Plants Development of Molten-Salt Heat Trasfer Fluid Technology for Parabolic Trough Solar ...

  4. Novel Molten Salts Thermal Energy Storage for Concentrating Solar...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

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

  5. Prototype Tests for the Recovery and Conversion of UF6Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

    SciTech Connect (OSTI)

    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.

  6. Prototype Tests for the Recovery and Conversion of UF6 Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

    SciTech Connect (OSTI)

    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.

  7. Evaluation of Salt Coolants for Reactor Applications

    SciTech Connect (OSTI)

    Williams, David F

    2008-01-01

    Molten fluorides were initially developed for use in the nuclear industry as the high-temperature fluid fuel for the Molten Salt Reactor (MSR). The U.S. Department of Energy Office of Nuclear Energy is exploring the use of molten salts as primary and secondary coolants in a new generation of solid-fueled, thermal-spectrum, hightemperature reactors. This paper provides a review of relevant properties for use in evaluation and ranking of salt coolants for high-temperature reactors. Nuclear, physical, and chemical properties were reviewed, and metrics for evaluation are recommended. Chemical properties of the salt were examined to identify factors that affect materials compatibility (i.e., corrosion). Some preliminary consideration of economic factors for the candidate salts is also presented.

  8. Project Profile: Modular and Scalable Baseload Molten Salt Plant Conceptual

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Design and Feasibility | Department of Energy Modular and Scalable Baseload Molten Salt Plant Conceptual Design and Feasibility Project Profile: Modular and Scalable Baseload Molten Salt Plant Conceptual Design and Feasibility eSolar logo eSolar, under the Baseload CSP FOA, designed a 100-MW, 75% capacity factor, molten salt power tower plant, based around a molten salt receiver and heliostat field module with a nominal thermal rating of 50 MWth. They used a modular approach, which can be

  9. Sandia Energy - Molten Nitrate Salt Initial Flow Testing is a...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Nitrate Salt Initial Flow Testing is a Tremendous Success Home Renewable Energy News Concentrating Solar Power Solar Molten Nitrate Salt Initial Flow Testing is a Tremendous...

  10. Thermal Characterization of Molten Salt Systems

    SciTech Connect (OSTI)

    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.

  11. Parametric Analyses of Single-zone Thorium-fueled Molten Salt...

    Office of Scientific and Technical Information (OSTI)

    Title: Parametric Analyses of Single-zone Thorium-fueled Molten Salt Reactor Fuel Cycle Options Authors: Powers, Jeffrey J 1 ; Worrall, Andrew 1 ; Gehin, Jess C 1 ; Harrison, ...

  12. Low temperature oxidation using support molten salt catalysts

    DOE Patents [OSTI]

    Weimer, Alan W.; Czerpak, Peter J.; Hilbert, Patrick M.

    2003-05-20

    Molten salt reactions are performed by supporting the molten salt on a particulate support and forming a fluidized bed of the supported salt particles. The method is particularly suitable for combusting hydrocarbon fuels at reduced temperatures, so that the formation NO.sub.x species is reduced. When certain preferred salts are used, such as alkali metal carbonates, sulfur and halide species can be captured by the molten salt, thereby reducing SO.sub.x and HCl emissions.

  13. Experimental studies of actinides in molten salts

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    Piyush Sabharwall; Michael George mckellar; Su-Jong Yoon

    2013-11-01

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

  15. Environmental health and safety plan for the Molten Salt Reactor Experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    SciTech Connect (OSTI)

    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.

  16. In Situ NDA Conformation Measurements Performed at Auxiliary Charcoal Bed and Other Main Charcoal Beds After Uranium Removal from Molten Salt Reactor Experiment ACB at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Haghighi, M. H.; Kring, C. T.; McGehee, J. T.; Jugan, M. R.; Chapman, J.; Meyer, K. E.

    2002-02-26

    The Molten Salt Reactor Experiment (MSRE) site is located in Tennessee, on the U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR). The MSRE was run by Oak Ridge National Laboratory (ORNL) to demonstrate the desirable features of the molten-salt concept in a practical reactor that could be operated safely and reliably. It introduced the idea of a homogeneous reactor using fuel salt media and graphite moderation for power and breeder reactors. The MSRE reactor and associated components are located in cells beneath the floor in the high-bay area of Building 7503. The reactor was operated from June 1965 to December 1969. When the reactor was shut down, fuel salt was drained from the reactor circuit to two drain tanks. A ''clean'' salt was then circulated through the reactor as a decontamination measure and drained to a third drain tank. When operations ceased, the fuel and flush salts were allowed to cool and solidify in the drain tanks. At shutdown, the MSRE facility complex was placed in a surveillance and maintenance program. Beginning in 1987, it was discovered that gaseous uranium (U-233/U-232) hexafluoride (UF6) had moved throughout the MSRE process systems. The UF6 had been generated when radiolysis in the fluorine salts caused the individual constituents to dissociate to their component atoms, including free fluorine. Some of the free fluorine combined with uranium fluorides (UF4) in the salt to produce UF6. UF6 is gaseous at slightly above ambient temperatures; thus, periodic heating of the fuel salts (which was intended to remedy the radiolysis problems) and simple diffusion had allowed the UF6 to move out of the salt and into the process systems of MSRE. One of the systems that UF6 migrated into due to this process was the offgas system which is vented to the MSRE main charcoal beds and MSRE auxiliary charcoal bed (ACB). Recently, the majority of the uranium laden-charcoal material residing within the ACB was safely and successfully removed using

  17. Molten salt battery having inorganic paper separator

    DOE Patents [OSTI]

    Walker, Jr., Robert D.

    1977-01-01

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

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

    SciTech Connect (OSTI)

    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

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

    SciTech Connect (OSTI)

    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

  20. Project Profile: Molten Salt-Carbon Nanotube Thermal Storage | Department

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    of Energy Molten Salt-Carbon Nanotube Thermal Storage Project Profile: Molten Salt-Carbon Nanotube Thermal Storage TEES logo Texas Engineering Experiment Station (TEES), under the Thermal Storage FOA, created a composite thermal energy storage material by embedding nanoparticles in a molten salt base material. Approach Graphic of a chart with dots and horizontal lines. TEES measured the specific heat using modulated digital scanning calorimetry and created a system performance and economic

  1. Project Profile: Novel Molten Salts Thermal Energy Storage for

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Concentrating Solar Power Generation | Department of Energy Novel Molten Salts Thermal Energy Storage for Concentrating Solar Power Generation Project Profile: Novel Molten Salts Thermal Energy Storage for Concentrating Solar Power Generation Alabama logo The University of Alabama, under the Thermal Storage FOA, is developing thermal energy storage (TES) media consisting of low melting point (LMP) molten salt with high TES density for sensible heat storage systems. Approach They will conduct

  2. Accelerator-driven subcritical fission in molten salt core: Closing...

    Office of Scientific and Technical Information (OSTI)

    Accelerator-driven subcritical fission in molten salt core: Closing the nuclear fuel cycle for green nuclear energy Citation Details In-Document Search Title: Accelerator-driven ...

  3. Molten Salt-Carbon Nanotube Thermal Energy Storage for Concentrating...

    Office of Scientific and Technical Information (OSTI)

    Concentrating Solar Power Systems Final Report Citation Details In-Document Search Title: Molten Salt-Carbon Nanotube Thermal Energy Storage for Concentrating Solar Power Systems ...

  4. Molten salt bath circulation design for an electrolytic cell

    DOE Patents [OSTI]

    Dawless, R.K.; LaCamera, A.F.; Troup, R.L.; Ray, S.P.; Hosler, R.B.

    1999-08-17

    An electrolytic cell for reduction of a metal oxide to a metal and oxygen has an inert anode and an upwardly angled roof covering the inert mode. The angled roof diverts oxygen bubbles into an upcomer channel, thereby agitating a molten salt bath in the upcomer channel and improving dissolution of a metal oxide in the molten salt bath. The molten salt bath has a lower velocity adjacent the inert anode in order to minimize corrosion by substances in the bath. A particularly preferred cell produces aluminum by electrolysis of alumina in a molten salt bath containing aluminum fluoride and sodium fluoride. 4 figs.

  5. Molten salt bath circulation design for an electrolytic cell

    DOE Patents [OSTI]

    Dawless, Robert K.; LaCamera, Alfred F.; Troup, R. Lee; Ray, Siba P.; Hosler, Robert B.

    1999-01-01

    An electrolytic cell for reduction of a metal oxide to a metal and oxygen has an inert anode and an upwardly angled roof covering the inert mode. The angled roof diverts oxygen bubbles into an upcomer channel, thereby agitating a molten salt bath in the upcomer channel and improving dissolution of a metal oxide in the molten salt bath. The molten salt bath has a lower velocity adjacent the inert anode in order to minimize corrosion by substances in the bath. A particularly preferred cell produces aluminum by electrolysis of alumina in a molten salt bath containing aluminum fluoride and sodium fluoride.

  6. Project Profile: Novel Molten Salts Thermal Energy Storage for...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Novel Molten Salts Thermal Energy Storage for Concentrating Solar Power Generation Project ... They will conduct detailed tests using a laboratory-scale TES system to: Graphic of a ...

  7. MOLTEN PLUTONIUM FUELED FAST BREEDER REACTOR

    DOE Patents [OSTI]

    Kiehn, R.M.; King, L.D.P.; Peterson, R.E.; Swickard, E.O. Jr.

    1962-06-26

    A description is given of a nuclear fast reactor fueled with molten plutonium containing about 20 kg of plutonium in a tantalum container, cooled by circulating liquid sodium at about 600 to 650 deg C, having a large negative temperature coefficient of reactivity, and control rods and movable reflector for criticality control. (AEC)

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

    SciTech Connect (OSTI)

    Pattrick Calderoni

    2010-09-01

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

  9. Molten salt electrolyte battery cell with overcharge tolerance

    DOE Patents [OSTI]

    Kaun, Thomas D.; Nelson, Paul A.

    1989-01-01

    A molten salt electrolyte battery having an increased overcharge tolerance employs a negative electrode with two lithium alloy phases of different electrochemical potential, one of which allows self-discharge rates which permits battery cell equalization.

  10. Molten Salt Heat Transfer Fluid (HTF) - Energy Innovation Portal

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Find More Like This Return to Search Molten Salt Heat Transfer Fluid (HTF) Sandia National ... Sandia has developed a heat transfer fluid (HTF) for use at elevated temperatures that has ...

  11. 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

    SciTech Connect (OSTI)

    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.

  12. Treatment of plutonium process residues by molten salt oxidation

    SciTech Connect (OSTI)

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

    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.

  13. LIFE Materails: Molten-Salt Fuels Volume 8

    SciTech Connect (OSTI)

    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.

  14. Enhanced molten salt purification by electrochemical methods: feasibility experiments with flibe

    SciTech Connect (OSTI)

    Alan K Wertsching; Brandon S Grover; Pattrick Calderoni

    2010-09-01

    Molten salts are considered within the Very High Temperature Reactor program as heat transfer media because of their intrinsically favorable thermo-physical properties at temperatures starting from 300 C and extending up to 1200 C. In this context two main applications of molten salt are considered, both involving fluoride-based materials: as primary coolants for a heterogeneous fuel reactor core and as secondary heat transport medium to a helium power cycle for electricity generation or other processing plants, such as hydrogen production. The reference design concept here considered is the Advanced High Temperature Reactor (AHTR), which is a large passively safe reactor that uses solid graphite-matrix coated-particle fuel (similar to that used in gas-cooled reactors) and a molten salt primary and secondary coolant with peak temperatures between 700 and 1000 C, depending upon the application. However, the considerations included in this report apply to any high temperature system employing fluoride salts as heat transfer fluid, including intermediate heat exchangers for gas-cooled reactor concepts and homogenous molten salt concepts, and extending also to fast reactors, accelerator-driven systems and fusion energy systems. The 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 (INL/EXT-10-18297) highlighted how thermo-physical properties of the materials that directly impact the heat transfer behavior are strongly correlated to the of 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 titled An experimental test plan for the characterization of molten salt thermo

  15. Separation of actinides from lanthanides utilizing molten salt electrorefining

    SciTech Connect (OSTI)

    Grimmett, D.L.; Fusselman, S.P.; Roy, J.J.; Gay, R.L.; Krueger, C.L.; Storvick, T.S.; Inoue, T.; Hijikata, T.; Takahashi, N.

    1996-10-01

    TRUMP-S (TRansUranic Management through Pyropartitioning Separation) is a pyrochemical process being developed to separate actinides form fission products in nuclear waste. A key process step involving molten salt electrorefining to separate actinides from lanthanides has been studied on a laboratory scale. Electrorefining of U, Np, Pu, Am, and lanthanide mixtures from molten cadmium at 450 C to a solid cathode utilizing a molten chloride electrolyte resulted in > 99% removal of actinides from the molten cadmium and salt phases. Removal of the last few percent of actinides is accompanied by lowered cathodic current efficiency and some lanthanide codeposition. Actinide/lanthanide separation ratios on the cathode are ordered U > Np > Pu > Am and are consistent with predictions based on equilibrium potentials.

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

    SciTech Connect (OSTI)

    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.

  17. Assessment of the Use of Nitrogen Trifluoride for Purifying Coolant and Heat Transfer Salts in the Fluoride Salt-Cooled High-Temperature Reactor

    SciTech Connect (OSTI)

    Scheele, Randall D.; Casella, Andrew M.

    2010-09-28

    This report provides an assessment of the use of nitrogen trifluoride for removing oxide and water-caused contaminants in the fluoride salts that will be used as coolants in a molten salt cooled reactor.

  18. Molten-Salt-Based Growth of Group III Nitrides

    DOE Patents [OSTI]

    Waldrip, Karen E.; Tsao, Jeffrey Y.; Kerley, Thomas M.

    2008-10-14

    A method for growing Group III nitride materials using a molten halide salt as a solvent to solubilize the Group-III ions and nitride ions that react to form the Group III nitride material. The concentration of at least one of the nitride ion or Group III cation is determined by electrochemical generation of the ions.

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

    SciTech Connect (OSTI)

    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 ..gamma..-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.

  20. Corrosion of aluminides by molten nitrate salt

    SciTech Connect (OSTI)

    Tortorelli, P.F.; Bishop, P.S.

    1990-01-01

    The corrosion of titanium-, iron-, and nickel-based aluminides by a highly aggressive, oxidizing NaNO{sub 3}(-KNO{sub 3})-Na{sub 2}O{sub 2} has been studied at 650{degree}C. It was shown that weight changes could be used to effectively evaluate corrosion behavior in the subject nitrate salt environments provided these data were combined with salt analyses and microstructural examinations. The studies indicated that the corrosion of relatively resistant aluminides by these nitrate salts proceeded by oxidation and a slow release from an aluminum-rich product layer into the salt at rates lower than that associated with many other types of metallic materials. The overall corrosion process and resulting rate depended on the particular aluminide being exposed. In order to minimize corrosion of nickel or iron aluminides, it was necessary to have aluminum concentrations in excess of 30 at. %. However, even at a concentration of 50 at. % Al, the corrosion resistance of TiAl was inferior to that of Ni{sub 3}Al and Fe{sub 3}Al. At higher aluminum concentrations, iron, nickel, and iron-nickel aluminides exhibited quite similar weight changes, indicative of the principal role of aluminum in controlling the corrosion process in NaNO{sub 3}(-KNO{sub 3})-Na{sub 2}O{sub 2} salts. 20 refs., 5 figs., 3 tabs.

  1. Molten Salt Fuel Version of Laser Inertial Fusion Fission Energy (LIFE)

    SciTech Connect (OSTI)

    Moir, R W; Shaw, H F; Caro, A; Kaufman, L; Latkowski, J F; Powers, J; Turchi, P A

    2008-10-24

    Molten salt with dissolved uranium is being considered for the Laser Inertial Confinement Fusion Fission Energy (LIFE) fission blanket as a backup in case a solid-fuel version cannot meet the performance objectives, for example because of radiation damage of the solid materials. Molten salt is not damaged by radiation and therefore could likely achieve the desired high burnup (>99%) of heavy atoms of {sup 238}U. A perceived disadvantage is the possibility that the circulating molten salt could lend itself to misuse (proliferation) by making separation of fissile material easier than for the solid-fuel case. The molten salt composition being considered is the eutectic mixture of 73 mol% LiF and 27 mol% UF{sub 4}, whose melting point is 490 C. The use of {sup 232}Th as a fuel is also being studied. ({sup 232}Th does not produce Pu under neutron irradiation.) The temperature of the molten salt would be {approx}550 C at the inlet (60 C above the solidus temperature) and {approx}650 C at the outlet. Mixtures of U and Th are being considered. To minimize corrosion of structural materials, the molten salt would also contain a small amount ({approx}1 mol%) of UF{sub 3}. The same beryllium neutron multiplier could be used as in the solid fuel case; alternatively, a liquid lithium or liquid lead multiplier could be used. Insuring that the solubility of Pu{sup 3+} in the melt is not exceeded is a design criterion. To mitigate corrosion of the steel, a refractory coating such as tungsten similar to the first wall facing the fusion source is suggested in the high-neutron-flux regions; and in low-neutron-flux regions, including the piping and heat exchangers, a nickel alloy, Hastelloy, would be used. These material choices parallel those made for the Molten Salt Reactor Experiment (MSRE) at ORNL. The nuclear performance is better than the solid fuel case. At the beginning of life, the tritium breeding ratio is unity and the plutonium plus {sup 233}U production rate is {approx}0

  2. Molten salt processing of mixed wastes with offgas condensation

    SciTech Connect (OSTI)

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

    1991-05-13

    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.

  3. Nuclear Hybrid Energy Systems: Molten Salt Energy Storage

    SciTech Connect (OSTI)

    P. Sabharwall; M. Green; S.J. Yoon; S.M. Bragg-Sitton; C. Stoots

    2014-07-01

    With growing concerns in the production of reliable energy sources, the next generation in reliable power generation, hybrid energy systems, are being developed to stabilize these growing energy needs. The hybrid energy system incorporates multiple inputs and multiple outputs. The vitality and efficiency of these systems resides in the energy storage application. Energy storage is necessary for grid stabilizing and storing the overproduction of energy to meet peak demands of energy at the time of need. With high thermal energy production of the primary nuclear heat generation source, molten salt energy storage is an intriguing option because of its distinct properties. This paper will discuss the different energy storage options with the criteria for efficient energy storage set forth, and will primarily focus on different molten salt energy storage system options through a thermodynamic analysis

  4. Atomistic Adaptive Ensemble Calculations of Eutectics of Molten Salt

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Mixtures | Argonne Leadership Computing Facility Atomistic Adaptive Ensemble Calculations of Eutectics of Molten Salt Mixtures PI Name: Saivenkataraman Jayaraman PI Email: sjayara@sandia.gov Institution: Sandia National Laboratories Allocation Program: INCITE Allocation Hours at ALCF: 10,000,000 Year: 2012 Research Domain: Energy Technologies New and improved heat-transfer media with higher operating temperature ranges promise to turn solar-thermal power into a competitively cost-effective

  5. Technical review of Molten Salt Oxidation

    SciTech Connect (OSTI)

    Not Available

    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.

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

    SciTech Connect (OSTI)

    Powers, J

    2008-10-23

    The Laser Inertial Confinement Fusion Fission Energy (LIFE) Program being developed at Lawrence Livermore National Laboratory (LLNL) aims to design a hybrid fission-fusion subcritical nuclear engine that uses a laser-driven Inertial Confinement Fusion (ICF) system to drive a subcritical fission blanket. This combined fusion-fission hybrid system could be used for generating electricity, material transmutation or incineration, or other applications. LIFE does not require enriched fuel since it is a sub-critical system and LIFE can sustain power operation beyond the burnup levels at which typical fission reactors need to be refueled. In light of these factors, numerous options have been suggested and are being investigated. Options being investigated include fueling LIFE engines with spent nuclear fuel to aid in disposal/incineration of commercial spent nuclear fuel or using depleted uranium or thorium fueled options to enhance proliferation resistance and utilize non-fissile materials [1]. LIFE engine blanket designs using a molten salt fuel system represent one area of investigation. Possible applications of a LIFE engine with a molten salt blanket include uses as a spent nuclear fuel burner, fissile fuel breeding platform, and providing a backup alternative to other LIFE engine blanket designs using TRISO fuel particles in case the TRISO particles are found to be unable to withstand the irradiation they will be subjected to. These molten salts consist of a mixture of LiF with UF{sub 4} or ThF{sub 4} or some combination thereof. Future systems could look at using PuF{sub 3} or PuF{sub 4} as well, though no work on such system with initial plutonium loadings has been performed for studies documented in this report. The purpose of this report is to document preliminary neutronics design studies performed to support the development of a molten salt blanket LIFE engine option, as part of the LIFE Program being performed at Lawrence Livermore National laboratory

  7. Materials corrosion in molten LiF-NaF-KF eutectic salt under different reduction-oxidation conditions

    SciTech Connect (OSTI)

    Sellers, R. S.; Cheng, W. J.; Anderson, M. H.; Sridharan, K.; Wang, C. J.; Allen, T. R.

    2012-07-01

    Molten fluoride salts such as FLiNaK (LiF-NaF-KF: 46.5-11.5-42 mol %) have been proposed for use as secondary reactor coolants, media for transfer of high temperature process heat from nuclear reactors to chemical plants, and for concentrated solar power thermal energy storage. In molten fluoride salts, passive oxide films are chemically unstable, and corrosion is driven largely by the thermodynamically driven dissolution of alloying elements into the molten salt environment. Two alloys, Hastelloy{sup R} N and 316L stainless steel were exposed to molten FLiNaK salt in a 316L stainless steel crucible under argon cover gas for 1000 hours at 850 deg. C. Graphite was present in some of the crucibles with the goal of studying corrosion behavior of relevant reactor material combinations. In addition, a technique to reduce alloy corrosion through modification of the reduction-oxidation state was tested by the inclusion of zirconium to the system. Corrosion of 316L stainless steel was noted to occur primarily through surface depletion of chromium, an effect that was enhanced by the presence of graphite. Hastelloy{sup R} N experienced weight gain through electrochemical plating of corrosion products derived from the 316L stainless steel crucible. In the presence of zirconium, both alloys gained weight through plating of zirconium and as a result formed intermetallic layers. (authors)

  8. Molten metal reactor and method of forming hydrogen, carbon monoxide and carbon dioxide using the molten alkaline metal reactor

    DOE Patents [OSTI]

    Bingham, Dennis N.; Klingler, Kerry M.; Turner, Terry D.; Wilding, Bruce M.

    2012-11-13

    A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

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

    DOE Patents [OSTI]

    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.

  10. Project Profile: Long-Shafted Molten Salt Pump | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Long-Shafted Molten Salt Pump Project Profile: Long-Shafted Molten Salt Pump Pratt Whitney Rocketdyne logo Pratt & Whitney Rocketdyne (PWR), under the CSP R&D FOA, is validating the manufacturability of a large-scale molten salt receiver panel and then confirming its operation in prototypic solar flux. This work is an important step in reducing the LCOE from a central receiver solar power plant. Approach Image of PWR's design for an advanced molten salt receiver panel for a large

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

  13. Molten salt considerations for accelerator-driven subcritical fission to close the nuclear fuel cycle

    SciTech Connect (OSTI)

    Sooby, Elizabeth; Baty, Austin; Gerity, James; McIntyre, Peter; Melconian, Karie; Pogue, Nathaniel; Sattarov, Akhdiyor; Adams, Marvin; Tsevkov, Pavel; Phongikaroon, Supathorn; Simpson, Michael; Tripathy, Prabhat

    2013-04-19

    The host salt selection, molecular modeling, physical chemistry, and processing chemistry are presented here for an accelerator-driven subcritical fission in a molten salt core (ADSMS). The core is fueled solely with the transuranics (TRU) and long-lived fission products (LFP) from used nuclear fuel. The neutronics and salt composition are optimized to destroy the transuranics by fission and the long-lived fission products by transmutation. The cores are driven by proton beams from a strong-focusing cyclotron stack. One such ADSMS system can destroy the transuranics in the used nuclear fuel produced by a 1GWe conventional reactor. It uniquely provides a method to close the nuclear fuel cycle for green nuclear energy.

  14. Chemistry control and corrosion mitigation of heat transfer salts for the fluoride salt reactor (FHR)

    SciTech Connect (OSTI)

    Kelleher, B. C.; Sellers, S. R.; Anderson, M. H.; Sridharan, K.; Scheele, R. D.

    2012-07-01

    The Molten Salt Reactor Experiment (MSRE) was a prototype nuclear reactor which operated from 1965 to 1969 at Oak Ridge National Laboratory. The MSRE used liquid fluoride salts as a heat transfer fluid and solvent for fluoride based {sup 235}U and {sup 233}U fuel. Extensive research was performed in order to optimize the removal of oxide and metal impurities from the reactor's heat transfer salt, 2LiF-BeF{sub 2} (FLiBe). This was done by sparging a mixture of anhydrous hydrofluoric acid and hydrogen gas through the FLiBe at elevated temperatures. The hydrofluoric acid reacted with oxides and hydroxides, fluorinating them while simultaneously releasing water vapor. Metal impurities such as iron and chromium were reduced by hydrogen gas and filtered out of the salt. By removing these impurities, the corrosion of reactor components was minimized. The Univ. of Wisconsin - Madison is currently researching a new chemical purification process for fluoride salts that make use of a less dangerous cleaning gas, nitrogen trifluoride. Nitrogen trifluoride has been predicted as a superior fluorinating agent for fluoride salts. These purified salts will subsequently be used for static and loop corrosion tests on a variety of reactor materials to ensure materials compatibility for the new FHR designs. Demonstration of chemistry control methodologies along with potential reduction in corrosion is essential for the use of a fluoride salts in a next generator nuclear reactor system. (authors)

  15. Recent advances in the molten salt destruction of energetic materials

    SciTech Connect (OSTI)

    Pruneda, C. O., LLNL

    1996-09-01

    We have demonstrated the use of the Molten Salt Destruction (MSD) Process for destroying explosives, liquid gun propellant, and explosives-contaminated materials on a 1.5 kg of explosive/hr bench- scale unit (1, 2, 3, 4, 5). In our recently constructed 5 kg/hr pilot- scale unit we have also demonstrated the destruction of a liquid gun propellant and simulated wastes containing HMX (octogen). 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 metallic particles, are retained in the molten salt. The destruction of energetic materials waste is accomplished by introducing it, together with air, into a vessel containing molten salt (a eutectic mixture of sodium, potassium, and lithium carbonates). The following pure explosives have been destroyed in our bench-scale experimental unit located at Lawrence Livermore National Laboratory`s (LLNL) High Explosives Applications Facility (HEAF): ammonium picrate, HMX, K- 6 (keto-RDX), NQ, NTO, PETN, RDX, TATB, and TNT. In addition, the following compositions were also destroyed: Comp B, LX- IO, LX- 1 6, LX- 17, PBX-9404, and XM46 (liquid gun propellant). In this 1.5 kg/hr bench-scale unit, the fractions of carbon converted to CO and of chemically bound nitrogen converted to NO{sub x} were found to be well below 1%. In addition to destroying explosive powders and compositions we have also destroyed materials that are typical of residues which result from explosives operations. These include shavings from machined pressed parts of plastic-bonded explosives and sump waste containing both explosives and non-explosive debris. Based on the process data obtained on the bench-scale unit we designed and constructed a next-generation 5 kg/hr pilot-scale unit, incorporating LLNL`s advanced chimney design. The pilot unit has completed process implementation operations and explosives safety reviews. To date, in this

  16. Method of removal of heavy metal from molten salt in IFR fuel pyroprocessing

    DOE Patents [OSTI]

    Gay, Eddie C.

    1995-01-01

    An electrochemical method of separating heavy metal values from a radioactive molten salt including Li halide at temperatures of about 500.degree. C. The method comprises positioning a solid Li--Cd alloy anode in the molten salt containing the heavy metal values, positioning a Cd-containing cathode or a solid cathode positioned above a catch crucible in the molten salt to recover the heavy metal values, establishing a voltage drop between the anode and the cathode to deposit material at the cathode to reduce the concentration of heavy metals in the salt, and controlling the deposition rate at the cathode by controlling the current between the anode and cathode.

  17. Electrochemical behavior of simulated debris from a severe accident using a molten salt system

    SciTech Connect (OSTI)

    Takahashi, Yuya; Nakamura, Hitoshi; Yamada, Akira; Mizuguchi, Koji; Fujita, Reiko

    2013-07-01

    In a severe nuclear accident, the fuel in the reactor may melt, forming debris, which contains a UO{sub 2}-ZrO{sub 2} stable oxide mixture and parts of the reactor, such as Zircaloy and iron components. Proper handling of the debris is a critically important issue. The debris does not have the same composition as spent fuel, and so it is impossible to apply conventional reprocessing technology directly. In this study, we successfully separated Zr and Fe from simulated debris using NaCl-KCl molten salt electrolysis, and we selectively recovered the Zr and Fe. The simulated debris was made from Zr, Fe, and CeO{sub 2}. The CeO{sub 2} was used for simulating stable UO{sub 2}-ZrO{sub 2}. With this approach, it should be possible to reduce the volume of the debris by recovering metals, which can then be treated as low level radioactive wastes.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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

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

    SciTech Connect (OSTI)

    Kruizenga, Alan Michael; Crocker, Robert W.

    2012-09-01

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

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

    SciTech Connect (OSTI)

    Turchi, Craig; Kurup, Parthiv; Akar, Sertac; Flores, Francisco

    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.

  2. COMPUTATIONAL THERMODYNAMIC MODELING OF HOT CORROSION OF ALLLOYS HAYNES 242 AND HASTELLOYTMN FOR MOLTEN SALT SERVICE

    SciTech Connect (OSTI)

    Michael V. Glazoff; Piyush Sabharwall; Akira Tokuhiro

    2014-09-01

    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.

  3. Corrosion of Ferritic Steels in High Temperature Molten Salt Coolants for Nuclear Applications

    SciTech Connect (OSTI)

    Farmer, J; El-Dasher, B; de Caro, M S; Ferreira, J

    2008-11-25

    Corrosion of ferritic steels in high temperature molten fluoride salts may limit the life of advanced reactors, including some hybrid systems that are now under consideration. In some cases, the steel may be protected through galvanic coupling with other less noble materials with special neutronic properties such a beryllium. This paper reports the development of a model for predicting corrosion rates for various ferritic steels, with and without oxide dispersion strengthening, in FLiBe (Li{sub 2}BeF{sub 4}) and FLiNaK (Li-Na-K-F) coolants at temperatures up to 800 C. Mixed potential theory is used to account for the protection of steel by beryllium, Tafel kinetics are used to predict rates of dissolution as a function of temperature and potential, and the thinning of the mass-transfer boundary layer with increasing Reynolds number is accounted for with dimensionless correlations. The model also accounts for the deceleration of corrosion as the coolants become saturated with dissolved chromium and iron. This paper also reports electrochemical impedance spectroscopy of steels at their corrosion potentials in high-temperature molten salt environments, with the complex impedance spectra interpreted in terms of the interfacial charge transfer resistance and capacitance, as well as the electrolyte conductivity. Such in situ measurement techniques provide valuable insight into the degradation of materials under realistic conditions.

  4. Method for converting UF5 to UF4 in a molten fluoride salt

    DOE Patents [OSTI]

    Bennett, Melvin R.; Bamberger, Carlos E.; Kelmers, A. Donald

    1977-01-01

    The reduction of UF.sub.5 to UF.sub.4 in a molten fluoride salt by sparging with hydrogen is catalyzed by metallic platinum. The reaction is also catalyzed by platinum alloyed with gold reaction equipment.

  5. Viscosity of multi-component molten nitrate salts : liquidus to 200 degrees C.

    SciTech Connect (OSTI)

    Bradshaw, Robert W.

    2010-03-01

    The viscosity of molten salts comprising ternary and quaternary mixtures of the nitrates of sodium, potassium, lithium and calcium was determined experimentally. Viscosity was measured over the temperature range from near the relatively low liquidus temperatures of he individual mixtures to 200C. Molten salt mixtures that do not contain calcium nitrate exhibited relatively low viscosity and an Arrhenius temperature dependence. Molten salt mixtures that contained calcium nitrate were relatively more viscous and viscosity increased as the roportion of calcium nitrate increased. The temperature dependence of viscosity of molten salts containing calcium nitrate displayed curvature, rather than linearity, when plotted in Arrhenius format. Viscosity data for these mixtures were correlated by the Vogel-Fulcher- ammann-Hesse equation.

  6. Sandia-AREVA Commission Solar Thermal/Molten Salt Energy-Storage...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AREVA Commission Solar ThermalMolten Salt Energy-Storage Demonstration - Sandia Energy Energy ... for storage, and removes the need for two sets of heat-exchangers in the system. ...

  7. Electrical double layers and differential capacitance in molten salts from density functional theory

    SciTech Connect (OSTI)

    Frischknecht, Amalie L.; Halligan, Deaglan O.; Parks, Michael L.

    2014-08-05

    Classical density functional theory (DFT) is used to calculate the structure of the electrical double layer and the differential capacitance of model molten salts. The DFT is shown to give good qualitative agreement with Monte Carlo simulations in the molten salt regime. The DFT is then applied to three common molten salts, KCl, LiCl, and LiKCl, modeled as charged hard spheres near a planar charged surface. The DFT predicts strong layering of the ions near the surface, with the oscillatory density profiles extending to larger distances for larger electrostatic interactions resulting from either lower temperature or lower dielectric constant. In conclusion, overall the differential capacitance is found to be bell-shaped, in agreement with recent theories and simulations for ionic liquids and molten salts, but contrary to the results of the classical Gouy-Chapman theory.

  8. Electrical double layers and differential capacitance in molten salts from density functional theory

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Frischknecht, Amalie L.; Halligan, Deaglan O.; Parks, Michael L.

    2014-08-05

    Classical density functional theory (DFT) is used to calculate the structure of the electrical double layer and the differential capacitance of model molten salts. The DFT is shown to give good qualitative agreement with Monte Carlo simulations in the molten salt regime. The DFT is then applied to three common molten salts, KCl, LiCl, and LiKCl, modeled as charged hard spheres near a planar charged surface. The DFT predicts strong layering of the ions near the surface, with the oscillatory density profiles extending to larger distances for larger electrostatic interactions resulting from either lower temperature or lower dielectric constant. Inmore » conclusion, overall the differential capacitance is found to be bell-shaped, in agreement with recent theories and simulations for ionic liquids and molten salts, but contrary to the results of the classical Gouy-Chapman theory.« less

  9. Molten Metal Treatment by Salt Fluxing with Low Environmental Emissions

    SciTech Connect (OSTI)

    Yogeshwar Sahai

    2007-07-31

    Abstract: Chlorine gas is traditionally used for fluxing of aluminum melt for removal of alkali and alkaline earth elements. However this results in undesirable emissions of particulate matter and gases such as HCl and chlorine, which are often at unacceptable levels. Additionally, chlorine gas is highly toxic and its handling, storage, and use pose risks to employees and the local community. Holding of even minimal amounts of chlorine necessitates extensive training for all plant employees. Fugitive emissions from chlorine usage within the plant cause accelerated corrosion of plant equipment. The Secondary Aluminum Maximum Achievable Control Technology (MACT) under the Clean Air Act, finalized in March 2000 has set very tough new limits on particulate matter (PM) and total hydrogen chloride emissions from aluminum melting and holding furnaces. These limits are 0.4 and 0.1 lbs per ton of aluminum for hydrogen chloride and particulate emissions, respectively. Assuming new technologies for meeting these limits can be found, additional requirements under the Clean Air Act (Prevention of Significant Deterioration and New Source Review) trigger Best Available Control Technology (BACT) for new sources with annual emissions (net emissions not expressed per ton of production) over specified amounts. BACT currently is lime coated bag-houses for control of particulate and HCl emissions. These controls are expensive, difficult to operate and maintain, and result in reduced American competitiveness in the global economy. Solid salt fluxing is emerging as a viable option for the replacement of chlorine gas fluxing, provided emissions can be consistently maintained below the required levels. This project was a cooperative effort between the Ohio State University and Alcoa to investigate and optimize the effects of solid chloride flux addition in molten metal for alkali impurity and non-metallic inclusion removal minimizing dust and toxic emissions and maximizing energy

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

    SciTech Connect (OSTI)

    Reddy, Ramana G.

    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 go

  11. Convective heat transfer in the laminar-turbulent transition region with molten salt in a circular tube

    SciTech Connect (OSTI)

    Yu-ting, Wu; Bin, Liu; Chong-fang, Ma; Hang, Guo [Key Laboratory of Enhanced Heat Transfer and Energy Conservation, Ministry of Education and Key Laboratory of Heat Transfer and Energy Conversion, Beijing municipality, College of Environmental and Energy Engineering, Beijing University of Technology, Beijing 100022 (China)

    2009-10-15

    In order to understand the heat transfer characteristics of molten salt and testify the validity of the well-known empirical convective heat transfer correlations, experimental study on transition convective heat transfer with molten salt in a circular tube was conducted. Molten salt circulations were realized and operated in a specially designed system over 1000 h. The average forced convective heat transfer coefficients of molten salt were determined by least-squares method based on the measured data of flow rates and temperatures. Finally, a heat transfer correlation of transition flow with molten salt in a circular tube was obtained and good agreement was observed between the experimental data of molten salt and the well-known correlations presented by Hausen and Gnielinski, respectively. (author)

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

    DOE Patents [OSTI]

    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.

  13. Method of removal of heavy metal from molten salt in IFR fuel pyroprocessing

    DOE Patents [OSTI]

    Gay, E.C.

    1995-10-03

    An electrochemical method is described for separating heavy metal values from a radioactive molten salt including Li halide at temperatures of about 500 C. The method comprises positioning a solid Li--Cd alloy anode in the molten salt containing the heavy metal values, positioning a Cd-containing cathode or a solid cathode positioned above a catch crucible in the molten salt to recover the heavy metal values, establishing a voltage drop between the anode and the cathode to deposit material at the cathode to reduce the concentration of heavy metals in the salt, and controlling the deposition rate at the cathode by controlling the current between the anode and cathode. 3 figs.

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

    SciTech Connect (OSTI)

    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.

  15. Molten-Salt Batteries for Medium and Large-Scale Energy Storage

    SciTech Connect (OSTI)

    Lu, Xiaochuan; Yang, Zhenguo

    2014-12-01

    This chapter discusses two types of molten salt batteries. Both of them are based on a beta-alumina solid electrolyte and molten sodium anode, i.e., sodium-sulfur (Na-S) battery and sodium-metal halide (ZEBRA) batteries. The chapter first reviews the basic electrochemistries and materials for various battery components. It then describes the performance of state-of-the-art batteries and future direction in material development for these batteries.

  16. Reversible electro-optic device employing aprotic molten salts and method

    DOE Patents [OSTI]

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

    2008-01-08

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

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

    DOE Patents [OSTI]

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

    2005-03-01

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

  18. Process for recovering tritium from molten lithium metal

    DOE Patents [OSTI]

    Maroni, Victor A.

    1976-01-01

    Lithium tritide (LiT) is extracted from molten lithium metal that has been exposed to neutron irradiation for breeding tritium within a thermonuclear or fission reactor. The extraction is performed by intimately contacting the molten lithium metal with a molten lithium salt, for instance, lithium chloride - potassium chloride eutectic to distribute LiT between the salt and metal phases. The extracted tritium is recovered in gaseous form from the molten salt phase by a subsequent electrolytic or oxidation step.

  19. Development of Molten-Salt Heat Trasfer Fluid Technology for Parabolic Trough Solar Power Plants

    Broader source: Energy.gov [DOE]

    "This PowerPoint presentation was originally given by Dylan Grogan, principal investigator at Abengoa Solar, during a SunShot Initiative Concentrating Solar Power program review on April 24, 2013. The project, Development of Molten-Salt Heat Transfer Fluid Technology for Parabolic Trough Solar Power Plants, seeks to determine whether the inorganic fluids (molten salts) offer a sufficient reduction in levelized energy costs to pursue further development, and to develop the components required for their use. The presentation focuses on presenting conclusions from Phase 1 of the program and looks ahead to review Phase 2 activities."

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

    SciTech Connect (OSTI)

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

    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.

  1. Effect of composition on the density of multi-component molten nitrate salts.

    SciTech Connect (OSTI)

    Bradshaw, Robert W.

    2009-12-01

    The density of molten nitrate salts was measured to determine the effects of the constituents on the density of multi-component mixtures. The molten salts consisted of various proportions of the nitrates of potassium, sodium, lithium and calcium. Density measurements ere performed using an Archimedean method and the results were compared to data reported in the literature for the individual constituent salts or simple combinations, such as the binary Solar Salt mixture of NaNO3 and KNO3. The addition of calcium nitrate generally ncreased density, relative to potassium nitrate or sodium nitrate, while lithium nitrate decreased density. The temperature dependence of density is described by a linear equation regardless of composition. The molar volume, and thereby, density of multi-component mixtures an be calculated as a function of temperature using a linear additivity rule based on the properties of the individual constituents.

  2. Oxygen production by molten alkali metal salts using multiple absorption-desorption cycles

    DOE Patents [OSTI]

    Cassano, Anthony A.

    1985-01-01

    A continuous chemical air separation is performed wherein oxygen is recovered with a molten alkali metal salt oxygen acceptor in a series of absorption zones which are connected to a plurality of desorption zones operated in separate parallel cycles with the absorption zones. A greater recovery of high pressure oxygen is achieved at reduced power requirements and capital costs.

  3. Oxygen production by molten alkali metal salts using multiple absorption-desorption cycles

    DOE Patents [OSTI]

    Cassano, A.A.

    1985-07-02

    A continuous chemical air separation is performed wherein oxygen is recovered with a molten alkali metal salt oxygen acceptor in a series of absorption zones which are connected to a plurality of desorption zones operated in separate parallel cycles with the absorption zones. A greater recovery of high pressure oxygen is achieved at reduced power requirements and capital costs. 3 figs.

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

    SciTech Connect (OSTI)

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

    1995-02-01

    Accelerator-driven transmutation technology (ADTT) refers to a concept for a system that uses a blanket assembly driven by a source of neutrons produced when high-energy protons from an accelerator strike a heavy metal target. One application for such a system is called Accelerator-Based Plutonium Conversion, or ABC. Currently, the version of this concept being proposed by the Los Alamos National Laboratory features a liquid lead target material and a blanket fuel of molten fluorides that contain plutonium. Thus, the materials to be used in such a system must have, in addition to adequate mechanical strength, corrosion resistance to molten lead, corrosion resistance to molten fluoride salts, and resistance to radiation damage. In this report the corrosion properties of liquid lead and the LiF-BeF{sub 2} molten salt system are reviewed in the context of candidate materials for the above application. Background information has been drawn from extensive past studies. The system operating temperature, type of protective environment, and oxidation potential of the salt are shown to be critical design considerations. Factors such as the generation of fission products and transmutation of salt components also significantly affect corrosion behavior, and procedures for inhibiting their effects are discussed. In view of the potential for extreme conditions relative to neutron fluxes and energies that can occur in an ADTT, a knowledge of radiation effects is a most important factor. Present information for potential materials selections is summarized.

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

    SciTech Connect (OSTI)

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

    1995-03-01

    Accelerator-driven transmutation technology (ADTT) refers to a concept for a system that uses a blanket assembly driven by a source of neutrons produced when high-energy protons from an accelerator strike a heavy metal target. One application for such a system is called Accelerator-Based Plutonium Conversion, or ABC. Currently, the version of this concept being proposed by the Los Alamos National Laboratory features a liquid lead target material and a blanket fuel of molten fluorides that contain plutonium. Thus, the materials to be used in such a system must have, in addition to adequate mechanical strength, corrosion resistance to molten lead, corrosion resistance to molten fluoride salts, and resistance to radiation damage. In this report the corrosion properties of liquid lead and the LiF-BeF{sub 2} molten salt system are reviewed in the context of candidate materials for the above application. Background information has been drawn from extensive past studies. The system operating temperature, type of protective environment, and oxidation potential of the salt are shown to be critical design considerations. Factors such as the generation of fission products and transmutation of salt components also significantly affect corrosion behavior, and procedures for inhibiting their effects are discussed. In view of the potential for extreme conditions relative to neutron fluxes and energies that can occur in an ADTT, a knowledge of radiation effects is a most important factor. Present information for potential materials selections is summarized.

  6. Uncertainty Studies of Real Anode Surface Area in Computational Analysis for Molten Salt Electrorefining

    SciTech Connect (OSTI)

    Sungyeol Choi; Jaeyeong Park; Robert O. Hoover; Supathorn Phongikaroon; Michael F. Simpson; Kwang-Rag Kim; Il Soon Hwang

    2011-09-01

    This study examines how much cell potential changes with five differently assumed real anode surface area cases. Determining real anode surface area is a significant issue to be resolved for precisely modeling molten salt electrorefining. Based on a three-dimensional electrorefining model, calculated cell potentials compare with an experimental cell potential variation over 80 hours of operation of the Mark-IV electrorefiner with driver fuel from the Experimental Breeder Reactor II. We succeeded to achieve a good agreement with an overall trend of the experimental data with appropriate selection of a mode for real anode surface area, but there are still local inconsistencies between theoretical calculation and experimental observation. In addition, the results were validated and compared with two-dimensional results to identify possible uncertainty factors that had to be further considered in a computational electrorefining analysis. These uncertainty factors include material properties, heterogeneous material distribution, surface roughness, and current efficiency. Zirconium's abundance and complex behavior have more impact on uncertainty towards the latter period of electrorefining at given batch of fuel. The benchmark results found that anode materials would be dissolved from both axial and radial directions at least for low burn-up metallic fuels after active liquid sodium bonding was dissolved.

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

    SciTech Connect (OSTI)

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

    2012-04-01

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

  8. Development of Molten-Salt Heat Trasfer Fluid Technology for...

    Office of Environmental Management (EM)

    molt en-salt HTF CSP plant show an LCOE of below 0.12kWhe (real 2009 ), w it h a 10% ITC Innovat ive Technology Solut ions f or Sustainability ABENGOA SOLAR Phase 1 Conclusions ...

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

    DOE Patents [OSTI]

    Willit, James L.

    2010-09-21

    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.

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

    DOE Patents [OSTI]

    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.

  11. Temperature effect on laser-induced breakdown spectroscopy spectra of molten and solid salts

    SciTech Connect (OSTI)

    Cynthia Hanson; Supathorn Phongikaroon; Jill R. Scott

    2014-07-01

    Laser-induced breakdown spectroscopy (LIBS) has been investigated as a potential analytical tool to improve operations and safeguards for electrorefiners, such as those used in processing spent nuclear fuel. This study set out to better understand the effect of sample temperature and physical state on LIBS spectra of molten and solid salts by building calibration curves of cerium and assessing self-absorption, plasma temperature, electron density, and local thermal equilibrium (LTE). Samples were composed of a LiCl–KCl eutectic salt, an internal standard of MnCl2, and varying concentrations of CeCl3 (0.1, 0.3, 0.5, 0.8, and 1.0 wt.% Ce) under different temperatures (773, 723, 673, 623, and 573 K). Analysis of salts in their molten form is preferred as plasma plumes from molten samples experienced less self-absorption, less variability in plasma temperature, and higher clearance of the minimum electron density required for local thermal equilibrium. These differences are attributed to plasma dynamics as a result of phase changes. Spectral reproducibility was also better in the molten state due to sample homogeneity.

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

    SciTech Connect (OSTI)

    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.

  13. Expedited demonstration of molten salt mixed waste treatment technology. Final report

    SciTech Connect (OSTI)

    1995-02-02

    This final report discusses the molten salt mixed waste project in terms of the various subtasks established. Subtask 1: Carbon monoxide emissions; Establish a salt recycle schedule and/or a strategy for off-gas control for MWMF that keeps carbon monoxide emission below 100 ppm on an hourly averaged basis. Subtask 2: Salt melt viscosity; Experiments are conducted to determine salt viscosity as a function of ash composition, ash concentration, temperature, and time. Subtask 3: Determine that the amount of sodium carbonate entrained in the off-gas is minimal, and that any deposited salt can easily be removed form the piping using a soot blower or other means. Subtask 4: The provision of at least one final waste form that meets the waste acceptance criteria of a landfill that will take the waste. This report discusses the progress made in each of these areas.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    Blanchard, T.P. Jr.

    1992-12-01

    Conductance and EMF studies of CaCl{sub 2}-based melts were performed in the temperature range 790--990 C. Conductivity data collected using magnesia tubes and capillaries showed deviations from the data recommended by the National Bureau of Standards. These deviations are attributed to the slow dissolution of magnesia by the CaCl{sub 2}-CaO melt. Conductivity data for molten CaCl{sub 2} using a pyrolytic boron nitride capillary were in reasonable agreement with the recommended data; however, undissolved CaO in CaCl{sub 2} may have caused blockage of the pyrolytic boron nitride capillary, resulting in fluctuations in the measured resistance. The utility of the AgCl/Ag reference electrode in CaCl{sub 2}-AgCl and CaCl{sub 2}-CaO-AgCl melts, using asbestos diaphragms and Vycor glass as reference half-cell membranes, was also investigated. Nernstian behavior was observed using both types of reference half-cell membranes in CaCl{sub 2}-AgCl melts. The AgCl/Ag reference electrode also exhibited Nernstian behavior in CaCl{sub 2}-CaO-AgCl melts using a Vycor reference half-cell membrane and a magnesia crucible. The use of CaCl{sub 2} as a solvent is of interest since it is used in plutonium metal purification, as well as various other commercial applications. 97 refs., 33 figs., 13 tabs.

  16. Design considerations for concentrating solar power tower systems employing molten salt.

    SciTech Connect (OSTI)

    Moore, Robert Charles; Siegel, Nathan Phillip; Kolb, Gregory J.; Vernon, Milton E.; Ho, Clifford Kuofei

    2010-09-01

    The Solar Two Project was a United States Department of Energy sponsored project operated from 1996 to 1999 to demonstrate the coupling of a solar power tower with a molten nitrate salt as a heat transfer media and for thermal storage. Over all, the Solar Two Project was very successful; however many operational challenges were encountered. In this work, the major problems encountered in operation of the Solar Two facility were evaluated and alternative technologies identified for use in a future solar power tower operating with a steam Rankine power cycle. Many of the major problems encountered can be addressed with new technologies that were not available a decade ago. These new technologies include better thermal insulation, analytical equipment, pumps and values specifically designed for molten nitrate salts, and gaskets resistant to thermal cycling and advanced equipment designs.

  17. Modeling and analysis of a molten salt electrowinning system with liquid cadmium cathode

    SciTech Connect (OSTI)

    Kim, K.R.; Ahn, D.H.; Paek, S.; Kwon, S.W.; Kim, S.H.; Shim, J.B.; Chung, H.; Kim, E.H.

    2007-07-01

    In the present work, an electrowinning process in the LiCl-KCl/Cd system is considered to model and analyze the equilibrium behavior and electro-transport of the actinide and rare-earth elements. Equilibrium distributions of the actinide and rare-earth elements in a molten salt and liquid cadmium system have been estimated for an infinite potentiostatic electrolysis from the thermodynamic data and material balance. A simple dynamic modeling of this process was performed by taking into account the material balances and diffusion-controlled electrochemical reactions in a diffusion layer at an electrode interface between the molten salt and liquid cadmium cathode. This model demonstrated a prediction of the concentration behaviors, a faradic current of each element and an electrochemical potential as function of the time up to the corresponding electro-transport satisfying a given applied current based on a galvano-static electrolysis. (authors)

  18. Sandia-AREVA Commission Solar Thermal/Molten Salt Energy-Storage

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Demonstration AREVA Commission Solar Thermal/Molten Salt Energy-Storage Demonstration - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery

  19. Customer Interface Document for the Molten Salt Test Loop at the NSTTF

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Customer Interface Document for the Molten Salt Test Loop at the NSTTF - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Energy

  20. An Evaluation of Molten-Salt Power Towers Including Results of the Solar Two Project

    SciTech Connect (OSTI)

    REILLY, HUGH E.; KOLB, GREGORY J.

    2001-11-01

    This report utilizes the results of the Solar Two project, as well as continuing technology development, to update the technical and economic status of molten-salt power towers. The report starts with an overview of power tower technology, including the progression from Solar One to the Solar Two project. This discussion is followed by a review of the Solar Two project--what was planned, what actually occurred, what was learned, and what was accomplished. The third section presents preliminary information regarding the likely configuration of the next molten-salt power tower plant. This section draws on Solar Two experience as well as results of continuing power tower development efforts conducted jointly by industry and Sandia National Laboratories. The fourth section details the expected performance and cost goals for the first commercial molten-salt power tower plant and includes a comparison of the commercial performance goals to the actual performance at Solar One and Solar Two. The final section summarizes the successes of Solar Two and the current technology development activities. The data collected from the Solar Two project suggest that the electricity cost goals established for power towers are reasonable and can be achieved with some simple design improvements.

  1. Thermal analysis of solar thermal energy storage in a molten-salt thermocline

    SciTech Connect (OSTI)

    Yang, Zhen; Garimella, Suresh V.

    2010-06-15

    A comprehensive, two-temperature model is developed to investigate energy storage in a molten-salt thermocline. The commercially available molten salt HITEC is considered for illustration with quartzite rocks as the filler. Heat transfer between the molten salt and quartzite rock is represented by an interstitial heat transfer coefficient. Volume-averaged mass and momentum equations are employed, with the Brinkman-Forchheimer extension to the Darcy law used to model the porous-medium resistance. The governing equations are solved using a finite-volume approach. The model is first validated against experiments from the literature and then used to systematically study the discharge behavior of thermocline thermal storage system. Thermal characteristics including temperature profiles and discharge efficiency are explored. Guidelines are developed for designing solar thermocline systems. The discharge efficiency is found to be improved at small Reynolds numbers and larger tank heights. The filler particle size strongly influences the interstitial heat transfer rate, and thus the discharge efficiency. (author)

  2. A view of treatment process of melted nuclear fuel on a severe accident plant using a molten salt system

    SciTech Connect (OSTI)

    Fujita, R.; Takahashi, Y.; Nakamura, H.; Mizuguchi, K.; Oomori, T.

    2013-07-01

    At severe accident such as Fukushima Daiichi Nuclear Power Plant Accident, the nuclear fuels in the reactor would melt and form debris which contains stable UO2-ZrO2 mixture corium and parts of vessel such as zircaloy and iron component. The requirements for solution of issues are below; -) the reasonable treatment process of the debris should be simple and in-situ in Fukushima Daiichi power plant, -) the desirable treatment process is to take out UO{sub 2} and PuO{sub 2} or metallic U and TRU metal, and dispose other fission products as high level radioactive waste; and -) the candidate of treatment process should generate the smallest secondary waste. Pyro-process has advantages to treat the debris because of the high solubility of the debris and its total process feasibility. Toshiba proposes a new pyro-process in molten salts using electrolysing Zr before debris fuel being treated.

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

    SciTech Connect (OSTI)

    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 600%C2%B0C were evaluated, which resulted in main steam temperatures of 553 and 580%C2%B0C, respectively. Also, the effects of final feedwater temperature (between 260 and 320%C2%B0C) were evaluated, which impacted the cold salt return temperature. The annual energy production and levelized cost of energy (LCOE) were calculated using the System Advisory Model on 165 MWe subcritical plants (baseline and advanced) and the most promising supercritical plants. It was concluded that the supercritical steam plants produced more annual energy than the baseline subcritical steam plant for the same-size heliostat field, receiver, and thermal storage system. Two supercritical steam plants had the highest annual performance and had nearly the same LCOE. Both operated at 230 bar-a main steam pressure. One was designed for a hot salt temperature of 600%C2%B0C and the other 565%C2%B0C. The LCOEs for these plants were about 10% lower than the baseline subcritical plant operating at 120 bar-a main steam pressure and a hot salt temperature of 565%C2%B0C. Based on the results of this study, it appears economically and technically feasible to incorporate supercritical steam turbines in molten-salt power tower plants.

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

    SciTech Connect (OSTI)

    Michael McDowell; Alan Schwartz

    2010-03-31

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

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

    SciTech Connect (OSTI)

    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. Customer interface document for the Molten Salt Test Loop (MSTL) system.

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

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

    2011-09-20

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

  10. Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control, Corrosion Mitigation, and Modeling

    SciTech Connect (OSTI)

    Anderson, Mark; Sridharan, Kumar; Morgan, Dane; Peterson, Per; Calderoni, Pattrick; Scheele, Randall; Casekka, Andrew; McNamara, Bruce

    2015-01-22

    The concept of a molten salt reactor has existed for nearly sixty years. Previously all work was done during a large collaborative effort at Oak Ridge National Laboratory, culminating in a research reactor which operated for 15,000 hours without major error. This technical success has garnished interest in modern, high temperature, reactor schemes. Research using molten fluoride salts for nuclear applications requires a steady supply of high grade molten salts. There is no bulk supplier of research grade fluoride salts in the world, so a facility which could provide all the salt needed for testing at the University of Wisconsin had to be produced. Two salt purification devices were made for this purpose, a large scale purifier, and a small scale purifier, each designed to clean the salts from impurities and reduce their corrosion potential. As of now, the small scale has performed with flibe salt, hydrogen, and hydrogen fluoride, yielding clean salt. This salt is currently being used in corrosion testing facilities at the Massachusetts Institute of Technology and the University of Wisconsin. Working with the beryllium based salts requires extensive safety measures and health monitoring to prevent the development of acute or chronic beryllium disease, two pulmonary diseases created by an allergic reaction to beryllium in the lungs. Extensive health monitoring, engineering controls, and environment monitoring had to be set up with the University of Wisconsin department of Environment, Health and Safety. The hydrogen fluoride required for purification was also an extreme health hazard requiring thoughtful planning and execution. These dangers have made research a slow and tedious process. Simple processes, such as chemical handling and clean-up, can take large amounts of ingenuity and time. Other work has complemented the experimental research at Wisconsin to advance high temperature reactor goals. Modeling work has been performed in house to re

  11. Corrosion resistance of stainless steels during thermal cycling in alkali nitrate molten salts.

    SciTech Connect (OSTI)

    Bradshaw, Robert W.; Goods, Steven Howard

    2001-09-01

    The corrosion behavior of three austenitic stainless steels was evaluated during thermal cycling in molten salt mixtures consisting of NaNO{sub 3} and KNO{sub 3}. Corrosion tests were conducted with Types 316, 316L and 304 stainless steels for more than 4000 hours and 500 thermal cycles at a maximum temperature of 565 C. Corrosion rates were determined by chemically descaling coupons. Metal losses ranged from 5 to 16 microns and thermal cycling resulted in moderately higher corrosion rates compared to isothermal conditions. Type 316 SS was somewhat more corrosion resistant than Type 304 SS in these tests. The effect of carbon content on corrosion resistance was small, as 316L SS corroded only slightly slower than 316 SS. The corrosion rates increased as the dissolved chloride content of the molten salt mixtures increased. Chloride concentrations approximating 1 wt.%, coupled with thermal cycling, resulted in linear weight loss kinetics, rather than parabolic kinetics, which described corrosion rates for all other conditions. Optical microscopy and electron microprobe analysis revealed that the corrosion products consisted of iron-chromium spinel, magnetite, and sodium ferrite, organized as separate layers. Microanalysis of the elemental composition of the corrosion products further demonstrated that the chromium content of the iron-chromium spinel layer was relatively high for conditions in which parabolic kinetics were observed. However, linear kinetics were observed when the spinel layer contained relatively little chromium.

  12. Electrochemistry of LiCl-Li2O-H2O Molten Salt Systems

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    Chang, Jui-Cheng; Sun, I-Wen; Yang, Cheng-Hsien; Yang, Hao-Hsun; Hsueh, Mao-Lin; Ho, Wen-Yueh; Chang, Jia-Yaw

    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)

  14. Product Recovery from HTGR Reactor Fuel Processing Salt Official...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Demonstration of Fuel and Fission Product Recovery from HTGR Reactor Fuel Processing Salt ... HTGR, MST, CST Retention: Permanent Demonstration of Fuel and Fission Product Recovery ...

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

    SciTech Connect (OSTI)

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

    2013-04-19

    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.

  16. Molten Salt-Carbon Nanotube Thermal Energy Storage for Concentrating Solar Power Systems Final Report

    SciTech Connect (OSTI)

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

    SciTech Connect (OSTI)

    Mathur, Anoop

    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

  18. Liquid fuel molten salt reactors for thorium utilization (Journal...

    Office of Scientific and Technical Information (OSTI)

    removal of fission products as well as introduction of fissile fuel and fertile materials ... Country of Publication: United States Language: English Subject: 21 SPECIFIC NUCLEAR ...

  19. I-NERI ANNUAL TECHNICAL PROGRESS REPORT: 2006-002-K, Separation of Fission Products from Molten LiCl-KCl Salt Used for Electrorefining of Metal Fuels

    SciTech Connect (OSTI)

    S. Frank

    2009-09-01

    An attractive alternative to the once-through disposal of electrorefiner salt is to selectively remove the active fission products from the salt and recycle the salt back to the electrorefiner (ER). This would allow salt reuse for some number of cycles before ultimate disposal of the salt in a ceramic waste form. Reuse of ER salt would, thus, greatly reduce the volume of ceramic waste produced during the pyroprocessing of spent nuclear fuel. This final portion of the joint I-NERI research project is to demonstrate the separation of fission products from molten ER salt by two methods previously selected during phase two (FY-08) of this project. The two methods selected were salt/zeolite contacting and rare-earth fission product precipitation by oxygen bubbling. The ER salt used in these tests came from the Mark-IV electrorefiner used to anodically dissolved driver fuel from the EBR-II reactor on the INL site. The tests were performed using the Hot Fuel Dissolution Apparatus (HFDA) located in the main cell of the Hot Fuels Examination Facility (HFEF) at the Materials and Fuels complex on the INL site. Results from these tests were evaluated during a joint meeting of KAERI and INL investigators to provide recommendations as to the future direction of fission product removal from electrorefiner salt that accumulate during spent fuel treatment. Additionally, work continued on kinetic measurements of surrogate quaternary salt systems to provide fundamental kinetics on the ion exchange system and to expand the equilibrium model system developed during the first two phases of this project. The specific objectives of the FY09 I-NERI research activities at the INL include the following: Perform demonstration tests of the selected KAERI precipitation and INL salt/zeolite contacting processes for fission product removal using radioactive, fission product loaded ER salt Continue kinetic studies of the quaternary Cs/Sr-LiCl-KCl system to determine the rate of ion exchange

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

    SciTech Connect (OSTI)

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

    2014-06-01

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

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

    SciTech Connect (OSTI)

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

    2012-07-01

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

  2. Preliminary study of the electrolysis of aluminum sulfide in molten salts

    SciTech Connect (OSTI)

    Minh, N.Q.; Loutfy, R.O.; Yao, N.P.

    1983-02-01

    A preliminary laboratory-scale study of the electrolysis of aluminum sulfide in molten salts investigated the (1) solubility of Al/sub 2/S/sub 3/ in molten salts, (2) electrochemical behavior of Al/sub 2/S/sub 3/, and (3) electrolysis of Al/sub 2/S/sub 3/ with the determination of current efficiency as a function of current density. The solubility measurements show that MgCl/sub 2/-NaCl-KCl eutectic electrolyte at 1023 K can dissolve up to 3.3 mol % sulfide. The molar ratio of sulfur to aluminum in the eutectic is about one, which suggests that some sulfur remains undissolved, probably in the form of MgS. The experimental data and thermodynamic calculations suggest that Al/sub 2/S/sub 3/ dissolves in the eutectic to form AlS/sup +/ species in solution. Addition of AlCl/sub 3/ to the eutectic enhances the solubility of Al/sub 2/S/sub 3/; the solubility increases with increasing AlCl/sub 3/ concentration. The electrode reaction mechanism for the electrolysis of Al/sub 2/S/sub 3/ was elucidated by using linear sweep voltammetry. The cathodic reduction of aluminum-ion-containing species to aluminum proceeds by a reversible, diffusion-controlled, three-electron reaction. The anodic reaction involves the two-electron discharge of sulfide-ion-containing species, followed by the fast dimerization of sulfur atoms to S/sub 2/. Electrolysis experiments show that Al/sub 2/S/sub 3/ dissolved in molten MgCl/sub 2/-NaCl-KCl eutectic or in eutectic containing AlCl/sub 3/ can be electrolyzed to produce aluminum and sulfur. In the eutectic at 1023 K, the electrolysis can be conducted up to about 300 mA/cm/sup 2/ for the saturation solubility of Al/sub 2/S/sub 3/. Although these preliminary results are promising, additional studies are needed to elucidate many critical operating parameters before the technical potential of the electrolysis can be accurately assessed. 20 figures, 18 tables.

  3. Process Heat Exchanger Options for Fluoride Salt High Temperature Reactor

    SciTech Connect (OSTI)

    Piyush Sabharwall; Eung Soo Kim; Michael McKellar; Nolan Anderson

    2011-04-01

    The work reported herein is a significant intermediate step in reaching the final goal of commercial-scale deployment and usage of molten salt as the heat transport medium for process heat applications. The primary purpose of this study is to aid in the development and selection of the required heat exchanger for power production and process heat application, which would support large-scale deployment.

  4. Molten salt coal gasification process development unit. Phase 1. Volume 1. PDU operations. Final report

    SciTech Connect (OSTI)

    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.

  5. Development of a control algorithm for a molten-salt solar central receiver in a cylindrical configuration

    SciTech Connect (OSTI)

    Kolb, G.J.

    1991-01-01

    A control algorithm is proposed for a molten-salt solar central receiver in a cylindrical configuration. The algorithm simultaneously regulates the receiver outlet temperature and limits thermal-fatigue damage of the receiver tubes to acceptable levels. The algorithm is similar to one that was successfully tested for a receiver in a cavity configuration at the Central Receiver Test Facility in 1988. Due to the differences in the way solar flux is introduced on the receivers during cloud-induced transients, the cylindrical receiver will be somewhat more difficult to control than the cavity receiver. However, simulations of a proposed cylindrical receiver at the Solar Two power plant have indicated that automatic control during severe cloud transients is feasible. This paper also provides important insights regarding receiver design and lifetime as well as a strategy for reducing the power consumed by the molten-salt pumps. 14 refs., 7 figs., 2 tabs.

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

    DOE Patents [OSTI]

    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.

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

    DOE Patents [OSTI]

    Maroni, V.A.; von Winbush, S.

    1987-05-01

    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 that about /minus/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.

  8. Economic evaluation of solar-only and hybrid power towers using molten salt technology

    SciTech Connect (OSTI)

    Kolb, G.J.

    1996-12-01

    Several hybrid and solar-only configurations for molten-salt power towers were evaluated with a simple economic model, appropriate for screening analysis. The solar specific aspects of these plants were highlighted. In general, hybrid power towers were shown to be economically superior to solar-only plants with the same field size. Furthermore, the power-booster hybrid approach was generally preferred over the fuel-saver hybrid approach. Using today`s power tower technology, economic viability for the solar power-boost occurs at fuel costs in the neighborhood of $2.60/MBtu to $4.40/ MBtu (low heating value) depending on whether coal-based or gas-turbine-based technology is being offset. The cost Of CO[sub 2] avoidance was also calculated for solar cases in which the fossil fuel cost was too low for solar to be economically viable. The avoidance costs are competitive with other proposed methods of removing CO[sub 2] from fossil-fired power plants.

  9. OSTIblog Articles in the reactor Topic | OSTI, US Dept of Energy...

    Office of Scientific and Technical Information (OSTI)

    Weinberg and his team's work led to the construction and operation of the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory in Oak Ridge, Tennessee, where ...

  10. Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for {sup 233}U fuel production: 1. Concept of a stationary Tokamak as a neutron source

    SciTech Connect (OSTI)

    Azizov, E. A.; Gladush, G. G. Dokuka, V. N.; Khayrutdinov, R. R.

    2015-12-15

    On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of {sup 233}U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.

  11. Materials compatibility during the chlorination of molten CaCl/sub 2/. CaO salts. [CaCl/sub 2/. CaO salt

    SciTech Connect (OSTI)

    Rense, C.E.C.; Fife, K.W.; Bowersox, D.F.; Ferran, M.D.

    1987-01-01

    As part of our effort to develop a semicontinuous PuO/sub 2/ reduction process, we are investigating promising materials for containing a 900/sup 0/C molten CaCl/sub 2/ . CaO chlorination reaction. We want the material to contain this reaction and to be reusable. We tested candidate materials in a simulated salt (no plutonium) using anhydrous HCl as the chlorinating agent. Data are presented on the performance of 36 metals and alloys, 9 ceramics, and 3 coatings.

  12. Annular core liquid-salt cooled reactor with multiple fuel and...

    Office of Scientific and Technical Information (OSTI)

    Annular core liquid-salt cooled reactor with multiple fuel and blanket zones Citation Details In-Document Search Title: Annular core liquid-salt cooled reactor with multiple fuel ...

  13. Method to Reduce Molten Salt Penetration into Bulk Vitrification Refractory Materials

    SciTech Connect (OSTI)

    Bagaasen, L.M.; Hrma, P.R.; Kim, D.S.; Schweiger, M.J.; Matyas, J.; Rodriguez, C.P. [Pacific Northwest National Laboratory, Richland WA (United States); Witwer, K.S. [AMEC Nuclear Holdings Ltd., GeoMelt Division, Richland, WA (United States)

    2008-07-01

    Bulk vitrification (BV) is a process that heats a feed material consisting of glass-forming solids and dried low-activity waste (LAW) in a disposable refractory-lined metal box using electrical power supplied through carbon electrodes. The feed is heated to the point that the LAW decomposes and combines with the solids to generate a vitreous waste form. However, the castable refractory block (CRB) portion of the refractory lining has sufficient porosity to allow the low-viscosity molten ionic salt (MIS), which contains technetium (Tc) in a soluble form, to penetrate the CRB. This limits the effectiveness of the final waste form. This paper describes tests conducted to develop a method aimed at reducing the quantities of soluble Tc in the CRB. Tests showed that MIS formed in significant quantities at temperatures above 300 deg. C, remained stable until roughly 550 deg. C where it began to thermally decompose, and was completely decomposed by 800 deg. C. The estimated volume fraction of MIS in the feed was greater than 40%, and the CRB material contained 11 to 15% open porosity, a combination allowing a large quantity of MIS to migrate through the feed and penetrate the open porosity of the CRB. If the MIS is decomposed at temperatures below 300 deg. C or can be contained in the feed until it fully decomposes by 800 deg. C, MIS migration into the CRB can be avoided. Laboratory and crucible-scale experiments showed that a variety of methods, individually or in combination, can decrease MIS penetration into the CRB. Modifying the CRB to block MIS penetration was not deemed practical as a method to prevent the large quantities of MIS penetration seen in the full-scale tests, but it may be useful to reduce the impacts of lower levels of MIS penetration. Modifying the BV feed materials to better contain the MIS proved to be more successful. A series of qualitative and quantitative crucible tests were developed that allowed screening of feed modifications that might be

  14. Zr electrorefining process for the treatment of cladding hull waste in LiCl-KCl molten salts

    SciTech Connect (OSTI)

    Lee, Chang Hwa; Lee, You Lee; Jeon, Min Ku; Kang, Kweon Ho; Choi, Yong Taek; Park, Geun Il

    2013-07-01

    Zr electrorefining for the treatment of Zircaloy-4 cladding hull waste is demonstrated in LiCl-KCl-ZrCl{sub 4} molten salts. Although a Zr oxide layer thicker than 5 μm strongly inhibits the Zr dissolution process, pre-treatment processes increases the dissolution kinetics. For 10 g-scale experiments, the purities of the recovered Zr were 99.54 wt.% and 99.74 wt.% for fresh and oxidized cladding tubes, respectively, with no electrical contact issue. The optimal condition for Zr electrorefining has been found to improve the morphological feature of the recovered Zr, which reduces the salt incorporation by examining the effect of the process parameters such as the ZrCl{sub 4} concentration and the applied potential.

  15. Transpiring wall supercritical water oxidation reactor salt deposition studies

    SciTech Connect (OSTI)

    Haroldsen, B.L.; Mills, B.E.; Ariizumi, D.Y.; Brown, B.G.

    1996-09-01

    Sandia National Laboratories has teamed with Foster Wheeler Development Corp. and GenCorp, Aerojet to develop and evaluate a new supercritical water oxidation reactor design using a transpiring wall liner. In the design, pure water is injected through small pores in the liner wall to form a protective boundary layer that inhibits salt deposition and corrosion, effects that interfere with system performance. The concept was tested at Sandia on a laboratory-scale transpiring wall reactor that is a 1/4 scale model of a prototype plant being designed for the Army to destroy colored smoke and dye at Pine Bluff Arsenal in Arkansas. During the tests, a single-phase pressurized solution of sodium sulfate (Na{sub 2}SO{sub 4}) was heated to supercritical conditions, causing the salt to precipitate out as a fine solid. On-line diagnostics and post-test observation allowed us to characterize reactor performance at different flow and temperature conditions. Tests with and without the protective boundary layer demonstrated that wall transpiration provides significant protection against salt deposition. Confirmation tests were run with one of the dyes that will be processed in the Pine Bluff facility. The experimental techniques, results, and conclusions are discussed.

  16. Controlled temperature expansion in oxygen production by molten alkali metal salts

    DOE Patents [OSTI]

    Erickson, Donald C.

    1985-06-04

    A continuous process is set forth for the production of oxygen from an oxygen containing gas stream, such as air, by contacting a feed gas stream with a molten solution of an oxygen acceptor to oxidize the acceptor and cyclically regenerating the oxidized acceptor by releasing oxygen from the acceptor wherein the oxygen-depleted gas stream from the contact zone is treated sequentially to temperature reduction by heat exchange against the feed stream so as to condense out entrained oxygen acceptor for recycle to the process, combustion of the gas stream with fuel to elevate its temperature and expansion of the combusted high temperature gas stream in a turbine to recover power.

  17. Controlled temperature expansion in oxygen production by molten alkali metal salts

    DOE Patents [OSTI]

    Erickson, D.C.

    1985-06-04

    A continuous process is set forth for the production of oxygen from an oxygen containing gas stream, such as air, by contacting a feed gas stream with a molten solution of an oxygen acceptor to oxidize the acceptor and cyclically regenerating the oxidized acceptor by releasing oxygen from the acceptor wherein the oxygen-depleted gas stream from the contact zone is treated sequentially to temperature reduction by heat exchange against the feed stream so as to condense out entrained oxygen acceptor for recycle to the process, combustion of the gas stream with fuel to elevate its temperature and expansion of the combusted high temperature gas stream in a turbine to recover power. 1 fig.

  18. Actinide removal from spent salts

    DOE Patents [OSTI]

    Hsu, Peter C.; von Holtz, Erica H.; Hipple, David L.; Summers, Leslie J.; Adamson, Martyn G.

    2002-01-01

    A method for removing actinide contaminants (uranium and thorium) from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents are added to precipitate the thorium as thorium oxide and/or the uranium as either uranium oxide or as a diuranate salt. The precipitated materials are filtered, dried and packaged for disposal as radioactive waste. About 90% of the thorium and/or uranium present is removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 0.1 ppm of thorium or uranium.

  19. Metals removal from spent salts

    DOE Patents [OSTI]

    Hsu, Peter C.; Von Holtz, Erica H.; Hipple, David L.; Summers, Leslie J.; Brummond, William A.; Adamson, Martyn G.

    2002-01-01

    A method and apparatus for removing metal contaminants from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents may be added to precipitate the metal oxide and/or the metal as either metal oxide, metal hydroxide, or as a salt. The precipitated materials are filtered, dried and packaged for disposal as waste or can be immobilized as ceramic pellets. More than about 90% of the metals and mineral residues (ashes) present are removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be spray-dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 1.0 ppm of contaminants.

  20. Salt Selection for the LS-VHTR

    SciTech Connect (OSTI)

    Williams, D.F.; Clarno, K.T.

    2006-07-01

    Molten fluorides were initially developed for use in the nuclear industry as the high temperature fluid-fuel for a Molten Salt Reactor (MSR). The Office of Nuclear Energy is exploring the use of molten fluorides as a primary coolant (rather than helium) in an Advanced High Temperature Reactor (AHTR) design, also know as the Liquid-Salt cooled Very High Temperature Reactor (LS-VHTR). This paper provides a review of relevant properties for use in evaluation and ranking of candidate coolants for the LS-VHTR. Nuclear, physical, and chemical properties were reviewed and metrics for evaluation are recommended. Chemical properties of the salt were examined for the purpose of identifying factors that effect materials compatibility (i.e., corrosion). Some preliminary consideration of economic factors for the candidate salts is also presented. (authors)

  1. Molten salt as a heat transfer fluid for heating a subsurface formation

    DOE Patents [OSTI]

    Nguyen, Scott Vinh (Houston, TX); Vinegar, Harold J. (Bellaire, TX)

    2010-11-16

    A heating system for a subsurface formation includes a conduit located in an opening in the subsurface formation. An insulated conductor is located in the conduit. A material is in the conduit between a portion of the insulated conductor and a portion of the conduit. The material may be a salt. The material is a fluid at operating temperature of the heating system. Heat transfers from the insulated conductor to the fluid, from the fluid to the conduit, and from the conduit to the subsurface formation.

  2. Molten Salt Coal Gasification Process Development Unit. Phase 2. Quarterly technical progress report No. 2, October-December 1980

    SciTech Connect (OSTI)

    Slater, M. H.

    1981-01-20

    This represents the second quarterly progress report on Phase 2 of the Molten Salt Coal Gasification Process Development Unit (PDU) Program. Phase 1 of this program started in March 1976 and included the design, construction, and initial operation of the PDU. On June 25, 1980, Phase 2 of the program was initiated. It covers a 1-year operations program utilizing the existing PDU and is planned to include five runs with a targeted total operating time of 9 weeks. During this report period, Run 6, the initial run of the Phase 2 program was completed. The gasification system was operated for a total of 95 h at pressures up to 10 atm. Average product gas HHV values of 100 Btu/scf were recorded during 10-atm operation, while gasifying coal at a rate of 1100 lb/h. The run was terminated when the melt overflow system plugged after 60 continuous hours of overflow. Following this run, melt withdrawal system revisions were made, basically by changing the orifice materials from Monofrax to an 80 Cobalt-20 Chromium alloy. By the end of the report period, the PDU was being prepared for Run 7.

  3. Gas Turbine/Solar Parabolic Trough Hybrid Design Using Molten Salt Heat Transfer Fluid: Preprint

    SciTech Connect (OSTI)

    Turchi, C. S.; Ma, Z.

    2011-08-01

    Parabolic trough power plants can provide reliable power by incorporating either thermal energy storage (TES) or backup heat from fossil fuels. This paper describes a gas turbine / parabolic trough hybrid design that combines a solar contribution greater than 50% with gas heat rates that rival those of natural gas combined-cycle plants. Previous work illustrated benefits of integrating gas turbines with conventional oil heat-transfer-fluid (HTF) troughs running at 390?C. This work extends that analysis to examine the integration of gas turbines with salt-HTF troughs running at 450 degrees C and including TES. Using gas turbine waste heat to supplement the TES system provides greater operating flexibility while enhancing the efficiency of gas utilization. The analysis indicates that the hybrid plant design produces solar-derived electricity and gas-derived electricity at lower cost than either system operating alone.

  4. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    SciTech Connect (OSTI)

    Qualls, A. L.; Betzler, Benjamin R.; Brown, Nicholas R.; Carbajo, Juan; Greenwood, Michael Scott; Hale, Richard Edward; Harrison, Thomas J.; Powers, Jeffrey J.; Robb, Kevin R.; Terrell, Jerry W.

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  5. Evaluation of Salt Coolants for Reactor Applications (Journal...

    Office of Scientific and Technical Information (OSTI)

    Some preliminary consideration of economic factors for the candidate salts is also presented. Authors: Williams, David F 1 + Show Author Affiliations ORNL Publication Date: ...

  6. A New Approach for Modeling and Analysis of Molten Salt Reactors...

    Office of Scientific and Technical Information (OSTI)

    on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, Sun Valley, ID, USA, 20130505, 20130509 Research Org: Oak Ridge National Laboratory ...

  7. A new approach for modeling and analysis of molten salt reactors...

    Office of Scientific and Technical Information (OSTI)

    depletion software packages (e.g., continuous online feed and removal of materials). ... The current implementation is through a utility code that uses the two-dimensional (2D) ...

  8. Experimental techniques to determine salt formation and deposition in supercritical water oxidation reactors

    SciTech Connect (OSTI)

    Chan, J.P.C.; LaJeunesse, C.A.; Rice, S.F.

    1994-08-01

    Supercritical Water Oxidation (SCWO) is an emerging technology for destroying aqueous organic waste. Feed material, containing organic waste at concentrations typically less than 10 wt % in water, is pressurized and heated to conditions above water`s critical point where the ability of water to dissolve hydrocarbons and other organic chemicals is greatly enhanced. An oxidizer, is then added to the feed. Given adequate residence time and reaction temperature, the SCWO process rapidly produces innocuous combustion products. Organic carbon and nitrogen in the feed emerge as CO{sub 2} and N{sub 2}; metals, heteroatoms, and halides appear in the effluent as inorganic salts and acids. The oxidation of organic material containing heteroatoms, such as sulfur or phosphorous, forms acid anions. In the presence of metal ions, salts are formed and precipitate out of the supercritical fluid. In a tubular configured reactor, these salts agglomerate, adhere to the reactor wall, and eventually interfere by causing a flow restriction in the reactor leading to an increase in pressure. This rapid precipitation is due to an extreme drop in salt solubility that occurs as the feed stream becomes supercritical. To design a system that can accommodate the formation of these salts, it is important to understand the deposition process quantitatively. A phenomenological model is developed in this paper to predict the time that reactor pressure begins to rise as a function of the fluid axial temperature profile and effective solubility curve. The experimental techniques used to generate effective solubility curves for one salt of interest, Na{sub 2}SO{sub 4}, are described, and data is generated for comparison. Good correlation between the model and experiment is shown. An operational technique is also discussed that allows the deposited salt to be redissolved in a single phase and removed from the affected portion of the reactor. This technique is demonstrated experimentally.

  9. An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor

    SciTech Connect (OSTI)

    Yoder Jr, Graydon L; Aaron, Adam M; Cunningham, Richard Burns; Fugate, David L; Holcomb, David Eugene; Kisner, Roger A; Peretz, Fred J; Robb, Kevin R; Wilgen, John B; Wilson, Dane F

    2014-01-01

    The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.

  10. Preliminary requirements for a Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR)

    SciTech Connect (OSTI)

    Massie, M.; Forsberg, C.; Forget, B.; Hu, L. W.

    2012-07-01

    A Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR) design is being developed at MIT to provide the first demonstration and test of a salt-cooled reactor using high-temperature fuel. The first step is to define the requirements. The top level requirements are (1) provide the confidence that a larger demonstration reactor is warranted and (2) develop the necessary data for a larger-scale reactor. Because requirements will drive the design of the FHTR, a significant effort is being undertaken to define requirements and understand the tradeoffs that will be required for a practical design. The preliminary requirements include specifications for design parameters and necessary tests of major reactor systems. Testing requirements include demonstration of components, systems, and procedures for refueling, instrumentation, salt temperature control to avoid coolant freezing, salt chemistry and volume control, tritium monitoring and control, and in-service inspection. Safety tests include thermal hydraulics, neutronics - including intrinsic core shutdown mechanisms such as Doppler feedback - and decay heat removal systems. Materials and coolant testing includes fuels (including mechanical wear and fatigue) and system corrosion behavior. Preliminary analysis indicates a thermal power output below 30 MW, an initial core using pebble-bed or prismatic-block fuel, peak outlet temperatures of at least 700 deg. C, and use of FLi{sup 7}Be ({sup 7}LiF-BeF{sub 2}) coolant. The option to change-out the reactor core, fuel type, and major components is being investigated. While the FHTR will be used for materials testing, its primary mission is as a reactor system performance test to enable the design and licensing of a FHR demonstration power reactor. (authors)

  11. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    DOE Patents [OSTI]

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  12. Considerations of Alloy N for Fluoride Salt-Cooled High-Temperature Reactor Applications

    SciTech Connect (OSTI)

    Ren, Weiju; Muralidharan, Govindarajan; Wilson, Dane F; Holcomb, David Eugene

    2011-01-01

    Fluoride Salt-Cooled High-Temperature Reactors (FHRs) are a promising new class of thermal-spectrum nuclear reactors. The reactor structural materials must possess high-temperature strength and chemical compatibility with the liquid fluoride salt as well as with a power cycle fluid such as supercritical water while remaining resistant to residual air within the containment. Alloy N was developed for use with liquid fluoride salts and it possesses adequate strength and chemical compatibility up to about 700 C. A distinctive property of FHRs is that their maximum allowable coolant temperature is restricted by their structural alloy maximum service temperature. As the reactor thermal efficiency directly increases with the maximum coolant temperature, higher temperature resistant alloys are strongly desired. This paper reviews the current status of Alloy N and its relevance to FHRs including its design principles, development history, high temperature strength, environmental resistance, metallurgical stability, component manufacturability, ASME codification status, and reactor service requirements. The review will identify issues and provide guidance for improving the alloy properties or implementing engineering solutions.

  13. 100 MWe Baseload Molten Salt Plant Phase 1 & 2 Summary Report: Summary of Conceptual Design, Preliminary Design, Commercialization and Risk Reduction Activities

    SciTech Connect (OSTI)

    Tyner, Craig; Kraft, Dave; Moursund, Carter; Santelmann, Ken; Greaney, Andy; Zillmer, Andrew; Heap, Andy; Sakadjian, Bartev; Hannemann, Chris; Rogers, Dale; Gross, David; Wasyluk, David; Fondriest, Ed; Soni, Gaurav; Bindra, Hitesh; Marshall, Jason; Risner, Jeremy; Pacheco, Jim; Martin, Joe; Montesano, Kevin; Foder, Matt; Zavodny, Maximillian; Slack, Mike; Donnellan, Nathan; Sage, William

    2012-11-27

    This document describes steps taken to develop our conceptual and preliminary designs of a modular concept for deploying a 75% capacity factor, 100-MWe solar power plant. The modular approach consists of 14 solar power towers interconnected by hot and cold salt piping leading back to a central power block where the salt storage tanks and power generation systems are located. The plant is described in several sections. First, the overall plant is described, including the general arrangement, process and heat flow diagrams, system interface definitions, and electrical description. Next, each system is described in detail following the flow of energy from incident sunlight, through the plant, to the grid. These systems include the solar collector system (SCS), solar receiver system (SRS), thermal storage system (TSS), steam generator system (SGS), and power generation system (PGS). Then, the plant control system (PCS) and balance of plant (BOP) are discussed as supporting entities. Each system of the plant is described in sufficient detail to allow for the following to be developed: material cost, erection cost, project schedule, EPC bids, detailed performance modeling, and operations and maintenance cost. Cost, schedule, and performance estimates are not described in this document. Two approaches to demonstration of the technology are presented: a single tower integrated into an existing power block and a four tower stand alone 50 MWe power plant. Various demonstration partners have expressed interested in both approaches. The process by which a detailed plant performance model was developed is described to support the development of accurate LCOE data. Information on material and instrument testing is also provided for critical materials and instruments required for molten salt service.

  14. Fluoride Salt-Cooled High-Temperature Reactor Technology Development and Demonstration Roadmap

    SciTech Connect (OSTI)

    Holcomb, David Eugene; Flanagan, George F; Mays, Gary T; Pointer, William David; Robb, Kevin R; Yoder Jr, Graydon L

    2013-11-01

    Fluoride salt-cooled High-temperature Reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics, and fully passive safety. This roadmap describes the principal remaining FHR technology challenges and the development path needed to address the challenges. This roadmap also provides an integrated overview of the current status of the broad set of technologies necessary to design, evaluate, license, construct, operate, and maintain FHRs. First-generation FHRs will not require any technology breakthroughs, but do require significant concept development, system integration, and technology maturation. FHRs are currently entering early phase engineering development. As such, this roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant, the lack of an approved licensing framework, the lack of qualified, salt-compatible structural materials, and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

  15. An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

    SciTech Connect (OSTI)

    Holcomb, David Eugene; Cetiner, Sacit M; Flanagan, George F; Peretz, Fred J; Yoder Jr, Graydon L

    2009-11-01

    This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

  16. Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors

    SciTech Connect (OSTI)

    Yoder Jr, Graydon L; Elkassabgi, Yousri M.; De Leon, Gerardo I.; Fetterly, Caitlin N.; Ramos, Jorge A.; Cunningham, Richard Burns

    2012-02-01

    Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental

  17. SiC Schottky Diode Detectors for Measurement of Actinide Concentrations from Alpha Activities in Molten Salt Electrolyte

    SciTech Connect (OSTI)

    Windl, Wolfgang; Blue, Thomas

    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.

  18. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    SciTech Connect (OSTI)

    Peterson, Per; Greenspan, Ehud

    2015-02-09

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X

  19. Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

    SciTech Connect (OSTI)

    Katoh, Yutai; Wilson, Dane F; Forsberg, Charles W

    2007-09-01

    The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) composites are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.

  20. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) for Power and Process Heat

    SciTech Connect (OSTI)

    Forsberg, Charles; Hu, Lin-wen; Peterson, Per; Sridharan, Kumar

    2015-01-21

    In 2011 the U.S. Department of Energy through its Nuclear Energy University Program (NEUP) awarded a 3- year integrated research project (IRP) to the Massachusetts Institute of Technology (MIT) and its partners at the University of California at Berkeley (UCB) and the University of Wisconsin at Madison (UW). The IRP included Westinghouse Electric Company and an advisory panel chaired by Regis Matzie that provided advice as the project progressed. The first sentence of the proposal stated the goals: The objective of this Integrated Research Project (IRP) is to develop a path forward to a commercially viable salt-cooled solid-fuel high-temperature reactor with superior economic, safety, waste, nonproliferation, and physical security characteristics compared to light-water reactors. This report summarizes major results of this research.

  1. Stability of Molten Core Materials

    SciTech Connect (OSTI)

    Layne Pincock; Wendell Hintze

    2013-01-01

    The purpose of this report is to document a literature and data search for data and information pertaining to the stability of nuclear reactor molten core materials. This includes data and analysis from TMI-2 fuel and INL’s LOFT (Loss of Fluid Test) reactor project and other sources.

  2. DECONTAMINATION OF NEUTRON-IRRADIATED REACTOR FUEL

    DOE Patents [OSTI]

    Buyers, A.G.; Rosen, F.D.; Motta, E.E.

    1959-12-22

    A pyrometallurgical method of decontaminating neutronirradiated reactor fuel is presented. In accordance with the invention, neutron-irradiated reactor fuel may be decontaminated by countercurrently contacting the fuel with a bed of alkali and alkaine fluorides under an inert gas atmosphere and inductively melting the fuel and tracking the resulting descending molten fuel with induction heating as it passes through the bed. By this method, a large, continually fresh surface of salt is exposed to the descending molten fuel which enhances the efficiency of the scrubbing operation.

  3. Pre-Conceptual Design of a Fluoride-Salt-Cooled Small Modular Advanced High Temperature Reactor (SmAHTR)

    SciTech Connect (OSTI)

    Greene, Sherrell R; Gehin, Jess C; Holcomb, David Eugene; Carbajo, Juan J; Ilas, Dan; Cisneros, Anselmo T; Varma, Venugopal Koikal; Corwin, William R; Wilson, Dane F; Yoder Jr, Graydon L; Qualls, A L; Peretz, Fred J; Flanagan, George F; Clayton, Dwight A; Bradley, Eric Craig; Bell, Gary L; Hunn, John D; Pappano, Peter J; Cetiner, Sacit M

    2011-02-01

    This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high temperature reactors (FHRs). A preliminary reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.

  4. Molten core retention assembly

    DOE Patents [OSTI]

    Lampe, Robert F.

    1976-06-22

    Molten fuel produced in a core overheating accident is caught by a molten core retention assembly consisting of a horizontal baffle plate having a plurality of openings therein, heat exchange tubes having flow holes near the top thereof mounted in the openings, and a cylindrical, imperforate baffle attached to the plate and surrounding the tubes. The baffle assembly is supported from the core support plate of the reactor by a plurality of hanger rods which are welded to radial beams passing under the baffle plate and intermittently welded thereto. Preferably the upper end of the cylindrical baffle terminates in an outwardly facing lip to which are welded a plurality of bearings having slots therein adapted to accept the hanger rods.

  5. Project Profile: Heat Transfer and Latent Heat Storage in Inorganic Molten

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Salts for CSP Plants | Department of Energy Heat Transfer and Latent Heat Storage in Inorganic Molten Salts for CSP Plants Project Profile: Heat Transfer and Latent Heat Storage in Inorganic Molten Salts for CSP Plants Terrafore logo Terrafore, under the Thermal Storage FOA, is developing an economically feasible thermal energy storage (TES) system based on phase change materials (PCMs), for CSP plants. Approach This diagram shows how Terrafore is using a molten salt slurry to improve the

  6. Synchrotron X-ray diffraction and Raman spectroscopy of Ln{sub 3}NbO{sub 7} (Ln=La, Pr, Nd, Sm-Lu) ceramics obtained by molten-salt synthesis

    SciTech Connect (OSTI)

    Siqueira, K.P.F.; Soares, J.C.; Granado, E.; Bittar, E.M.; Paula, A.M. de; Moreira, R.L.; Dias, A.

    2014-01-15

    Ln{sub 3}NbO{sub 7} (Ln=La, Pr, Nd, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, and Lu) ceramics were obtained by molten-salt synthesis and their structures were systematically investigated by synchrotron X-ray diffraction (SXRD), second harmonic generation (SHG) and Raman spectroscopy. It was observed that ceramics with the largest ionic radii (La, Pr, Nd) crystallized into the Pmcn space group, while the ceramics with intermediate ionic radii (Sm-Gd) exhibited a different crystal structure belonging to the Ccmm space group. For this last group of ceramics, this result was corroborated by SHG and Raman scattering and ruled out any possibility for the non-centrosymmetric C 222{sub 1} space group, solving a recent controversy in the literature. Finally, according to SXRD, Tb-Lu containing samples exhibited an average defect fluorite structure (Fm3{sup ¯}m space group). Nonetheless, broad scattering at forbidden Bragg reflections indicates the presence of short-range domains with lower symmetry. Vibrational spectroscopy showed the presence of six Raman-active modes, inconsistent with the average cubic fluorite structure, and in line with the existence of lower-symmetry nano-domains immersed in the average fluorite structure of these ceramics. - Graphical abstract: Raman spectrum for Sm{sub 3}NbO{sub 7} ceramics showing their 27 phonon modes adjusted through Lorentzian lines. According to synchrotron X-ray diffraction and Raman scattering, this material belongs to the space group Cmcm. Display Omitted - Highlights: • Ln{sub 3}NbO{sub 7} ceramics were obtained by molten-salt synthesis. • SXRD, SHG and Raman scattering confirmed orthorhombic and cubic structures. • Ccmm instead of C222{sub 1} is the correct structure for Sm–Gd ceramics. • Pmcn space group was confirmed for La-, Pr- and Nd-based ceramics. • For Tb–Lu ceramics, ordered domains of a pyrochlore structure were observed.

  7. Project Profile: Fundamental Corrosion Studies in High-Temperature Molten

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Salt Systems for Next-Generation CSP Systems | Department of Energy Fundamental Corrosion Studies in High-Temperature Molten Salt Systems for Next-Generation CSP Systems Project Profile: Fundamental Corrosion Studies in High-Temperature Molten Salt Systems for Next-Generation CSP Systems Savannah River National Laboratory logo -- This project is inactive -- The Savannah River National Laboratory (SRNL), under the National Laboratory R&D competitive funding opportunity, is working with

  8. Challenges in the Development of Advanced Reactors

    SciTech Connect (OSTI)

    P. Sabharwall; M.C. Teague; S.M. Bragg-Sitton; M.W. Patterson

    2012-08-01

    Past generations of nuclear reactors have been successively developed and the next generation is currently being developed, demonstrating the constant progress and technical and industrial vitality of nuclear energy. In 2000 US Department of Energy launched Generation IV International Forum (GIF) which is one of the main international frameworks for the development of future nuclear systems. The six systems that were selected were: sodium cooled fast reactor, lead cooled fast reactor, supercritical water cooled reactor, very high temperature gas cooled reactor (VHTR), gas cooled fast reactor and molten salt reactor. This paper discusses some of the proposed advanced reactor concepts that are currently being researched to varying degrees in the United States, and highlights some of the major challenges these concepts must overcome to establish their feasibility and to satisfy licensing requirements.

  9. Selective Separation of Cs and Sr from LiCl-Based Salt for Electrochemical Processing of Oxide Spent Nuclear Fuel

    SciTech Connect (OSTI)

    P Sachdev

    2008-07-01

    Electrochemical processing technology is currently being used for the treatment of metallic spent fuel from the Experimental Breeder Reactor-II at Idaho National Laboratory. The treatment of oxide-based spent nuclear fuel via electrochemical processing is possible provided there is a front-end oxide reduction step. During this reduction process, certain fission products, including Cs and Sr, partition into the salt phase and form chlorides. Both solid state and molten LiCl-zeolite-A ion exchange tests were conducted for selectively removing Cs and Sr from LiCl-based salt. The solid-state tests produced in excess of 99% removal of Cs and Sr. The molten state tests failed due to phase transformation of the zeolite structure when in contact with the molten LiCl salt.

  10. Measurement of the Melting Point Temperature of Several Lithium-Sodium-Beryllium Fluoride Salt (Flinabe) Mixtures

    SciTech Connect (OSTI)

    McDonald, J.M; Nygren, R.E.; Lutz, T.J.; Tanaka, T.J; Ulrickson, M.A.; Boyle, T.J.; Troncosa, K.P.

    2005-04-15

    The molten salt Flibe, a combination of lithium and beryllium fluorides studied for molten salt fission reactors, has been proposed as a breeder and coolant for fusion applications. The melting points of 2LiF-BeF{sub 2} and LiF-BeF{sub 2} are 460 deg. C and 363 deg. C, but LiF-BeF{sub 2} is rather viscous and has less lithium for breeding. In the Advanced Power Extraction (APEX) Program, concepts with a free flowing liquid for the first wall and blanket were investigated. Flinabe (a mixture of LiF, BeF{sub 2} and NaF) was selected for a molten salt design because a melting temperature below 350 deg. C appeared possible and this provided an attractive operating temperature window for a reactor. To confirm that a ternary salt with a low melting temperature existed, several combinations of the fluoride salts, LiF, NaF and BeF{sub 2}, were melted in a stainless steel crucible under vacuum. One had an apparent melting temperature of 305 deg. C. The test system, preparation of the mixtures, melting procedures and temperature curves for the melting and cooling are presented along with the apparent melting points. Thermal modeling of the salt pool and crucible is reported in an accompanying paper.

  11. Supported molten-metal catalysts

    DOE Patents [OSTI]

    Datta, Ravindra; Singh, Ajeet; Halasz, Istvan; Serban, Manuela

    2001-01-01

    An entirely new class of catalysts called supported molten-metal catalysts, SMMC, which can replace some of the existing precious metal catalysts used in the production of fuels, commodity chemicals, and fine chemicals, as well as in combating pollution. SMMC are based on supporting ultra-thin films or micro-droplets of the relatively low-melting (<600.degree. C.), inexpensive, and abundant metals and semimetals from groups 1, 12, 13, 14, 15 and 16, of the periodic table, or their alloys and intermetallic compounds, on porous refractory supports, much like supported microcrystallites of the traditional solid metal catalysts. It thus provides orders of magnitude higher surface area than is obtainable in conventional reactors containing molten metals in pool form and also avoids corrosion. These have so far been the chief stumbling blocks in the application of molten metal catalysts.

  12. Development of strong-sense validation benchmarks for the fluoride salt-cooled high-temperature reactor

    SciTech Connect (OSTI)

    Blandford, E. D.

    2012-07-01

    The Fluoride salt-cooled High-temperature Reactor (FHR) is a class of reactor concepts currently under development for the U. S. Dept. of Energy. The FHR is defined as a Generation IV reactor that features low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. Recent experimental work using simulant fluids have been performed to demonstrate key 'proof of principle' FHR concepts and have helped inform the reactor design process. An important element of developing FHR technology is to sufficiently validate the predictive accuracy of the computer codes used to model system response. This paper presents a set of thermal-hydraulics experiments, defined as Strong-Sense Benchmarks (SSB's), which will help establish the FHR validation domain for simulant fluid suitability. These SSB's are more specifically designed to investigate single-phase natural circulation which is the dominant mode of FHR decay heat removal during off-normal conditions. SSB s should be viewed as engineering reference standards and differ from traditional confirmatory experiments in the sense that they are more focused on fundamental physics as opposed to reproducing high levels of physical similarity with the prototypical design. (authors)

  13. High Temperature Fluoride Salt Test Loop

    SciTech Connect (OSTI)

    Aaron, Adam M.; Cunningham, Richard Burns; Fugate, David L.; Holcomb, David Eugene; Kisner, Roger A.; Peretz, Fred J.; Robb, Kevin R.; Wilson, Dane F.; Yoder, Jr, Graydon L.

    2015-12-01

    Effective high-temperature thermal energy exchange and delivery at temperatures over 600°C has the potential of significant impact by reducing both the capital and operating cost of energy conversion and transport systems. It is one of the key technologies necessary for efficient hydrogen production and could potentially enhance efficiencies of high-temperature solar systems. Today, there are no standard commercially available high-performance heat transfer fluids above 600°C. High pressures associated with water and gaseous coolants (such as helium) at elevated temperatures impose limiting design conditions for the materials in most energy systems. Liquid salts offer high-temperature capabilities at low vapor pressures, good heat transport properties, and reasonable costs and are therefore leading candidate fluids for next-generation energy production. Liquid-fluoride-salt-cooled, graphite-moderated reactors, referred to as Fluoride Salt Reactors (FHRs), are specifically designed to exploit the excellent heat transfer properties of liquid fluoride salts while maximizing their thermal efficiency and minimizing cost. The FHR s outstanding heat transfer properties, combined with its fully passive safety, make this reactor the most technologically desirable nuclear power reactor class for next-generation energy production. Multiple FHR designs are presently being considered. These range from the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) [1] design originally developed by UC-Berkeley to the Small Advanced High-Temperature Reactor (SmAHTR) and the large scale FHR both being developed at ORNL [2]. The value of high-temperature, molten-salt-cooled reactors is also recognized internationally, and Czechoslovakia, France, India, and China all have salt-cooled reactor development under way. The liquid salt experiment presently being developed uses the PB-AHTR as its focus. One core design of the PB-AHTR features multiple 20 cm diameter, 3.2 m long fuel channels

  14. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) with Silicon-Carbide-Matrix Coated-Particle Fuel

    SciTech Connect (OSTI)

    Forsberg, C. W.; Terrani, Kurt A; Snead, Lance Lewis; Katoh, Yutai

    2012-01-01

    The FHR is a new reactor concept that uses coated-particle fuel and a low-pressure liquid-salt coolant. Its neutronics are similar to a high-temperature gas-cooled reactor (HTGR). The power density is 5 to 10 times higher because of the superior cooling properties of liquids versus gases. The leading candidate coolant salt is a mixture of {sup 7}LiF and BeF{sub 2} (FLiBe) possessing a boiling point above 1300 C and the figure of merit {rho}C{sub p} (volumetric heat capacity) for the salt slightly superior to water. Studies are underway to define a near-term base-line concept while understanding longer-term options. Near-term options use graphite-matrix coated-particle fuel where the graphite is both a structural component and the primary neutron moderator. It is the same basic fuel used in HTGRs. The fuel can take several geometric forms with a pebble bed being the leading contender. Recent work on silicon-carbide-matrix (SiCm) coated-particle fuel may create a second longer-term fuel option. SiCm coated-particle fuels are currently being investigated for use in light-water reactors. The replacement of the graphite matrix with a SiCm creates a new family of fuels. The first motivation behind the effort is to take advantage of the superior radiation resistance of SiC compared to graphite in order to provide a stable matrix for hosting coated fuel particles. The second motivation is a much more rugged fuel under accident, repository, and other conditions.

  15. An Overview of Liquid Fluoride Salt Heat Transport Technology

    SciTech Connect (OSTI)

    Cetiner, Mustafa Sacit; Holcomb, David Eugene

    2010-01-01

    Liquid fluoride salts are a leading candidate heat transport medium for high-temperature applications. This report provides an overview of the current status of liquid salt heat transport technology. The report includes a high-level, parametric evaluation of liquid fluoride salt heat transport loop performance to allow intercomparisons between heat-transport fluid options as well as providing an overview of the properties and requirements for a representative loop. Much of the information presented here derives from the earlier molten salt reactor program and a significant advantage of fluoride salts, as high temperature heat transport media is their consequent relative technological maturity. The report also includes a compilation of relevant thermophysical properties of useful heat transport fluoride salts. Fluoride salts are both thermally stable and with proper chemistry control can be relatively chemically inert. Fluoride salts can, however, be highly corrosive depending on the container materials selected, the salt chemistry, and the operating procedures used. The report also provides an overview of the state-of-the-art in reduction-oxidation chemistry control methodologies employed to minimize salt corrosion as well as providing a general discussion of heat transfer loop operational issues such as start-up procedures and freeze-up vulnerability.

  16. Reactor apparatus

    DOE Patents [OSTI]

    Echtler, J. Paul

    1981-01-01

    A reactor apparatus for hydrocracking a polynuclear aromatic hydrocarbonaceous feedstock to produce lighter hydrocarbon fuels by contacting the hydrocarbonaceous feedstock with hydrogen in the presence of a molten metal halide catalyst.

  17. FUSED SALT PROCESS FOR RECOVERY OF VALUES FROM USED NUCLEAR REACTOR FUELS

    DOE Patents [OSTI]

    Moore, R.H.

    1960-08-01

    A process is given for recovering plutonium from a neutron-irradiated uranium mass (oxide or alloy) by dissolving the mass in an about equimolar alkali metalaluminum double chloride, adding aluminum metal to the mixture obtained at a temperature of between 260 and 860 deg C, and separating a uranium-containing metal phase and a plutonium-chloride- and fission-product chloridecontaining salt phase. Dissolution can be expedited by passing carbon tetrachloride vapors through the double salt. Separation without reduction of plutonium from neutron- bombarded uranium and that of cerium from uranium are also discussed.

  18. Energy Department Completes Salt Coolant Material Transfer to Czech Republic for Advanced Reactor Research

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy recently joined with the U.S. Embassy in Prague and the Czech Republic’s Ministry of Industry and Trade to complete the transfer of 75 kilograms of fluoride salt from the Department’s Oak Ridge National Laboratory to the Czech Nuclear Research Institute Řež.

  19. Thermal-Hydraulic Design of a Fluoride High-Temperature Demonstration Reactor

    SciTech Connect (OSTI)

    Carbajo, Juan J; Qualls, A L

    2016-01-01

    INTRODUCTION The Fluoride High-Temperature Reactor (FHR) named the Demonstration Reactor (DR) is a novel reactor concept using molten salt coolant and TRIstructural ISOtropic (TRISO) fuel that is being developed at Oak Ridge National Laboratory (ORNL). The objective of the FHR DR is to advance the technology readiness level of FHRs. The FHR DR will demonstrate technologies needed to close remaining gaps to commercial viability. The FHR DR has a thermal power of 100 MWt, very similar to the SmAHTR, another FHR ORNL concept (Refs. 1 and 2) with a power of 125 MWt. The FHR DR is also a small version of the Advanced High Temperature Reactor (AHTR), with a power of 3400 MWt, cooled by a molten salt and also being developed at ORNL (Ref. 3). The FHR DR combines three existing technologies: (1) high-temperature, low-pressure molten salt coolant, (2) high-temperature coated-particle TRISO fuel, (3) and passive decay heat cooling systems by using Direct Reactor Auxiliary Cooling Systems (DRACS). This paper presents FHR DR thermal-hydraulic design calculations.

  20. Challenges in the Development of High Temperature Reactors

    SciTech Connect (OSTI)

    Piyush Sabharwall; Shannon M. Bragg-Sitton; Carl Stoots

    2013-10-01

    Advanced reactor designs offer potentially significant improvements over currently operating light water reactors including improved fuel utilization, increased efficiency, higher temperature operation (enabling a new suite of non-electric industrial process heat applications), and increased safety. As with most technologies, these potential performance improvements come with a variety of challenges to bringing advanced designs to the marketplace. There are technical challenges in material selection and thermal hydraulic and power conversion design that arise particularly for higher temperature, long life operation (possibly >60 years). The process of licensing a new reactor design is also daunting, requiring significant data collection for model verification and validation to provide confidence in safety margins associated with operating a new reactor design under normal and off-normal conditions. This paper focuses on the key technical challenges associated with two proposed advanced reactor concepts: the helium gas cooled Very High Temperature Reactor (VHTR) and the molten salt cooled Advanced High Temperature Reactor (AHTR).

  1. Separation and Recovery of Uranium Metal from Spent Light Water Reactor Fuel via Electrolytic Reduction and Electrorefining

    SciTech Connect (OSTI)

    S. D. Herrmann; S. X. Li

    2010-09-01

    A series of bench-scale experiments was performed in a hot cell at Idaho National Laboratory to demonstrate the separation and recovery of uranium metal from spent light water reactor (LWR) oxide fuel. The experiments involved crushing spent LWR fuel to particulate and separating it from its cladding. Oxide fuel particulate was then converted to metal in a series of six electrolytic reduction runs that were performed in succession with a single salt loading of molten LiCl 1 wt% Li2O at 650 C. Analysis of salt samples following the series of electrolytic reduction runs identified the diffusion of select fission products from the spent fuel to the molten salt electrolyte. The extents of metal oxide conversion in the post-test fuel were also quantified, including a nominal 99.7% conversion of uranium oxide to metal. Uranium metal was then separated from the reduced LWR fuel in a series of six electrorefining runs that were performed in succession with a single salt loading of molten LiCl-KCl-UCl3 at 500 C. Analysis of salt samples following the series of electrorefining runs identified additional partitioning of fission products into the molten salt electrolyte. Analyses of the separated uranium metal were performed, and its decontamination factors were determined.

  2. Liquid surface skimmer apparatus for molten lithium and method

    DOE Patents [OSTI]

    Robinson, Samuel C.; Pollard, Roy E.; Thompson, William F.; Stark, Marshall W.; Currin, Jr., Robert T.

    1995-01-01

    This invention relates to an apparatus for separating two fluids having different specific gravities. The invention also relates to a method for using the separating apparatus of the present invention. This invention particularly relates to the skimming of molten lithium metal from the surface of a fused salt electrolyte in the electrolytic production of lithium metal from a mixed fused salt.

  3. Production of chlorine from chloride salts

    DOE Patents [OSTI]

    Rohrmann, Charles A. (Kennewick, WA)

    1981-01-01

    A process for converting chloride salts and sulfuric acid to sulfate salts and elemental chlorine is disclosed. A chloride salt and sulfuric acid are combined in a furnace where they react to produce a sulfate salt and hydrogen chloride. Hydrogen chloride from the furnace contacts a molten salt mixture containing an oxygen compound of vanadium, an alkali metal sulfate and an alkali metal pyrosulfate to recover elemental chlorine. In the absence of an oxygen-bearing gas during the contacting, the vanadium is reduced, but is regenerated to its active higher valence state by separately contacting the molten salt mixture with an oxygen-bearing gas.

  4. Integral and Separate Effects Tests for Thermal Hydraulics Code Validation for Liquid-Salt Cooled Nuclear Reactors

    SciTech Connect (OSTI)

    Peterson, Per

    2012-10-30

    The objective of the 3-year project was to collect integral effects test (IET) data to validate the RELAP5-3D code and other thermal hydraulics codes for use in predicting the transient thermal hydraulics response of liquid salt cooled reactor systems, including integral transient response for forced and natural circulation operation. The reference system for the project is a modular, 900-MWth Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a specific type of Fluoride salt-cooled High temperature Reactor (FHR). Two experimental facilities were developed for thermal-hydraulic integral effects tests (IETs) and separate effects tests (SETs). The facilities use simulant fluids for the liquid fluoride salts, with very little distortion to the heat transfer and fluid dynamics behavior. The CIET Test Bay facility was designed, built, and operated. IET data for steady state and transient natural circulation was collected. SET data for convective heat transfer in pebble beds and straight channel geometries was collected. The facility continues to be operational and will be used for future experiments, and for component development. The CIET 2 facility is larger in scope, and its construction and operation has a longer timeline than the duration of this grant. The design for the CIET 2 facility has drawn heavily on the experience and data collected on the CIET Test Bay, and it was completed in parallel with operation of the CIET Test Bay. CIET 2 will demonstrate start-up and shut-down transients and control logic, in addition to LOFC and LOHS transients, and buoyant shut down rod operation during transients. Design of the CIET 2 Facility is complete, and engineering drawings have been submitted to an external vendor for outsourced quality controlled construction. CIET 2 construction and operation continue under another NEUP grant. IET data from both CIET facilities is to be used for validation of system codes used for FHR modeling, such as RELAP5-3D. A set of

  5. Method for continuously recovering metals using a dual zone chemical reactor

    DOE Patents [OSTI]

    Bronson, Mark C. (Livermore, CA)

    1995-01-01

    A dual zone chemical reactor continuously processes metal-containing materials while regenerating and circulating a liquid carrier. The starting materials are fed into a first reaction zone of a vessel containing a molten salt carrier. The starting materials react to form a metal product and a by-product that dissolves in the molten salt that flows to a second reaction zone in the reaction vessel. The second reaction zone is partitioned from, but in fluid communication with, the first reaction zone. The liquid carrier continuously circulates along a pathway between the first reaction zone and the second reaction zone. A reactive gas is introduced into the second reaction zone to react with the reaction by-product to generate the molten salt. The metal product, the gaseous waste products, and the excess liquid carrier are removed without interrupting the operation of the reactor. The design of the dual zone reactor can be adapted to combine a plurality of liquid carrier regeneration zones in a multiple dual zone chemical reactor for production scale processing.

  6. Method for continuously recovering metals using a dual zone chemical reactor

    DOE Patents [OSTI]

    Bronson, M.C.

    1995-02-14

    A dual zone chemical reactor continuously processes metal-containing materials while regenerating and circulating a liquid carrier. The starting materials are fed into a first reaction zone of a vessel containing a molten salt carrier. The starting materials react to form a metal product and a by-product that dissolves in the molten salt that flows to a second reaction zone in the reaction vessel. The second reaction zone is partitioned from, but in fluid communication with, the first reaction zone. The liquid carrier continuously circulates along a pathway between the first reaction zone and the second reaction zone. A reactive gas is introduced into the second reaction zone to react with the reaction by-product to generate the molten salt. The metal product, the gaseous waste products, and the excess liquid carrier are removed without interrupting the operation of the reactor. The design of the dual zone reactor can be adapted to combine a plurality of liquid carrier regeneration zones in a multiple dual zone chemical reactor for production scale processing. 6 figs.

  7. Molten Glass for Thermal Storage: Advanced Molten Glass for Heat Transfer and Thermal Energy Storage

    SciTech Connect (OSTI)

    2012-01-01

    HEATS Project: Halotechnics is developing a high-temperature thermal energy storage system using a new thermal-storage and heat-transfer material: earth-abundant and low-melting-point molten glass. Heat storage materials are critical to the energy storage process. In solar thermal storage systems, heat can be stored in these materials during the day and released at night—when the sun is not out—to drive a turbine and produce electricity. In nuclear storage systems, heat can be stored in these materials at night and released to produce electricity during daytime peak-demand hours. Halotechnics new thermal storage material targets a price that is potentially cheaper than the molten salt used in most commercial solar thermal storage systems today. It is also extremely stable at temperatures up to 1200°C—hundreds of degrees hotter than the highest temperature molten salt can handle. Being able to function at high temperatures will significantly increase the efficiency of turning heat into electricity. Halotechnics is developing a scalable system to pump, heat, store, and discharge the molten glass. The company is leveraging technology used in the modern glass industry, which has decades of experience handling molten glass.

  8. Liquid Salts as Media for Process Heat Transfer from VHTR's: Forced Convective Channel Flow Thermal Hydraulics, Materials, and Coating

    SciTech Connect (OSTI)

    Sridharan, Kumar; Anderson, Mark; Allen, Todd; Corradini, Michael

    2012-01-30

    The goal of this NERI project was to perform research on high temperature fluoride and chloride molten salts towards the long-term goal of using these salts for transferring process heat from high temperature nuclear reactor to operation of hydrogen production and chemical plants. Specifically, the research focuses on corrosion of materials in molten salts, which continues to be one of the most significant challenges in molten salts systems. Based on the earlier work performed at ORNL on salt properties for heat transfer applications, a eutectic fluoride salt FLiNaK (46.5% LiF-11.5%NaF-42.0%KF, mol.%) and a eutectic chloride salt (32%MgCl2-68%KCl, mole %) were selected for this study. Several high temperature candidate Fe-Ni-Cr and Ni-Cr alloys: Hastelloy-N, Hastelloy-X, Haynes-230, Inconel-617, and Incoloy-800H, were exposed to molten FLiNaK with the goal of understanding corrosion mechanisms and ranking these alloys for their suitability for molten fluoride salt heat exchanger and thermal storage applications. The tests were performed at 850C for 500 h in sealed graphite crucibles under an argon cover gas. Corrosion was noted to occur predominantly from dealloying of Cr from the alloys, an effect that was particularly pronounced at the grain boundaries Alloy weight-loss due to molten fluoride salt exposure correlated with the initial Cr-content of the alloys, and was consistent with the Cr-content measured in the salts after corrosion tests. The alloys weight-loss was also found to correlate to the concentration of carbon present for the nominally 20% Cr containing alloys, due to the formation of chromium carbide phases at the grain boundaries. Experiments involving molten salt exposures of Incoloy-800H in Incoloy-800H crucibles under an argon cover gas showed a significantly lower corrosion for this alloy than when tested in a graphite crucible. Graphite significantly accelerated alloy corrosion due to the reduction of Cr from solution by graphite and formation

  9. Modular & Scalable Molten Salt Plant Design

    Broader source: Energy.gov [DOE]

    This presentation was delivered at the SunShot Concentrating Solar Power (CSP) Program Review 2013, held April 23–25, 2013 near Phoenix, Arizona.

  10. Inherent structure of a molten salt

    SciTech Connect (OSTI)

    La Violette, Randall A.; Budzien, Joanne L.; Stillinger, Frank H.

    2000-05-08

    We calculated the inherent structure of a model melt of zinc (II) bromide over a wide range of densities. Stable, metastable, and unstable branches were obtained for the zero temperature pressure-volume isotherm of the inherent structure. The pressure-volume isotherm, the void distribution, and the structure factor were used to identify the spinodal, independent of any model equation of state. (c) 2000 American Institute of Physics.

  11. Molten salt fuels with high plutonium solubility

    DOE Patents [OSTI]

    Moir, Ralph W; Turchi, Patrice E.A.; Shaw, Henry F; Kaufman, Larry

    2013-08-13

    The present invention includes a composition of LiF--ThF.sub.4--UF.sub.4--PuF.sub.3 for use as a fuel in a nuclear engine.

  12. Electrodeposition of molten silicon

    DOE Patents [OSTI]

    De Mattei, Robert C.; Elwell, Dennis; Feigelson, Robert S.

    1981-01-01

    Silicon dioxide is dissolved in a molten electrolytic bath, preferably comprising barium oxide and barium fluoride. A direct current is passed between an anode and a cathode in the bath to reduce the dissolved silicon dioxide to non-alloyed silicon in molten form, which is removed from the bath.

  13. An Overview of Liquid Fluoride Salt Heat Transport Systems

    SciTech Connect (OSTI)

    Holcomb, David Eugene; Cetiner, Sacit M

    2010-09-01

    Heat transport is central to all thermal-based forms of electricity generation. The ever increasing demand for higher thermal efficiency necessitates power generation cycles transitioning to progressively higher temperatures. Similarly, the desire to provide direct thermal coupling between heat sources and higher temperature chemical processes provides the underlying incentive to move toward higher temperature heat transfer loops. As the system temperature rises, the available materials and technology choices become progressively more limited. Superficially, fluoride salts at {approx}700 C resemble water at room temperature being optically transparent and having similar heat capacity, roughly three times the viscosity, and about twice the density. Fluoride salts are a leading candidate heat-transport material at high temperatures. Fluoride salts have been extensively used in specialized industrial processes for decades, yet they have not entered widespread deployment for general heat transport purposes. This report does not provide an exhaustive screening of potential heat transfer media and other high temperature liquids such as alkali metal carbonate eutectics or chloride salts may have economic or technological advantages. A particular advantage of fluoride salts is that the technology for their use is relatively mature as they were extensively studied during the 1940s-1970s as part of the U.S. Atomic Energy Commission's program to develop molten salt reactors (MSRs). However, the instrumentation, components, and practices for use of fluoride salts are not yet developed sufficiently for commercial implementation. This report provides an overview of the current understanding of the technologies involved in liquid salt heat transport (LSHT) along with providing references to the more detailed primary information resources. Much of the information presented here derives from the earlier MSR program. However, technology has evolved over the intervening years, and

  14. Flibe Use in Fusion Reactors - An Initial Safety Assessment

    SciTech Connect (OSTI)

    Cadwallader, Lee Charles; Longhurst, Glen Reed

    1999-04-01

    This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF2) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material.

  15. Flibe use in fusion reactors -- An initial safety assessment

    SciTech Connect (OSTI)

    Cadwallader, L.C.; Longhurst, G.R.

    1999-03-01

    This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF{sub 2}) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material.

  16. PYROCHEMICAL DECONTAMINATION METHOD FOR REACTOR FUEL

    DOE Patents [OSTI]

    Buyers, A.G.

    1959-06-30

    A pyro-chemical method is presented for decontaminating neutron irradiated uranium and separating plutonium therefrom by contact in the molten state with a metal chloride salt. Uranium trichloride and uranium tetrachloride either alone or in admixture with alkaline metal and alkaline eanth metal fluorides under specified temperature and specified phase ratio conditions extract substantially all of the uranium from the irradiated uranium fuel together with certain fission products. The phases are then separated leaving purified uranium metal. The uranium and plutonium in the salt phase can be reduced to forin a highly decontaminated uraniumplutonium alloy. The present method possesses advantages for economically decontaminating irradiated nuclear fuel elements since irradiated fuel may be proccessed immediately after withdrawal from the reactor and the uranium need not be dissolved and later reduced to the metallic form. Accordingly, the uranium may be economically refabricated and reinserted into the reactor.

  17. Development of high temperature transport technology for LiCl-KCl eutectic salt in pyroprocessing

    SciTech Connect (OSTI)

    Lee, Sung Ho; Lee, Hansoo; Kim, In Tae; Kim, Jeong-Guk

    2013-07-01

    The development of high-temperature transport technologies for molten salt is a prerequisite and a key issue in the industrialization of pyro-reprocessing for advanced fuel cycle scenarios. The solution of a molten salt centrifugal pump was discarded because of the high corrosion power of a high temperature molten salt, so the suction pump solution was selected. An apparatus for salt transport experiments by suction was designed and tested using LiC-KCl eutectic salt. The experimental results of lab-scale molten salt transport by suction showed a 99.5% transport rate (ratio of transported salt to total salt) under a vacuum range of 100 mtorr - 10 torr at 500 Celsius degrees. The suction system has been integrated to the PRIDE (pyroprocessing integrated inactive demonstration) facility that is a demonstrator using non-irradiated materials (natural uranium and surrogate materials). The performance of the suction pump for the transport of molten salts has been confirmed.

  18. Vitrification of IFR and MSBR halide salt reprocessing wastes

    SciTech Connect (OSTI)

    Siemer, D.D.

    2013-07-01

    Both of the genuinely sustainable (breeder) nuclear fuel cycles (IFR - Integral Fast Reactor - and MSBR - Molten Salt Breeder Reactor -) studied by the USA's national laboratories would generate high level reprocessing waste (HLRW) streams consisting of a relatively small amount ( about 4 mole %) of fission product halide (chloride or fluoride) salts in a matrix comprised primarily (about 95 mole %) of non radioactive alkali metal halide salts. Because leach resistant glasses cannot accommodate much of any of the halides, most of the treatment scenarios previously envisioned for such HLRW have assumed a monolithic waste form comprised of a synthetic analog of an insoluble crystalline halide mineral. In practice, this translates to making a 'substituted' sodalite ('Ceramic Waste Form') of the IFR's chloride salt-based wastes and fluoroapatite of the MSBR's fluoride salt-based wastes. This paper discusses my experimental studies of an alternative waste management scenario for both fuel cycles that would separate/recycle the waste's halide and immobilize everything else in iron phosphate (Fe-P) glass. It will describe both how the work was done and what its results indicate about how a treatment process for both of those wastes should be implemented (fluoride and chloride behave differently). In either case, this scenario's primary advantages include much higher waste loadings, much lower overall cost, and the generation of a product (glass) that is more consistent with current waste management practices. (author)

  19. Electrochromic Salts, Solutions, and Devices

    DOE Patents [OSTI]

    Burrell, Anthony K.; Warner, Benjamin P.; McClesky, T. Mark

    2008-11-11

    Electrochromic salts. Electrochromic salts of dicationic viologens such as methyl viologen and benzyl viologen associated with anions selected from bis(trifluoromethylsulfonyl)imide, bis(perfluoroethylsulfonyl)imide, and tris(trifluoromethylsulfonyl)methide are produced by metathesis with the corresponding viologen dihalide. They are highly soluble in molten quarternary ammonium salts and together with a suitable reductant provide electrolyte solutions that are used in electrochromic windows.

  20. Electrochromic Salts, Solutions, and Devices

    DOE Patents [OSTI]

    Burrell, Anthony K.; Warner, Benjamin P.; McClesky, T. Mark

    2008-10-14

    Electrochromic salts. Electrochromic salts of dicationic viologens such as methyl viologen and benzyl viologen associated with anions selected from bis(trifluoromethylsulfonyl)imide, bis(perfluoroethylsulfonyl)imide, and tris(trifluoromethylsulfonyl)methide are produced by metathesis with the corresponding viologen dihalide. They are highly soluble in molten quarternary ammonium salts and together with a suitable reductant provide electrolyte solutions that are used in electrochromic windows.

  1. Electrochromic salts, solutions, and devices

    DOE Patents [OSTI]

    Burrell, Anthony K.; Warner, Benjamin P.; McClesky,7,064,212 T. Mark

    2006-06-20

    Electrochromic salts. Electrochromic salts of dicationic viologens such as methyl viologen and benzyl viologen associated with anions selected from bis(trifluoromethylsulfonyl)imide, bis(perfluoroethylsulfonyl)imide, and tris(trifluoromethylsulfonyl)methide are produced by metathesis with the corresponding viologen dihalide. They are highly soluble in molten quarternary ammonium salts and together with a suitable reductant provide electrolyte solutions that are used in electrochromic windows.

  2. NON-NRC FUNDED RELAP5-3D VERSION 4.x.x SOFTWARE REACTOR EXCURSION AND LEAK ANALYSIS PACKAGE - THREE DIMENSIONAL

    SciTech Connect (OSTI)

    2012-03-26

    The RELAP5-3D Version 3.x code has been developed for best-estimate transient simulation of nuclear reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems including pressurized water reactors, boiling water reactors, Soviet-designed reactors, heavy water reactors, gas-cooled reactors, liquid metal and molten salt cooled reactors, and even fusion reactors. Numerical models include multi-dimensional hydrodynamics, 1- and 2-D heat transfer in metal walls, 0-, 1-, 2-, and 3-D neutron kinetics, trips, and control systems. Secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems.

  3. NON-NRC FUNDED RELAP5-3D VERSION 4.x.x SOFTWARE REACTOR EXCURSION AND LEAK ANALYSIS PACKAGE - THREE DIMENSIONAL

    Energy Science and Technology Software Center (OSTI)

    2012-03-26

    The RELAP5-3D Version 3.x code has been developed for best-estimate transient simulation of nuclear reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems including pressurized watermore » reactors, boiling water reactors, Soviet-designed reactors, heavy water reactors, gas-cooled reactors, liquid metal and molten salt cooled reactors, and even fusion reactors. Numerical models include multi-dimensional hydrodynamics, 1- and 2-D heat transfer in metal walls, 0-, 1-, 2-, and 3-D neutron kinetics, trips, and control systems. Secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems.« less

  4. Engineering Database of Liquid Salt Thermophysical and Thermochemical Properties

    SciTech Connect (OSTI)

    Manohar S. Sohal; Matthias A. Ebner; Piyush Sabharwall; Phil Sharpe

    2010-03-01

    The purpose of this report is to provide a review of thermodynamic and thermophysical properties of candidate molten salt coolants, which may be used as a primary coolant within a nuclear reactor or heat transport medium from the Next Generation Nuclear Plant (NGNP) to a processing plant, for example, a hydrogen-production plant. Thermodynamic properties of four types of molten salts, including LiF-BeF2 (67 and 33 mol%, respectively; also known as FLiBe), LiF-NaF-KF (46.5, 11.5, and 52 mol%, also known as FLiNaK), and KCl-MgCl2 (67 and 33 mol%), and sodium nitrate-sodium nitrite-potassium nitrate (NaNO3–NaNO2–KNO3, (7-49-44 or 7-40-53 mol%) have been investigated. Limitations of existing correlations to predict density, viscosity, specific heat capacity, surface tension, and thermal conductivity, were identified. The impact of thermodynamic properties on the heat transfer, especially Nusselt number was also discussed. Stability of the molten salts with structural alloys and their compatibility with the structural alloys was studied. Nickel and alloys with dense Ni coatings are effectively inert to corrosion in fluorides but not so in chlorides. Of the chromium containing alloys, Hastelloy N appears to have the best corrosion resistance in fluorides, while Haynes 230 was most resistant in chloride. In general, alloys with increasing carbon and chromium content are increasingly subject to corrosion by the fluoride salts FLiBe and FLiNaK, due to attack and dissolution of the intergranular chromium carbide. Future research to obtain needed information was identified.

  5. Impact of corrosion test container material in molten fluorides

    SciTech Connect (OSTI)

    Olson, Luke C.; Fuentes, Roderick E.; Martinez-Rodriguez, Michael J.; Ambrosek, James W.; Sridharan, Kumar; Anderson, Mark H.; Garcia-Diaz, Brenda L.; Gray, Joshua; Allen, Todd R.

    2015-10-15

    The effects of crucible material choice on alloy corrosion rates in immersion tests in molten LiF–NaF–KF (46.5–11.5-42 mol. %) salt held at 850 °C for 500 hrs are described. Four crucible materials were studied. Molten salt exposures of Incoloy-800H in graphite, Ni, Incoloy-800H, and pyrolytic boron nitride (PyBN) crucibles all led to weight-loss in the Incoloy-800H coupons. Alloy weight loss was ~30 times higher in the graphite and Ni crucibles in comparison to the Incoloy-800H and PyBN crucibles. It is hypothesized galvanic coupling between the alloy coupons and crucible materials contributed to the higher corrosion rates. Alloy salt immersion in graphite and Ni crucibles had similar weight-loss hypothesized to occur due to the rate limiting out diffusion of Cr in the alloys to the surface where it reacts with and dissolves into the molten salt, followed by the reduction of Cr from solution at the molten salt and graphite/Ni interfaces. As a result, both the graphite and the Ni crucibles provided sinks for the Cr, in the formation of a Ni–Cr alloy in the case of the Ni crucible, and Cr carbide in the case of the graphite crucible.

  6. Impact of corrosion test container material in molten fluorides

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Olson, Luke C.; Fuentes, Roderick E.; Martinez-Rodriguez, Michael J.; Ambrosek, James W.; Sridharan, Kumar; Anderson, Mark H.; Garcia-Diaz, Brenda L.; Gray, Joshua; Allen, Todd R.

    2015-10-15

    The effects of crucible material choice on alloy corrosion rates in immersion tests in molten LiF–NaF–KF (46.5–11.5-42 mol. %) salt held at 850 °C for 500 hrs are described. Four crucible materials were studied. Molten salt exposures of Incoloy-800H in graphite, Ni, Incoloy-800H, and pyrolytic boron nitride (PyBN) crucibles all led to weight-loss in the Incoloy-800H coupons. Alloy weight loss was ~30 times higher in the graphite and Ni crucibles in comparison to the Incoloy-800H and PyBN crucibles. It is hypothesized galvanic coupling between the alloy coupons and crucible materials contributed to the higher corrosion rates. Alloy salt immersion inmore » graphite and Ni crucibles had similar weight-loss hypothesized to occur due to the rate limiting out diffusion of Cr in the alloys to the surface where it reacts with and dissolves into the molten salt, followed by the reduction of Cr from solution at the molten salt and graphite/Ni interfaces. As a result, both the graphite and the Ni crucibles provided sinks for the Cr, in the formation of a Ni–Cr alloy in the case of the Ni crucible, and Cr carbide in the case of the graphite crucible.« less

  7. Fundamental Properties of Salts

    SciTech Connect (OSTI)

    Toni Y Gutknecht; Guy L Fredrickson

    2012-11-01

    Thermal properties of molten salt systems are of interest to electrorefining operations, pertaining to both the Fuel Cycle Research & Development Program (FCR&D) and Spent Fuel Treatment Mission, currently being pursued by the Department of Energy (DOE). The phase stability of molten salts in an electrorefiner may be adversely impacted by the build-up of fission products in the electrolyte. Potential situations that need to be avoided, during electrorefining operations, include (i) fissile elements build up in the salt that might approach the criticality limits specified for the vessel, (ii) electrolyte freezing at the operating temperature of the electrorefiner due to changes in the liquidus temperature, and (iii) phase separation (non-homogenous solution). The stability (and homogeneity) of the phases can be monitored by studying the thermal characteristics of the molten salts as a function of impurity concentration. Simulated salt compositions consisting of the selected rare earth and alkaline earth chlorides, with a eutectic mixture of LiCl-KCl as the carrier electrolyte, were studied to determine the melting points (thermal characteristics) using a Differential Scanning Calorimeter (DSC). The experimental data were used to model the liquidus temperature. On the basis of the this data, it became possible to predict a spent fuel treatment processing scenario under which electrorefining could no longer be performed as a result of increasing liquidus temperatures of the electrolyte.

  8. An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

    SciTech Connect (OSTI)

    Rosenthal, Murray Wilford

    2009-08-01

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

  9. Hot filament technique for measuring the thermal conductivity of molten lithium fluoride

    SciTech Connect (OSTI)

    Jaworske, D.A.; Perry, W.D.

    1990-01-01

    Molten salts, such as lithium fluoride, are attractive candidates for thermal energy storage in solar dynamic space power systems because of their high latent heat of fusion. However, these same salts have poor thermal conductivities which inhibit the transfer of heat into the solid phase and out of the liquid phase. One concept for improving the thermal conductivity of the thermal energy storage system is to add a conductive filler material to the molten salt. High thermal conductivity pitch-based graphite fibers are being considered for this application. Although there is some information available on the thermal conductivity of lithium fluoride solid, there is very little information on lithium fluoride liquid, and no information on molten salt graphite fiber composites. This paper describes a hot filament technique for determining the thermal conductivity of molten salts. The hot filament technique was used to find the thermal conductivity of molten lithium fluoride at 930 C, and the thermal conductivity values ranged from 1.2 to 1.6 W/mK. These values are comparable to the slightly larger value of 5.0 W/mK for lithium fluoride solid. In addition, two molten salt graphite fiber composites were characterized with the hot filament technique and these results are also presented.

  10. Management of Salt Waste from Electrochemical Processing of Used Nuclear Fuel

    SciTech Connect (OSTI)

    Michael F. Simpson; Michael N. Patterson; Joon Lee; Yifeng Wang; Joshua Versey; Ammon Williams; Supathorn Phongikaroon; James Allensworth; Man-Sung Yim

    2013-10-01

    Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electrorefiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form.

  11. Management of salt waste from electrochemical processing of used nuclear fuel

    SciTech Connect (OSTI)

    Simpson, M.F.; Patterson, M.N.; Lee, J.; Wang, Y.; Versey, J.; Phongikaroon, S.

    2013-07-01

    Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electro-refiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form. (authors)

  12. ASME Material Challenges for Advanced Reactor Concepts

    SciTech Connect (OSTI)

    Piyush Sabharwall; Ali Siahpush

    2013-07-01

    This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at higher temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.

  13. Molten uranium dioxide structure and dynamics

    SciTech Connect (OSTI)

    Skinner, L. B.; Parise, J. B.; Benmore, C. J.; Weber, J. K.R.; Williamson, M. A.; Tamalonis, A.; Hebden, A.; Wiencek, T.; Alderman, O. L.G.; Guthrie, M.; Leibowitz, L.

    2014-11-21

    Uranium dioxide (UO2) is the major nuclear fuel component of fission power reactors. A key concern during severe accidents is the melting and leakage of radioactive UO2 as it corrodes through its zirconium cladding and steel containment. Yet, the very high temperatures (>3140 kelvin) and chemical reactivity of molten UO2 have prevented structural studies. In this work, we combine laser heating, sample levitation, and synchrotron x-rays to obtain pair distribution function measurements of hot solid and molten UO2. The hot solid shows a substantial increase in oxygen disorder around the lambda transition (2670 K) but negligible U-O coordination change. On melting, the average U-O coordination drops from 8 to 6.7 ± 0.5. Molecular dynamics models refined to this structure predict higher U-U mobility than 8-coordinated melts.

  14. Molten uranium dioxide structure and dynamics

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Skinner, L. B.; Parise, J. B.; Benmore, C. J.; Weber, J. K.R.; Williamson, M. A.; Tamalonis, A.; Hebden, A.; Wiencek, T.; Alderman, O. L.G.; Guthrie, M.; et al

    2014-11-21

    Uranium dioxide (UO2) is the major nuclear fuel component of fission power reactors. A key concern during severe accidents is the melting and leakage of radioactive UO2 as it corrodes through its zirconium cladding and steel containment. Yet, the very high temperatures (>3140 kelvin) and chemical reactivity of molten UO2 have prevented structural studies. In this work, we combine laser heating, sample levitation, and synchrotron x-rays to obtain pair distribution function measurements of hot solid and molten UO2. The hot solid shows a substantial increase in oxygen disorder around the lambda transition (2670 K) but negligible U-O coordination change. Onmore » melting, the average U-O coordination drops from 8 to 6.7 ± 0.5. Molecular dynamics models refined to this structure predict higher U-U mobility than 8-coordinated melts.« less

  15. REACTOR

    DOE Patents [OSTI]

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  16. REACTOR

    DOE Patents [OSTI]

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  17. Apparatus for controlling molten core debris

    DOE Patents [OSTI]

    Golden, Martin P. [Trafford, PA; Tilbrook, Roger W. [Monroeville, PA; Heylmun, Neal F. [Pittsburgh, PA

    1977-07-19

    Apparatus for containing, cooling, diluting, dispersing and maintaining subcritical the molten core debris assumed to melt through the bottom of a nuclear reactor pressure vessel in the unlikely event of a core meltdown. The apparatus is basically a sacrificial bed system which includes an inverted conical funnel, a core debris receptacle including a spherical dome, a spherically layered bed of primarily magnesia bricks, a cooling system of zig-zag piping in graphite blocks about and below the bed and a cylindrical liner surrounding the graphite blocks including a steel shell surrounded by firebrick. Tantalum absorber rods are used in the receptacle and bed.

  18. Apparatus for controlling molten core debris. [LMFBR

    DOE Patents [OSTI]

    Golden, M.P.; Tilbrook, R.W.; Heylmun, N.F.

    1977-07-19

    Disclosed is an apparatus for containing, cooling, diluting, dispersing and maintaining subcritical the molten core debris assumed to melt through the bottom of a nuclear reactor pressure vessel in the unlikely event of a core meltdown. The apparatus is basically a sacrificial bed system which includes an inverted conical funnel, a core debris receptacle including a spherical dome, a spherically layered bed of primarily magnesia bricks, a cooling system of zig-zag piping in graphite blocks about and below the bed and a cylindrical liner surrounding the graphite blocks including a steel shell surrounded by firebrick. Tantalum absorber rods are used in the receptacle and bed. 9 claims, 22 figures.

  19. Presence of Li clusters in molten LiCl-Li

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Merwin, Augustus; Phillips, William C.; Williamson, Mark A.; Willit, James L.; Motsegood, Perry N.; Chidambaram, Dev

    2016-05-05

    Molten mixtures of lithium chloride and metallic lithium are of significant interest in various metal oxide reduction processes. These solutions have been reported to exhibit seemingly anomalous physical characteristics that lack a comprehensive explanation. ln the current work, the physical chemistry of molten solutions of lithium chloride and metallic lithium, with and without lithium oxide, was investigated using in situ Raman spectroscopy. The Raman spectra obtained from these solutions were in agreement with the previously reported spectrum of the lithium cluster, Li8. Furthermore, this observation is indicative of a nanofluid type colloidal suspension of Li8, in a molten salt matrix.more » It is suggested that the formation and suspension of lithium clusters in lithium chloride is the cause of various phenomena exhibited by these solutions that were previously unexplainable.« less

  20. Multi-Purpose Thermal Hydraulic Loop: Advanced Reactor Technology Integral System Test (ARTIST) Facility for Support of Advanced Reactor Technologies

    SciTech Connect (OSTI)

    James E. O'Brien; Piyush Sabharwall; SuJong Yoon

    2001-11-01

    Effective and robust high temperature heat transfer systems are fundamental to the successful deployment of advanced reactors for both power generation and non-electric applications. Plant designs often include an intermediate heat transfer loop (IHTL) with heat exchangers at either end to deliver thermal energy to the application while providing isolation of the primary reactor system. In order to address technical feasibility concerns and challenges a new high-temperature multi-fluid, multi-loop test facility “Advanced Reactor Technology Integral System Test facility” (ARTIST) is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water. Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed

  1. REACTOR

    DOE Patents [OSTI]

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  2. REACTORS

    DOE Patents [OSTI]

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  3. Molten carbonate fuel cell

    DOE Patents [OSTI]

    Kaun, Thomas D.; Smith, James L.

    1987-01-01

    A molten electrolyte fuel cell with an array of stacked cells and cell enclosures isolating each cell except for access to gas manifolds for the supply of fuel or oxidant gas or the removal of waste gas, the cell enclosures collectively providing an enclosure for the array and effectively avoiding the problems of electrolyte migration and the previous need for compression of stack components, the fuel cell further including an inner housing about and in cooperation with the array enclosure to provide a manifold system with isolated chambers for the supply and removal of gases. An external insulated housing about the inner housing provides thermal isolation to the cell components.

  4. Molten carbonate fuel cell

    DOE Patents [OSTI]

    Kaun, T.D.; Smith, J.L.

    1986-07-08

    A molten electrolyte fuel cell is disclosed with an array of stacked cells and cell enclosures isolating each cell except for access to gas manifolds for the supply of fuel or oxidant gas or the removal of waste gas. The cell enclosures collectively provide an enclosure for the array and effectively avoid the problems of electrolyte migration and the previous need for compression of stack components. The fuel cell further includes an inner housing about and in cooperation with the array enclosure to provide a manifold system with isolated chambers for the supply and removal of gases. An external insulated housing about the inner housing provides thermal isolation to the cell components.

  5. Molten carbonate fuel cell cathode with mixed oxide coating

    DOE Patents [OSTI]

    Hilmi, Abdelkader; Yuh, Chao-Yi

    2013-05-07

    A molten carbonate fuel cell cathode having a cathode body and a coating of a mixed oxygen ion conductor materials. The mixed oxygen ion conductor materials are formed from ceria or doped ceria, such as gadolinium doped ceria or yttrium doped ceria. The coating is deposited on the cathode body using a sol-gel process, which utilizes as precursors organometallic compounds, organic and inorganic salts, hydroxides or alkoxides and which uses as the solvent water, organic solvent or a mixture of same.

  6. SEPARATION OF METAL SALTS BY ADSORPTION

    DOE Patents [OSTI]

    Gruen, D.M.

    1959-01-20

    It has been found that certain metal salts, particularly the halides of iron, cobalt, nickel, and the actinide metals, arc readily absorbed on aluminum oxide, while certain other salts, particularly rare earth metal halides, are not so absorbed. Use is made of this discovery to separate uranium from the rare earths. The metal salts are first dissolved in a molten mixture of alkali metal nitrates, e.g., the eutectic mixture of lithium nitrate and potassium nitrate, and then the molten salt solution is contacted with alumina, either by slurrying or by passing the salt solution through an absorption tower. The process is particularly valuable for the separation of actinides from lanthanum-group rare earths.

  7. Liquid Salt Heat Exchanger Technology for VHTR Based Applications

    SciTech Connect (OSTI)

    Anderson, Mark; Sridhara, Kumar; Allen, Todd; Peterson, Per

    2012-10-11

    The objective of this research is to evaluate performance of liquid salt fluids for use as a heat carrier for transferring high-temperature process heat from the very high-temperature reactor (VHTR) to chemical process plants. Currently, helium is being considered as the heat transfer fluid; however, the tube size requirements and the power associated with pumping helium may not be economical. Recent work on liquid salts has shown tremendous potential to transport high-temperature heat efficiently at low pressures over long distances. This project has two broad objectives: To investigate the compatibility of Incoloy 617 and coated and uncoated SiC ceramic composite with MgCl2-KCl molten salt to determine component lifetimes and aid in the design of heat exchangers and piping; and, To conduct the necessary research on the development of metallic and ceramic heat exchangers, which are needed for both the helium-to-salt side and salt-to-process side, with the goal of making these heat exchangers technologically viable. The research will consist of three separate tasks. The first task deals with material compatibility issues with liquid salt and the development of techniques for on-line measurement of corrosion products, which can be used to measure material loss in heat exchangers. Researchers will examine static corrosion of candidate materials in specific high-temperature heat transfer salt systems and develop an in situ electrochemical probe to measure metallic species concentrations dissolved in the liquid salt. The second task deals with the design of both the intermediate and process side heat exchanger systems. Researchers will optimize heat exchanger design and study issues related to corrosion, fabrication, and thermal stresses using commercial and in-house codes. The third task focuses integral testing of flowing liquid salts in a heat transfer/materials loop to determine potential issues of using the salts and to capture realistic behavior of the salts in a

  8. The Development of an INL Capability for High Temperature Flow, Heat Transfer, and Thermal Energy Storage with Applications in Advanced Small Modular Reactors, High Temperature Heat Exchangers, Hybrid Energy Systems, and Dynamic Grid Energy Storage C

    SciTech Connect (OSTI)

    Sun, Xiaodong; Zhang, Xiaoqin; Kim, Inhun; O'Brien, James; Sabharwall, Piyush

    2014-10-01

    The overall goal of this project is to support Idaho National Laboratory in developing a new advanced high temperature multi fluid multi loop test facility that is aimed at investigating fluid flow and heat transfer, material corrosion, heat exchanger characteristics and instrumentation performance, among others, for nuclear applications. Specifically, preliminary research has been performed at The Ohio State University in the following areas: 1. A review of fluoride molten salts’ characteristics in thermal, corrosive, and compatibility performances. A recommendation for a salt selection is provided. Material candidates for both molten salt and helium flow loop have been identified. 2. A conceptual facility design that satisfies the multi loop (two coolant loops [i.e., fluoride molten salts and helium]) multi purpose (two operation modes [i.e., forced and natural circulation]) requirements. Schematic models are presented. The thermal hydraulic performances in a preliminary printed circuit heat exchanger (PCHE) design have been estimated. 3. An introduction of computational methods and models for pipe heat loss analysis and cases studies. Recommendations on insulation material selection have been provided. 4. An analysis of pipe pressure rating and sizing. Preliminary recommendations on pipe size selection have been provided. 5. A review of molten fluoride salt preparation and chemistry control. An introduction to the experience from the Molten Salt Reactor Experiment at Oak Ridge National Laboratory has been provided. 6. A review of some instruments and components to be used in the facility. Flowmeters and Grayloc connectors have been included. This report primarily presents the conclusions drawn from the extensive review of literatures in material selections and the facility design progress at the current stage. It provides some useful guidelines in insulation material and pipe size selection, as well as an introductory review of facility process and components.

  9. Reactor

    DOE Patents [OSTI]

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  10. Design report on SCDAP/RELAP5 model improvements - debris bed and molten pool behavior

    SciTech Connect (OSTI)

    Allison, C.M.; Rempe, J.L.; Chavez, S.A.

    1994-11-01

    the SCDAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and in combination with VICTORIA, fission product release and transport during severe accidents. Improvements for existing debris bed and molten pool models in the SCDAP/RELAP5/MOD3.1 code are described in this report. Model improvements to address (a) debris bed formation, heating, and melting; (b) molten pool formation and growth; and (c) molten pool crust failure are discussed. Relevant data, existing models, proposed modeling changes, and the anticipated impact of the changes are discussed. Recommendations for the assessment of improved models are provided.

  11. Liquid salt environment stress-rupture testing

    DOE Patents [OSTI]

    Ren, Weiju; Holcomb, David E.; Muralidharan, Govindarajan; Wilson, Dane F.

    2016-03-22

    Disclosed herein are systems, devices and methods for stress-rupture testing selected materials within a high-temperature liquid salt environment. Exemplary testing systems include a load train for holding a test specimen within a heated inert gas vessel. A thermal break included in the load train can thermally insulate a load cell positioned along the load train within the inert gas vessel. The test specimen can include a cylindrical gage portion having an internal void filled with a molten salt during stress-rupture testing. The gage portion can have an inner surface area to volume ratio of greater than 20 to maximize the corrosive effect of the molten salt on the specimen material during testing. Also disclosed are methods of making a salt ingot for placement within the test specimen.

  12. Molten metal injector system and method

    DOE Patents [OSTI]

    Meyer, Thomas N.; Kinosz, Michael J.; Bigler, Nicolas; Arnaud, Guy

    2003-04-01

    Disclosed is a molten metal injector system including a holder furnace, a casting mold supported above the holder furnace, and a molten metal injector supported from a bottom side of the mold. The holder furnace contains a supply of molten metal having a metal oxide film surface. The bottom side of the mold faces the holder furnace. The mold defines a mold cavity for receiving the molten metal from the holder furnace. The injector projects into the holder furnace and is in fluid communication with the mold cavity. The injector includes a piston positioned within a piston cavity defined by a cylinder for pumping the molten metal upward from the holder furnace and injecting the molten metal into the mold cavity under pressure. The piston and cylinder are at least partially submerged in the molten metal when the holder furnace contains the molten metal. The cylinder further includes a molten metal intake for receiving the molten metal into the piston cavity. The molten metal intake is located below the metal oxide film surface of the molten metal when the holder furnace contains the molten metal. A method of injecting molten metal into a mold cavity of a casting mold is also disclosed.

  13. Analysis of fluid fuel flow to the neutron kinetics on molten...

    Office of Scientific and Technical Information (OSTI)

    Data of reactivity, neutron flux, and the macroscopic fission cross section for ... NEUTRONS; NUCLEAR DATA COLLECTIONS; REACTIVITY; REACTOR OPERATION; SALTS; THERMAL ...

  14. A Feasibility Study of Steelmaking by Molten Oxide Electrolysis (TRP9956)

    SciTech Connect (OSTI)

    Donald R. Sadoway; Gerbrand Ceder

    2009-12-31

    Molten oxide electrolysis (MOE) is an extreme form of molten salt electrolysis, a technology that has been used to produce tonnage metals for over 100 years - aluminum, magnesium, lithium, sodium and the rare earth metals specifically. The use of carbon-free anodes is the distinguishing factor in MOE compared to other molten salt electrolysis techniques. MOE is totally carbon-free and produces no CO or CO2 - only O2 gas at the anode. This project is directed at assessing the technical feasibility of MOE at the bench scale while determining optimum values of MOE operating parameters. An inert anode will be identified and its ability to sustain oxygen evalution will be demonstrated.

  15. Advanced Reactors Thermal Energy Transport for Process Industries

    SciTech Connect (OSTI)

    P. Sabharwall; S.J. Yoon; M.G. McKellar; C. Stoots; George Griffith

    2014-07-01

    The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as liquid fuel production, district heating, desalination, hydrogen production, and other process heat applications, etc. Some of the major technology challenges that must be overcome before the advanced reactors could be licensed on the reactor side are qualification of next generation of nuclear fuel, materials that can withstand higher temperature, improvement in power cycle thermal efficiency by going to combined cycles, SCO2 cycles, successful demonstration of advanced compact heat exchangers in the prototypical conditions, and from the process side application the challenge is to transport the thermal energy from the reactor to the process plant with maximum efficiency (i.e., with minimum temperature drop). The main focus of this study is on doing a parametric study of efficient heat transport system, with different coolants (mainly, water, He, and molten salts) to determine maximum possible distance that can be achieved.

  16. Noncentrosymmetric salt inclusion oxides: Role of salt lattices and counter ions in bulk polarity

    SciTech Connect (OSTI)

    West, J. Palmer; Hwu, Shiou-Jyh

    2012-11-15

    The synthesis and structural features of a newly emerged class of salt-inclusion solids (SISs) are reviewed. The descriptive chemistry with respect to the role of ionic salt and its correlation with bulk noncentrosymmetricity and polarity of the covalent oxide lattice in question is discussed by means of structure analysis. These unprecedented discoveries have opened doors to novel materials synthesis via the utilities of salt-inclusion chemistry (SIC) that are otherwise known as the molten-salt approach. The result of these investigations prove that the bulk acentricity, or cancellation of which, can be accounted for from the perspective of ionic and/or salt lattices. Highlights: Black-Right-Pointing-Pointer Synthesis and structure of newly emerged salt-inclusion solids are reviewed. Black-Right-Pointing-Pointer Salt lattice and its symmetry correlation with polar framework are discussed. Black-Right-Pointing-Pointer Preservation of acentricity is accounted for from the perspective of ionic and salt lattices.

  17. DEVELOPMENT OF A MULTI-LOOP FLOW AND HEAT TRANSFER FACILITY FOR ADVANCED NUCLEAR REACTOR THERMAL HYDRAULIC AND HYBRID ENERGY SYSTEM STUDIES

    SciTech Connect (OSTI)

    James E. O'Brien; Piyush Sabharwall; SuJong Yoon

    2001-09-01

    A new high-temperature multi-fluid, multi-loop test facility for advanced nuclear applications is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water. Molten salts have been identified as excellent candidate heat transport fluids for primary or secondary coolant loops, supporting advanced high temperature and small modular reactors (SMRs). Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed. A preliminary design configuration will be presented, with the required characteristics of the various components. The loop will utilize advanced high-temperature compact printed-circuit heat exchangers (PCHEs) operating at prototypic intermediate heat exchanger (IHX) conditions. The initial configuration will include a high-temperature (750°C), high-pressure (7 MPa) helium loop thermally integrated with a molten fluoride salt (KF-ZrF4) flow loop operating at low pressure (0.2 MPa) at a temperature of ~450°C. Experiment design challenges include identification of suitable materials and components that will withstand the required loop operating conditions. Corrosion and high temperature creep behavior are major considerations. The facility will include a thermal energy storage capability designed to support scaled process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will also provide important data for code ve

  18. Analyses of High Pressure Molten Debris Dispersion for a Typical PWR Plant

    SciTech Connect (OSTI)

    Osamu KAawabata; Mitsuhiro Kajimoto [Japan Nuclear Energy Safety Organization (Japan)

    2006-07-01

    In such severe core damage accident, as small LOCAs with no ECCS injection or station blackout, in which the primary reactor system remains pressurized during core melt down, certain modes of vessel failure would lead to a high pressure ejection of molten core material. In case of a local failure of the lower head, the molten materials would initially be ejected into the cavity beneath the pressure vessel may subsequently be swept out from the cavity to the containment atmosphere and it might cause the early containment failure by direct contact of containment steel liner with core debris. When the contribution of a high-pressure scenario in a core damage frequency increases, early conditional containment failure probability may become large. In the present study, the verification analysis of PHOENICS code and the combining analysis with MELCOR and PHOENICS codes were performed to examine the debris dispersion behavior during high pressure melt ejection. The PHOENICS code which can treat thermal hydraulic phenomena, was applied to the verification analysis for melt dispersion experiments conducted by the Purdue university in the United States. A low pressure melt dispersion experiment at initial pressure 1.4 MPas used metal woods as a molten material was simulated. The analytical results with molten debris dispersion mostly from the model reactor cavity compartment showed an agreement with the experimental result, but the analysis result of a volumetric median diameter of the airborne debris droplets was estimated about 1.5 times of the experimental result. The injection rates of molten debris and steam after reactor vessel failure for a typical PWR plant were analyzed using the MELCOR code. In addition, PHOENICS was applied to a 3D analysis for debris dispersion with low primary pressure at the reactor vessel failure. The analysis result showed that almost all the molten debris were dispersed from the reactor vessel cavity compartment by about 45 seconds after the

  19. Recovery of protactinium from molten fluoride nuclear fuel compositions

    DOE Patents [OSTI]

    Baes, C.F. Jr.; Bamberger, C.; Ross, R.G.

    1973-12-25

    A method is provided for separating protactinium from a molten fluonlde salt composition consisting essentially of at least one alkali and alkaline earth metal fluoride and at least one soluble fluoride of uranium or thorium which comprises oxidizing the protactinium in said composition to the + 5 oxidation state and contacting said composition with an oxide selected from the group consisting of an alkali metal oxide, an alkaline earth oxide, thorium oxide, and uranium oxide, and thereafter isolating the resultant insoluble protactinium oxide product from said composition. (Official Gazette)

  20. Anodic oxidation of sulfide ions in molten lithium fluoride

    SciTech Connect (OSTI)

    Lloyd, C.L.; Gilbert, J.B. II . Applied Research Lab.)

    1994-10-01

    The study of sulfur and sulfide oxidation in molten salt systems is of current interest in high energy battery, and metallurgical applications. Cyclic voltammetry experiments have been performed on lithium sulfide in a lithium fluoride electrolyte at 1,161 K using a graphite working electrode and a platinum quasi-reference electrode. Two distinct oxidation mechanisms are observed for the sulfide ions. The first oxidation produces sulfur and at a higher potential a disulfide species is proposed to have formed. Both oxidations appear to be reversible and diffusion controlled.

  1. Cathode preparation method for molten carbonate fuel cell

    DOE Patents [OSTI]

    Smith, James L.; Sim, James W.; Kucera, Eugenia H.

    1988-01-01

    A method of preparing a porous cathode structure for use in a molten carbonate fuel cell begins by providing a porous integral plaque of sintered nickel oxide particles. The nickel oxide plaque can be obtained by oxidizing a sintered plaque of nickel metal or by compacting and sintering finely divided nickel oxide particles to the desired pore structure. The porous sintered nickel oxide plaque is contacted with a lithium salt for a sufficient time to lithiate the nickel oxide structure and thus enhance its electronic conductivity. The lithiation can be carried out either within an operating fuel cell or prior to assembling the plaque as a cathode within the fuel cell.

  2. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High...

    Office of Scientific and Technical Information (OSTI)

    and Reactivity Control for Salt-Cooled High Temperature Reactors Citation Details In-Document Search Title: Pebble Fuel Handling and Reactivity Control for Salt-Cooled High ...

  3. Molten salt considerations for accelerator-driven subcritical...

    Office of Scientific and Technical Information (OSTI)

    It uniquely provides a method to close the nuclear fuel cycle for green nuclear energy. Authors: Sooby, Elizabeth ; Baty, Austin ; Gerity, James ; McIntyre, Peter ; Melconian, ...

  4. Multi-functional sensor system for molten salt technologies

    DOE Patents [OSTI]

    Redey, Laszlo; Gourishankar, Karthick; Williamson, Mark A.

    2009-12-15

    The present invention relates to a multi-functional sensor system that simultaneously measures cathode and anode electrode potentials, dissolved ion (i.e. oxide) concentration, and temperatures in an electrochemical cell. One embodiment of the invented system generally comprises: a reference(saturated) electrode, a reference(sensing) electrode, and a data acquisition system. Thermocouples are built into the two reference electrodes to provide important temperature information.

  5. Project Profile: Modular and Scalable Baseload Molten Salt Plant...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    (receiver, field piping, thermal storage, and steam generator) and their integration with eSolar's heliostat technology and a conventional reheat steam turbine power block. ...

  6. THERMODYNAMIC PROPERTIES OF MOLTEN NITRATE SALTS Joseph G. Cordaro

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the world as heat transfer fluids and phase change materials for concentrated solar power (CSP) production and storage. In order to design large production facilities,...

  7. Molten Salt-Carbon Nanotube Thermal Energy Storage for Concentrating...

    Office of Scientific and Technical Information (OSTI)

    ... Close Cite: Bibtex Format Close 0 pages in this document matching the terms "" Search For Terms: Enter terms in the toolbar above to search the full text of this document for ...

  8. Molten Nitrate Salt Development for Thermal Energy Storage in...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    An important component of thermal energy storage system optimization is selecting the working fluid used as the storage media andor heat transfer fluid. Large quantities of the ...

  9. Plutonium and minor actinides utilization in Thorium molten salt...

    Office of Scientific and Technical Information (OSTI)

    The original FUJI-12 design considers Thsup 233U or ThPu as main fuel. In accordance with the currently suggestion to stay away from the separation of Pu and minor actinides ...

  10. Pulsed deuterium lithium nuclear reactor

    SciTech Connect (OSTI)

    Fischer, A.G.

    1980-01-08

    A nuclear reactor that burns hydrogen bomb material 6-lithium deuterotritide to helium in successive microexplosions which are ignited electrically and enclosed by this same molten material, and that permits the conversion of the reaction heat into useful electrical power. A specially-constructed high-current pulse machine is discharged via a thermally-preformed highly conducting path through a mass of the molten salt 6lid1-xtx (0

  11. International Subcommittee Report

    Energy Savers [EERE]

    China also has a very aggressive R&D program on advanced reactors, including sodium fast reactors, high temperature gas cooled reactors, and molten salt reactors Continued ...

  12. International Subcommittee Report

    Broader source: Energy.gov (indexed) [DOE]

    China also has a very aggressive RD&D program on advanced reactors, including sodium fast reactors, high temperature gas cooled reactors, and molten salt reactors Chinese ...

  13. Molten carbonate fuel cell matrices

    DOE Patents [OSTI]

    Vogel, Wolfgang M.; Smith, Stanley W.

    1985-04-16

    A molten carbonate fuel cell including a cathode electrode of electrically conducting or semiconducting lanthanum containing material and an electrolyte containing matrix of an electrically insulating lanthanum perovskite. In addition, in an embodiment where the cathode electrode is LaMnO.sub.3, the matrix may include LaAlO.sub.3 or a lithium containing material such as LiAlO.sub.2 or Li.sub.2 TiO.sub.3.

  14. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  15. Technology Development Roadmap for the Advanced High Temperature Reactor Secondary Heat Exchanger

    SciTech Connect (OSTI)

    P. Sabharwall; M. McCllar; A. Siahpush; D. Clark; M. Patterson; J. Collins

    2012-09-01

    This Technology Development Roadmap (TDRM) presents the path forward for deploying large-scale molten salt secondary heat exchangers (MS-SHX) and recognizing the benefits of using molten salt as the heat transport medium for advanced high temperature reactors (AHTR). This TDRM will aid in the development and selection of the required heat exchanger for: power production (the first anticipated process heat application), hydrogen production, steam methane reforming, methanol to gasoline production, or ammonia production. This TDRM (a) establishes the current state of molten salt SHX technology readiness, (b) defines a path forward that systematically and effectively tests this technology to overcome areas of uncertainty, (c) demonstrates the achievement of an appropriate level of maturity prior to construction and plant operation, and (d) identifies issues and prioritizes future work for maturing the state of SHX technology. This study discusses the results of a preliminary design analysis of the SHX and explains the evaluation and selection methodology. An important engineering challenge will be to prevent the molten salt from freezing during normal and off-normal operations because of its high melting temperature (390°C for KF ZrF4). The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The need for efficiency, compactness, and safety challenge the capabilities of existing heat exchanger technology. The description of potential heat exchanger configurations or designs (such as printed circuit, spiral or helical coiled, ceramic, plate and fin, and plate type) were covered in an earlier report (Sabharwall et al. 2011). Significant future work, much of which is suggested in this report, is needed before the benefits and full potential of the AHTR can be realized. The execution of this TDRM will focuses

  16. Molten carbonate fuel cell separator

    DOE Patents [OSTI]

    Nickols, Richard C.

    1986-09-02

    In a stacked array of molten carbonate fuel cells, a fuel cell separator is positioned between adjacent fuel cells to provide isolation as well as a conductive path therebetween. The center portion of the fuel cell separator includes a generally rectangular, flat, electrical conductor. Around the periphery of the flat portion of the separator are positioned a plurality of elongated resilient flanges which form a gas-tight seal around the edges of the fuel cell. With one elongated flange resiliently engaging a respective edge of the center portion of the separator, the sealing flanges, which are preferably comprised of a noncorrosive material such as an alloy of yttrium, iron, aluminum or chromium, form a tight-fitting wet seal for confining the corrosive elements of the fuel cell therein. This arrangement permits a good conductive material which may be highly subject to corrosion and dissolution to be used in combination with a corrosion-resistant material in the fuel cell separator of a molten carbonate fuel cell for improved fuel cell conductivity and a gas-tight wet seal.

  17. Molten carbonate fuel cell separator

    DOE Patents [OSTI]

    Nickols, R.C.

    1984-10-17

    In a stacked array of molten carbonate fuel cells, a fuel cell separator is positioned between adjacent fuel cells to provide isolation as well as a conductive path therebetween. The center portion of the fuel cell separator includes a generally rectangular, flat, electrical conductor. Around the periphery of the flat portion of the separator are positioned a plurality of elongated resilient flanges which form a gas-tight seal around the edges of the fuel cell. With one elongated flange resiliently engaging a respective edge of the center portion of the separator, the sealing flanges, which are preferably comprised of a noncorrosive material such as an alloy of yttrium, iron, aluminum or chromium, form a tight-fitting wet seal for confining the corrosive elements of the fuel cell therein. This arrangement permits a good conductive material which may be highly subject to corrosion and dissolution to be used in combination with a corrosion-resistant material in the fuel cell separator of a molten carbonate fuel cell for improved fuel cell conductivity and a gas-tight wet seal.

  18. Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

    SciTech Connect (OSTI)

    Gougar, Hans David

    2015-10-01

    The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

  19. A vacuum disengager for tritium removal from HYLIFE-II Reactor Flibe

    SciTech Connect (OSTI)

    Dolan, T.J.; Longhurst, G.R.; Garcia-Otero, E.

    1992-09-01

    We have designed a vacuum disengager system to remove tritium from the Flibe (Li{sub 2}BeF{sub 4}) molten salt coolant of the HYLIFE-II fusion reactor. There is a two-stage vacuum disengager in each of three intermediate heat exchanger (IHX) loops. Each stage consists of a vacuum chamber 4 m in diameter and 7 m tall. As 0.2 mm diameter molten salt droplets fall vertically downward into the vacuum, most of the tritium diffuses out of the droplets and is pumped away. A fraction {Phi} {approximately}10{sup {minus}5} of the 8.6 MCi/day tritium source (from breeding in the Flibe and from unburned fuel) remains in the Flibe as it leaves the vacuum disengagers, and about 21% of that permeates into the intermediate coolant loop, so about 20 Ci/day leak into the steam system. With Flibe primary coolant and a vacuum disengager, it appears that an intermediate coolant loop is not needed to prevent tritium from leaking into the steam system. An experiment is needed to demonstrate Flibe vacuum disengager operation.

  20. Liquid Fluoride Salt Experimentation Using a Small Natural Circulation Cell

    SciTech Connect (OSTI)

    Yoder Jr, Graydon L; Heatherly, Dennis Wayne; Williams, David F; Elkassabgi, Yousri M.; Caja, Joseph; Caja, Mario; Jordan, John; Salinas, Roberto

    2014-04-01

    A small molten fluoride salt experiment has been constructed and tested to develop experimental techniques for application in liquid fluoride salt systems. There were five major objectives in developing this test apparatus: Allow visual observation of the salt during testing (how can lighting be introduced, how can pictures be taken, what can be seen) Determine if IR photography can be used to examine components submerged in the salt Determine if the experimental configuration provides salt velocity sufficient for collection of corrosion data for future experimentation Determine if a laser Doppler velocimeter can be used to quantify salt velocities. Acquire natural circulation heat transfer data in fluoride salt at temperatures up to 700oC All of these objectives were successfully achieved during testing with the exception of the fourth: acquiring velocity data using the laser Doppler velocimeter. This paper describes the experiment and experimental techniques used, and presents data taken during natural circulation testing.

  1. Carbide Coatings for Nickel Alloys, Graphite and Carbon/Carbon Composites to be used in Fluoride Salt Valves

    SciTech Connect (OSTI)

    Nagle, Denis; Zhang, Dajie

    2015-10-22

    The focus of this research was concerned with developing materials technology that supports the evolution of Generation IV Advanced High Temperature Reactor (AHTR) concepts. Specifically, we investigate refractory carbide coatings for 1) nickel alloys, and 2) commercial carbon-carbon composites (CCCs). Numerous compelling reasons have driven us to focus on carbon and carbide materials. First, unlike metals, the strength and modulus of CCCs increase with rising temperature. Secondly, graphite and carbon composites have been proven effective for resisting highly corrosive fluoride melts such as molten cryolite [Na₃AlF₆] at ~1000°C in aluminum reduction cells. Thirdly, graphite and carbide materials exhibit extraordinary radiation damage tolerance and stability up to 2000°C. Finally, carbides are thermodynamically more stable in liquid fluoride salt than the corresponding metals (i.e. Cr and Zr) found in nickel based alloys.

  2. Cathode for molten carbonate fuel cell

    DOE Patents [OSTI]

    Kaun, Thomas D.; Mrazek, Franklin C.

    1990-01-01

    A porous sintered cathode for a molten carbonate fuel cell and method of making same, the cathode including a skeletal structure of a first electronically conductive material slightly soluble in the electrolyte present in the molten carbonate fuel cell covered by fine particles of a second material of possibly lesser electronic conductivity insoluble in the electrolyte present in the molten carbonate fuel cell, the cathode having a porosity in the range of from about 60% to about 70% at steady-state cell operating conditions consisting of both macro-pores and micro-pores.

  3. Recirculating Molten Metal Supply System And Method

    DOE Patents [OSTI]

    Kinosz, Michael J.; Meyer, Thomas N.

    2003-07-01

    The melter furnace includes a heating chamber (16), a pump chamber (18), a degassing chamber (20), and a filter chamber (22). The pump chamber (18) is located adjacent the heating chamber (16) and houses a molten metal pump (30). The degassing chamber (20) is located adjacent and in fluid communication with the pump chamber (18), and houses a degassing mechanism (36). The filter chamber (22) is located adjacent and in fluid communication with the degassing chamber (20). The filter chamber (22) includes a molten metal filter (38). The melter furnace (12) is used to supply molten metal to an externally located holder furnace (14), which then recirculates molten metal back to the melter furnace (12).

  4. Electrode for molten carbonate fuel cell

    DOE Patents [OSTI]

    Iacovangelo, Charles D.; Zarnoch, Kenneth P.

    1983-01-01

    A sintered porous electrode useful for a molten carbonate fuel cell is produced which is composed of a plurality of 5 wt. % to 95 wt. % nickel balance copper alloy encapsulated ceramic particles sintered together by the alloy.

  5. Stainless steel corrosion by molten nitrates : analysis and lessons learned.

    SciTech Connect (OSTI)

    Kruizenga, Alan Michael

    2011-09-01

    A secondary containment vessel, made of stainless 316, failed due to severe nitrate salt corrosion. Corrosion was in the form of pitting was observed during high temperature, chemical stability experiments. Optical microscopy, scanning electron microscopy and energy dispersive spectroscopy were all used to diagnose the cause of the failure. Failure was caused by potassium oxide that crept into the gap between the primary vessel (alumina) and the stainless steel vessel. Molten nitrate solar salt (89% KNO{sub 3}, 11% NaNO{sub 3} by weight) was used during chemical stability experiments, with an oxygen cover gas, at a salt temperature of 350-700 C. Nitrate salt was primarily contained in an alumina vessel; however salt crept into the gap between the alumina and 316 stainless steel. Corrosion occurred over a period of approximately 2000 hours, with the end result of full wall penetration through the stainless steel vessel; see Figures 1 and 2 for images of the corrosion damage to the vessel. Wall thickness was 0.0625 inches, which, based on previous data, should have been adequate to avoid corrosion-induced failure while in direct contact with salt temperature at 677 C (0.081-inch/year). Salt temperatures exceeding 650 C lasted for approximately 14 days. However, previous corrosion data was performed with air as the cover gas. High temperature combined with an oxygen cover gas obviously drove corrosion rates to a much higher value. Corrosion resulted in the form of uniform pitting. Based on SEM and EDS data, pits contained primarily potassium oxide and potassium chromate, reinforcing the link between oxides and severe corrosion. In addition to the pitting corrosion, a large blister formed on the side wall, which was mainly composed of potassium, chromium and oxygen. All data indicated that corrosion initiated internally and moved outward. There was no evidence of intergranular corrosion nor were there any indication of fast pathways along grain boundaries. Much of the

  6. Method for making a uranium chloride salt product

    DOE Patents [OSTI]

    Miller, William E.; Tomczuk, Zygmunt

    2004-10-05

    The subject apparatus provides a means to produce UCl.sub.3 in large quantities without incurring corrosion of the containment vessel or associated apparatus. Gaseous Cl is injected into a lower layer of Cd where CdCl.sub.2 is formed. Due to is lower density, the CdCl.sub.2 rises through the Cd layer into a layer of molten LiCl--KCL salt where a rotatable basket containing uranium ingots is suspended. The CdCl.sub.2 reacts with the uranium to form UCl.sub.3 and Cd. Due to density differences, the Cd sinks down to the liquid Cd layer and is reused. The UCl.sub.3 combines with the molten salt. During production the temperature is maintained at about 600.degree. C. while after the uranium has been depleted the salt temperature is lowered, the molten salt is pressure siphoned from the vessel, and the salt product LiCl--KCl-30 mol % UCl.sub.3 is solidified.

  7. Method and apparatus for atomization and spraying of molten metals

    DOE Patents [OSTI]

    Hobson, D.O.; Alexeff, I.; Sikka, V.K.

    1988-07-19

    A method and device for dispersing molten metal into fine particulate spray, the method comprises applying an electric current through the molten metal and simultaneously applying a magnetic field to the molten metal in a plane perpendicular to the electric current, whereby the molten metal is caused to form into droplets at an angle perpendicular to both the electric current and the magnetic field. The device comprises a structure for providing a molten metal, appropriately arranged electrodes for applying an electric current through the molten metal, and a magnet for providing a magnetic field in a plane perpendicular to the electric current. 11 figs.

  8. A method of measuring a molten metal liquid pool volume

    DOE Patents [OSTI]

    Garcia, G.V.; Carlson, N.M., Donaldson, A.D.

    1990-12-12

    A method of measuring a molten metal liquid pool volume and in particular molten titanium liquid pools, including the steps of (a) generating an ultrasonic wave at the surface of the molten metal liquid pool, (b) shining a light on the surface of a molten metal liquid pool, (c) detecting a change in the frequency of light, (d) detecting an ultrasonic wave echo at the surface of the molten metal liquid pool, and (e) computing the volume of the molten metal liquid. 3 figs.

  9. Sodium-tetravalent sulfur molten chloroaluminate cell

    DOE Patents [OSTI]

    Mamantov, Gleb (Knoxville, TN)

    1985-04-02

    A sodium-tetravalent sulfur molten chloroaluminate cell with a .beta."-alumina sodium ion conductor having a S-Al mole ratio of above about 0.15 in an acidic molten chloroaluminate cathode composition is disclosed. The cathode composition has an AlCl.sub.3 -NaCl mole percent ratio of above about 70-30 at theoretical full charge. The cell provides high energy densities at low temperatures and provides high energy densities and high power densities at moderate temperatures.

  10. Updated Generation IV Reactors Integrated Materials Technology Program Plan, Revision 2

    SciTech Connect (OSTI)

    Corwin, William R; Burchell, Timothy D; Halsey, William; Hayner, George; Katoh, Yutai; Klett, James William; McGreevy, Timothy E; Nanstad, Randy K; Ren, Weiju; Snead, Lance Lewis; Stoller, Roger E; Wilson, Dane F

    2005-12-01

    The Department of Energy's (DOE's) Generation IV Nuclear Energy Systems Program will address the research and development (R&D) necessary to support next-generation nuclear energy systems. Such R&D will be guided by the technology roadmap developed for the Generation IV International Forum (GIF) over two years with the participation of over 100 experts from the GIF countries. The roadmap evaluated over 100 future systems proposed by researchers around the world. The scope of the R&D described in the roadmap covers the six most promising Generation IV systems. The effort ended in December 2002 with the issue of the final Generation IV Technology Roadmap [1.1]. The six most promising systems identified for next generation nuclear energy are described within the roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor - SCWR and the Very High Temperature Reactor - VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor - GFR, the Lead-cooled Fast Reactor - LFR, and the Sodium-cooled Fast Reactor - SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides, and may provide an alternative to accelerator-driven systems. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. Accordingly, DOE has identified materials as one of the focus areas for Gen IV technology development.

  11. Molten Carbonate and Phosphoric Acid Stationary Fuel Cells: Overview and

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Gap Analysis | Department of Energy Molten Carbonate and Phosphoric Acid Stationary Fuel Cells: Overview and Gap Analysis Molten Carbonate and Phosphoric Acid Stationary Fuel Cells: Overview and Gap Analysis This report describes the technical and cost gap analysis performed to identify pathways for reducing the costs of molten carbonate fuel cell (MCFC) and phosphoric acid fuel cell (PAFC) stationary fuel cell power plants. Molten Carbonate and Phosphoric Acid Stationary Fuel Cells:

  12. Method for removing copper from molten metal with a molten slag and for recovering the copper from the slag

    SciTech Connect (OSTI)

    Oden, L.L.

    1993-12-31

    The present invention relates generally to a method for removing impurity metal from a molten metal such as iron and steel. It is a method for removing copper from molten iron and steel with a molten slag and thereafter recovering the copper from the slag.

  13. In-Vessel Retention of Molten Corium: Lessons Learned and Outstanding Issues

    SciTech Connect (OSTI)

    J.L. Rempe; K.Y. Suh; F. B. Cheung; S. B. Kim

    2008-03-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Advanced 600 MWe Pressurized Water Reactor (PWR) designed by Westinghouse (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing Light Water Reactors (LWRs). However, it is not clear that the ERVC proposed for the AP600 could provide sufficient heat removal for higher-power reactors (up to 1500 MWe) without additional enhancements. This paper reviews efforts made and results reported regarding the enhancement of IVR in LWRs. Where appropriate, the paper identifies what additional data or analyses are needed to demonstrate that there is sufficient margin for successful IVR in high power thermal reactors.

  14. Advanced High Temperature Reactor Systems and Economic Analysis

    SciTech Connect (OSTI)

    Holcomb, David Eugene; Peretz, Fred J; Qualls, A L

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience with

  15. The MSFR as a flexible CR reactor: the viewpoint of safety

    SciTech Connect (OSTI)

    Fiorina, C.; Cammi, A.; Franceschini, F.; Krepel, J.

    2013-07-01

    In this paper, the possibility has first been discussed of using the liquid-fuelled Molten Salt Fast Reactor (MSFR) as a flexible conversion ratio (CR) reactor without design modification. By tuning the reprocessing rate it is possible to determine the content of fission products in the core, which in turn can significantly affect the neutron economy without incurring in solubility problems. The MSFR can thus be operated as U-233 breeder (CR>1), iso-breeder (CR=1) and burner reactor (CR<1). In particular a 40 year doubling time can be achieved, as well as a considerable Transuranics and MA (minor actinide) burning rate equal to about 150 kg{sub HN}/GWE-yr. The safety parameters of the MSFR have then been evaluated for different fuel cycle strategies. Th use and a softer spectrum combine to give a strong Doppler coefficient, one order of magnitude higher compared to traditional fast reactors (FRs). The fuel expansion coefficient is comparable to the Doppler coefficient and is only mildly affected by core compositions, thus assisting the fuel cycle flexibility of the MSFR. ?eff and generation time are comparable to the case of traditional FRs, if a static fuel is assumed. A notable reduction of ?eff is caused by salt circulation, but a low value of this parameter is a limited concern in the MSFR thanks to the lack of a burnup reactivity swing and of positive feedbacks. A simple approach has also been developed to evaluate the MSFR capabilities to withstand all typical double-fault accidents, for different fuel cycle options.

  16. Nuclear reactor melt-retention structure to mitigate direct containment heating

    DOE Patents [OSTI]

    Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

    1991-01-01

    A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

  17. The Advanced High-Temperature Reactor (AHTR) for Producing Hydrogen to Manufacture Liquid Fuels

    SciTech Connect (OSTI)

    Forsberg, C.W.; Peterson, P.F.; Ott, L.

    2004-10-06

    Conventional world oil production is expected to peak within a decade. Shortfalls in production of liquid fuels (gasoline, diesel, and jet fuel) from conventional oil sources are expected to be offset by increased production of fuels from heavy oils and tar sands that are primarily located in the Western Hemisphere (Canada, Venezuela, the United States, and Mexico). Simultaneously, there is a renewed interest in liquid fuels from biomass, such as alcohol; but, biomass production requires fertilizer. Massive quantities of hydrogen (H2) are required (1) to convert heavy oils and tar sands to liquid fuels and (2) to produce fertilizer for production of biomass that can be converted to liquid fuels. If these liquid fuels are to be used while simultaneously minimizing greenhouse emissions, nonfossil methods for the production of H2 are required. Nuclear energy can be used to produce H2. The most efficient methods to produce H2 from nuclear energy involve thermochemical cycles in which high-temperature heat (700 to 850 C) and water are converted to H2 and oxygen. The peak nuclear reactor fuel and coolant temperatures must be significantly higher than the chemical process temperatures to transport heat from the reactor core to an intermediate heat transfer loop and from the intermediate heat transfer loop to the chemical plant. The reactor temperatures required for H2 production are at the limits of practical engineering materials. A new high-temperature reactor concept is being developed for H2 and electricity production: the Advanced High-Temperature Reactor (AHTR). The fuel is a graphite-matrix, coated-particle fuel, the same type that is used in modular high-temperature gas-cooled reactors (MHTGRs). The coolant is a clean molten fluoride salt with a boiling point near 1400 C. The use of a liquid coolant, rather than helium, reduces peak reactor fuel and coolant temperatures 100 to 200 C relative to those of a MHTGR. Liquids are better heat transfer fluids than gases

  18. Method and apparatus for spraying molten materials

    DOE Patents [OSTI]

    Glovan, Ronald J. (Butte, MT); Tierney, John C. (Butte, MT); McLean, Leroy L. (Butte, MT); Johnson, Lawrence L. (Butte, MT); Nelson, Gordon L. (Butte, MT); Lee, Ying-Ming (Butte, MT)

    1996-01-01

    A metal spray apparatus is provided with a supersonic nozzle. Molten metal is injected into a gas stream flowing through the nozzle under pressure. By varying the pressure of the injected metal, the droplet can be made in various selected sizes with each selected size having a high degree of size uniformity. A unique one piece graphite heater provides easily controlled uniformity of temperature in the nozzle and an attached tundish which holds the pressurized molten metal. A unique U-shaped gas heater provides extremely hot inlet gas temperatures to the nozzle. A particularly useful application of the spray apparatus is coating of threads of a fastener with a shape memory alloy. This permits a fastener to be easily inserted and removed but provides for a secure locking of the fastener in high temperature environments.

  19. Method and apparatus for spraying molten materials

    DOE Patents [OSTI]

    Glovan, R.J.; Tierney, J.C.; McLean, L.L.; Johnson, L.L.; Nelson, G.L.; Lee, Y.M.

    1996-06-25

    A metal spray apparatus is provided with a supersonic nozzle. Molten metal is injected into a gas stream flowing through the nozzle under pressure. By varying the pressure of the injected metal, the droplet can be made in various selected sizes with each selected size having a high degree of size uniformity. A unique one piece graphite heater provides easily controlled uniformity of temperature in the nozzle and an attached tundish which holds the pressurized molten metal. A unique U-shaped gas heater provides extremely hot inlet gas temperatures to the nozzle. A particularly useful application of the spray apparatus is coating of threads of a fastener with a shape memory alloy. This permits a fastener to be easily inserted and removed but provides for a secure locking of the fastener in high temperature environments. 12 figs.

  20. Molten metal holder furnace and casting system incorporating the molten metal holder furnace

    DOE Patents [OSTI]

    Kinosz, Michael J.; Meyer, Thomas N.

    2003-02-11

    A bottom heated holder furnace (12) for containing a supply of molten metal includes a storage vessel (30) having sidewalls (32) and a bottom wall (34) defining a molten metal receiving chamber (36). A furnace insulating layer (42) lines the molten metal receiving chamber (36). A thermally conductive heat exchanger block (54) is located at the bottom of the molten metal receiving chamber (36) for heating the supply of molten metal. The heat exchanger block (54) includes a bottom face (65), side faces (66), and a top face (67). The heat exchanger block (54) includes a plurality of electrical heaters (70) extending therein and projecting outward from at least one of the faces of the heat exchanger block (54), and further extending through the furnace insulating layer (42) and one of the sidewalls (32) of the storage vessel (30) for connection to a source of electrical power. A sealing layer (50) covers the bottom face (65) and side faces (66) of the heat exchanger block (54) such that the heat exchanger block (54) is substantially separated from contact with the furnace insulating layer (42).

  1. Molten-Caustic-Leaching System Integration Project

    SciTech Connect (OSTI)

    Not Available

    1992-01-01

    The objective of this project is to modify an existing molten-caustic-leaching (MCL) system for coal upgrading so that it operates in an integrated continuous manner. The overall strategy consists of several tasks, but only a few are discussed here. Tasks discussed are: MCL circuit component testing (coal sample procurement), final circuit modifications for integrated operation, coal product handling/waste disposal (coal inventory disposal, MCL solid waste disposal), project management and control. (VC)

  2. Alternative Waste Forms for Electro-Chemical Salt Waste

    SciTech Connect (OSTI)

    Crum, Jarrod V.; Sundaram, S. K.; Riley, Brian J.; Matyas, Josef; Arreguin, Shelly A.; Vienna, John D.

    2009-10-28

    This study was undertaken to examine alternate crystalline (ceramic/mineral) and glass waste forms for immobilizing spent salt from the Advanced Fuel Cycle Initiative (AFCI) electrochemical separations process. The AFCI is a program sponsored by U.S. Department of Energy (DOE) to develop and demonstrate a process for recycling spent nuclear fuel (SNF). The electrochemical process is a molten salt process for the reprocessing of spent nuclear fuel in an electrorefiner and generates spent salt that is contaminated with alkali, alkaline earths, and lanthanide fission products (FP) that must either be cleaned of fission products or eventually replaced with new salt to maintain separations efficiency. Currently, these spent salts are mixed with zeolite to form sodalite in a glass-bonded waste form. The focus of this study was to investigate alternate waste forms to immobilize spent salt. On a mole basis, the spent salt is dominated by alkali and Cl with minor amounts of alkaline earth and lanthanides. In the study reported here, we made an effort to explore glass systems that are more compatible with Cl and have not been previously considered for use as waste forms. In addition, alternate methods were explored with the hope of finding a way to produce a sodalite that is more accepting of as many FP present in the spent salt as possible. This study was done to investigate two different options: (1) alternate glass families that incorporate increased concentrations of Cl; and (2) alternate methods to produce a mineral waste form.

  3. Supported Molten Metal Membranes for Hydrogen Separation

    SciTech Connect (OSTI)

    Datta, Ravindra; Ma, Yi Hua; Yen, Pei-Shan; Deveau, Nicholas; Fishtik, Ilie; Mardilovich, Ivan

    2013-09-30

    We describe here our results on the feasibility of a novel dense metal membrane for hydrogen separation: Supported Molten Metal Membrane, or SMMM.1 The goal in this work was to develop these new membranes based on supporting thin films of low-melting, non- precious group metals, e.g., tin (Sn), indium (In), gallium (Ga), or their alloys, to provide a flux and selectivity of hydrogen that rivals the conventional but substantially more expensive palladium (Pd) or Pd alloy membranes, which are susceptible to poisoning by the many species in the coal-derived syngas, and further possess inadequate stability and limited operating temperature range. The novelty of the technology presented numerous challenges during the course of this project, however, mainly in the selection of appropriate supports, and in the fabrication of a stable membrane. While the wetting instability of the SMMM remains an issue, we did develop an adequate understanding of the interaction between molten metal films with porous supports that we were able to find appropriate supports. Thus, our preliminary results indicate that the Ga/SiC SMMM at 550 ºC has a permeance that is an order of magnitude higher than that of Pd, and exceeds the 2015 DOE target. To make practical SMM membranes, however, further improving the stability of the molten metal membrane is the next goal. For this, it is important to better understand the change in molten metal surface tension and contact angle as a function of temperature and gas-phase composition. A thermodynamic theory was, thus, developed, that is not only able to explain this change in the liquid-gas surface tension, but also the change in the solid-liquid surface tension as well as the contact angle. This fundamental understanding has allowed us to determine design characteristics to maintain stability in the face of changing gas composition. These designs are being developed. For further progress, it is also important to understand the nature of solution and

  4. Salt Waste Processing Initiatives

    Office of Environmental Management (EM)

    Patricia Suggs Salt Processing Team Lead Assistant Manager for Waste Disposition Project Office of Environmental Management Savannah River Site Salt Waste Processing Initiatives 2 ...

  5. PROCESS OF DISSOLVING FUEL ELEMENTS OF NUCLEAR REACTORS

    DOE Patents [OSTI]

    Wall, E.M.V.; Bauer, D.T.; Hahn, H.T.

    1963-09-01

    A process is described for dissolving stainless-steelor zirconium-clad uranium dioxide fuel elements by immersing the elements in molten lead chloride, adding copper, cuprous chloride, or cupric chloride as a catalyst and passing chlorine through the salt mixture. (AEC)

  6. Process of making electrolyte structure for molten carbonate fuel cells

    DOE Patents [OSTI]

    Arendt, R.H.; Curran, M.J.

    1980-08-05

    An electrolyte structure is produced by forming matrix material powder into a blank at room temperature and impregnating the resulting matrix blank with molten electrolyte.

  7. Molten Carbonate and Phosphoric Acid Stationary Fuel Cells: Overview...

    Broader source: Energy.gov (indexed) [DOE]

    This report describes the technical and cost gap analysis performed to identify pathways for reducing the costs of molten carbonate fuel cell (MCFC) and phosphoric acid fuel cell ...

  8. Electrochemical cell utilizing molten alkali metal electrode-reactant

    DOE Patents [OSTI]

    Virkar, Anil V.; Miller, Gerald R.

    1983-11-04

    An improved electrochemical cell comprising an additive-modified molten alkali metal electrode-reactant and/or electrolyte is disclosed. Various electrochemical cells employing a molten alkali metal, e.g., sodium, electrode in contact with a cationically conductive ceramic membrane experience a lower resistance and a lower temperature coefficient of resistance whenever small amounts of selenium are present at the interface of the electrolyte and the molten alkali metal. Further, cells having small amounts of selenium present at the electrolyte-molten metal interface exhibit less degradation of the electrolyte under long term cycling conditions.

  9. The application of electrorefining for recovery and purification of fuel discharged from the Integral Fast Reactor

    SciTech Connect (OSTI)

    Burris, L.; Steunenberg, R.K.; Miller, W.E.

    1986-01-01

    An electrorefining process employing a molten salt electrolyte and a molten cadmium anode is proposed for the separation of uranium and plutonium from fission products and cladding material in discharged IFR driver fuel. The use of a liquid cadmium anode, which is the unique feature of the process, permits selective dissolution of the fuel from the cladding and prevents electrolytic corrosion of the steel container and contamination of the product by noble metal fission products.

  10. Hydrodynamic simulation of a lithium chloride salt system.

    SciTech Connect (OSTI)

    Eberle, C. S.; Herrmann, S. D.; Knighton, G. C.

    1999-02-12

    A fused lithium chloride salt system's constitutive properties were evaluated and compared to a number of fluid properties, and water was shown to be an excellent simulant of lithium chloride salt. With a simple flow model, the principal scaling term was shown to be a function of the kinematic viscosity. A water mock-up of the molten salt was also shown to be within a {+-}3% error in the scaling analysis. This made it possible to consider developing water scaled tests of the molten salt system. Accurate flow velocity and pressure measurements were acquired by developing a directional velocity probe. The device was constructed and calibrated with a repeatable accuracy of {+-}15%. This was verified by a detailed evaluation of the probe. Extensive flow measurements of the engineering scale mockup were conducted, and the results were carefully compared to radial flow patterns of a straight blade stirrer. The flow measurements demonstrated an anti-symmetric nature of the stirring, and many additional effects were also identified. The basket design was shown to prevent fluid penetration into the fuel baskets when external stirring was the flow mechanism.

  11. Electrolyte salts for power sources

    DOE Patents [OSTI]

    Doddapaneni, Narayan; Ingersoll, David

    1995-01-01

    Electrolyte salts for power sources comprising salts of phenyl polysulfonic acids and phenyl polyphosphonic acids. The preferred salts are alkali and alkaline earth metal salts, most preferably lithium salts.

  12. Closing nuclear fuel cycle with fast reactors: problems and prospects

    SciTech Connect (OSTI)

    Shadrin, A.; Dvoeglazov, K.; Ivanov, V.

    2013-07-01

    The closed nuclear fuel cycle (CNFC) with fast reactors (FR) is the most promising way of nuclear energetics development because it prevents spent nuclear fuel (SNF) accumulation and minimizes radwaste volume due to minor actinides (MA) transmutation. CNFC with FR requires the elaboration of safety, environmentally acceptable and economically effective methods of treatment of SNF with high burn-up and low cooling time. The up-to-date industrially implemented SNF reprocessing technologies based on hydrometallurgical methods are not suitable for the reprocessing of SNF with high burn-up and low cooling time. The alternative dry methods (such as electrorefining in molten salts or fluoride technologies) applicable for such SNF reprocessing have not found implementation at industrial scale. So the cost of SNF reprocessing by means of dry technologies can hardly be estimated. Another problem of dry technologies is the recovery of fissionable materials pure enough for dense fuel fabrication. A combination of technical solutions performed with hydrometallurgical and dry technologies (pyro-technology) is proposed and it appears to be a promising way for the elaboration of economically, ecologically and socially accepted technology of FR SNF management. This paper deals with discussion of main principle of dry and aqueous operations combination that probably would provide safety and economic efficiency of the FR SNF reprocessing. (authors)

  13. Electrically conductive containment vessel for molten aluminum

    DOE Patents [OSTI]

    Holcombe, Cressie E.; Scott, Donald G.

    1985-01-01

    The present invention is directed to a containment vessel which is particularly useful in melting aluminum. The vessel of the present invention is a multilayered vessel characterized by being electrically conductive, essentially nonwettable by and nonreactive with molten aluminum. The vessel is formed by coating a tantalum substrate of a suitable configuration with a mixture of yttria and particulate metal borides. The yttria in the coating inhibits the wetting of the coating while the boride particulate material provides the electrical conductivity through the vessel. The vessel of the present invention is particularly suitable for use in melting aluminum by ion bombardment.

  14. Electrically conductive containment vessel for molten aluminum

    DOE Patents [OSTI]

    Holcombe, C.E.; Scott, D.G.

    1984-06-25

    The present invention is directed to a containment vessel which is particularly useful in melting aluminum. The vessel of the present invention is a multilayered vessel characterized by being electrically conductive, essentially nonwettable by and nonreactive with molten aluminum. The vessel is formed by coating a tantalum substrate of a suitable configuration with a mixture of yttria and particulate metal 10 borides. The yttria in the coating inhibits the wetting of the coating while the boride particulate material provides the electrical conductivity through the vessel. The vessel of the present invention is particularly suitable for use in melting aluminum by ion bombardment.

  15. Anode composite for molten carbonate fuel cell

    DOE Patents [OSTI]

    Iacovangelo, Charles D.; Zarnoch, Kenneth P.

    1983-01-01

    An anode composite useful for a molten carbonate fuel cell comprised of a porous sintered metallic anode component having a porous bubble pressure barrier integrally sintered to one face thereof, said barrier being comprised of metal coated ceramic particles sintered together and to said anode by means of said metal coating, said metal coating enveloping said ceramic particle and being selected from the group consisting of nickel, copper and alloys thereof, the median pore size of the barrier being significantly smaller than that of the anode.

  16. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    SciTech Connect (OSTI)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  17. Method of making molten carbonate fuel cell ceramic matrix tape

    DOE Patents [OSTI]

    Maricle, Donald L.; Putnam, Gary C.; Stewart, Jr., Robert C.

    1984-10-23

    A method of making a thin, flexible, pliable matrix material for a molten carbonate fuel cell is described. The method comprises admixing particles inert in the molten carbonate environment with an organic polymer binder and ceramic particle. The composition is applied to a mold surface and dried, and the formed compliant matrix material removed.

  18. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    SciTech Connect (OSTI)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  19. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, J.P.

    1992-03-17

    A process is described for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

  20. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, John P.

    1992-01-01

    A process for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

  1. Pump for molten metal or other fluid

    DOE Patents [OSTI]

    Horton, James A.; Brown, Donald L.

    1994-01-01

    A pump having no moving parts which can be used to pump high temperature molten metal or other fluids in a vacuum or low pressure environment, and a method for pumping such fluids. The pump combines elements of a bubble pump with a trap which isolates the vacuum or low pressure region from the gas used to create the bubbles. When used in a vacuum the trap prevents the pumping gas from escaping into the isolated region and thereby reducing the quality of the vacuum. The pump includes a channel in which a pumping gas is forced under pressure into a cavity where bubbles are formed. The cavity is in contact with a reservoir which contains the molten metal or other fluid which is to be pumped. The bubbles rise up into a column (or pump tube) carrying the fluid with them. At the top of the column is located a deflector which causes the bubbles to burst and the drops of pumped fluid to fall into a trap. The fluid accumulates in the trap, eventually forcing its way to an outlet. A roughing pump can be used to withdraw the pumping gas from the top of the column and assist with maintaining the vacuum or low pressure environment.

  2. Numerical Study on Crossflow Printed Circuit Heat Exchanger for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Yoon, Su-Jong; Sabharwall, Piyush; Kim, Eung-Soo

    2014-03-01

    Various fluids such as water, gases (helium), molten salts (FLiNaK, FLiBe) and liquid metal (sodium) are used as a coolant of advanced small modular reactors (SMRs). The printed circuit heat exchanger (PCHE) has been adopted as the intermediate and/or secondary heat exchanger of SMR systems because this heat exchanger is compact and effective. The size and cost of PCHE can be changed by the coolant type of each SMR. In this study, the crossflow PCHE analysis code for advanced small modular reactor has been developed for the thermal design and cost estimation of the heat exchanger. The analytical solution of single pass, both unmixed fluids crossflow heat exchanger model was employed to calculate a two dimensional temperature profile of a crossflow PCHE. The analytical solution of crossflow heat exchanger was simply implemented by using built in function of the MATLAB program. The effect of fluid property uncertainty on the calculation results was evaluated. In addition, the effect of heat transfer correlations on the calculated temperature profile was analyzed by taking into account possible combinations of primary and secondary coolants in the SMR systems. Size and cost of heat exchanger were evaluated for the given temperature requirement of each SMR.

  3. Analytical methods for determining the reactivity of pyrochemical salts

    SciTech Connect (OSTI)

    Phillips, A.G.; Stakebake, J.L.

    1994-05-01

    Pyrochemical processes used for the purification of plutonium have generated quantities of residue that contain varying amounts of reactive metals such as potassium, sodium, calcium, and magnesium. These residues are currently considered hazardous and are being managed under RCRA because of the reactivity characteristic. This designation is based solely on process knowledge. Currently there is no approved procedure for determining the reactivity of a solid with water. A method is being developed to rapidly evaluate the reactivity of pyrochemical salts with water by measuring the rate of hydrogen generation. The method was initially tested with a magnesium containing pyrochemical salt. A detection limit of approximately 0.004 g of magnesium was established. A surrogate molten salt extraction residue was also tested. Extrapolation of test data resulted in a hydrogen generation rate of 4.4 mg/(g min).

  4. Examination of Liquid Fluoride Salt Heat Transfer

    SciTech Connect (OSTI)

    Yoder Jr, Graydon L

    2014-01-01

    The need for high efficiency power conversion and energy transport systems is increasing as world energy use continues to increase, petroleum supplies decrease, and global warming concerns become more prevalent. There are few heat transport fluids capable of operating above about 600oC that do not require operation at extremely high pressures. Liquid fluoride salts are an exception to that limitation. Fluoride salts have very high boiling points, can operate at high temperatures and low pressures and have very good heat transfer properties. They have been proposed as coolants for next generation fission reactor systems, as coolants for fusion reactor blankets, and as thermal storage media for solar power systems. In each case, these salts are used to either extract or deliver heat through heat exchange equipment, and in order to design this equipment, liquid salt heat transfer must be predicted. This paper discusses the heat transfer characteristics of liquid fluoride salts. Historically, heat transfer in fluoride salts has been assumed to be consistent with that of conventional fluids (air, water, etc.), and correlations used for predicting heat transfer performance of all fluoride salts have been the same or similar to those used for water conventional fluids an, water, etc). A review of existing liquid salt heat transfer data is presented, summarized, and evaluated on a consistent basis. Less than 10 experimental data sets have been found in the literature, with varying degrees of experimental detail and measured parameters provided. The data has been digitized and a limited database has been assembled and compared to existing heat transfer correlations. Results vary as well, with some data sets following traditional correlations; in others the comparisons are less conclusive. This is especially the case for less common salt/materials combinations, and suggests that additional heat transfer data may be needed when using specific salt eutectics in heat transfer

  5. Casting Apparatus Including A Gas Driven Molten Metal Injector And Method

    DOE Patents [OSTI]

    Trudel, David R.; Meyer, Thomas N.; Kinosz, Michael J.; Arnaud, Guy; Bigler, Nicolas

    2003-06-17

    The filtering molten metal injector system includes a holder furnace, a casting mold supported above the holder furnace, and at least one molten metal injector supported from a bottom side of the casting mold. The holder furnace contains a supply of molten metal. The mold defines a mold cavity for receiving the molten metal from the holder furnace. The molten metal injector projects into the holder furnace. The molten metal injector includes a cylinder defining a piston cavity housing a reciprocating piston for pumping the molten metal upward from the holder furnace to the mold cavity. The cylinder and piston are at least partially submerged in the molten metal when the holder furnace contains the molten metal. The cylinder or the piston includes a molten metal intake for receiving the molten metal into the piston cavity when the holder furnace contains molten metal. A conduit connects the piston cavity to the mold cavity. A molten metal filter is located in the conduit for filtering the molten metal passing through the conduit during the reciprocating movement of the piston. The molten metal intake may be a valve connected to the cylinder, a gap formed between the piston and an open end of the cylinder, an aperture defined in the sidewall of the cylinder, or a ball check valve incorporated into the piston. A second molten metal filter preferably covers the molten metal intake to the injector.

  6. Electrolyte paste for molten carbonate fuel cells

    DOE Patents [OSTI]

    Bregoli, Lawrance J.; Pearson, Mark L.

    1995-01-01

    The electrolyte matrix and electrolyte reservoir plates in a molten carbonate fuel cell power plant stack are filled with electrolyte by applying a paste of dry electrolyte powder entrained in a dissipatable carrier to the reactant flow channels in the current collector plate. The stack plates are preformed and solidified to final operating condition so that they are self sustaining and can be disposed one atop the other to form the power plant stack. Packing the reactant flow channels with the electrolyte paste allows the use of thinner electrode plates, particularly on the anode side of the cells. The use of the packed electrolyte paste provides sufficient electrolyte to fill the matrix and to entrain excess electrolyte in the electrode plates, which also serve as excess electrolyte reservoirs. When the stack is heated up to operating temperatures, the electrolyte in the paste melts, the carrier vaporizes, or chemically decomposes, and the melted electrolyte is absorbed into the matrix and electrode plates.

  7. Single ion dynamics in molten sodium bromide

    SciTech Connect (OSTI)

    Alcaraz, O.; Trullas, J.; Demmel, F.

    2014-12-28

    We present a study on the single ion dynamics in the molten alkali halide NaBr. Quasielastic neutron scattering was employed to extract the self-diffusion coefficient of the sodium ions at three temperatures. Molecular dynamics simulations using rigid and polarizable ion models have been performed in parallel to extract the sodium and bromide single dynamics and ionic conductivities. Two methods have been employed to derive the ion diffusion, calculating the mean squared displacements and the velocity autocorrelation functions, as well as analysing the increase of the line widths of the self-dynamic structure factors. The sodium diffusion coefficients show a remarkable good agreement between experiment and simulation utilising the polarisable potential.

  8. Slime-busting Salt

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    past issues All Issues submit Slime-busting Salt A potential new treatment gets bacteria deep in their hiding places May 1, 2015 Slime-busting Salt Biofilms are made of...

  9. Ancient Salt Beds

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Ancient Salt Beds Dr. Jack Griffith The key to the search for life on other planets may go through WIPP's ancient salt beds. In 2008, a team of scientists led by Jack Griffith, from the University of North Carolina, Chapel Hill, retrieved salt samples from the WIPP underground and studied them with a transmission electron microscopy lab at the Lineberger Comprehensive Cancer Center of the University of North Carolina School of Medicine. In examining fluid inclusions in the salt and solid halite

  10. Method for the regeneration of spent molten zinc chloride

    DOE Patents [OSTI]

    Zielke, Clyde W.; Rosenhoover, William A.

    1981-01-01

    In a process for regenerating spent molten zinc chloride which has been used in the hydrocracking of coal or ash-containing polynuclear aromatic hydrocarbonaceous materials derived therefrom and which contains zinc chloride, zinc oxide, zinc oxide complexes and ash-containing carbonaceous residue, by incinerating the spent molten zinc chloride to vaporize the zinc chloride for subsequent condensation to produce a purified molten zinc chloride: an improvement comprising the use of clay in the incineration zone to suppress the vaporization of metals other than zinc. Optionally water is used in conjunction with the clay to further suppress the vaporization of metals other than zinc.

  11. Recovery of the actinides by electrochemical methods in molten chlorides using solid aluminium cathode

    SciTech Connect (OSTI)

    Malmbeck, R.; Mendes, E.; Serp, J.; Soucek, P.; Glatz, J.P.; Cassayre, L.

    2007-07-01

    An electrorefining process in molten chloride salts is being developed at ITU to reprocess the spent nuclear fuel. According to the thermochemical properties of the system, aluminium is the most promising electrode material for the separation of actinides (An) from lanthanides (Ln). The actinides are selectively reduced from the fission products and stabilized by the formation of solid and compact actinide-aluminium alloys with the reactive cathode material. In this work, the maximum loading of aluminium with actinides was investigated by potentiostatic and galvano-static electrorefining of U-Pu- Zr alloys. A very high aluminium capacity was achieved, as the average loading was 1.6 g of U and Pu into 1 g of aluminium and the maximum achieved loading was 2.3 g. For recovery of the actinides from aluminium, a process based on chlorination and a subsequent sublimation of AlCl{sub 3} is proposed. (authors)

  12. Degassing of molten alloys with the assistance of ultrasonic vibration

    DOE Patents [OSTI]

    Han, Qingyou; Xu, Hanbing; Meek, Thomas T.

    2010-03-23

    An apparatus and method are disclosed in which ultrasonic vibration is used to assist the degassing of molten metals or metal alloys thereby reducing gas content in the molten metals or alloys. High-intensity ultrasonic vibration is applied to a radiator that creates cavitation bubbles, induces acoustic streaming in the melt, and breaks up purge gas (e.g., argon or nitrogen) which is intentionally introduced in a small amount into the melt in order to collect the cavitation bubbles and to make the cavitation bubbles survive in the melt. The molten metal or alloy in one version of the invention is an aluminum alloy. The ultrasonic vibrations create cavitation bubbles and break up the large purge gas bubbles into small bubbles and disperse the bubbles in the molten metal or alloy more uniformly, resulting in a fast and clean degassing.

  13. Method for removing semiconductor layers from salt substrates

    DOE Patents [OSTI]

    Shuskus, Alexander J.; Cowher, Melvyn E.

    1985-08-27

    A method is described for removing a CVD semiconductor layer from an alkali halide salt substrate following the deposition of the semiconductor layer. The semiconductor-substrate combination is supported on a material such as tungsten which is readily wet by the molten alkali halide. The temperature of the semiconductor-substrate combination is raised to a temperature greater than the melting temperature of the substrate but less than the temperature of the semiconductor and the substrate is melted and removed from the semiconductor by capillary action of the wettable support.

  14. WATER BOILER REACTOR

    DOE Patents [OSTI]

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  15. Pressurized tundish for controlling a continuous flow of molten metal

    DOE Patents [OSTI]

    Lewis, T.W.; Hamill, P.E. Jr.; Ozgu, M.R.; Padfield, R.C.; Rego, D.N.; Brita, G.P.

    1990-07-24

    A pressurized tundish for controlling a continuous flow of molten metal is characterized by having a pair of principal compartments, one being essentially unpressurized and receiving molten metal introduced thereto, and the other being adapted for maintaining a controlled gaseous pressure over the surface of the fluid metal therein, whereby, by controlling the pressure within the pressurized chamber, metal exiting from the tundish is made to flow continually and at a controlled rate. 1 fig.

  16. Pressurized tundish for controlling a continuous flow of molten metal

    DOE Patents [OSTI]

    Lewis, Thomas W.; Hamill, Jr., Paul E.; Ozgu, Mustafa R.; Padfield, Ralph C.; Rego, Donovan N.; Brita, Guido P.

    1990-01-01

    A pressurized tundish for controlling a continous flow of molten metal characterized by having a pair of principal compartments, one being essentially unpressurized and receiving molten metal introduced thereto, and the other being adapted for maintaining a controlled gaseous pressure over the surface of the fluid metal therein, whereby, by controlling the pressure within the pressurized chamber, metal exiting from the tundish is made to flow continually and at a controlled rate.

  17. Solid tags for identifying failed reactor components

    DOE Patents [OSTI]

    Bunch, Wilbur L.; Schenter, Robert E.

    1987-01-01

    A solid tag material which generates stable detectable, identifiable, and measurable isotopic gases on exposure to a neutron flux to be placed in a nuclear reactor component, particularly a fuel element, in order to identify the reactor component in event of its failure. Several tag materials consisting of salts which generate a multiplicity of gaseous isotopes in predetermined ratios are used to identify different reactor components.

  18. Analysis of molten fuel-coolant interaction during a reactivity-initiated accident experiment. [BWR; PWR

    SciTech Connect (OSTI)

    El-Genk, M.S.; Hobbins, R.R.

    1981-01-01

    The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO/sub 2/ fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm/sup 3//s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an energetic MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, or to UO/sub 2/ fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only 0.3%.

  19. Search for: thorium | DOE PAGES

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Liquid fuel molten salt reactors for thorium utilization ... (thermal, intermediate, fast, and mixed-spectrum zoned ... Soil samples are fused with sodium hydroxide at 600 C in ...

  20. The Title Goes Here

    Office of Scientific and Technical Information (OSTI)

    water, high temperature gas cooled, fast spectrum sodium, molten salt, etc.), advanced ... in the United States is in currently operating light water reactors (LWRs). ...

  1. OSTI, US Dept of Energy Office of Scientific and Technical Information...

    Office of Scientific and Technical Information (OSTI)

    Weinberg and his team's work led to the construction and operation of the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory in Oak Ridge, Tennessee, where ...

  2. OSTIblog Articles in the thorium Topic | OSTI, US Dept of Energy...

    Office of Scientific and Technical Information (OSTI)

    Weinberg and his team's work led to the construction and operation of the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory in Oak Ridge, Tennessee, where ...

  3. PatentAwarenessProgram_FY2014_09.2015_download.indd

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    ... The double well mass filter has applications in nuclear waste remediation and nuclear fuel reprocessing. 4 M-871 Hybrid Molten Salt Reactor (HMSR): Method and System to Fully ...

  4. Search for: All records | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    ... USDOE Office of Civilian Radioactive Waste Management (RW) (United States) USDOE Office of ... to conventional critical thorium reactors, for both solid and molten salt fuels. ...

  5. Fuel salt and container material studies for MOSART transforming system

    SciTech Connect (OSTI)

    Ignatiev, V.; Feynberg, O.; Merzlyakov, A.; Surenkov, A.; Zagnitko, A.; Afonichkin, V.; Bovet, A.; Khokhlov, V.; Subbotin, V.; Gordeev, M.; Panov, A.; Toropov, A.

    2013-07-01

    A study is under progress to examine the feasibility of single stream Molten Salt Actinide Recycling and Transmuting system without and with Th support (MOSART) fuelled with different compositions of actinide tri-fluorides (AnF{sub 3}) from used LWR fuel. New fast-spectrum design options with homogeneous core and fuel salts with high enough solubility for AnF{sub 3} are being examined because of new goals. The flexibility of single fluid MOSART concept with Th support is underlined, particularly, possibility of its operation in self-sustainable mode (Conversion Ratio: CR=1) using different loadings and make up. The paper summarizes the most current status of fuel salt and container material data for the MOSART concept received within ISTC-3749 and ROSATOM-MARS projects. Key physical and chemical properties of various fluoride fuel salts are reported. The issues like salt purification, the electroreduction of U(IV) to U(III) in LiF-ThF{sub 4} and the electroreduction of Yb(III) to Yb(II) in LiF-NaF are detailed.

  6. The Effect of Accident Conditions on the Molten Core Material Relocation into the Lower Head of a PWR Vessel with Application to TMI-2

    SciTech Connect (OSTI)

    An Xuegao; Dhir, Vijay K.; Okrent, David

    2000-11-15

    The damage progression of the reactor core and the slumping mechanism of molten material to the lower head of the reactor vessel were examined through simulation of severe accident scenarios that lead to large-scale core damage. The calculations were carried out on a Three Mile Island Unit 2 configuration using the computer code SCDAP/RELAP5/MOD3.2.Different accident scenarios were simulated. The high-pressure injection and makeup flow rates were changed. The extreme case with no water being added during the accident was examined. Reflood by restart of coolant pump 2B was also studied. Finally, the size of the power-operated relief valve opening was also changed. The effects of these accident scenarios on the accident progression and the core damage process were studied.It is concluded that, according to code MOD3.2, the molten material slumped to the lower head of the reactor vessel when the junction of the top and side crusts failed after the molten pool had reached the periphery of the core. When the effective stress caused by pressure imbalance inside and outside of the crust became larger than the ultimate strength of the crust, the crust failed mechanically.

  7. Bulk Vitrification Performance Enhancement: Refractory Lining Protection Against Molten Salt Penetration

    SciTech Connect (OSTI)

    Hrma, Pavel R.; Bagaasen, Larry M.; Schweiger, Michael J.; Evans, Michael B.; Smith, Benjamin T.; Arrigoni, Benjamin M.; Kim, Dong-Sang; Rodriguez, Carmen P.; Yokuda, Satoru T.; Matyas, Josef; Buchmiller, William C.; Gallegos, Autumn B.; Fluegel, Alexander

    2007-08-06

    Bulk vitrification (BV) is a process that heats a feed material that consists of glass-forming solids and dried low-activity waste (LAW) in a disposable refractory-lined metal box using electrical power supplied through carbon electrodes. The feed is heated to the point that the LAW decomposes and combines with the solids to generate a vitreous waste form. This study supports the BV design and operations by exploring various methods aimed at reducing the quantities of soluble Tc in the castable refractory block portion of the refractory lining, which limits the effectiveness of the final waste form.

  8. Exchange Reactions Between a Molten Salt and a Solution of Tri...

    Office of Scientific and Technical Information (OSTI)

    LIQUIDS; LITHIUM NITRATES; MAGNESIUM; MAGNESIUM COMPLEXES; MAGNESIUM COMPOUNDS; MERCURY COMPOUNDS; MINERAL ACIDS; MIXING; NICKEL; NICKEL COMPLEXES; NICKEL COMPOUNDS; ...

  9. Corrosion in Very High-Temperature Molten Salt for Next Generation...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Systems for High Performance High Temperature Heat Transfer Fluids Degradation Mechanisms and Development of Protective Coatings for TES and HTF Containment Materials - F13 Q1

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

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Domestic Health Studies and Activities Domestic Health Studies and Activities Purpose The Atomic Energy Act of 1957 - Section 8(a) requires research and development activities relating to the protection of health during research and production activities. The requirement is fulfilled by conducting and supporting health studies and other research activities to determine if DOE workers and people living in communities near DOE sites are adversely affected by exposures to hazardous materials from

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

    Office of Energy Efficiency and Renewable Energy (EERE)

    This presentation was delivered at the SunShot Concentrating Solar Power (CSP) Program Review 2013, held April 23–25, 2013 near Phoenix, Arizona.

  12. Corrosion Studies in High-Temperature Molten Salt Systems for CSP Applications- FY13 Q1

    Broader source: Energy.gov [DOE]

    This document summarizes the progress of this Savannah River National Laboratory project, funded by SunShot, for the first quarter of fiscal year 2013.

  13. Corrosion in Very High-Temperature Molten Salt for Next Generation CSP Systems

    Broader source: Energy.gov [DOE]

    This presentation was delivered at the SunShot Concentrating Solar Power (CSP) Program Review 2013, held April 23–25, 2013 near Phoenix, Arizona.

  14. Relative electrochemical stability of lithium and aluminum salts and their solvents

    SciTech Connect (OSTI)

    Ciemiecki, K.T.; Auborn, J.J.

    1983-10-01

    The stability series were determined by cyclic voltammetric measurements at platinum electrodes in dry acetonitrile. These results are applicable to the electrochemical synthesis of new organic electrodes and to the development of an ambient temperature rechargeable battery. Cyclic voltammetry was also used to establish electrochemical windows for a class of room temperature chloroaluminate molten salts from which aluminum can be reversibly electrodeposited. Enhancement of oxidative stability by increased Lewis acidity was observed in both the melts and the more conventional electrolytes.

  15. Dosimetry using silver salts

    DOE Patents [OSTI]

    Warner, Benjamin P.

    2003-06-24

    The present invention provides a method for detecting ionizing radiation. Exposure of silver salt AgX to ionizing radiation results in the partial reduction of the salt to a mixture of silver salt and silver metal. The mixture is further reduced by a reducing agent, which causes the production of acid (HX) and the oxidized form of the reducing agent (R). Detection of HX indicates that the silver salt has been exposed to ionizing radiation. The oxidized form of the reducing agent (R) may also be detected. The invention also includes dosimeters employing the above method for detecting ionizing radiation.

  16. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  17. Process Heat Exchanger Options for the Advanced High Temperature Reactor

    SciTech Connect (OSTI)

    Piyush Sabharwall; Eung Soo Kim; Michael McKellar; Nolan Anderson

    2011-06-01

    The work reported herein is a significant intermediate step in reaching the final goal of commercial-scale deployment and usage of molten salt as the heat transport medium for process heat applications. The primary purpose of this study is to aid in the development and selection of the required heat exchanger for power production and process heat application, which would support large-scale deployment.

  18. Hydroxycarboxylic acids and salts

    SciTech Connect (OSTI)

    Kiely, Donald E; Hash, Kirk R; Kramer-Presta, Kylie; Smith, Tyler N

    2015-02-24

    Compositions which inhibit corrosion and alter the physical properties of concrete (admixtures) are prepared from salt mixtures of hydroxycarboxylic acids, carboxylic acids, and nitric acid. The salt mixtures are prepared by neutralizing acid product mixtures from the oxidation of polyols using nitric acid and oxygen as the oxidizing agents. Nitric acid is removed from the hydroxycarboxylic acids by evaporation and diffusion dialysis.

  19. Water purification using organic salts

    DOE Patents [OSTI]

    Currier, Robert P.

    2004-11-23

    Water purification using organic salts. Feed water is mixed with at least one organic salt at a temperature sufficiently low to form organic salt hydrate crystals and brine. The crystals are separated from the brine, rinsed, and melted to form an aqueous solution of organic salt. Some of the water is removed from the aqueous organic salt solution. The purified water is collected, and the remaining more concentrated aqueous organic salt solution is reused.

  20. Metals processing control by counting molten metal droplets

    DOE Patents [OSTI]

    Schlienger, Eric; Robertson, Joanna M.; Melgaard, David; Shelmidine, Gregory J.; Van Den Avyle, James A.

    2000-01-01

    Apparatus and method for controlling metals processing (e.g., ESR) by melting a metal ingot and counting molten metal droplets during melting. An approximate amount of metal in each droplet is determined, and a melt rate is computed therefrom. Impedance of the melting circuit is monitored, such as by calculating by root mean square a voltage and current of the circuit and dividing the calculated current into the calculated voltage. Analysis of the impedance signal is performed to look for a trace characteristic of formation of a molten metal droplet, such as by examining skew rate, curvature, or a higher moment.

  1. Porous electrolyte retainer for molten carbonate fuel cell

    DOE Patents [OSTI]

    Singh, Raj N.; Dusek, Joseph T.

    1983-06-21

    A porous tile for retaining molten electrolyte within a fuel cell is prepared by sintering particles of lithium aluminate into a stable structure. The tile is assembled between two porous metal plates which serve as electrodes with fuels gases such as H.sub.2 and CO opposite to oxidant gases such as O.sub.2 and CO.sub.2. The tile is prepared with a porosity of 55-65% and a pore size distribution selected to permit release of sufficient molten electrolyte to wet but not to flood the adjacent electrodes.

  2. Porous electrolyte retainer for molten carbonate fuel cell. [lithium aluminate

    DOE Patents [OSTI]

    Singh, R.N.; Dusek, J.T.

    1979-12-27

    A porous tile for retaining molten electrolyte within a fuel cell is prepared by sintering particles of lithium aluminate into a stable structure. The tile is assembled between two porous metal plates which serve as electrodes with fuels gases such as H/sub 2/ and CO opposite to oxidant gases such as O/sub 2/ and CO/sub 2/. The tile is prepared with a porosity of 55 to 65% and a pore size distribution selected to permit release of sufficient molten electrolyte to wet but not to flood the adjacent electrodes.

  3. Thermal conditions and functional requirements for molten fuel containment

    SciTech Connect (OSTI)

    Kang, C.S.; Torri, A.

    1980-05-01

    This paper discusses the configuration and functional requirements for the molten fuel containment system (MFCS) in the GCFR demonstration plant design. Meltdown conditions following a loss of shutdown cooling (LOSC) accident were studied to define the core debris volume for a realistic meltdown case. Materials and thicknesses of the molten fuel container were defined. Stainless steel was chosen as the sacrificial material and magnesium oxide was chosen as the crucible material. Thermal conditions for an expected quasi-steady state were analyzed. Highlights of the functional requirements which directly affect the MFCS design are discussed.

  4. Fabrication of catalytic electrodes for molten carbonate fuel cells

    DOE Patents [OSTI]

    Smith, James L.

    1988-01-01

    A porous layer of catalyst material suitable for use as an electrode in a molten carbonate fuel cell includes elongated pores substantially extending across the layer thickness. The catalyst layer is prepared by depositing particulate catalyst material into polymeric flocking on a substrate surface by a procedure such as tape casting. The loaded substrate is heated in a series of steps with rising temperatures to set the tape, thermally decompose the substrate with flocking and sinter bond the catalyst particles into a porous catalytic layer with elongated pores across its thickness. Employed as an electrode, the elongated pores provide distribution of reactant gas into contact with catalyst particles wetted by molten electrolyte.

  5. Molten metal containment vessel with rare earth oxysulfide protective coating thereon and method of making same

    DOE Patents [OSTI]

    Krikorian, Oscar H.; Curtis, Paul G.

    1992-01-01

    An improved molten metal containment vessel is disclosed in which wetting of the vessel's inner wall surfaces by molten metal is inhibited by coating at least the inner surfaces of the containment vessel with one or more rare earth oxysulfide or rare earth sulfide compounds to inhibit wetting and or adherence by the molten metal to the surfaces of the containment vessel.

  6. Amine salts of nitroazoles

    DOE Patents [OSTI]

    Kienyin Lee; Stinecipher, M.M.

    1993-10-26

    Compositions of matter, a method of providing chemical energy by burning said compositions, and methods of making said compositions are described. These compositions are amine salts of nitroazoles. 1 figure.

  7. Electrodialysis technology for salt recovery from aluminum salt cake

    SciTech Connect (OSTI)

    Hryn, J. N.; Krumdick, G.; Graziano, D.; Sreenivasarao, K.

    2000-02-02

    Electrodialysis technology for recovering salt from aluminum salt cake is being developed at Argonne National Laboratory. Salt cake, a slag-like aluminum-industry waste stream, contains aluminum metal, salt (NaCl and KCl), and nonmetallics (primarily aluminum oxide). Salt cake can be recycled by digesting with water and filtering to recover the metal and oxide values. A major obstacle to widespread salt cake recycling is the cost of recovering salt from the process brine. Electrodialysis technology developed at Argonne appears to be a cost-effective approach to handling the salt brines, compared to evaporation or disposal. In Argonne's technology, the salt brine is concentrated until salt crystals are precipitated in the electrodialysis stack; the crystals are recovered downstream. The technology is being evaluated on the pilot scale using Eurodia's EUR 40-76-5 stack.

  8. BOILING REACTORS

    DOE Patents [OSTI]

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  9. Molten metal feed system controlled with a traveling magnetic field

    DOE Patents [OSTI]

    Praeg, Walter F.

    1991-01-01

    A continuous metal casting system in which the feed of molten metal is controlled by means of a linear induction motor capable of producing a magnetic traveling wave in a duct that connects a reservoir of molten metal to a caster. The linear induction motor produces a traveling magnetic wave in the duct in opposition to the pressure exerted by the head of molten metal in the reservoir so that p.sub.c =p.sub.g -p.sub.m where p.sub.c is the desired pressure in the caster, p.sub.g is the gravitational pressure in the duct exerted by the force of the head of molten metal in the reservoir, and p.sub.m is the electromagnetic pressure exerted by the force of the magnetic field traveling wave produced by the linear induction motor. The invention also includes feedback loops to the linear induction motor to control the casting pressure in response to measured characteristics of the metal being cast.

  10. Sulfur tolerant molten carbonate fuel cell anode and process

    DOE Patents [OSTI]

    Remick, Robert J.

    1990-01-01

    Molten carbonate fuel cell anodes incorporating a sulfur tolerant carbon monoxide to hydrogen water-gas-shift catalyst provide in situ conversion of carbon monoxide to hydrogen for improved fuel cell operation using fuel gas mixtures of over about 10 volume percent carbon monoxide and up to about 10 ppm hydrogen sulfide.

  11. Salt Repository Research,

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Salt Lake City Gamma Shield Thunder Exercise Concludes National Nuclear Security Administration (NNSA) and the FBI announced today the completion of the Gamma Shield Thunder counterterrorism table-top exercise at LDS Hospital. The exercise is part of NNSA's Silent Thunder table-top series, which is aimed at giving federal, state and local

    6 th US/German Workshop on Salt Repository Research, Design, and Operation Hotel Pullmann Dresden Newa Dresden September 7 - 9, 2015 September 7- Monday

  12. Salt Selected (FINAL)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    WHY SALT WAS SELECTED AS A DISPOSAL MEDIUM Waste Isolation Pilot Plant U.S. Department Of Energy Government officials and scientists chose the Waste Isolation Pilot Plant (WIPP) site through a selection process that started in the 1950s. At that time, the National Academy of Sciences conducted a nationwide search for geological formations stable enough to contain radioactive wastes for thousands of years. In 1955, after extensive study, salt deposits were recommended as a promising medium for

  13. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

  14. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    SciTech Connect (OSTI)

    Corwin, William R; Burchell, Timothy D; Katoh, Yutai; McGreevy, Timothy E; Nanstad, Randy K; Ren, Weiju; Snead, Lance Lewis; Wilson, Dane F

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural

  15. CONVECTION REACTOR

    DOE Patents [OSTI]

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  16. Casting Apparatus Including A Gas Driven Molten Metal Injector And Method

    DOE Patents [OSTI]

    Meyer, Thomas N.

    2004-06-01

    The casting apparatus (50) includes a holding vessel (10) for containing a supply of molten metal (12) and a casting mold (52) located above the holding vessel (10) and having a casting cavity (54). A molten metal injector (14) extends into the holding vessel (10) and is at least partially immersed in the molten metal (12) in the holding vessel (10). The molten metal injector (14) is in fluid communication with the casting cavity (54). The molten metal injector (14) has an injector body (16) defining an inlet opening (24) for receiving molten metal into the injector body (16). A gas pressurization source (38) is in fluid communication with the injector body (16) for cyclically pressurizing the injector body (16) and inducing molten metal to flow from the injector body (16) to the casting cavity (54). An inlet valve (42) is located in the inlet opening (24) in the injector body (16) for filling molten metal into the injector body (16). The inlet valve (42) is configured to prevent outflow of molten metal from the injector body (16) during pressurization and permit inflow of molten metal into the injector body (16) after pressurization. The inlet valve (42) has an inlet valve actuator (44) located above the surface of the supply of molten metal (12) and is operatively connected to the inlet valve (42) for operating the inlet valve (42) between open and closed positions.

  17. Status of the direct absorption receiver panel research experiment: Salt flow and solar test requirements and plans

    SciTech Connect (OSTI)

    Tyner, C.E.

    1989-03-01

    The Panel Research Experiment (PRE) is the first large-scale solar test of the molten nitrate salt direct absorption receiver (DAR) concept. The purpose of the PRE is to demonstrate the engineering feasibility and practicality of the DAR. We will conduct the test at the Central Receiver Test Facility in Albuquerque in two phases: salt flow testing and solar testing. This is a working document to define PRE test objectives and requirements, document the test hardware design, and define test plans. 13 refs., 12 figs., 1 tab.

  18. Gas releases from salt

    SciTech Connect (OSTI)

    Ehgartner, B.; Neal, J.; Hinkebein, T.

    1998-06-01

    The occurrence of gas in salt mines and caverns has presented some serious problems to facility operators. Salt mines have long experienced sudden, usually unexpected expulsions of gas and salt from a production face, commonly known as outbursts. Outbursts can release over one million cubic feet of methane and fractured salt, and are responsible for the lives of numerous miners and explosions. Equipment, production time, and even entire mines have been lost due to outbursts. An outburst creates a cornucopian shaped hole that can reach heights of several hundred feet. The potential occurrence of outbursts must be factored into mine design and mining methods. In caverns, the occurrence of outbursts and steady infiltration of gas into stored product can effect the quality of the product, particularly over the long-term, and in some cases renders the product unusable as is or difficult to transport. Gas has also been known to collect in the roof traps of caverns resulting in safety and operational concerns. The intent of this paper is to summarize the existing knowledge on gas releases from salt. The compiled information can provide a better understanding of the phenomena and gain insight into the causative mechanisms that, once established, can help mitigate the variety of problems associated with gas releases from salt. Outbursts, as documented in mines, are discussed first. This is followed by a discussion of the relatively slow gas infiltration into stored crude oil, as observed and modeled in the caverns of the US Strategic Petroleum Reserve. A model that predicts outburst pressure kicks in caverns is also discussed.

  19. Deep Eutectic Salt Formulations Suitable as Advanced Heat Transfer Fluids

    SciTech Connect (OSTI)

    Raade, Justin; Roark, Thomas; Vaughn, John; Bradshaw, Robert

    2013-07-22

    Concentrating solar power (CSP) facilities are comprised of many miles of fluid-filled pipes arranged in large grids with reflective mirrors used to capture radiation from the sun. Solar radiation heats the fluid which is used to produce steam necessary to power large electricity generation turbines. Currently, organic, oil-based fluid in the pipes has a maximum temperature threshold of 400 °C, allowing for the production of electricity at approximately 15 cents per kilowatt hour. The DOE hopes to foster the development of an advanced heat transfer fluid that can operate within higher temperature ranges. The new heat transfer fluid, when used with other advanced technologies, could significantly decrease solar electricity cost. Lower costs would make solar thermal electricity competitive with gas and coal and would offer a clean, renewable source of energy. Molten salts exhibit many desirable heat transfer qualities within the range of the project objectives. Halotechnics developed advanced heat transfer fluids (HTFs) for application in solar thermal power generation. This project focused on complex mixtures of inorganic salts that exhibited a high thermal stability, a low melting point, and other favorable characteristics. A high-throughput combinatorial research and development program was conducted in order to achieve the project objective. Over 19,000 candidate formulations were screened. The workflow developed to screen various chemical systems to discover salt formulations led to mixtures suitable for use as HTFs in both parabolic trough and heliostat CSP plants. Furthermore, salt mixtures which will not interfere with fertilizer based nitrates were discovered. In addition for use in CSP, the discovered salt mixtures can be applied to electricity storage, heat treatment of alloys and other industrial processes.

  20. Salt Repository Research,

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    th US/German Workshop on Salt Repository Research, Design, and Operation Hotel Pullman Dresden Newa September 7 - 9, 2015 September 7- Monday 08:00-08:30 Registration 08:30-08:50 Welcome by the organizers T. Lautsch, DBE F. Hansen, SNL W. Steininger, PTKA 08:50-09:15 Welcome by BMWi U. Borak, BMWi 09:15-09:30 Welcome by USDOE N. Buschman, US DOE 09:30-10:00 NEA Salt Club J. Mönig, GRS SAFETY CASE ISSUES 10:00-10:30 WIPP recovery F. Hansen, SNL 10:30-11:00 Coffee break and photo event

  1. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fraas, A.P.; Mills, C.B.

    1961-11-21

    A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

  2. Electromagnetic valve for controlling the flow of molten, magnetic material

    DOE Patents [OSTI]

    Richter, Tomas

    1998-01-01

    An electromagnetic valve for controlling the flow of molten, magnetic material is provided, which comprises an induction coil for generating a magnetic field in response to an applied alternating electrical current, a housing, and a refractory composite nozzle. The nozzle is comprised of an inner sleeve composed of an erosion resistant refractory material (e.g., a zirconia ceramic) through which molten, magnetic metal flows, a refractory outer shell, and an intermediate compressible refractory material, e.g., unset, high alumina, thermosetting mortar. The compressible refractory material is sandwiched between the inner sleeve and outer shell, and absorbs differential expansion stresses that develop within the nozzle due to extreme thermal gradients. The sandwiched layer of compressible refractory material prevents destructive cracks from developing in the refractory outer shell.

  3. Electromagnetic valve for controlling the flow of molten, magnetic material

    DOE Patents [OSTI]

    Richter, T.

    1998-06-16

    An electromagnetic valve for controlling the flow of molten, magnetic material is provided, which comprises an induction coil for generating a magnetic field in response to an applied alternating electrical current, a housing, and a refractory composite nozzle. The nozzle is comprised of an inner sleeve composed of an erosion resistant refractory material (e.g., a zirconia ceramic) through which molten, magnetic metal flows, a refractory outer shell, and an intermediate compressible refractory material, e.g., unset, high alumina, thermosetting mortar. The compressible refractory material is sandwiched between the inner sleeve and outer shell, and absorbs differential expansion stresses that develop within the nozzle due to extreme thermal gradients. The sandwiched layer of compressible refractory material prevents destructive cracks from developing in the refractory outer shell. 5 figs.

  4. Laser Acoustic Molten Metal Depth Sensing in Titanium

    SciTech Connect (OSTI)

    J. B. Walter; K. L. Telschow; R. E. Haun

    1999-09-22

    A noncontacting ultrasonic method has been investigated for probing the solidification front in molten titanium for the purposes of profiling the channel depth in a plasma hearth re-melter. The method, known as Laser Ultrasonics, utilized a pulsed laser for generation of ultrasonic waves at the surface of a molten metal pool. The ultrasonic waves propagated into the liquid titanium reflected from the solidification front and the boundaries of the solid plug. A Fabry-Perot interferometer, driven by a second laser, demodulated the small displacements caused by the ultrasonic wave motion at the liquid surface. The method and results of measurements taken within a small research plasma melting furnace will be described. Successful results were obtained even directly beneath the plasma arc using this all-optical approach.

  5. Laser Acoustic Molten Metal Depth Sensing in Titanium

    SciTech Connect (OSTI)

    Walter, John Bradley; Telschow, Kenneth Louis; Haun, R.E.

    1999-08-01

    A noncontacting ultrasonic method has been investigated for probing the solidification front in molten titanium for the purposes of profiling the channel depth in plasma hearth re-melter. The method, known as Laser Ultrasonics, utilized a pulsed laser for generation of ultrasonic waves at the surface of a molten metal pool. The ultrasonic waves propagated into the liquid titanium reflected from the solidification front and the boundaries of the solid plug. A Fabry-Perot interferometer, driven by a second laser, demodulated the small displacements caused by the ultrasonic wave motion at the liquid surface. The method and results of measurements taken within a small research plasma melting furnace will be described. Successful results were obtained even directly beneath the plasma arc using this all optical approach.

  6. REACTOR COOLING

    DOE Patents [OSTI]

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  7. Non-segregating electrolytes for molten carbonate fuel cells

    SciTech Connect (OSTI)

    Krumpelt, M.; Kaun, T.; Lanagan, M.

    1996-08-01

    Current MCFCs use a Li/K carbonate mixture; the segregation increases the K concentration near the cathode, leading to increase cathode solubility and performance decline. ANL is developing molten carbonates that have minimal segregation; the approach is using Li-Na carbonates. In screening tests, fully developed potential distributions were obtained for 4 Li/Na compositions, and performance data were used to compare these.

  8. Molten carbonate fuel cell reduction of nickel deposits

    DOE Patents [OSTI]

    Smith, James L. (Lemont, IL); Zwick, Stanley A. (Darien, IL)

    1987-01-01

    A molten carbonate fuel cell with anode and cathode electrodes and an eleolyte formed with two tile sections, one of the tile sections being adjacent the anode and limiting leakage of fuel gas into the electrolyte with the second tile section being adjacent the cathode and having pores sized to permit the presence of oxygen gas in the electrolyte thereby limiting the formation of metal deposits caused by the reduction of metal compositions migrating into the electrolyte from the cathode.

  9. METHOD OF PROTECTING TANTALUM CRUCIBLES AGAINST REACTION WITH MOLTEN URANIUM

    DOE Patents [OSTI]

    Feder, H.M.; Chellew, N.R.

    1960-08-16

    Tantalum crucibles against reaction with molten uranium by contacting the surfaces to be protected with metallic boron (as powder, vapor, or suspension in a liquid-volatilenonreacting medium, such as acetone and petroleum oil) at about 1800 deg C in vacuum, discontinuing contact with the boron, and heating the crucibles to a temperature of between 1800 aad 2000 deg C, whereby the tantalum boride formed in the first heating step is converted to tantalum monoboride.

  10. High current density cathode for electrorefining in molten electrolyte

    DOE Patents [OSTI]

    Li, Shelly X.

    2010-06-29

    A high current density cathode for electrorefining in a molten electrolyte for the continuous production and collection of loose dendritic or powdery deposits. The high current density cathode eliminates the requirement for mechanical scraping and electrochemical stripping of the deposits from the cathode in an anode/cathode module. The high current density cathode comprises a perforated electrical insulated material coating such that the current density is up to 3 A/cm.sup.2.

  11. Molten-Caustic-Leaching (Gravimelt) System Integration Project, Phase 2

    SciTech Connect (OSTI)

    Not Available

    1993-03-01

    This is a report of the maintenance, refurbishment, modifications, and off-line circuit component testing of the integrated test circuit of the Molten-Caustic-Leaching (MCL or Gravimelt) process for the desulfurization and demineralization of coal. The project is sponsored by the Pittsburgh Energy Technology Center of the US Department of Energy under Contract No. DE-AC22-86-PC91257.

  12. Electrolytic orthoborate salts for lithium batteries (Patent...

    Office of Scientific and Technical Information (OSTI)

    Electrolytic orthoborate salts for lithium batteries Title: Electrolytic orthoborate salts for lithium batteries Orthoborate salts suitable for use as electrolytes in lithium ...

  13. Salt Wells Geothermal Area | Open Energy Information

    Open Energy Info (EERE)

    Salt Wells Geothermal Area Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Salt Wells Geothermal Area Contents 1 Area Overview 2 History and Infrastructure 2.1 Salt...

  14. Electromagnetic confinement for vertical casting or containing molten metal

    DOE Patents [OSTI]

    Lari, Robert J.; Praeg, Walter F.; Turner, Larry R.

    1991-01-01

    An apparatus and method adapted to confine a molten metal to a region by means of an alternating electromagnetic field. As adapted for use in the present invention, the alternating electromagnetic field given by B.sub.y =(2.mu..sub.o .rho.gy).sup.1/2 (where B.sub.y is the vertical component of the magnetic field generated by the magnet at the boundary of the region; y is the distance measured downward form the top of the region, .rho. is the metal density, g is the acceleration of gravity and .mu..sub.o is the permeability of free space) induces eddy currents in the molten metal which interact with the magnetic field to retain the molten metal with a vertical boudnary. As applied to an apparatus for the continuous casting of metal sheets or rods, metal in liquid form can be continuously introduced into the region defined by the magnetic field, solidified and conveyed away from the magnetic field in solid form in a continuous process.

  15. Salt repository design approach

    SciTech Connect (OSTI)

    Matthews, S.C.

    1983-01-01

    This paper presents a summary discussion of the approaches that have been and will be taken in design of repository facilities for use with disposal of radioactive wastes in salt formations. Since specific sites have yet to be identified, the discussion is at a general level, supplemented with illustrative examples where appropriate. 5 references, 1 figure.

  16. MOLTEN CARBONATE FUEL CELL PRODUCT DESIGN IMPROVEMENT

    SciTech Connect (OSTI)

    H.C. Maru; M. Farooque

    2002-02-01

    The carbonate fuel cell promises highly efficient, cost-effective and environmentally superior power generation from pipeline natural gas, coal gas, biogas, and other gaseous and liquid fuels. FuelCell Energy, Inc. has been engaged in the development of this unique technology, focusing on the development of the Direct Fuel Cell (DFC{reg_sign}). The DFC{reg_sign} design incorporates the unique internal reforming feature which allows utilization of a hydrocarbon fuel directly in the fuel cell without requiring any external reforming reactor and associated heat exchange equipment. This approach upgrades waste heat to chemical energy and thereby contributes to a higher overall conversion efficiency of fuel energy to electricity with low levels of environmental emissions. Among the internal reforming options, FuelCell Energy has selected the Indirect Internal Reforming (IIR)--Direct Internal Reforming (DIR) combination as its baseline design. The IIR-DIR combination allows reforming control (and thus cooling) over the entire cell area. This results in uniform cell temperature. In the IIR-DIR stack, a reforming unit (RU) is placed in between a group of fuel cells. The hydrocarbon fuel is first fed into the RU where it is reformed partially to hydrogen and carbon monoxide fuel using heat produced by the fuel cell electrochemical reactions. The reformed gases are then fed to the DIR chamber, where the residual fuel is reformed simultaneously with the electrochemical fuel cell reactions. FuelCell Energy plans to offer commercial DFC power plants in various sizes, focusing on the subMW as well as the MW-scale units. The plan is to offer standardized, packaged DFC power plants operating on natural gas or other hydrocarbon-containing fuels for commercial sale. The power plant design will include a diesel fuel processing option to allow dual fuel applications. These power plants, which can be shop-fabricated and sited near the user, are ideally suited for distributed power

  17. Hydrocracking with molten zinc chloride catalyst containing 2-12% ferrous chloride

    DOE Patents [OSTI]

    Zielke, Clyde W.; Bagshaw, Gary H.

    1981-01-01

    In a process for hydrocracking heavy aromatic polynuclear carbonaceous feedstocks to produce hydrocarbon fuels boiling below about 475.degree. C. by contacting the feedstocks with hydrogen in the presence of a molten zinc chloride catalyst and thereafter separating at least a major portion of the hydrocarbon fuels from the spent molten zinc chloride catalyst, an improvement comprising: adjusting the FeCl.sub.2 content of the molten zinc chloride to from about 2 to about 12 mol percent based on the mixture of ferrous chloride and molten zinc chloride.

  18. Search for: thorium | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    ... 1 eV and 105 eV, perform as well as fast spectrum systems in this fuel cycle. ... continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. ...

  19. Sustainable Thorium Nuclear Fuel Cycles: A Comparison of Intermediate...

    Office of Scientific and Technical Information (OSTI)

    1 eV and 105 eV, perform as well as fast spectrum systems in this fuel cycle. ... continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. ...

  20. Microsoft Word - Sustainable_Thorium_v8.docx

    Office of Scientific and Technical Information (OSTI)

    1 eV and 105 eV, perform as well as fast spectrum systems in this fuel cycle. ... continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. ...

  1. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  2. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  3. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Wigner, E.P.

    1960-11-22

    A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

  4. REACTOR SHIELD

    DOE Patents [OSTI]

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  5. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  6. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  7. Tunable molten oxide pool assisted plasma-melter vitrification systems

    DOE Patents [OSTI]

    Titus, Charles H.; Cohn, Daniel R.; Surma, Jeffrey E.

    1998-01-01

    The present invention provides tunable waste conversion systems and apparatus which have the advantage of highly robust operation and which provide complete or substantially complete conversion of a wide range of waste streams into useful gas and a stable, nonleachable solid product at a single location with greatly reduced air pollution to meet air quality standards. The systems provide the capability for highly efficient conversion of waste into high quality combustible gas and for high efficiency conversion of the gas into electricity by utilizing a high efficiency gas turbine or an internal combustion engine. The solid product can be suitable for various commercial applications. Alternatively, the solid product stream, which is a safe, stable material, may be disposed of without special considerations as hazardous material. In the preferred embodiment, the arc plasma furnace and joule heated melter are formed as a fully integrated unit with a common melt pool having circuit arrangements for the simultaneous independently controllable operation of both the arc plasma and the joule heated portions of the unit without interference with one another. The preferred configuration of this embodiment of the invention utilizes two arc plasma electrodes with an elongated chamber for the molten pool such that the molten pool is capable of providing conducting paths between electrodes. The apparatus may additionally be employed with reduced use or without further use of the gases generated by the conversion process. The apparatus may be employed as a net energy or net electricity producing unit where use of an auxiliary fuel provides the required level of electricity production. Methods and apparatus for converting metals, non-glass forming waste streams and low-ash producing inorganics into a useful gas are also provided. The methods and apparatus for such conversion include the use of a molten oxide pool having predetermined electrical, thermal and physical

  8. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  9. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  10. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

  11. Electrochemical devices utilizing molten alkali metal electrode-reactant

    DOE Patents [OSTI]

    Hitchcock, David C.; Mailhe, Catherine C.; De Jonghe, Lutgard C.

    1986-01-01

    Electrochemical cells are provided with a reactive metal to reduce the oxide of the alkali metal electrode-reactant. Cells employing a molten alkali metal electrode, e.g., sodium, in contact with a ceramic electrolyte, which is a conductor of the ions of the alkali metal forming the electrode, exhibit a lower resistance when a reactive metal, e.g., vanadium, is allowed to react with and reduce the alkali metal oxide. Such cells exhibit less degradation of the electrolyte and of the glass seals often used to joining the electrolyte to the other components of the cell under cycling conditions.

  12. Electrochemical devices utilizing molten alkali metal electrode-reactant

    DOE Patents [OSTI]

    Hitchcock, D.C.; Mailhe, C.C.; De Jonghe, L.C.

    1985-07-10

    Electrochemical cells are provided with a reactive metal to reduce the oxide of the alkali metal electrode-reactant. Cells employing a molten alkali metal electrode, e.g., sodium, in contact with a ceramic electrolyte, which is a conductor of the ions of the alkali metal forming the electrode, exhibit a lower resistance when a reactive metal, e.g., vanadium, is allowed to react with and reduce the alkali metal oxide. Such cells exhibit less degradation of the electrolyte and of the glass seals often used to joining the electrolyte to the other components of the cell under cycling conditions.

  13. All ceramic structure for molten carbonate fuel cell

    DOE Patents [OSTI]

    Smith, James L.; Kucera, Eugenia H.

    1992-01-01

    An all-ceramic molten carbonate fuel cell having a composition formed of a multivalent metal oxide or oxygenate such as an alkali metal, transition metal oxygenate. The structure includes an anode and cathode separated by an electronically conductive interconnect. The electrodes and interconnect are compositions ceramic materials. Various combinations of ceramic compositions for the anode, cathode and interconnect are disclosed. The fuel cell exhibits stability in the fuel gas and oxidizing environments. It presents reduced sealing and expansion problems in fabrication and has improved long-term corrosion resistance.

  14. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  15. Salt Repository Research,

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    on Salt Repository Research, Design, and Operation La Fonda Hotel Santa Fe, New Mexico September 7 - 11, 2014 Please join us Sunday September 7, 2014 for a welcome and reception at the La Fonda Hotel hosted by Sandia National Laboratories beginning at 6:00 PM. Day 1 Technical Agenda September 8 - Monday 08:00-08:45 Sign-in and distribution of meeting materials 08:45-09:45 Welcome addresses H.C. Pape (BMWi) US-DOE Offices Highlights of US/German Collaboration F. Hansen (SNL) W. Steininger (PTKA)

  16. Device for controlling the pouring of molten materials

    DOE Patents [OSTI]

    Moore, A.F.; Duncan, A.L.

    1994-02-15

    A device is described for controlling the pouring of a molten material from a crucible or other container. The device includes an annular retainer ring for mounting in the drain opening in the bottom of a conventional crucible, the retainer ring defining a opening there through. The device also includes a plug member having an annular forward end portion for force-fit reception in the opening of the retainer ring to selectively seal the opening and for being selectively forced through the opening. The plug member has a rear end portion for being positioned within the crucible, the rear end portion including stop means for prohibiting the rear end portion from passing through the opening in the retainer ring when the forward end portion is selectively forced through the opening. The plug member defines at least one, and preferably a plurality of flutes, each extending from a point rearward the annular forward end portion of the plug member, and forward the stop means, to a point rearward of the stop means. The flutes permit fluid communication between the interior and exterior of the crucible when the forward end portion of the plug member is forced through the opening in the retaining ring such that the molten material is allowed to flow from the crucible. 5 figures.

  17. Device for controlling the pouring of molten materials

    DOE Patents [OSTI]

    Moore, Alan F. (Knoxville, TN); Duncan, Alfred L. (Clinton, TN)

    1994-01-01

    A device for controlling the pouring of a molten material from a crucible or other container. The device (10) includes an annular retainer ring (12) for mounting in the drain opening in the bottom of a conventional crucible (16), the retainer ring defining a opening (14) therethrough. The device (10) also includes a plug member (22) having an annular forward end portion (24) for force-fit reception in the opening (14) of the retainer ring (12) to selectively seal the opening (14) and for being selectively forced through the opening (14). The plug member (22) has a rear end portion (26) for being positioned within the crucible (16), the rear end portion (26) including stop means for prohibiting the rear end portion from passing through the opening (14) in the retainer ring (12) when the forward end portion (24) is selectively forced through the opening. The plug member (22) defines at least one, and preferably a plurality of flutes (32), each extending from a point rearward the annular forward end portion (24) of the plug member (22), and forward the stop means, to a point rearward of the stop means. The flutes (32) permit fluid communication between the interior and exterior of the crucible (16) when the forward end portion (24) of the plug member (22) is forced through the opening (14) in the retaining ring (12) such that the molten material is allowed to flow from the crucible (16).

  18. Electrolyte salts for nonaqueous electrolytes

    DOE Patents [OSTI]

    Amine, Khalil; Zhang, Zhengcheng; Chen, Zonghai

    2012-10-09

    Metal complex salts may be used in lithium ion batteries. Such metal complex salts not only perform as an electrolyte salt in a lithium ion batteries with high solubility and conductivity, but also can act as redox shuttles that provide overcharge protection of individual cells in a battery pack and/or as electrolyte additives to provide other mechanisms to provide overcharge protection to lithium ion batteries. The metal complex salts have at least one aromatic ring. The aromatic moiety may be reversibly oxidized/reduced at a potential slightly higher than the working potential of the positive electrode in the lithium ion battery. The metal complex salts may also be known as overcharge protection salts.

  19. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

    1958-09-01

    This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

  20. Study on LiCl waste salt treatment process by layer melt crystallization

    SciTech Connect (OSTI)

    Cho, Yung-Zun; Lee, Tae-Kyo; Choi, Jung-Hoon; Eun, Hee-Chul; Park, Hwan-Seo; Kim, In-Tae; Park, Geun-Il

    2013-07-01

    Layer melt crystallization operated in a static mode has been applied to separate Group I and II chlorides from surrogate LiCl waste salt. The effects of operating conditions such as crystal growing rate(or flux) and initial impurity concentration on separation (or concentration) of cesium, strontium and barium involved in a LiCl melts were analyzed. In a layer crystallization process, separation was impaired by occlusion of impurities and by residual melt adhering to LiCl crystal after at the end of the process. The crystal growth rate strongly affects the crystal structure, therefore the separation efficiency, while the effect of the initial Cs and Sr concentration in LiCl molten salt was nearly negligible. (authors)

  1. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    SciTech Connect (OSTI)

    Bahri, Che Nor Aniza Che Zainul Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-29

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  2. Catalytic reactor

    DOE Patents [OSTI]

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  3. POWER REACTOR

    DOE Patents [OSTI]

    Zinn, W.H.

    1958-07-01

    A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

  4. REACTOR CONTROL

    DOE Patents [OSTI]

    Fortescue, P.; Nicoll, D.

    1962-04-24

    A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

  5. Bioconversion reactor

    DOE Patents [OSTI]

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  6. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  7. Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control...

    Office of Scientific and Technical Information (OSTI)

    and health monitoring to prevent the development of acute or chronic beryllium disease, two pulmonary diseasesmore created by an allergic reaction to beryllium in the lungs. ...

  8. Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point...

    Office of Scientific and Technical Information (OSTI)

    ... Authors: Qualls, A. L. 1 ; Betzler, Benjamin R. 1 ; Carbajo, Juan 1 ; Hale, Richard Edward 1 ; Harrison, Thomas J. 1 ; Powers, Jeffrey J. 1 ; Robb, Kevin R. 1 ; ...

  9. (Reactor dosimetry)

    SciTech Connect (OSTI)

    West, C.D.

    1990-09-13

    The lead in most aspects of research reactor design and use passed from the USA about 15 years ago, soon after the construction of the HFIR and HFBR. The Europeans have consistently upgraded and improved their existing facilities and have built new ones including the HFR at Grenoble and ORPHEE at Saclay. They studied ultra-high flux concepts ({approximately}10{sup 20}/m{sup {minus}2}{center dot}s{sup {minus}1}) about 10 years ago, and are in the design phase of a new, highly efficient medium flux reactor to be built at Garching, near Munich in Germany. A visit was made to Interatom, the firm -- the equivalent of the Architect/Engineer for the ANS project -- responsible, under contract to the Technical University of Munich, for the new Munich reactor design. There are many similarities to the ANS design, and we reviewed and discussed technical and safety aspects of the two reactors. A request was made for some new, hitherto proprietary, experimental data on reactor thermal hydraulics and cooling that will be very valuable to the ANS project. I presented a seminar on the ANS project. A visit was made to Kernforschungszentrum Karlsruhe and knowledge was gained from Dr. Kuchle, a true pioneer of ultra-high flux reactor concepts, of their work. Dr. Kuchle kindly reviewed the ANS reference core and cooling system design (with favorable conclusions). I then talked with researchers working on materials irradiation damage and activation of structural materials by neutron irradiation, both key issues for the ANS. I was shown some new techniques they have developed for testing materials irradiation effects at high fluences, in a short time, using accelerated particle beams.

  10. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Szilard, L.

    1957-09-24

    Reactors of the type employing plates of natural uranium in a moderator are discussed wherein the plates are um-formly disposed in parallel relationship to each other thereby separating the moderator material into distinct and individual layers. Each plate has an uninterrupted sunface area substantially equal to the cross-sectional area of the active portion of the reactor, the particular size of the plates and the volume ratio of moderator to uranium required to sustain a chain reaction being determinable from the known purity of these materials and other characteristics such as the predictable neutron losses due to the formation of radioactive elements of extremely high neutron capture cross section.

  11. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  12. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Young, G.

    1963-01-01

    This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

  13. Baseload Nitrate Salt Central Receiver Power Plant Design Final Report

    SciTech Connect (OSTI)

    Tilley, Drake; Kelly, Bruce; Burkholder, Frank

    2014-12-12

    The objectives of the work were to demonstrate that a 100 MWe central receiver plant, using nitrate salt as the receiver coolant, thermal storage medium, and heat transport fluid in the steam generator, can 1) operate, at full load, for 6,400 hours each year using only solar energy, and 2) satisfy the DOE levelized energy cost goal of $0.09/kWhe (real 2009 $). To achieve these objectives the work incorporated a large range of tasks relating to many different aspects of a molten salt tower plant. The first Phase of the project focused on developing a baseline design for a Molten Salt Tower and validating areas for improvement. Tasks included a market study, receiver design, heat exchanger design, preliminary heliostat design, solar field optimization, baseline system design including PFDs and P&IDs and detailed cost estimate. The baseline plant met the initial goal of less than $0.14/kWhe, and reinforced the need to reduce costs in several key areas to reach the overall $0.09/kWhe goal. The major improvements identified from Phase I were: 1) higher temperature salt to improve cycle efficiency and reduce storage requirements, 2) an improved receiver coating to increase the efficiency of the receiver, 3) a large receiver design to maximize storage and meet the baseload hours objective, and 4) lower cost heliostat field. The second Phase of the project looked at advancing the baseline tower with the identified improvements and included key prototypes. To validate increasing the standard solar salt temperature to 600 °C a dynamic test was conducted at Sandia. The results ultimately proved the hypothesis incorrect and showed high oxide production and corrosion rates. The results lead to further testing of systems to mitigate the oxide production to be able to increase the salt temperature for a commercial plant. Foster Wheeler worked on the receiver design in both Phase I and Phase II looking at both design and lowering costs utilizing commercial fossil boiler

  14. DOE - Office of Legacy Management -- Salt_Lake

    Office of Legacy Management (LM)

    Salt_Lake Salt Lake City Sites ut_map Salt Lake City Disposal Site Salt Lake City Processing Site Last Updated: 12/14/2015

  15. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  16. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Starr, C.

    1963-01-01

    This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

  17. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

    1959-10-27

    BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

  18. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J.

    1976-08-17

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

  19. Neutronic reactor

    DOE Patents [OSTI]

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  20. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Anderson, H.L.

    1958-10-01

    The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

  1. Plant salt-tolerance mechanisms

    SciTech Connect (OSTI)

    Deinlein, Ulrich; Stephan, Aaron B.; Horie, Tomoaki; Luo, Wei; Xu, Guohua; Schroeder, Julian I.

    2014-06-01

    Crop performance is severely affected by high salt concentrations in soils. To engineer more salt-tolerant plants it is crucial to unravel the key components of the plant salt-tolerance network. Here we review our understanding of the core salt-tolerance mechanisms in plants. Recent studies have shown that stress sensing and signaling components can play important roles in regulating the plant salinity stress response. We also review key Na+ transport and detoxification pathways and the impact of epigenetic chromatin modifications on salinity tolerance. In addition, we discuss the progress that has been made towards engineering salt tolerance in crops, including marker-assisted selection and gene stacking techniques. We also identify key open questions that remain to be addressed in the future.

  2. Plant salt-tolerance mechanisms

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Deinlein, Ulrich; Stephan, Aaron B.; Horie, Tomoaki; Luo, Wei; Xu, Guohua; Schroeder, Julian I.

    2014-06-01

    Crop performance is severely affected by high salt concentrations in soils. To engineer more salt-tolerant plants it is crucial to unravel the key components of the plant salt-tolerance network. Here we review our understanding of the core salt-tolerance mechanisms in plants. Recent studies have shown that stress sensing and signaling components can play important roles in regulating the plant salinity stress response. We also review key Na+ transport and detoxification pathways and the impact of epigenetic chromatin modifications on salinity tolerance. In addition, we discuss the progress that has been made towards engineering salt tolerance in crops, including marker-assisted selectionmore » and gene stacking techniques. We also identify key open questions that remain to be addressed in the future.« less

  3. OECD MCCI project long-term 2-D molten core concrete interaction test design report, Rev. 0. September 30, 2002.

    SciTech Connect (OSTI)

    Farmer, M. T.; Kilsdonk, D. J.; Lomperski, S.; Aeschliman, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following two technical objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of the first program objective, the Small-Scale Water Ingression and Crust Strength (SSWICS) test series has been initiated to provide fundamental information on the ability of water to ingress into cracks and fissures that form in the debris during quench, thereby augmenting the otherwise conduction-limited heat transfer process. A test plan for Melt Eruption Separate Effects Tests (MESET) has also been developed to provide information on the extent of crust growth and melt eruptions as a function of gas sparging rate under well-controlled experiment conditions. In terms of the second program objective, the project Management Board (MB) has approved startup activities required to carry out

  4. Method for producing hydrocarbon fuels from heavy polynuclear hydrocarbons by use of molten metal halide catalyst

    DOE Patents [OSTI]

    Gorin, Everett

    1979-01-01

    In a process for hydrocracking heavy polynuclear carbonaceous feedstocks to produce lighter hydrocarbon fuels by contacting the heavy feedstocks with hydrogen in the presence of a molten metal halide catalyst, thereafter separating at least a substantial portion of the carbonaceous material associated with the reaction mixture from the spent molten metal halide and thereafter regenerating the metal halide catalyst, an improvement comprising contacting the spent molten metal halide catalyst after removal of a major portion of the carbonaceous material therefrom with an additional quantity of hydrogen is disclosed.

  5. Vehicle Technologies Office Merit Review 2015: Efficient Rechargeable Li/O2 Batteries Utilizing Stable Inorganic Molten Salt Electrolytes

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation given by Liox at 2015 DOE Hydrogen and Fuel Cells Program and Vehicle Technologies Office Annual Merit Review and Peer Evaluation Meeting about efficient rechargeable Li/O2 batteries...

  6. Fundamental Corrosion Studies in High-Temperature Molten Salt Systems for Next-Generation CSP Systems- FY13 Q2

    Broader source: Energy.gov [DOE]

    This document summarizes the progress of this SRNL project, funded by SunShot, for the second quarter of fiscal year 2013.

  7. Structural Interactions within Lithium Salt Solvates: Acyclic...

    Office of Scientific and Technical Information (OSTI)

    Structural Interactions within Lithium Salt Solvates: Acyclic Carbonates and Esters Citation Details In-Document Search Title: Structural Interactions within Lithium Salt Solvates: ...

  8. Enterprise Assessments Salt Waste Processing Facility Construction...

    Office of Environmental Management (EM)

    Salt Waste Processing Facility Construction Quality and Fire Protection Systems Follow-up Review at the Savannah River Site - January 2016 Enterprise Assessments Salt Waste ...

  9. REACTOR MONITORING

    DOE Patents [OSTI]

    Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

    1959-02-01

    A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

  10. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Treshow, M.

    1958-08-19

    A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

  11. REACTOR UNLOADING

    DOE Patents [OSTI]

    Leverett, M.C.

    1958-02-18

    This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

  12. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wade, E.J.

    1958-09-16

    This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

  13. Nuclear reactor

    DOE Patents [OSTI]

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  14. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  15. NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  16. Neutronic reactor

    DOE Patents [OSTI]

    Lewis, Warren R.

    1978-05-30

    A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

  17. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  18. REACTOR CONTROL

    DOE Patents [OSTI]

    Ruano, W.J.

    1957-12-10

    This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

  19. Molten Carbonate and Phosphoric Acid Stationary Fuel Cells: Overview and Gap Analysis

    Fuel Cell Technologies Publication and Product Library (EERE)

    This report details technical and cost gap analyses of molten carbonate fuel cell and phosphoric acid fuel cell stationary fuel cell power plants and identifies pathways for reducing costs.

  20. Molten Carbonate and Phosphoric Acid Stationary Fuel Cells. Overview and Gap Analysis

    SciTech Connect (OSTI)

    Remick, Robert; Wheeler, Douglas

    2010-09-01

    This report details technical and cost gap analyses of molten carbonate fuel cell and phosphoric acid fuel cell stationary fuel cell power plants and identifies pathways for reducing costs.