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Office of Scientific and Technical Information
Accelerators for Subcritical Molten-Salt Reactors
SciTech Connect
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.
Johnson, Roland (Muons, Inc.)
Fast Spectrum Molten Salt Reactor Options
DOE Green Energy (OSTI)
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.
Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Patton, Bruce W [ORNL; Howard, Rob L [ORNL; Harrison, Thomas J [ORNL
ABS> Fuel reprocessing methods are being investigated for molten salt nuclear reactors which use LiF--BeF/sub 2/ salt as a solvent for UF/sub 4/ and ThF/sub 4/. A liquid HF dissolution procedure coupled with fluorination has been developed for recovery of the uranium and LiF- BeF/sub 2/ solvent salt which is highly enriched in Li/sup 7/. The recovered salt is decontaminated in the process from the major reactor poisons; namely, rare earths and neptunium. A brief investigation of alternate methods, including oxide precipitation, partial freezing, and metal reduction, indicated that such methods may give some separation of the solvent salt from reactor poisons, but they do not appear to be sufficiently quantitative for a simple processing operation. Solubilities of LiF and BeF/sub 2/ in aqueous 70t0 100% HF are presented. The BeF/sub 2/ solubility is appreciably increased in the presence of water and large amounts of LiF. Salt solubilities of 150 g/liter are attainable. Tracer experiments indicate that rare earth solubilities, relative to LiF-- BeF/sub 2/ solvent salt solubility, increase from about 10/sup -4/ mole% in 98% HF to 0.003 mole% in 80% HF. Fluorination of uranium from LiF--BeF/sub 2/ salt was demonstrated. This appears feasible also for the recovery of the relatively small ccncentration of uranium produced in the LiF- BeF/sub 2/ThF/sub 4/ blanket. A proposed chemical flowsheet is presented on the basis of this exploratory work as applied to the semicontinuous processing of a 600 Mw power reactor. (auth)
Campbell, D.O.; Cathers, G.I.
Developments in Molten Salt and Liquid-Salt-Cooled Reactors
Science Conference Proceedings (OSTI)
In the last 5 years, there has been a rapid growth in interest in the use of high-temperature (700 to 1000 deg C) molten and liquid fluoride salts as coolants in nuclear systems. This renewed interest is a consequence of new applications for high-temperature heat and the development of new reactor concepts. Fluoride salts have melting points between 350 and 500 deg C; thus, they are of use only in high-temperature systems. Historically, steam cycles with temperature limits of {approx}550 deg C have been the only efficient method to convert heat to electricity. This limitation produced few incentives to develop high-temperature reactors for electricity production. However, recent advances in Brayton gas turbine technology now make it possible to convert higher-temperature heat efficiency into electricity on an industrial scale and thus have created the enabling technology for more efficient nuclear reactors. Simultaneously, there is a growing interest in using high-temperature nuclear heat for the production of hydrogen and shale oil. Five nuclear-related applications are being investigated: (1) liquid-salt heat-transport systems in hydrogen and shale oil production systems; (2) the advanced high-temperature reactor, which uses a graphite-matrix coated-particle fuel and a liquid salt coolant; (3) the liquid-salt-cooled fast reactor which uses metal-clad fuel and a liquid salt coolant; (4) the molten salt reactor, with the fuel dissolved in the molten salt coolant; and (5) fusion energy systems. The reasons for the new interest in liquid salt coolants, the reactor concepts, and the relevant programs are described. (author)
Forsberg, Charles W. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6165 (United States)
Fuel-salt volumes external to the core of a molten-salt reactor are calculated for a system in which the fuel salt circulates through the core and primary exchanger by free convection. In the calculation of these volumes, the exchanger heights above the core top range from 5 to 20 ft. Coolants considered for the primary exchanger are a second molten salt and helium. External fuel holdup is found to be the same with either coolant. Two sets of terminal temperatures are selected for the helium. The first combination permits steam generation at 850 psia, 900 deg F. The second set is selected for a closed gas turbine cycle with an 1100 deg F turbine inlet temperature. Specific power (thermal kw/kg 235) is found to be about 900 Mv/kg, based on initial, clean conditions and a 60 Mw (thermal) output. A specific power of 1275 kw/kg is estimated for a forced convection system of the same rating. (auth)
Romie, F.E.; Kinyon, B.W.
The Thorium Molten Salt Reactor Moving on from the MSBR
E-Print Network (OSTI)
A re-evaluation of the Molten Salt Breeder Reactor concept has revealed problems related to its safety and to the complexity of the reprocessing considered. A reflection is carried out anew in view of finding innovative solutions leading to the Thorium Molten Salt Reactor concept. Several main constraints are established and serve as guides to parametric evaluations. These then give an understanding of the influence of important core parameters on the reactor's operation. The aim of this paper is to discuss this vast research domain and to single out the Molten Salt Reactor configurations that deserve further evaluation.
Mathieu, L; Brissot, R; Le Brun, C; Liatard, E; Loiseaux, J M; M�plan, O; Merle-Lucotte, E; Nuttin, A; Wilson, J; Garzenne, C; Lecarpentier, D; Walle, E
The Thorium Molten Salt Reactor : Moving on from the MSBR
L. Mathieu; D. Heuer; R. Brissot; C. Le Brun; E. Liatard; J. M. Loiseaux; O. M�plan; E. Merle-Lucotte; A. Nuttin; J. Wilson; C. Garzenne; D. Lecarpentier; E. Walle; the GEDEPEON Collaboration
System Requirements Document for the Molten Salt Reactor Experiment
The purpose of the conversion process is to convert the {sup 233}U fluoride compounds that are being extracted from the Molten Salt Reactor Experiment (MSRE) equipment to a stable oxide for long-term storage at Bldg. 3019.
Aigner, R.D.
Molten Salt Breeder Reactors Academia Sinica, ITRI, NTHU
Power 4/8/12 Frank H. Shu Gen IV MSBR/LFTR Liquid fuel (molten salt) Molten salt coolant (unpopulated
Wang, Ming-Jye
A preliminary study of molten salt pcwer reactors was made. The most promising fuel carrier salts were the fluorides and chlorides of the alkali metals, zirconium, and beryllium. The chlorides were found to have lower melting points but were less stable and more corrosive than the fluorides. A Li/sup 7/ F- - BeF/sub 2/ mixture with ThF/sub 4/ and UF/sub 4/appeared to perform best. Of the numerous alloys tested as container material, Inconel and a nickel-- molybdenum alloy INOR-8 appeared to be the most resistant to corrosion. To study the performance, safety, economics, and construction costs of a typical molten salt reactor, a reactor of specific type and size was chosen for study. The reference design reactor was a two-region homogeneous converter with a core salt of 70 mole% Li/sup 7/F and 30% BeF/sub 2. ThF/sub 4/ and enough VF/sub 4/ for criticality were added. Study in- dicated that a molten salt reactor would prcduce economical power, but the problem of developing a salt core and a container metal which would last for mamy years of operation needed further study. (M.C.G.)
MacPherson, H.G.; Alexander, L.G.; Carrison, D.A.; Estabrook, J.Y.; Kinyon, B.W.; Mann, L.A.; Roberts, J.T.; Romie, F.E.; VonderLage, F.C.
Fission product behavior in the Molten Salt Reactor Experiment
Essentially all the fission product data for numerous and varied samples taken during operation of the Molten Salt Reactor Experiment or as part of the examination of specimens removed after particular phases of operation are reported, together with the appropriate inventory or other basis of comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct chemical groups. Evidence for fission product behavior during operation over a period of 26 months with $sup 235$U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using $sup 233$U fuel over a period of about 15 months (more than 5100 effective full- power hours). (auth)
Compere, E.L.; Kirslis, S.S.; Bohlmann, E.G.; Blankenship, F.F.; Grimes, W.R.
Removal of uranium and salt from the Molten Salt Reactor Experiment
In 1994, migration of {sup 233}U was discovered to have occurred at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). This paper describes the actions now underway to remove uranium from the off-gas piping and the charcoal bed, to remove and stabilize the salts, and to convert the uranium to a stable oxide for long-term storage.
Peretz, F.J.; Rushton, J.E.; Faulkner, R.L.; Walker, K.L.; Del Cul, G.D.
A rough survey of the nuclear charactenistics of graphite-moderated molten-salt reactors utilizing an initial complement of low enrichment uranium fuel has been made. Reactors can be constructed with initial enrichinents as low as 1.25% U-235; initial conversion ratios of as high as 0.8 can be obtained with enrichinent of less than 2%. Highly enriched uraninm would be added as make-up fuel, and such reactors could probably be operated for bunnups as high as 60,000 Mwd/ton before buildup of fission preducts wpuld make replacement of the fuel desirable. A typical circulating fuel reactor of this class might contain an initial inventory of 3600 tons of 1.8% enriched uranium, operated at 640 Mw (thermal), and generate a net of 260 Mw (electrical). The total fuel cycle cost would be approximately 1.3 mills/kwhr, of which 1.0 mill is bunnup of enniched U- 235. (auth)
MacPherson, H.G.
Conceptual Design of Molten Salt Loop Experiment for MIT Research Reactor
Molten salt is a promising coolant candidate for Advanced High Temperature Reactor (AHTR) Gen-IV designs. The low neutron absorption, high thermal capacity, chemical inertness, and high boiling point at low pressure of ...
Bean, Malcolm K.
Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system
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.
Harto, Andang Widi [Engineering Physics Department, Faculty of Engineering, Gadjah Mada University (Indonesia)
Investigation of Thermal Feedback Design for Improved Load-Following Capability of Thorium Molten Salt Reactors
The increasing deployment of renewable energy sources has raised concerns about the ramp-rate limitations of conventional steam and combustion turbines in providing load following during solar photovoltaic transients. As one of the promising Generation ... Keywords: molten salt reactors, thorium
Andrew M. Dodson, Roy A. Mccann
Molten salt electrolyte separator
DOE Patents (OSTI)
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.
Kaun, Thomas D. (New Lenox, IL)
Preliminary safety calculations to improve the design of Molten Salt Fast Reactor
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)
Brovchenko, M.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Capellan, N.; Ghetta, V.; Laureau, A. [LPSC, CNRS/IN2P3, Grenoble INP, 53,rue des Martyrs, 38026 Grenoble Cedex (France)
Plutonium and minor actinides utilization in Thorium molten salt reactor
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.
Waris, Abdul; Aji, Indarta K.; Novitrian,; Kurniadi, Rizal; Su'ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jalan Ganesa 10 Bandung 40132 (Indonesia)
A method for selectively precipitating protactinium from a neutron- irradiated fused fluoride salt composition comprising at least one metal fluoride selected from the group consisting of an alkali metal fluoride and an alkaline earth metal fluoride containing dissolved thorium-232 values is presented. An inorganic metal oxide corresponding to any of the metal fluorides of the composition is also added. (AEC)
Shaffer, J.H.; Strain, J.E.; Cuneo, D.R.; Kelly, M.J.
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.
Kaun, T.D.
Engineering Evaluation of Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiement for the Oak Ridge National Laboratory
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 - There are likely multiple phases of material in the salt (metal or compound), either suspended through the salt matrix, layered in the bottom of the tank, or both. These phases may contribute to plugging during any planned transfer. There is not enough data to know for sure. (4) Probe heat trace - The alternate transfer method does not include heat tracing of the bottom of the probe. There is a concern that this may cool the salt and other phases of materials present enough to block the flow of salt. (5) Stress-corrosion cracking - Additionally, there is a concern regarding moisture that may have been introduced into the tanks. Due to time constraints, this concern was not validated. However, if moisture was introduced into the tanks and not removed during heating the tanks before HF and F2 sparging, there would be an additional concern regarding the potential for stress-corrosion cracking of the tank walls.
Carlberg, Jon A.; Roberts, Kenneth T.; Kollie, Thomas G.; Little, Leslie E.; Brady, Sherman D.
Molten salt test loop
The objective of the Molten Salt Test Loop Project was to design, construct, and demonstrate operation of an outdoor high temperature molten salt test facility. This facility is operational, and can now be used to evaluate materials and components, and the design features and operating procedures required for molten salt heat transport systems. The initial application of the loop was to demonstrate the feasibility of using molten salt as the heat transport medium for a high temperature distributed collector system. A commercially available eutectic salt blend is used as the heat transfer fluid. This salt has a composition of 40% NaNO/sub 2/, 7% NaNO/sub 3/, and 53% KNO/sub 3/ and is marketed under the trade name Hitec. It has a freezing (solidifying) point of 142/sup 0/C (288/sup 0/F) and has been satisfactorily used at temperatures as high as 594/sup 0/C (1100/sup 0/F). General Atomic (GA) installed a row of Fixed Mirror Solar Concentrators (FMSC's) in the loop. The system was started up and a test program conducted. Startup went smoothly, with the exception of some burned-out trace heaters. Salt temperatures as high as 571/sup 0/C (1060/sup 0/F) were achieved.
Schuster, J.R.; Eggers, G.H.
KC! /FeS 2 cell lithium-silicon magnesium oxide molten-saltmolten-salt cells Na/Na glass/Na:z.Sn-S cell Na/NazO�xA!Symposium on Molten Salts, Physical Electrochemistry
Cairns, Elton J.
Molten salt destruction of energetic waste materials
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.
Brummond, William A. (Livermore, CA); Upadhye, Ravindra S. (Pleasanton, CA); Pruneda, Cesar O. (Livermore, CA)
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.
Brummond, W.A.; Upadhye, R.S.; Pruneda, C.O.
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.
Barton, C.J.; Grimes, W.R.
Molten Salts, Magnesium and Aluminum
Mar 1, 2011 ... Chloride 2011: Practice and Theory of Chloride-Based Metallurgy: Molten Salts, Magnesium and Aluminum Sponsored by: The Minerals,�...
Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel
The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [{approx}1000 Mw(e)] with passive safety systems to provide the potential for improved economics.
Forsberg, C.W.
Molten salt lithium cells
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.
Raistrick, I.D.; Poris, J.; Huggins, R.A.
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).
Raistrick, Ian D. (Menlo Park, CA); Poris, Jaime (Portola Valley, CA); Huggins, Robert A. (Stanford, CA)
Extracting information from the molten salt database
Molten salt technology is a catchall phrase that includes some very diverse ... nologies are linked by the general characteristics of molten salts that can function
Molten Salt Mixture Properties (KF-ZrF4 and KCl-MgCl2) for Use in RELAP5-3D for High Temperature Reactor Application
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.
N. A. Anderson; P. Sabharwall
Identification and evaluation of alternatives for the disposition of fluoride fuel and flush salts from the molten salt reactor experiment at Oak Ridge National Laboratory, Oak Ridge, Tennessee
This document presents an initial identification and evaluation of the alternatives for disposition of the fluoride fuel and flush salts stored in the drain tanks at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). It will serve as a resource for the U.S. Department of Energy contractor preparing the feasibility study for this activity under the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA). This document will also facilitate further discussion on the range of credible alternatives, and the relative merits of alternatives, throughout the time that a final alternative is selected under the CERCLA process.
Program management plan for the Molten Salt Reactor Experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee
The primary mission of the Molten Salt Reactor Experiment (MSRE) Remediation Project is to effectively implement the risk-reduction strategies and technical plans to stabilize and prevent further migration of uranium within the MSRE facility, remove the uranium and fuel salts from the system, and dispose of the fuel and flush salts by storage in appropriate depositories to bring the facility to a surveillance and maintenance condition before decontamination and decommissioning. This Project Management Plan (PMP) for the MSRE Remediation Project details project purpose; technical objectives, milestones, and cost objectives; work plan; work breakdown structure (WBS); schedule; management organization and responsibilities; project management performance measurement planning, and control; conduct of operations; configuration management; environmental, safety, and health compliance; quality assurance; operational readiness reviews; and training.
Batteries using molten salt electrolyte
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.
Guidotti, Ronald A. (Albuquerque, NM)
Cathode for molten salt batteries
A molten salt electrochemical system for battery applications comprises tetravalent sulfur as the active cathode material with a molten chloroaluminate solvent comprising a mixture of AlCl.sub.3 and MCl having a molar ratio of AlCl.sub.3 /MCl from greater than 50.0/50.0 to 80/20.
Mamantov, Gleb (Knoxville, TN); Marassi, Roberto (Camerino, IT)
Delivery system for molten salt oxidation of solid waste
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.
Brummond, William A. (Livermore, CA); Squire, Dwight V. (Livermore, CA); Robinson, Jeffrey A. (Manteca, CA); House, Palmer A. (Walnut Creek, CA)
Molten metal reactors
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.
Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M
Measurement of the axial distribution of radioactivity in the auxiliary charcoal bed of the Molten Salt Reactor Experiment at ORNL
The Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory commenced operation in 1964 and was shut down in 1969. It was fueled with {sup 233}UF{sub 4} in a carrier salt of LiF-BeF{sub 2}-ZrF{sub 4}, and it operated at 1,200 F. After it was shut down, the fuel was heated annually to 200 C to recombine fluorine (with the fuel) released due to radiation-induced reactions in the fuel salt. However, a competing reaction oxidized uranium to UF{sub 6}, which was released (along with F{sub 2}) from the fuel and trapped in one of four charcoal filters in the auxiliary charcoal bed (ACB). One of the tasks for decommissioning of the MSRE requires that at least 90% of the estimated 3 kg of {sup 233}U, and radioactive decay products, in this filter be removed, and one of the proposed methods is to vacuum the charcoal above a specified axial position in the filter. This requires that the axial distribution of activity in the filter be measured in a 60 rad/h radiation field to determine where this penetration can be made. To accomplish this, the shielded detector with a pinhole collimator, and with a laser positioning capability, was remotely translated to various axial positions to accomplish these measurements. Activities in the steel screen, and various regions of the charcoal bed, are estimated, and uncertainties in these estimates are generally {lt}1%. Results from this analysis are used for continued operational decisions for decommissioning of the MSRE.
Miller, L.F.; Buckner, M.; Buchanan, M.
Nuclear characteristics of a 1000-MW(e) molten-salt breeder reactor
Included are brief discussions of the MSBR reference concept, core neutronics, core design, breeding and resource utilization, and reactor kinetics. (DG)
Engel, J.R.; Kerr, H.T.; Allen, E.J.
Preliminary Design For Conventional and Compact Secondary Heat Exchanger in a Molten Salt Reactor
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
Piyush Sabharwall; Mike Patterson; Ali Siahpush; Eung Soo Kim
Modeling of Porous Electrodes in Molten-Salt Systems
of Porous Electrodes in Molten-Salt Systems^ John Newmanon High-Temperature Molten Salt B a t - teries, Argonneby the modeling of molten-salt cells, including some
Newman, John
Dysprosium Extraction Using Molten Salt Electrolysis Process
AlCl3 was used as a chlorinating agent in order to enable an efficient dissolution of metal in the molten salt phase in the salt bath. The metal chloride which is�...
Corrosion of High Temperature Alloys in Molten Salts
Fluoride and chloride salts are among the candidates for this application. However, materials corrosion is an issue in these molten salts, particularly in molten�...
Injector nozzle for molten salt destruction of energetic waste materials
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.
Brummond, W.A.; Upadhye, R.S.
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.
Brummond, William A. (Livermore, CA); Upadhye, Ravindra S. (Pleasanton, CA)
Molten salt safety study. Final report
The considerations concerning safety in using molten salt (40% potassium nitrate, 60% sodium nitrate) in a solar central receiver plant are addressed. The considerations are of a general nature and do not cover any details of equipment or plant operation. The study includes salt chemical reaction, experiments with molten salt, dry storage and handling constraints, and includes data from the National Fire Protection Association. The contents of this report were evaluated by two utility companies and they concluded that no major safety problems exist in using a molten salt solar system.
Not Available
Molten salt/metal extractions for recovery of transuranic elements
The integral fast reactor (EFR) is an advanced reactor concept that incorporates metallic driver and blanket fuels, an inherently safe, liquid-sodium-cooled, pool-type, reactor design, and on-site pyrochemical reprocessing (including electrorefining) of spent fuels and wastes. This paper describes a pyrochemical method that is being developed at Argonne National Laboratory to recover transuranic elements from the EFR electrorefiner process salt. The method uses multistage extractions between molten chloride salts and cadmium metal at high temperatures. The chemical basis of the salt extraction method, the test equipment, and a test plan are discussed.
Chow, L.S.; Basco, J.K.; Ackerman, J.P.; Johnson, T.R.
Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors
K. T. Assessment of Candidate Molten Salt Coolants for theK. T. Assessment of Candidate Molten Salt Coolants for thebeginning efforts for a molten salt reactor (MSR) program.
Scarlat, Raluca Olga
Natural Convection Fluoride Salt High Temperature Reactor Process ...
... oil shale processing, hydrogen production, and production of synfuels from coal. The new nuclear reactor design employs a molten salt coolant in a natural ...
Prototype Tests for the Recovery and Conversion of UF6 Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project
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.
Del Cul, G.D.; Icenhour, A.S.; Simmons, D.W.
Prototype Tests for the Recovery and Conversion of UF6Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project
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.
Del Cul, G.D.
Transmutation and inventory analysis in an ATW molten salt system
As an extension of earlier work to determine the equilibrium state of an ATW molten salt, power producing, reactor/transmuter, the WAIT code provides a time dependent view of material inventories and reactor parameters. By considering several cases, we infer that devices of this type do not reach equilibrium for dozens of years, and that equilibrium design calculations are inapplicable over most of the reactor life. Fissile inventory and keff both vary by factors of 1.5 or more between reactor startup and ultimate convergence to equilibrium.
Sisolak, J. E.; Truebenbach, M. T.; Henderson, D. L. [Department of Nuclear Engineering and Engineering Physics University of Wisconsin-Madison, Madison, Wisconsin 53706-1687 (United States)
Evaluation of Salt Coolants for Reactor Applications
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.
Williams, David F [ORNL
Colloidal stability of magnetic nanoparticles in molten salts
Molten salts are important heat transfer fluids used in nuclear, solar and other high temperature engineering systems. Dispersing nanoparticles in molten salts can enhance the heat transfer capabilities of the fluid. High ...
Somani, Vaibhav (Vaibhav Basantkumar)
Continuous extraction of molten chloride salts with liquid cadmium alloys
A pyrochemical method is being developed at Argonne National Laboratory (ANL) to provide contnuous multistage extractions between molten chloride salts and liquid cadmium alloys at 500{degrees}C. The extraction method will be used to recover transuranic (TRU) elements from the process salt in the electroretiner used in the pyrochemical reprocessing of spent fuel from the Integral Fast Reactor (IFR). The IFR is one of the Department of Energy`s advanced power reactor concepts. The recovered TRU elements are returned to the electrorefiner. The extracted salt undergoes further processing to remove rare earths and other fission products so that most of the purified salt can also be returned to the electrorefiner, thereby extending the useful life of the process salt many times.
Applications of molten salts in plutonium processing
Plutonium is efficiently recovered from scrap at Los Alamos by a series of chemical reactions and separations conducted at temperatures ranging from 700 to 900/sup 0/C. These processes usually employ a molten salt or salt eutectic as a heat sink and/or reaction medium. Salts for these operations were selected early in the development cycle. The selection criteria are being reevaluated. In this article we describe the processes now in use at Los Alamos and our studies of alternate salts and eutectics.
Bowersox, D.F.; Christensen, D.C.; Williams, J.D.
Molten Salt Heat Transfer Fluid (HTF) - Energy Innovation Portal
Solar Thermal Industrial Technologies Energy Storage Molten Salt Heat Transfer Fluid (HTF) Sandia National Laboratories. Contact SNL About This ...
CO2 Emission Reduction through Innovative Molten Salt Electrolysis ...
Electrochemical metallurgy especially through high temperature molten salt electrolysis with renewable electricity stands for a great opportunity for producing
Thermal Characterization of Molten Salt Systems
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.
Toni Y. Gutknecht; Guy L. Fredrickson
Helium-cooled molten-salt fusion breeder
We present a new conceptual design for a fusion reactor blanket that is intended to produce fissile material for fission power plants. Fast fission is suppressed by using beryllium instead of uranium to multiply neutrons. Thermal fission is suppressed by minimizing the fissile inventory. The molten-salt breeding medium (LiF + BeF/sub 2/ + ThF/sub 4/) is circulated through the blanket and to the on-line processing system where /sup 233/U and tritium are continuously removed. Helium cools the blanket and the austenitic steel tubes that contain the molten salt. Austenitic steel was chosen because of its ease of fabrication, adequate radiation-damage lifetime, and low corrosion by molten salt. We estimate that a breeder having 3000 MW of fusion power will produce 6500 kg of /sup 233/U per year. This amount is enough to provide makeup for 20 GWe of light-water reactors per year or twice that many high-temperature gas-cooled reactors or Canadian heavy-water reactors. Safety is enhanced because the afterheat is low and blanket materials do not react with air or water. The fusion breeder based on a pre-MARS tandem mirror is estimated to cost $4.9B or 2.35 times a light-water reactor of the same power. The estimated cost of the /sup 233/U produced is $40/g for fusion plants costing 2.35 times that of a light-water reactor if utility owned or $16/g if government owned.
Moir, R.W.; Lee, J.D.; Fulton, F.J.; Huegel, F.; Neef, W.S. Jr.; Sherwood, A.E.; Berwald, D.H.; Whitley, R.H.; Wong, C.P.C.; Devan, J.H.
Molten nitrate salt technology development status report
Recognizing thermal energy storage as potentially critical to the successful commercialization of solar thermal power systems, the Department of Energy (DOE) has established a comprehensive and aggressive thermal energy storage technology development program. Of the fluids proposed for heat transfer and energy storage molten nitrate salts offer significant economic advantages. The nitrate salt of most interest is a binary mixture of NaNO/sub 3/ and KNO/sub 3/. Although nitrate/nitrite mixtures have been used for decades as heat transfer and heat treatment fluids the use has been at temperatures of about 450/sup 0/C and lower. In solar thermal power systems the salts will experience a temperature range of 350 to 600/sup 0/C. Because central receiver applications place more rigorous demands and higher temperatures on nitrate salts a comprehensive experimental program has been developed to examine what effects, if any, the new demands and temperatures have on the salts. The experiments include corrosion testing, environmental cracking of containment materials, and determinations of physical properties and decomposition mechanisms. This report details the work done at Sandia National Laboratories in each area listed. In addition, summaries of the experimental programs at Oak Ridge National Laboratory, the University of New York, EIC Laboratories, Inc., and the Norwegian Institute of Technology on molten nitrate salts are given. Also discussed is how the experimental programs will influence the near-term central receiver programs such as utility repowering/industrial retrofit and cogeneration. The report is designed to provide easy access to the latest information and data on molten NaNO/sub 3//KNO/sub 3/ for the designers and engineers of future central receiver projects.
Carling, R.W.; Kramer, C.M.; Bradshaw, R.W.; Nissen, D.A.; Goods, S.H.; Mar, R.W.; Munford, J.W.; Karnowsky, M.M.; Biefeld, R.N.; Norem, N.J.
Molten salt battery having inorganic paper separator
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.
Walker, Jr., Robert D. (Gainesville, FL)
Diffusion Welding of Alloys for Molten Salt Service - Status Report
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. (RR E) A different set of alloys, such as Alloy N and 242, are needed to handle molten salts at this temperature. The diffusion welding development work described here builds on techniques developed during the NGNP work, as applied to these alloys. There is also the matter of dissimilar metal welding, since alloys suitable for salt service are generally not suited for service in gaseous oxidizing environments, and vice versa, and welding is required for the Class I boundaries in these systems, as identified in the relevant ASME codes.
Denis Clark; Ronald Mizia
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. (RR E) A different set of alloys, such as Alloy N and 242, are needed to handle molten salts at this temperature. The diffusion welding development work described here builds on techniques developed during the NGNP work, as applied to these alloys. There is also the matter of dissimilar metal welding, since alloys suitable for salt service are generally not suited for service in gaseous oxidizing environments, and vice versa, and welding is required for the Class I boundaries in these systems, as identified in the relevant ASME codes.
Denis Clark; Ronald Mizia; Piyush Sabharwall
Electrochemical Behavior of Calcium-Bismuth Alloys in Molten Salt ...
Abstract Scope, The electrochemical properties of calcium-bismuth alloys were investigated to ... Behavior of Silicon Electrodepositing in Fluoride Molten Salts.
Molten salt bath circulation design for an electrolytic cell
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.
Dawless, Robert K. (Monroeville, PA); LaCamera, Alfred F. (Trafford, PA); Troup, R. Lee (Murrysville, PA); Ray, Siba P. (Murrysville, PA); Hosler, Robert B. (Sarver, PA)
Ion Beam Experiment to Simulate Simultaneous Molten Salt ...
Experiments to expose candidate materials to simultaneous molten salt corrosion and ion-beam damage are staged at the Ion Beam Materials Laboratory at Los�...
Molten Salt Electrolysis for the Synthesis of Elemental Boron
An alternative method using molten salt electrolysis was developed in this work. The electrolyte system evaluated was MgF2-NaF-LiF with�...
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.
Dawless, R.K.; LaCamera, A.F.; Troup, R.L.; Ray, S.P.; Hosler, R.B.
Sensor Technology for Real Time Monitoring of Molten Salt ...
Presentation Title, Sensor Technology for Real Time Monitoring of Molten Salt Electrolytes During Nuclear Fuel Electrorefining. Author(s), Michael F. Simpson,�...
Molten salts database for energy applications
The growing interest in energy applications of molten salts is justified by several of their properties. Their possibilities of usage as a coolant, heat transfer fluid or heat storage substrate, require thermo-hydrodynamic refined calculations. Many researchers are using simulation techniques, such as Computational Fluid Dynamics (CFD) for their projects or conceptual designs. The aim of this work is providing a review of basic properties (density, viscosity, thermal conductivity and heat capacity) of the most common and referred salt mixtures. After checking data, tabulated and graphical outputs are given in order to offer the most suitable available values to be used as input parameters for other calculations or simulations. The reviewed values show a general scattering in characterization, mainly in thermal properties. This disagreement suggests that, in several cases, new studies must be started (and even new measurement techniques should be developed) to obtain accurate values.
Serrano-L�pez, Roberto; Cuesta-L�pez, Santiago
Materials considerations for molten salt accelerator-based plutonium conversion systems
A Molten-Salt Reactor Program for power applications was initiated at the Oak Ridge National Laboratory in 1956. In 1965 the Molten Salt Reactor Experiment (MSRE) went critical and was successfully operated for several years. Operation of the MSRE revealed two deficiencies in the Hastelloy N alloy that had been developed specifically for molten-salt systems. The alloy embrittled at elevated temperatures as a result of exposure to thermal neutrons (radiation damage) and grain boundary embrittlement occurred in materials to fuel salt. Intergranular cracking was found to be associated with fission products, viz. tellurium. An improved Hastelloy N composition was subsequently developed that had better resistance to both of these problems. However, the discovery that fission product cracking could be significantly decreased by making the salt sufficiently reducing offers the prospect of improved compatibility with molten salts containing fission products and resistance to radiation damage in ABC applications. Recommendations are made regarding the types of corrosion tests and mechanistic studies needed to qualify materials for operation with PuF{sub 3}-containing molten salts.
DeVan, J.H.; DiStefano, J.R.; Eatherly, W.P.; Keiser, J.R.; Klueh, R.L.
A Molten-Salt Reactor Program for civilian power applications was initiated at the Oak Ridge National Laboratory in 1956. In 1965 the Molten Salt Reactor Experiment (MSRE) went critical and was successfully operated for several years. Operation of the MSRE revealed two deficiencies in the Hastelloy N alloy that had been developed specifically for molten-salt systems. The alloy embrittled at elevated temperatures as a result of exposure to thermal neutrons (radiation damage) and grain boundary embrittlement occurred in materials exposed to fuel salt. Intergranular cracking was found to be associated with fission products, viz. tellurium. An improved Hastelloy N composition was subsequently developed that had better resistance to both of these problems. However, the discovery that fission product cracking could be significantly decreased by making the salt sufficiently reducing offers the prospect of improved compatibility with molten salts containing fission products and resistance to radiation damage in ABC applications. Recommendations are made regarding the types of corrosion tests and mechanistic studies needed to qualify materials for operation with PuF3-containing molten salts.
De Van, J. H.; Di Stefano, J. R.; Eatherly, W. P.; Keiser, J. R.; Klueh, R. L. [Oak Ridge National Laboratory P.O. Box 2008 Oak Ridge, Tennessee 37831 (United States)
Use of Activated Charcoal for Rn-220 Adsorption for Operations Associated with the Uranium Deposit in the Auxiliary Charcoal Bed at the Molten Salt Reactor Experiment Facility
Measurements have been collected with the purpose of evaluating the effectiveness of activated charcoal for the removal of {sup 220}Rn from process off-gas at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory. A series of bench-scale tests were performed at superficial flow velocities of 10, 18, 24, and 33 cm s{sup -1} (20, 35, 47, and 65 ft min{sup -1}) with a continuous input concentration of {sup 220}Rn in the range of 9 x 10{sup 3} pCi L{sup -1}. In addition, two tests were performed at the MSRE facility by flowing helium through the auxiliary charcoal bed uranium deposit. These tests were performed so that the adsorptive effectiveness could be evaluated with a relatively high concentration of {sup 220}Rn. In addition to measuring the effectiveness of activated charcoal as a {sup 220}Rn adsorption media, the source term for available {sup 220}Rn and gaseous fission products was evaluated and compared to what is believed to be present in the deposit. The results indicate that only a few percent of the total {sup 220}Rn in the deposit is actually available for removal and that the relative activity of fission gases is very small when compared to {sup 220}Rn. The measurement data were then used to evaluate the expected effectiveness of a proposed charcoal adsorption bed consisting of a right circular cylinder having a diameter of 43 cm and a length of 91 cm (17 in. I.D. x 3 ft.). The majority of the measurement data predicts an overall {sup 220}Rn activity reduction factor of about 1 x 10{sup 9} for such a design; however, two measurements collected at a flow velocity of 18 cm s{sup -1} (35 ft min{sup -1}) indicated that the reduction factor could be as low as 1 x 10{sup 6}. The adsorptive capacity of the proposed trap was also evaluated to determine the expected life prior to degradation of performance. Taking a conservative vantage point during analysis, it was estimated that the adsorption effectiveness should not begin to deteriorate until a {sup 220}Rn activity on the order of 10{sup 10} Ci has been processed. It was therefore concluded that degradation of performance would most likely occur as the result of causes other than filling by radon progeny.
Coleman, R.L.
Use of Activated Charcoal for {sup 220}Rn Adsorption for Operations Associated with the Uranium Deposit in the Auxiliary Charcoal Bed at the Molten Salt Reactor Experiment Facility
Measurements have been collected with the purpose of evaluating the effectiveness of activated charcoal for the removal of {sup 220}Rn from process off-gas at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory. A series of bench-scale tests were performed at superficial flow velocities of 10, 18, 24, and 33 cm/s (20, 35, 47, and 65 ft/min) with a continuous input concentration of {sup 220}Rn in the range of 9 x 10{sup 3} pCi/L. In addition, two tests were performed at the MSRE facility by flowing helium through the auxiliary charcoal bed uranium deposit. These tests were performed so that the adsorptive effectiveness could be evaluated with a relatively high concentration of {sup 220}Rn. In addition to measuring the effectiveness of activated charcoal as a {sup 220}Rn adsorption media, the source term for available {sup 220}Rn in the deposit is actually available for removal and that the relative activity of fission gases is very small when compared to {sup 220}Rn. The measurement data were then used to evaluate the expected effectiveness of a proposed charcoal adsorption bed consisting of a right circular cylinder having a diameter of 43 cm and a length of 91 cm (17 in. I.D. x 3 ft.). The majority of the measurement data predicts an overall 220Rn activity reduction factor of about 1 x 10{sup 9} for such a design; however, two measurements collected at a flow velocity of 18 cm/s (35 ft/min) indicated that the reduction factor could be as low as 1 x 10{sup 6}. The adsorptive capacity of the proposed trap was also evaluated to determine the expected life prior to degradation of performance. Taking a conservative vantage point during analysis, it was estimated that the adsorption effectiveness should not begin to deteriorate until a {sup 220}Rn activity on the order of 10{sup 10} Ci has been processed. It was therefore concluded that degradation of performance would likely occur as the result of causes other than filling by radon progeny.
Molten salt thermal energy storage systems: salt selection
A research program aimed at the development of a molten salt thermal energy storage system commenced in June 1976. This topical report describes Work performed under Task I: Salt Selection is described. A total of 31 inorganic salts and salt mixtures, including 9 alkali and alkaline earth carbonate mixtures, were evaluated for their suitability as heat-of-fusion thermal energy storage materials at temperatures of 850 to 1000/sup 0/F. Thermophysical properties, safety hazards, corrosion, and cost of these salts were compared on a common basis. We concluded that because alkali carbonate mixtures show high thermal conductivity, low volumetric expansion on melting, low corrosivity and good stability, they are attractive as heat-of-fusion storage materials in this temperature range. A 35 wt percent Li/sub 2/CO/sub 3/-65 wt percent K/sub 2/CO/sub 3/ (50 mole percent Li/sub 2/CO/sub 3/-50 mole percent K/sub 2/CO/sub 3/) mixture was selected as a model system for further experimental work. This is a eutectoid mixture having a heat of fusion of 148 Btu/lb (82 cal/g) that forms an equimolar compound, LiKCO/sub 3/. The Li/sub 2/CO/sub 3/-K/sub 2/CO/sub 3/ mixture is intended to serve as a model system to define heat transfer characteristics, potential problems, and to provide ''first-cut'' engineering data required for the prototype system. The cost of a thermal energy storage system containing this mixture cannot be predicted until system characteristics are better defined. However, our comparison of different salts indicated that alkali and alkaline earth chlorides may be more attractive from a salt cost point of view. The long-term corrosion characteristics and the effects of volume change on melting for the chlorides should be investigated to determine their overall suitability as a heat-of-fusion storage medium.
Maru, H.C.; Dullea, J.F.; Huang, V.S.
Molten salt electrolyte battery cell with overcharge tolerance
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.
Kaun, Thomas D. (New Lenox, IL); Nelson, Paul A. (Wheaton, IL)
Modeling of Molten Salt Mixtures: Thermodynamic Assessment of ...
Presentation Title, Modeling of Molten Salt Mixtures: Thermodynamic Assessment of CeBr3 and MBr-CeBr3 Systems (M=Li, Na, K, Rb). Author(s), Yue Wu,�...
An experimental test plan for the characterization of molten salt thermochemical properties in heat transport systems
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 same project [1]. However, this work focuses on two materials: the LiF-BeF2 eutectic (67 and 33 mol%, respectively, also known as flibe) as primary coolant and the LiF-NaF-KF eutectic (46.5, 11.5, and 52 mol%, respectively, also known as flinak) as secondary heat transport fluid. At first common issues are identified, involving the preparation and purification of the materials as well as the development of suitable diagnostics. Than issues specific to each material and its application are considered, with focus on the compatibility with structural materials and the extension of the existing properties database.
Pattrick Calderoni
Molten Salt Solar-Electric Experiment: Volumes 1 and 2
The Molten Salt Electric Experiment assembled and tested the first full-system experiment of a solar central receiver plant employing molten nitrate salt as the heat transport fluid and thermal storage medium. This report focuses on the last two phases of the project: testing/operation and evaluation. Overall project data will help utilities evaluate the central receiver concept's technical status, development requirements, and potential as a renewable source of electricity.
Treatment of plutonium process residues by molten salt oxidation
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.
Stimmel, J.; Wishau, R.; Ramsey, K.B.; Montoya, A.; Brock, J. [Los Alamos National Lab., NM (United States); Heslop, M. [Naval Surface Warfare Center (United States). Indian Head Div.; Wernly, K. [Molten Salt Oxidation Corp. (United States)
LIFE Materails: Molten-Salt Fuels Volume 8
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.
Moir, R; Brown, N; Caro, A; Farmer, J; Halsey, W; Kaufman, L; Kramer, K; Latkowski, J; Powers, J; Shaw, H; Turchi, P
Enhanced molten salt purification by electrochemical methods: feasibility experiments with flibe
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-chemistry properties in heat transport systems� describes the options available to reach such objectives and contains extended references to published work. The report highlights how electrochemical methods are the most promising techniques for the development of instrumentation aimed at the measurement of melts composition and for enhanced purification systems. The purpose of this work is to summarize preliminary experimental activities performed at the INL Safety and Tritium Applied Research facility in support of the development of electrochemistry based instrumentation and purification systems. The experiments have been focused on the LiF-BeF2 eutectic (67 and 33 mol%, respectively), also known as flibe.
Alan K Wertsching; Brandon S Grover; Pattrick Calderoni
New Opportunities for Metals Extraction and Waste Treatment by Electrochemical Processing in Molten Salts
Molten salt electrolysis is a proven technology for the extraction of metals -- all the world's primary aluminum is produced in this manner. The unique properties of molten salts also make them
Sadoway, Donald R.
Assessment of the Use of Nitrogen Trifluoride for Purifying Coolant and Heat Transfer Salts in the Fluoride Salt-Cooled High-Temperature Reactor
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.
Scheele, Randall D.; Casella, Andrew M.
Integrated demonstration of molten salt oxidation with salt recycle for mixed waste treatment
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.
Hsu, P.C.
Rapid Removal of Chlorine in Molten Salt Electrolysis of Magnesium ...
However, experimental data and modeling results in this study indicate that the ... bubbles on the current efficiency and the cell potential were investigated. ... High- Chloride Circuit for the Starfield Resources' Ferguson Lake Project � Direct Synthesis of Niobium Aluminides Powders by Sodiothermic Reduction in Molten Salts.
Development of the Process for the Recovery and Conversion of {sup 233}UF{sub 6} Chemisorbed in NaF Traps from the Molten Salt Reactor Remediation Project
The Molten Salt Reactor Experiment (MSRE) site at Oak Ridge National Laboratory is being cleaned up and remediated. The removal of {approx}37 kg of fissile {sup 233}U is the main activity. Of that inventory, {approx}23 kg has already been removed as UF{sub 6} from the piping system and chemisorbed in 25 NaF traps. This material is in temporary storage while it awaits conversion to a stable oxide. 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 uranium oxide (U{sub 3}O{sub 8}), which is suitable for long-term storage.The conversion of the MSRE material into an oxide presents unique problems, such as criticality concerns, a large radiation field caused by the daughters of {sup 232}U (an impurity isotope in the {sup 233}U), and the possible spread of the high-radiation field from the release of {sup 220}Rn gas. To overcome these problems, a novel process was conceived and developed. This process was specially tailored for providing remote operations inside a hot cell while maintaining full containment at all times to avoid the spread of contamination. This process satisfies criticality concerns, maximizes the recovery of uranium, minimizes any radiation exposure to operators, and keeps waste disposal to a minimum.
Cul, Guillermo D. del; Icenhour, Alan S.; Simmons, Darrell W. [Oak Ridge National Laboratory (United States)
Molten Salt Fuel Version of Laser Inertial Fusion Fission Energy (LIFE)
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}0.6 atoms per 14.1 MeV neutron.
Moir, R W; Shaw, H F; Caro, A; Kaufman, L; Latkowski, J F; Powers, J; Turchi, P A
Corrosion of aluminides by molten nitrate salt
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.
Tortorelli, P.F.; Bishop, P.S.
Industrial use of molten nitrate/nitrite salts
Nitrate salts have been used for years as a high-temperature heat transfer medium in the chemical and metal industries. This experience is often cited as an argument for the use of these salts in large-scale solar energy systems. However, this industrial experience has not been well documented and a study was carried out to provide such information to the solar community and to determine the applicability of this data base. Seven different industrial plants were visited and the plant operators were interviewed with regard to operating history and experience. In all cases the molten salt systems operate without problems. However, it is not possible to apply the base of industrial experience directly to solar thermal energy applications because of differences in operating temperature, salt composition, alloys used, and thermal/mechanical conditions.
Carling, R.W.; Mar, R.W.
Technical review of Molten Salt Oxidation
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.
Preliminary Neutronics Design Studies for a Molten Salt Blanket LIFE Engine
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. Preliminary design studies looking at fast ignition and hot spot ignition fusion options are documented, along with limited scoping studies performed to investigate other options of interest that surfaced during the main design effort. Lastly, side studies that were not part of the main design effort but may alter future work performed on LIFE engine designs are shown. The majority of all work reported in this document was performed during the Molten Salt Fast Ignition Moderator Study (MSFIMS) which sought to optimize the amount of moderator mixed into the molten salt region in order to produce the most compelling design. The studies in this report are of a limited scope and are intended to provide a preliminary neutronics analysis of the design concepts described herein to help guide decision processes and explore various options that a LIFE engine with a molten salt blanket might enable. None of the designs shown in this report, even reference cases selected for detailed description and analysis, have been fully optimized. The analyses were performed primarily as a neutronics study, though some consultation was made regarding thermal-hydraulic and structural concerns during both scoping out an initial model and subsequent to identifying a neutronics-based reference case to ensure that the design work contained no glaring mechanical or thermal issues that would preclude its feasibility. Any analyses and recommendations made in this report are either primarily or solely from the point of view of LIFE neutronics and ignore other fundamental issues related to molten salt fuel blankets such as chemical processing feasibility and political feasibility of a molten salt system.
Powers, J
SunShot Initiative: Modular and Scalable Baseload Molten Salt Plant
NLE Websites -- All DOE Office Websites (Extended Search)
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Oxidation of hydrogen halides to elemental halogens with catalytic molten salt mixtures
A process for oxidizing hydrogen halides by means of a catalytically active molten salt is disclosed. The subject hydrogen halide is contacted with a molten salt containing an oxygen compound of vanadium and alkali metal sulfates and pyrosulfates to produce an effluent gas stream rich in the elemental halogen. The reduced vanadium which remains after this contacting is regenerated to the active higher valence state by contacting the spent molten salt with a stream of oxygen-bearing gas.
Rohrmann, Charles A. (Kennewick, WA)
Multi-Physics Modeling of Molten Salt Transport in Solid Oxide ...
In both processes, electrolysis and/or electrorefining take place in the crucible, where raw material is continuously fed into the molten salt electrolyte, producing
NaNO3-KNO3 Ternary Molten Salts for Parabolic Trough
Presentation Title, Thermodynamic Properties of Novel Low Melting Point LiNO3- NaNO3-KNO3 Ternary Molten Salts for Parabolic Trough Solar Power�...
Thermodynamic Assessment of Hot Corrosion Mechanisms of Superalloys Hastelloy N and Haynes 242 in Eutectic Mixture of Molten Salts KF and ZrF4
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.
Michael V. Glazoff
Molten Salt Test Loop (MSTL) system customer interface document.
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.
Gill, David Dennis; Kolb, William J.; Briggs, Ronald D.
Method for removal of heavy metal from molten salt in IFR fuel pyroprocessing
This report details the pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor (IFR) which involves electrorefining spent fuel in a molten salt electrolyte (LiCl-KCI-U/PuCl{sub 3}) at 500{degree}C. The total heavy metal chloride concentration in the salt will be about 2 mol %. At some point, the concentrations of alkali, alkaline earth, and rare earth fission products in the salt must be reduced to lower the amount of heat generated in the electrorefiner. The heavy metal concentration in the salt must be reduced before removing the fission products from the salt. The operation uses a lithium-cadmium alloy anode that is solid at 500{degree}C, a solid mandrel cathode with a ceramic catch crucible below to collect heavy metal that falls off it, and a liquid cadmium cathode. The design criteria that had to be met by this equipment included the following: (1) control of the reduction rate by lithium, (2) good separation between heavy metal and rare earths, and (3) the capability to collect heavy metal and rare earths over a wide range of salt compositions. In tests conducted in an engineering-scale electrorefiner (10 kg uranium per cathode), good separation was achieved while removing uranium and rare earths from the salt. Only 13% of the rare earths was removed, while 99.9% of the uranium in the salt was removed; subsequently, the rare earths were also reduced to low concentrations. The uranium concentration in the salt was reduced to 0.05 ppm after uranium and rare earths were transferred from the salt to a solid mandrel cathode with a catch crucible. Rare earth concentrations in the salt were reduced to less than 0.01 wt % in these operations. Similar tests are planned to remove plutonium from the salt in a laboratory-scale (100--300 g heavy metal) electrorefiner.
Gay, E.C.; Miller, W.E.; Laidler, J.J.
Molten metal reactor and method of forming hydrogen, carbon monoxide and carbon dioxide using the molten alkaline metal reactor
Bingham, Dennis N.; Klingler, Kerry M.; Turner, Terry D.; Wilding, Bruce M.
Method of removal of heavy metal from molten salt in IFR fuel pyroprocessing
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.
Gay, E.C.
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.
Gay, Eddie C. (Park Forest, IL)
Sulfide ceramics in molten-salt electrolyte batteries
Sulfide ceramics are finding application in the manufacture of advanced batteries with molten salt electrolyte. Use of these ceramics as a peripheral seal component has permitted development of bipolar Li/FeS{sub 2} batteries. This bipolar battery has a molten lithium halide electrolyte and operates at 400 to 450C. Initial development and physical properties evaluations indicate the ability to form metal/ceramic bonded seal (13-cm ID) components for use in high-temperature corrosive environments. These sealants are generally CaAl{sub 2}S{sub 4}-based ceramics. Structural ceramics (composites with oxide or nitride fillers), highly wetting sealant formulations, and protective coatings are also being developed. Sulfide ceramics show great promise because of their relatively low melting point, high-temperature viscous flow, chemical stability, high-strength bonding, and tailored coefficients of thermal expansion. Our methodology of generating laminated metal/ceramic pellets (e.g., molybdenum/sulfide ceramic/molybdenum) with which to optimize materials formulation and seal processing is described.
Kaun, T.D.; Hash, M.C.; Simon, D.R.
Solar two: A molten salt power tower demonstration
A consortium of United States utility concerns led by the Southern California Edison Company (SCE) is conducting a cooperative project with the US Department of Energy (DOE), Sandia National Laboratories, and industry to convert the 10-MW Solar One Power Tower Pilot Plant to molten nitrate salt technology. The conversion involves installation of a new receiver, a new thermal storage system, and a new steam generator; it utilizes Solar One`s heliostat field and turbine generator. Successful operation of the converted plant, called Solar Two, will reduce economic risks in building initial commercial power tow projects and accelerate the commercial acceptance of this promising renewable energy technology. The estimated cost of Solar Two, including its three-year test period, is $48.5 million. The plant will begin operation in early 1996.
Tyner, C.E. [Sandia National Labs., Albuquerque, NM (United States); Sutherland, J.P. [Southern California Edison, Rosemead, CA (United States); Gould, W.R. Jr. [Bechtel Corp., San Francisco, CA (United States)
Advanced Thermal Storage System with Novel Molten Salt: December 8, 2011 - April 30, 2013
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.
Jonemann, M.
Plutonium and americium recovery from spent molten-salt-extraction salts with aluminum-magnesium alloys
Development work was performed to determine the feasibility of removing plutonium and americium from spent molten-salt-extraction (MSE) salts using Al-Mg alloys. If the product buttons from this process are compatible with subsequent aqueous processing, the complex chloride-to-nitrate aqueous conversion step which is presently required for these salts may be eliminated. The optimum alloy composition used to treat spent 8 wt % MSE salts in the past yielded poor phase-disengagement characteristics when applied to 30 mol % salts. After a limited investigation of other alloy compositions in the Al-Mg-Pu-Am system, it was determined that the Al-Pu-Am system could yield a compatible alloy. In this system, experiments were performed to investigate the effects of plutonium loading in the alloy, excess magnesium, age of the spent salt on actinide recovery, phase disengagement, and button homogeneity. Experimental results indicate that 95 percent plutonium recoveries can be attained for fresh salts. Further development is required for backlog salts generated prior to 1981. A homogeneous product alloy, as required for aqueous processing, could not be produced.
Cusick, M.J.; Sherwood, W.G.; Fitzpatrick, R.F.
Modeling Solute Thermokinetics in LiCI-KCI Molten Salt for Nuclear Waste Separation
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 at the eutectic composition (58 mol% LiCl, 42 mol% KCl), which is used for treating spent EBR-II fuel. The same process being used for EBRII fuel is currently being studied for widespread international implementation. The methods will focus on first-principles and first- principles derived interatomic potential based simulations, primarily using molecular dynamics. Results will be validated against existing literature and parallel ongoing experimental efforts. The simulation results will be of value for interpreting experimental results, validating analytical models, and for optimizing waste separation by potentially developing new salt configurations and operating conditions.
Morgan, Dane; Eapen, Jacob
Application of molten salts in pyrochemical processing of reactive metals
Various mixes of chloride and fluoride salts are used as the media for conducting pyrochemical processes in the production and purification of reactive metals. These processes generate a significant amount of contaminated waste that has to be treated for recycling or disposal. Molten calcium chloride based salt systems have been used in this work to electrolytically regenerate calcium metal from calcium oxide for the in situ reduction of reactive metal oxides. The recovery of calcium is characterized by the process efficiency to overcome back reactions in the electrowinning cell. A thermodynamic analysis, based on fundamental rate theory, has been performed to understand the process parameters controlling the metal deposition, rate, behavior of the ceramic anode-sheath and influence of the back-reactions. It has been observed that the deposition of calcium is dependent on the ionic diffusion through the sheath. It has also been evidenced that the recovered calcium is completely lost through the back-reactions in the absence of a sheath. A practical scenario has also been presented where the electrowon metal can be used in situ as a reductant to reduce another reactive metal oxide.
Mishra, B.; Olson, D.L. (Colorado School of Mines, Golden, CO (United States). Kroll Inst. for Extractive Metallurgy); Averill, W.A. (EG and G Rocky Flats, Inc., Golden, CO (United States). Rocky Flats Plant)
This report supersedes K-1-3425 and LA-2327(Prelim). The first experiment (LAMPRE I) in a program to develop molten plutonium fuels for fast reactors is described and the hazards associated with reactor operation are discussed and evaluated. The reactor desc=iption includes fuel element design, core configuration, sodium coolant system control, safety systems, fuel capsule charger, cover gas system, and shielding. Information of the site comprises population in surrounding areas, meteorological data, geology, and details of the reactor building. The hazmalfunction of the several elements comprising the reactor system. A calculation on the effect of fuel element bowiing appears in an appendix. (auth)
Swickard, E.O. comp.
Overview on Use of a Molten Salt HTF in a Trough Solar Field (Presentation)
This presentation discusses the utilization of molten salt as the heat transfer fluid in a parabolic trough solar field to improve system performance and to reduce the levelized electricity.
Kearney, D.; Kelly, B.; Cable, R.; Potrovitza, N.; Herrmann, U.; Nava, P.; Mahoney, R.; Pacheco, J.; Blake, D.; Price, H.
Method for converting UF5 to UF4 in a molten fluoride salt
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.
Bennett, Melvin R. (Oak Ridge, TN); Bamberger, Carlos E. (Oak Ridge, TN); Kelmers, A. Donald (Oak Ridge, TN)
The design and testing of a molten salt steam generator for solar application
This paper describes the design and testing of the Steam Generator Subsystem (SGS) for the Molten Salt Electric Experiment at Sandia Laboratories in Albuquerque, New Mexico. The Molten Salt Electric Experiment (MSEE) has been established to demonstrate the feasibility of the molten salt central receiver concept. The experiment is capable of generating 0.75 megawatts of electric power from solar energy, with the capability of storing seven megawatt-hours of thermal energy. The steam generator subsystem transfers sensible heat from the solar-heated molten nitrate salt to produce steam to drive a conventional turbine. This paper discusses the design requirements dictated by the steam generator application and also reviews the process conditions. Details of each of the SGS components are given, featuring the aspects of the design and performance unique to the solar application. The paper concludes with a summary of the test results confirming the overall design of the subsystem.
Allman, W.A.; Smith, D.C.; Kakarala, C.R.
Design and validation of an air window for a molten salt solar thermal receiver
This thesis contributes to the development of Concentrating Solar Power (CSP) receivers and focuses on the design of an efficient aperture. An air window is proposed for use as the aperture of a CSP molten salt receiver ...
Paxson, Adam Taylor
Luminescent properties of Y2O3:Eu3+ nanocrystals prepared by molten salt synthesis
A series of red phosphors Y2O3:Eu3+ were prepared by the molten salt method with different surfactants. Their structures, morphologies, and the photoluminescent properties were investigated at room temperature. The particles ...
Lijun Luo, Fenfen Hu, Li Xiong, Xiaofan Li, Meili Zhou, Zhengliang Wang
Molten salt synthesis and localized surface plasmon resonance study of vanadium dioxide nanopowders
Rutile-type vanadium dioxide nanopowders with four different sizes were successfully synthesized by carbothermal reducing V{sub 2}O{sub 5} in KCl-LiCl molten salt. XRD and TEM characterizations suggested that vanadium dioxide particles formed by a broken and reunited process of vanadium oxide. Molten salt and organic carbon sources are crucial to the size of final particles. In the presence of the molten salt, the organic carbon with a shorter chain length would induce smaller particles. The UV-VIS-IR spectral measurements for as-prepared vanadium dioxide announced an obvious localized surface plasmon resonance band in the near infrared region at 90 deg. C. - Graphical abstract: Schematic illustration of the formation mechanism of VO{sub 2}(M) nanoparticles in molten salt, particles size can be controlled by choosing organic carbon sources with different chain length.
Wang Fu [Key Laboratory of Photochemical Conversion and Optoelectronic Materials of Technical Institute of Physics and Chemistry, Chinese Academy of Sciences, Zhongguancun, Beijing 100190 (China); Graduate School of the Chinese Academy of Sciences, Beijing 100806 (China); Liu Yun [Key Laboratory of Photochemical Conversion and Optoelectronic Materials of Technical Institute of Physics and Chemistry, Chinese Academy of Sciences, Zhongguancun, Beijing 100190 (China); Liu Chunyan, E-mail: email@example.com [Key Laboratory of Photochemical Conversion and Optoelectronic Materials of Technical Institute of Physics and Chemistry, Chinese Academy of Sciences, Zhongguancun, Beijing 100190 (China)
Molten Salt Multi-anode Reactive Alloy Coating(MARC) of Ta-W ...
In this study, Ta-W coated samples (Ta-7.31W, Ta-4.12W and Ta-1.92W) were prepared by multi-anode reactive alloy coating (MARC) process in molten salt�...
Recovery of plutonium from molten salt extraction residues
Savannah River Laboratory (SRL), Savannah River Plant (SRP), and Rocky Flats Plant (RFP) are jointly developing a process to recover plutonium from molten salt extraction residues. These NaCl, KCl, MgCl/sub 2/ residues, which are generated in the pyrochemical extraction of /sup 241/Am from aged plutonium metal, contain up to 25 wt % dissolved PUCl/sub 3/ and up to 2 wt % AmCl/sub 3/. The objective is to develop a process to convert these residues to plutonium metal product and discardable waste. The first step of the conceptual process is to convert the actinides to a heterogenous scrub alloy with aluminum and magnesium. This step, performed at RFP, effectively separates the actinides from the bulk of the chloride. This scrub alloy will then be dissolved in a HNO/sub 3/-HF solution at SRP. Residual chloride will be removed by precipitation with Hg/sub 2/(NO/sub 3/)/sub 2/ followed by centrifugation. Plutonium and americium will be separated using the Purex solvent extraction process. The /sup 241/Am will be diverted to the solvent extraction waste stream where it can either be discarded to the waste farm or recovered. The plutonium will be finished via PuF/sub 3/ precipitation, oxidation to a mixture of PUF/sub 4/ and PuO/sub 2/, followed by reduction to plutonium metal with calcium.
Gray, L.W.; Holcomb, H.P.
Molten Metal Treatment by Salt Fluxing with Low Environmental Emissions
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 conservation. In this program, the salt metal interactions were studies and the emissions at laboratory scale at OSU were monitored. The goal of the project was to obtain a fundamental understanding, based on first principles, of the pollutant formation that occurs when the salts are used in furnaces. This information will be used to control process parameters so that emissions are consistently below the required levels. The information obtained in these experiments will be used in industrial furnaces at aluminum plants and which will help in optimizing the process.
Yogeshwar Sahai
Convective heat transfer in the laminar-turbulent transition region with molten salt in a circular tube
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)
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)
Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes
This paper presents the technical and economic feasibility of molten salt oxidation technology as a volume reduction and recovery process for {sup 238}Pu contaminated waste. Combustible low-level waste material contaminated with {sup 238}Pu residue is destroyed by oxidation in a 900 C molten salt reaction vessel. The combustible waste is destroyed creating carbon dioxide and steam and a small amount of ash and insoluble {sup 2328}Pu in the spent salt. The valuable {sup 238}Pu is recycled using aqueous recovery techniques. Experimental test results for this technology indicate a plutonium recovery efficiency of 99%. Molten salt oxidation stabilizes the waste converting it to a non-combustible waste. Thus installation and use of molten salt oxidation technology will substantially reduce the volume of {sup 238}Pu contaminated waste. Cost-effectiveness evaluations of molten salt oxidation indicate a significant cost savings when compared to the present plans to package, or re-package, certify and transport these wastes to the Waste Isolation Pilot Plant for permanent disposal. Clear and distinct cost advantages exist for MSO when the monetary value of the recovered {sup 238}Pu is considered.
Wishau, R.; Ramsey, K.B.; Montoya, A.
Selective Adsorption of Sodium Aluminum Fluoride Salts from Molten Aluminum
Aluminum is produced in electrolytic reduction cells where alumina feedstock is dissolved in molten cryolite (sodium aluminum fluoride) along with aluminum and calcium fluorides. The dissolved alumina is then reduced by electrolysis and the molten aluminum separates to the bottom of the cell. The reduction cell is periodically tapped to remove the molten aluminum. During the tapping process, some of the molten electrolyte (commonly referred as �bath� in the aluminum industry) is carried over with the molten aluminum and into the transfer crucible. The carryover of molten bath into the holding furnace can create significant operational problems in aluminum cast houses. Bath carryover can result in several problems. The most troublesome problem is sodium and calcium pickup in magnesium-bearing alloys. Magnesium alloying additions can result in Mg-Na and Mg-Ca exchange reactions with the molten bath, which results in the undesirable pickup of elemental sodium and calcium. This final report presents the findings of a project to evaluate removal of molten bath using a new and novel micro-porous filter media. The theory of selective adsorption or removal is based on interfacial surface energy differences of molten aluminum and bath on the micro-porous filter structure. This report describes the theory of the selective adsorption-filtration process, the development of suitable micro-porous filter media, and the operational results obtained with a micro-porous bed filtration system. The micro-porous filter media was found to very effectively remove molten sodium aluminum fluoride bath by the selective adsorption-filtration mechanism.
Leonard S. Aubrey; Christine A. Boyle; Eddie M. Williams; David H. DeYoung; Dawid D. Smith; Feng Chi
Molten Salt-Carbon Nanotube Thermal Energy Storage for Concentrating Solar Power Systems
Office of Scientific and Technical Information (OSTI)
Molten Salt-Carbon Nanotube Thermal Energy Storage for Concentrating Solar Power Systems Molten Salt-Carbon Nanotube Thermal Energy Storage for Concentrating Solar Power Systems Final Report March 31, 2012 Michael Schuller, Frank Little, Darren Malik, Matt Betts, Qian Shao, Jun Luo, Wan Zhong, Sandhya Shankar, Ashwin Padmanaban The Space Engineering Research Center Texas Engineering Experiment Station Texas A&M University Abstract 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,
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.
Molten Salt Heat Transport Loop: Materials Corrosion and Heat Transfer Phenomena
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.
Dr. Kumar Sridharan; Dr. Mark Anderson; Dr. Michael Corradini; Dr. Todd Allen; Luke Olson; James Ambrosek; Daniel Ludwig
Thorium-Fueled Underground Power Plant Based on Molten Salt Technology
This paper addresses the problems posed by running out of oil and gas supplies and the environmental problems that are due to greenhouse gases by suggesting the use of the energy available in the resource thorium, which is much more plentiful than the conventional nuclear fuel uranium. We propose the burning of this thorium dissolved as a fluoride in molten salt in the minimum viscosity mixture of LiF and BeF{sub 2} together with a small amount of {sup 235}U or plutonium fluoride to initiate the process to be located at least 10 m underground. The fission products could be stored at the same underground location. With graphite replacement or new cores and with the liquid fuel transferred to the new cores periodically, the power plant could operate for up to 200 yr with no transport of fissile material to the reactor or of wastes from the reactor during this period. Advantages that include utilization of an abundant fuel, inaccessibility of that fuel to terrorists or for diversion to weapons use, together with good economics and safety features such as an underground location will diminish public concerns. We call for the construction of a small prototype thorium-burning reactor.
Moir, Ralph W.; Teller, Edward [Lawrence Livermore National Laboratory (United States)
The Molten Salt Electrolytic Winning of Oxygen and Metal from ...
Cathodic Behavior of Silicon (?) in BaF2-CaF2 �SiO2 Melts ... Electrochemical Impedance Spectroscopy of Uranium Chloride in Molten LiCl-KCl Eutectic.
Electrolytic Production of Metals from Oxides Dissolved in Molten Salts
Potentiometric Sensor for Real-Time Monitoring of Multivalent Ion Concentrations in Molten Salt
Electrorefining of spent metallic nuclear fuel in high temperature molten salt systems is a core technology in pyroprocessing, which in turn plays a critical role in the development of advanced fuel cycle technologies. In electrorefining, spent nuclear fuel is treated electrochemically in order to effect separations between uranium, noble metals, and active metals, which include the transuranics. The accumulation of active metals in a lithium chloride-potassium chloride (LiCl-KCl) eutectic molten salt electrolyte occurs at the expense of the UCl3-oxidant concentration in the electrolyte, which must be periodically replenished. Our interests lie with the accumulation of active metals in the molten salt electrolyte. The real-time monitoring of actinide concentrations in the molten salt electrolyte is highly desirable for controlling electrochemical operations and assuring materials control and accountancy. However, real-time monitoring is not possible with current methods for sampling and chemical analysis. A new solid-state electrochemical sensor is being developed for real-time monitoring of actinide ion concentrations in a molten salt electrorefiner. The ultimate function of the sensor is to monitor plutonium concentrations during electrorefining operations, but in this work gadolinium was employed as a surrogate material for plutonium. In a parametric study, polycrystalline sodium beta double-prime alumina (Na-�?-alumina) discs and tubes were subject to vapor-phase exchange with gadolinium ions (Gd3+) using a gadolinium chloride salt (GdCl3) as a precursor to produce gadolinium beta double-prime alumina (Gd-�?-alumina) samples. Electrochemical impedance spectroscopy and microstructural analysis were performed on the ion-exchanged discs to determine the relationship between ion exchange and Gd3+ ion conductivity. The ion-exchanged tubes were configured as potentiometric sensors in order to monitor real-time Gd3+ ion concentrations in mixtures of gadolinium chloride (GdCl3) in LiCl-KCl eutectic molten salts through measurement of the potential difference between a reference and working electrode.
Peter A. Zink; Jan-Fong Jue; Brenda E. Serrano; Guy L. Fredrickson; Ben F. Cowan; Steven D. Herrmann; Shelly X. Li
Process for recovering tritium from molten lithium metal
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.
Maroni, Victor A. (Naperville, IL)
Thermal Properties of LiCl-KCl Molten Salt for Nuclear Waste Separation
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°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.
Kumar Sridharan; Todd Allen; Mark Anderson
Molecular Dynamics Simulation of the Transport Properties of Molten Transuranic Chloride Salts
The Accelerator Research Laboratory at Texas A&M is proposing a design for accelerator-driven subcritical fission in molten salt (ADSMS), a system that destroys the transuranic elements in used nuclear fuel. The transuranics (TRU) are the most enduring hazard of nuclear power. TRU contain high radiotoxicity and have half-lives of a thousand to a million years. The ADSMS core is fueled by a homogeneous chloride-based molten salt mixture containing TRUCl3 and NaCl. Certain thermodynamic properties are critical to modeling both the neutronics and heat transfer of an ADSMS system. There is a lack of experimental data on the density, heat capacity, electrical and thermal conductivities, and viscosity of TRUCl3 salt systems. Molecular dynamics simulations using a polarizable ion model (PIM) are employed to determine the density and heat capacity of these melts as a function of temperature. Green-Kubo methods are implemented to calculate the electrical conductivity, thermal conductivity, and viscosity of the salt using the outputs of the simulations. Results for pure molten salt systems are compared to experimental data when possible to validate the potentials used. Here I discuss chloride salt systems of interest, their calculated properties, and possible sources of error for our simulations.
Baty, Austin Alan
Oxygen production by molten alkali metal salts using multiple absorption-desorption cycles
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.
Cassano, Anthony A. (Allentown, PA)
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.
Cassano, A.A.
Solar Two: A Molten Salt Power Tower Demonstration* Craig E.Tyner
Solar Two: A Molten Salt Power Tower Demonstration* Craig E.Tyner Sandia National Laboratories.S. Department of Energy (DOE),Sandia National Laboratories, and industry to convert the 10-MwSolar One Power, is $48.5 million. The plant will begin operation in early 1996. Introduction A solar power tower plant
Laughlin, Robert B.
Thermal Analysis of Surrogate Simulated Molten Salts with Metal Chloride Impurities for Electrorefining Used Nuclear Fuel
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 previously known the liquidus temperature of the molten salt would change as spent fuel is processed through the Mk-IV electrorefiner. However, the extent of the increase in liquidus temperature was not known. This work is first of its kind in determining thermodynamic properties of a molten salt electrolyte containing transuranics, fission products and bond sodium. Experimental data concluded that the melting temperature of the electrolyte will become greater than the operating temperature of the Mk-IV ER during current fuel processing campaigns. Collected data also helps predict when the molten salt electrolyte will no longer be able to support electrorefining operations.
Toni Y. Gutknecht; Guy L. Fredrickson; Vivek Utgikar
Uncertainty Studies of Real Anode Surface Area in Computational Analysis for Molten Salt Electrorefining
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.
Sungyeol Choi; Jaeyeong Park; Robert O. Hoover; Supathorn Phongikaroon; Michael F. Simpson; Kwang-Rag Kim; Il Soon Hwang
Porous membrane electrochemical cell for uranium and transuranic recovery from molten salt electrolyte
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.
Willit, James L. (Batavia, IL)
Willit, James L. (Ratavia, IL)
Assessment of molten-salt solar central-receiver freeze-up and recovery events
Molten salt used as a heat transfer fluid in central-receiver so ar power plants has a high freezing point (430{degrees}F (221{degrees}C)). It is very likely during the life of the plant that the receiver will accidentally freeze up due to equipment malfunction or operator error. Experiments were conducted to measure the effects of a molten salt receiver freeze-up and recovery event and methods to thaw the receiver. In addition, simulated freeze/thaw experiments were conducted to determine what happens when salt freezes and is thawed in receiver tubes and to quantify the damage caused to candidate receiver tube materials. Fourteen tube samples of various materials, diameters and wall thicknesses were tested to destruction. Results of these tests are presented in this paper.
Pacheco, J.E.; Dunkin, S.R.
Materials Corrosion in Molten Fluoride Salts - Programmaster.org
About this Abstract. Meeting, 2013 TMS Annual Meeting & Exhibition. Symposium , Materials and Fuels for the Current and Advanced Nuclear Reactors II.
Molten Salt Nanomaterials for Thermal Energy Storage and Concentrated Solar Power Applications
The thermal efficiency of concentrated solar power (CSP) system depends on the maximum operating temperature of the system which is determined by the operating temperature of the TES device. Organic materials (such as synthetic oil, fatty acid, or paraffin wax) are typically used for TES. This limits the operating temperature of CSP units to below 400 degrees C. Increasing the operating temperature to 560 degrees C (i.e., the creeping temperature of stainless steel), can enhance the theoretical thermal efficiency from 54 percent to 63 percent. However, very few thermal storage materials are compatible for these high temperatures. Molten salts are thermally stable up to 600 degrees C and beyond. Using the molten salts as the TES materials confers several benefits, which include: (1) Higher operating temperature can significantly increase the overall cycle efficiency and resulting costs of power production. (2) Low cost of the molten salt materials can drastically reduce the cost. (3) The molten salts, which are environmentally safe, can also reduce the potential environmental impact. However, these materials suffer from poor thermo-physical properties. Impregnating these materials with nanoparticles can enhance these properties. Solvents doped with nanoparticles are termed as nanofluids. Nanofluids have been reported in the literature for the anomalous enhancement of their thermo-physical properties. In this study, the poor thermal properties of the molten salts were enhanced dramatically on mixing with nanoparticles. For example the specific heat capacity of these molten salt eutectics was found to be enhanced by as much as ~ 26 percent on mixing with nanoparticles at a mass fraction of ~ 1 percent. The resultant properties of these nanomaterials were found to be highly sensitive to small variations in the synthesis protocols. Computational models were also developed in this study to explore the fundamental transport mechanisms on the molecular scale for elucidating the anomalous enhancements in the thermo-physical properties that were measured in these experiments. This study is applicable for thermal energy storage systems utilized for other energy conversion technologies � such as geothermal energy, nuclear energy and a combination of energy generation technologies.
Shin, Donghyun
Expedited demonstration of molten salt mixed waste treatment technology. Final report
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.
An evaluation of possible next-generation high temperature molten-salt power towers.
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.
Kolb, Gregory J.
Effect of chloride content of molten nitrate salt on corrosion of A516 carbon steel.
The corrosion behavior of A516 carbon steel was evaluated to determine the effect of the dissolved chloride content in molten binary Solar Salt. Corrosion tests were conducted in a molten salt consisting of a 60-40 weight ratio of NaNO{sub 3} and KNO{sub 3} at 400{sup o}C and 450{sup o}C for up to 800 hours. Chloride concentrations of 0, 0.5 and 1.0 wt.% were investigated to determine the effect on corrosion of this impurity, which can be present in comparable amounts in commercial grades of the constituent salts. Corrosion rates were determined by descaled weight losses, corrosion morphology was examined by metallographic sectioning, and the types of corrosion products were determined by x-ray diffraction. Corrosion proceeded by uniform surface scaling and no pitting or intergranular corrosion was observed. Corrosion rates increased significantly as the concentration of dissolved chloride in the molten salt increased. The adherence of surface scales, and thus their protective properties, was degraded by dissolved chloride, fostering more rapid corrosion. Magnetite was the only corrosion product formed on the carbon steel specimens, regardless of chloride content or temperature.
Bradshaw, Robert W.; Clift, W. Miles
Novel Ternary Molten Salt Electrolytes for intermediate-temperature sodium/nickel chloride batteries
The sodium-nickel chloride (ZEBRA) battery is typically operated at relatively high temperature (250~350�C) to achieve adequate electrochemical performance. Reducing the operating temperature in the range of 150 to 200�C can lead to enhanced cycle life by suppressing temperature related degradation mechanisms. The reduced temperature range also allows for lower cost materials of construction such as elastomeric sealants and gaskets. To achieve adequate electrochemical performance at lower operating temperatures requires an overall reduction in ohmic losses associated with temperature. This includes reducing the ohmic resistance of ?�-alumina solid electrolyte (BASE) and the incorporation of low melting point molten salt as the secondary electrolyte. In present work, planar-type Na/NiCl2 cells with a thin flat plate BASE (600 ?m) and low melting point secondary electrolyte were evaluated at reduced temperatures. Molten salt formulation for use as secondary electrolytes were fabricated by the partial replace of NaCl in the standard secondary electrolyte (NaAlCl4) with other lower melting point alkali metal salts such as NaBr, LiCl, and LiBr. Electrochemical characterization of the ternary molten salts demonstrated , improved ionic conductivity, and sufficient electrochemical window at reduced temperatures. Furthermore, Na/NiCl2 cells with 50 mol% NaBr-containing secondary electrolyte exhibited reduced polarizations at 175�C compared to the cell with the standard NaAlCl4 catholyte. The cells also exhibited stable cycling performance even at 150oC.
Li, Guosheng; Lu, Xiaochuan; Coyle, Christopher A.; Kim, Jin Yong; Lemmon, John P.; Sprenkle, Vincent L.; Yang, Zhenguo
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.
Wishau, R.
Design considerations for concentrating solar power tower systems employing molten salt.
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.
Moore, Robert Charles; Siegel, Nathan Phillip; Kolb, Gregory J.; Vernon, Milton E.; Ho, Clifford Kuofei
Molten salt thermal energy storage systems. Project 8981, final report
The feasibility of storing thermal energy at temperatures of 450/sup 0/ to 535/sup 0/C (850/sup 0/ to 1000/sup 0/F) in the form of latent heat of fusion has been examined for over 30 inorganic salts and salt mixtures. Alkali carbonate mixtures are attractive as phase-change storage materials in this temperature range because of their relatively high storage capacity and thermal conductivity, moderate cost, low volumetric expansion upon melting, low corrosivity, and good chemical stability. An equimolar mixture of Li/sub 2/CO/sub 3/ and K/sub 2/CO/sub 3/, which melts at 505/sup 0/C with a latent heat of 148 Btu/lb, was chosen for experimental study. The cyclic charge/discharge behavior of laboratory- and engineering-scale systems was determined and compared with predictions based on a mathematical heat-transfer model that was developed during this program. The thermal performance of one engineering-scale unit remained very stable during 1400 hours of cyclic operation. Several means of improving heat conduction through the solid salt were explored. Areas requiring further investigation have been identified.
Maru, H.C.; Dullea, J.F.; Kardas, A.; Paul, L.
Direct-contact air/molten salt heat exchange for solar-thermal systems
Heat exchangers employing direct contact between molten draw salt and air were studied for use in solar industrial process heat (IPH) systems. Direct-contact systems consisting of a fin-tube preheater and a spray or packed column were compared to conventional heat exchangers. Direct contact reduced the IPH system cost by 5% to 10%. The direct-contact heat exchangers cost only 15% to 30% as much as comparable conventional exchangers. However, the rate of salt degradation by CO/sup 2/ and H/sub 2/O must be determined to see if it is acceptable.
Wright, J.D.; d'Agincourt, C.
Uranium oxides in a molten eutectic mixture of NaClKCl were chlorinated by bubbling chlorine gas through the mixture. The reaction product, uranyl chloride. was soluble in the molten salt. Although UO/sub 2/ was the most common oxide used, the reaction was similar in the other oxides. Phosgene and aluminum chloride were also used as chlorinating agents. A dense, crystalline precipitate of pure UO/sub 2/ was prepared by the reduction of the uranyl chloride contained in the molten salt solution. The reduction was accomplished by contacting the salt solution with any of several metals, by reaction with hydrogen or dry ammonia gas, or by electrolysis. Several kilograms of UO/sub 2/ were prepared by electrolysis using graphite electrodes. The physical properties of the material made it potentially useful as a ceramic fuel material. The initial high particle density of the "as-produced" UO/sub 2/ was considered of great potential advantage for adapting this process to the refabrication of irradiated UO/sub 2/ into recycle fuel elements. (M.C.G.)
Lyon, W.L.; Voiland, E.E.
An Evaluation of Molten-Salt Power Towers Including Results of the Solar Two Project
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.
Thermal analysis of solar thermal energy storage in a molten-salt thermocline
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)
Yang, Zhen; Garimella, Suresh V. [Cooling Technologies Research Center, NSF I/UCRC, School of Mechanical Engineering, Purdue University, West Lafayette, IN 47907-2088 (United States)
Incorporating supercritical steam turbines into molten-salt power tower plants : feasibility and performance.
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.
Pacheco, James Edward; Wolf, Thorsten [Siemens Energy, Inc., Orlando, FL; Muley, Nishant [Siemens Energy, Inc., Orlando, FL
Concentrating Solar Power - Molten Salt Pump Development, Final Technical Report (Phase 1)
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 design a pump for a trough solar power plant system, the intent is for the design to be extensible to a solar power tower application. This can be accomplished by adding pumping stages to increase the discharge pressure to the levels necessary for a solar power tower application. This report incorporates all available conceptual design information completed for this project in Phase I.
Michael McDowell; Alan Schwartz
Development of Molten-Salt Heat Transfer Fluid Technology for Parabolic Trough Solar Power Plants - Public Final Technical Report
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�/kWhe , which achieved the Phase 2 Go/No Go target of less than 0.12�/kWhe. Abengoa Solar has high confidence that the primary risk areas have been addressed in the project and a commercial plant utilizing molten salt is economically and technically feasible. The strong results from the Phase 1 and 2 research, testing, and analyses, summarized in this report, led Abengoa Solar to recommend that the project proceed to Phase 3. However, a commercially viable collector interconnection was not fully validated by the end of Phase 2, combined with the uncertainty in the federal budget, forced the DOE and Abengoa Solar to close the project. Thus the resources required to construct and operate a molten salt pilot plant will be solely supplied by Abengoa Solar.
Grogan, Dylan C. P.
Customer interface document for the Molten Salt Test Loop (MSTL) system.
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.
Pettit, Kathleen; Kolb, William J.; Gill, David Dennis; Briggs, Ronald D.
MOLTEN SALT CORROSION OF SUPERHEATERS IN BLACK LIQUOR RECOVERY BOILERS John Bohling, University of Tennessee Georgia Tech SURF 2010 Fellow
MOLTEN SALT CORROSION OF SUPERHEATERS IN BLACK LIQUOR RECOVERY BOILERS John Bohling, University Goodman Introduction In the papermaking industry, black liquor recovery boilers burn black liquor into the superheater region of the boiler, where the salt-deposit, or smelt, forms a scale on the superheater tubes.1
Li, Mo
Conceptual Design of Forced Convection Molten Salt Heat Transfer Testing Loop
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 be performed to develop probability of confidence in what is measured in the test loop. Overall, the testing loop will allow development of needed heat transfer related thermophysical parameters for all the salts, validate existing correlations, validate measuring instruments under harsh environment, and have extensive corrosion testing of materials of construction.
Manohar S. Sohal; Piyush Sabharwall; Pattrick Calderoni; Alan K. Wertsching; S. Brandon Grover
Molten Salt Power Tower Cost Model for the System Advisor Model (SAM)
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.
Turchi, C. S.; Heath, G. A.
Conceptual Design of a 100 MWe Modular Molten Salt Power Tower Plant
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.
James E. Pacheco; Carter Moursund, Dale Rogers, David Wasyluk
Glovebox design requirements for molten salt oxidation processing of transuranic waste
This paper presents an overview of potential technologies for stabilization of {sup 238}Pu-contaminated combustible waste. Molten salt oxidation (MSO) provides a method for removing greater than 99.999% of the organic matrix from combustible waste. Implementation of MSO processing at the Los Alamos National Laboratory (LANL) Plutonium Facility will eliminate the combustible matrix from {sup 238}Pu-contaminated waste and consequently reduce the cost of TRU waste disposal operations at LANL. The glovebox design requirements for unit operations including size reduction and MSO processing will be presented.
Ramsey, K.B.; Acosta, S.V. [Los Alamos National Lab., NM (United States); Wernly, K.D. [Molten Salt Oxidation Corp., Bensalem, PA (United States)
Electrochemical corrosion of iron-magnesium-alumina spinel (FMAS) in molten potassium salts and coal slag
Iron, magnesium-alumina spinel (FMAS) (0.25 Fe/sub 3/O/sub 4/ . 0.75 MgAl/sub 2/O/sub 4/) has been considered for use as an electrode in magnetohydrodynamic (MHD) generator channels. Predominantly an electronic conductor, FMAS has adequate electrical conductivity (>1 S/m) above 520/sup 0/K. In addition, FMAS can be easily fabricated into a form and sintered in air to >90% theoretical density and has a melting point of 2124 +- 20/sup 0/K. Laboratory tests to measure both the electrochemical and chemical corrosion of FMAS in molten K/sub 2/CO/sub 3/, K/sub 2/SO/sub 4/ and coal slags were developed at the Pacific Northwest Laboratory to evaluate the relative corrosion of FMAS. Under isothermal conditions, a direct electric current was passed between an anode and a cathode through a molten electrolyte. The molten coal slags were synthetic high-calcium, low-iron Montana Rosebud and low-calcium, high-iron Illinois No. 6. Evaluations of electrochemical corrosion were made as functions of current density, temperature, and slag composition. These results were compared to those of FMAS tested without electric current. The corrosion rates and reaction products were investigated by optical microscopy and scanning electron microscopy. Overall, FMAS has too-high an electrochemical corrosion rate to be considered as MHD electrodes in Montana Rosebud coal slag or in systems where only molten potassium salts are present. However, FMAS may be considered for use in high-iron coal slags although the corrosion rates are still quite high even in these slags.
Marchant, D.D.; Griffin, C.W.; Bates, J.L.
Effect of the graphite electrode material on the characteristics of molten salt electrolytically produced carbon nanomaterials
The electrochemical erosion of a graphite cathode during the electrolysis of molten lithium chloride salt may be used for the preparation of nano-structured carbon materials. It has been found that the structures and morphologies of these carbon nanomaterials are dependent on those of the graphite cathodes employed. A combination of tubular and spherical carbon nanostructures has been produced from a graphite with a microstructure of predominantly planar micro-sized grains and a minor fraction of more irregular nano-sized grains, whilst only spherical carbon nanostructures have been produced from a graphite with a microstructure of primarily nano-sized grains. Based on the experimental results, a best-fit regression equation is proposed that relates the crystalline domain size of the graphite reactants and the carbon products. The carbon nanomaterials prepared possess a fairly uniform mesoporosity with a sharp peak in pore size distribution at around 4 nm. The results are of crucial importance to the production of carbon nanomaterials by way of the molten salt electrolytic method. - Highlights: {yields} Carbon nanomaterials are synthesised by LiCl electrolysis with graphite electrodes. {yields} The degree of crystallinity of graphite reactant and carbon product are related. {yields} A graphite reactant is identified that enables the preparation of carbon nanotubes. {yields} The carbon products possess uniform mesoporosity with narrow pore size distribution.
Kamali, Ali Reza, E-mail: firstname.lastname@example.com; Schwandt, Carsten; Fray, Derek J.
Electrochemistry of LiCl-Li2O-H2O Molten Salt Systems
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.
Natalie J. Gese; Batric Pesic
Molten Salt-Carbon Nanotube Thermal Energy Storage for Concentrating Solar Power Systems Final Report
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.
Michael Schuller; Frank Little; Darren Malik; Matt Betts; Qian Shao; Jun Luo; Wan Zhong; Sandhya Shankar; Ashwin Padmanaban
Heat Transfer and Latent Heat Storage in Inorganic Molten Salts for Concentrating Solar Power Plants
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 experimentation.
Mathur, Anoop [Terrafore Inc.] [Terrafore Inc.
I-NERI ANNUAL TECHNICAL PROGRESS REPORT: 2006-002-K, Separation of Fission Products from Molten LiCl-KCl Salt Used for Electrorefining of Metal Fuels
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 during the salt/zeolite contacting process � Compare the adsorption models to experimentally obtained, ER salt results � Evaluate results obtained from the oxygen precipitation and salt/zeolite ion exchange studies to determine the best processes for selective fission-product removal from electrorefiner salt.
S. Frank
Potentiometric Sensor for Real-Time Remote Surveillance of Actinides in Molten Salts
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.
Natalie J. Gese; Jan-Fong Jue; Brenda E. Serrano; Guy L. Fredrickson
Investigation of cold filling receiver panels and piping in molten-nitrate-salt central-receiver solar power plants
Cold filling refers to flowing a fluid through piping or tubes that are at temperatures below the fluid`s freezing point. Since the piping and areas of the receiver in a molten-nitrate salt central-receiver solar power plant must be electrically heated to maintain their temperatures above the nitrate salt freezing point (430{degrees}F, 221{degrees}C), considerable energy could be used to maintain such temperatures during nightly shut down and bad weather. Experiments and analyses have been conducted to investigate cold filling receiver panels and piping as a way of reducing parasitic electrical power consumption and increasing the availability of the plant. The two major concerns with cold filling are: (1) how far can the molten salt penetrate cold piping before freezing closed and (2) what thermal stresses develop during the associated thermal shock. Cold fill experiments were conducted by flowing molten salt at 550{degrees}F (288{degrees}C) through cold panels, manifolds, and piping to determine the feasibility of cold filling the receiver and piping. The transient thermal responses were measured and heat transfer coefficients were calculated from the data. Nondimensional analysis is presented which quantifies the thermal stresses in a pipe or tube undergoing thermal shock. In addition, penetration distances were calculated to determine the distance salt could flow in cold pipes prior to freezing closed.
Pacheco, J.E.; Ralph, M.E.; Chavez, J.M.
A nightly conditioning method to reduce parasitic power consumption in molten-salt central-receiver solar-power plants
A method to reduce nightly parasitic power consumption in a molten salt central receiver is discussed where salt is drained from the piping and heat tracing is turned off to allow the piping to cool to ambient overnight, then in the morning the pipes are filled while they are cold. Since the piping and areas of the receiver in a molten-nitrate salt central-receiver solar power plant must be electrically heated to maintain their temperatures above the nitrate salt freezing point (430{degrees}F, 221{degrees}C), considerable energy could be used to maintain such temperatures during nightly shut down and bad weather. Experiments and analyses have been conducted to investigate cold filling receiver panels and piping as a way of reducing parasitic electrical power consumption and increasing the availability of the plant. The two major concerns with cold filling are: (1) how far can the molten salt penetrate cold piping before freezing closed and (2) what thermal stresses develop during the associated thermal shock. Experiments and analysis are discussed.
Pacheco, J.E.
Mixed-convective, conjugate heat transfer during molten salt quenching of small parts
It is common in free quenching immersion heat treatment calculations to locally apply constant or surface-averaged heat-transfer coefficients obtained from either free or forced steady convection over simple shapes with small temperature differences from the ambient fluid. This procedure avoids the solution of highly transient, non-Boussinesq conjugate heat transfer problems which often involve mixed convection, but it leaves great uncertainty about the general adequacy of the results. In this paper we demonstrate for small parts (dimensions of the order of inches rather than feet) quenched in molten salt, that it is feasible to calculate such nonuniform surface heat transfer from first principles without adjustable empirical parameters. We use literature physical property salt data from the separate publications of Kirst et al., Nissen, Carling, and Teja, et al. for T800 F is not considered to be important due to the short time the surface temperature exceeds that value for small parts. Similarly, for small parts, the local Reynolds and Rayleigh numbers are below the corresponding critical values for most if not all of the quench, so that we see no evidence of the existence of significant turbulence effects, only some large scale unsteadiness for brief periods. The experimental data comparisons from the open literature include some probe cooling-rate results of Foreman, as well as some cylinder thermal histories of Howes.
Chenoweth, D.R.
It is common in free quenching immersion heat treatment calculations to locally apply constant or surface-averaged heat-transfer coefficients obtained from either free or forced steady convection over simple shapes with small temperature differences from the ambient fluid. This procedure avoids the solution of highly transient, non-Boussinesq conjugate heat transfer problems which often involve mixed convection, but it leaves great uncertainty about the general adequacy of the results. In this paper we demonstrate for small parts (dimensions of the order of inches rather than feet) quenched in molten salt, that it is feasible to calculate such nonuniform surface heat transfer from first principles without adjustable empirical parameters. We use literature physical property salt data from the separate publications of Kirst et al., Nissen, Carling, and Teja, et al. for T<1000 F, and then extrapolate it to the initial part temperature. The reported thermal/chemical breakdown of NaNO{sub 2} for T>800 F is not considered to be important due to the short time the surface temperature exceeds that value for small parts. Similarly, for small parts, the local Reynolds and Rayleigh numbers are below the corresponding critical values for most if not all of the quench, so that we see no evidence of the existence of significant turbulence effects, only some large scale unsteadiness for brief periods. The experimental data comparisons from the open literature include some probe cooling-rate results of Foreman, as well as some cylinder thermal histories of Howes.
Process Heat Exchanger Options for Fluoride Salt High Temperature Reactor
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.
Piyush Sabharwall; Eung Soo Kim; Michael McKellar; Nolan Anderson
Liquefaction Chemistry B. Molten Salt Catalysis RationaleUsed Equipment and Procedure Molten Salt a. b. c. Treat~entEquipment and Procedure Molten Salt Treatment a. Equipment
Holten, R.R.
Salt-Zeolite Ion Exchange Equilibrium Studies for Complete Set of Fission Products in Molten LiCl-KCl
This paper presents results on LiCl-KCl based molten salts/zeolite-A contact experiments and the associated equilibrium ion exchange model. Experiments examine the contact behaviors of various ternary salts (LiCl-KCl-YCl3, LiCl-KCl-LaCl3, and LiCl-KCl-PrCl3) and quaternary salts (LiCl-KCl-CsCl-NdCl3 and LiCl-KCl-CsCl-SrCl2) with the zeolite-A. The developed equilibrium model assumes that there are ion exchange and occlusion sites, both of which are in equilibrium with the molten salt phase. A systematic approach in estimating total occlusion capacity of the zeolite-A is developed. The parameters of the model, including the total occlusion capacity of the zeolite-A, were determined from fitting experimental data collected via multiple independent studies including the ones reported in this paper. Experiments involving ternary salts were used for estimating the parameters of the model, while those involving quaternary salts were used to validate the model.
Tae-Sic Yoo; Steven M. Frank; Michael F. Simpson; Paula A. Hahn; Terry J. Battisti; Supathorn Phongikaroon
Thermal Storage Commercial Plant Design Study for a 2-Tank Indirect Molten Salt System: Final Report, 13 May 2002 - 31 December 2004
Subcontract report by Nexant, Inc., and Kearney and Associates regarding a study of a solar parabolic trough commercial plant design with 2-tank indirect molten salt thermal storage system.
Kelly, B.; Kearney, D.
Novel concepts in electrochemical solar cells. First quarterly progress report, May 15-July 15, 1979. [Molten salt electrolytes
Emphasis has been directed toward the development and evaluation of an efficient, operational photoelectrochemical cell (PEC) system involving (i) molten salt (and/or highly concentrated, inorganic, non-aqueous) electrolytes, and (ii) the promising semiconductor electrode materials such as CuTnS/sub 2/, CuInSe/sub 2/, MoS/sub 2/, MoSe/sup 2/ etc. As a direct consequence, the stages of the work program that are most critical at this time are the electrode fabrication and characterization and the electrolyte preparation and characterization phases. It has been demonstrated for the first time that a semiconductor electrode exhibits quite large photoeffects in a cell containing a molten salt electrolyte. Detailed studies are underway to explore the constraints and advantages of this type of electrolyte, from the standpoint of efficiency, corrosion of the electrode, and the chemical and physical properties related to overall cell performance. Progress is reported. (WHK)
DuBow, J.
Molten salt extraction process for the recovery of valued transition metals from land-based and deep-sea minerals
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.
Maroni, Victor A. (Naperville, IL); von Winbush, Samuel (Huntington, NY)
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.
Maroni, V.A.; von Winbush, S.
Novel concepts in electrochemical solar cells. Second quarterly progress report, August 15, 1979-October 15, 1979. [Molten salt electrolytes
It is considered that the short term stability of n-GaAs PEC's in a ferrocene-based, ambient temperature molten salt electrolyte is reasonably good. However, longer term evaluation is required to determine the extent and significance of corrosion, stability, etc. Extremely few fundamental studies have been made of the semiconductor/molten salt interphase and experiments in this area would be most useful. Indeed, even the design parameters for PECs of any kind have not been quantitatively delineated and present consideration will be given to models for PEC solar cells and limitations caused by ion transport in the electrolyte. The MoSe/sub 2/ and MoS/sub 2/ electrodes appear to have substrate reproducibility and transport limitations that make them unsuitable candidates for efficient PEC's at this time. Similarly, the lack of availability of high quality CuInSe/sub 2/ and CuInS/sub 2/ substrates limits the quantitative experimental evaluation of their utility for PEC applications. We are presently focusing attention on CdSe/CdTe mixtures and CdS as electrodes as well as Si and GaAs in molten salt and polyelectrolyte solutions. The system for solar cell evaluation and network analysis of substrates and cells was mode operational. Preliminary work on economic and theoretical modelling was begun. Progress is reported. (WHK)
DuBow, J.; Job, R.; Krishnan, R.; Gale, B.
Molten salt thermal energy storage systems: system design. [LiKCO/sub 3/ mixture
A five-task research program aimed at the development of molten salt thermal energy storage systems commenced in June 1976. The first topical report, covering Task 1, the selection of suitable salt systems for storage at 850 to 1000/sup 0/F, was issued in August 1976. It was concluded that a 35 Wt percent Li/sub 2/CO/sub 3/-65 Wt percent K/sub 2/CO/sub 3/ (LiKCO/sub 3/) mixture was most suitable for the purpose. Interrelationships between various design parameters were examined using the available solutions, and an engineering-scale storage unit was designed. This unit has an annular configuration with a 1-ft OD, 1.5-ft high, 2-in. dia heat transfer well. Preliminary experiments on a pilot size (3-in. OD) unit showed that temperature profiles and progress of the solid-liquid interface agreed with those predicted theoretically. Also, no supercooling was observed during cooldown, and the presence of significant convective mixing was indicated by negligible temperature gradients. Use of a lithium aluminate volume-change suppressor was investigated, but it appears to be nonessential because of the low volume-change in the LiKCO/sub 3/ system. Consideration of the relative heat-transfer resistances under practical conditions suggested that the use of a conductivity promoter will enhance the heat-transfer rates, thereby requiring smaller heat-transfer areas. Different configurations and materials were considered for this application; an aluminum wool appears to be most suitable. The corrosion resistance of various construction materials was investigated. Stainless steels and aluminum appear to be suitable construction materials for carbonates in the 850 to 1000/sup 0/F range. Testing of the engineering-scale system (Task 3) and verification of the conclusions derived under Task 2 are in progress.
Maru, H.C.; Kardas, A.; Huang, V.M.; Dullea, J.F.; Paul, L.; Marianowski, L.G.
Economic evaluation of solar-only and hybrid power towers using molten salt technology
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.
Kolb, G.J.
Electrodeposition of cobalt and cobalt-aluminum alloys from a room temperature chloroaluminate molten salt
The electrodeposition of magnetic cobalt-aluminum alloys was investigated in the Lewis acidic aluminum chloride-1-methyl-3-ethylimidazolium chloride [60.0--40.0 mole percent (m/o)] molten salt containing electrogenerated Co(II) at 25 C. rotating disk electrode voltammetry indicated that it is possible to produce alloy deposits containing up to 62 atomic (a/o) aluminum at potentials positive of that for the bulk deposition of aluminum. The onset of the underpotential-driven aluminum codeposition process occurred at around 0.40 V vs. the Al/Al(III) couple in a 5.00 mmol/liter Co(II) solution but decreased as the Co(II) concentration increased. The Co-Al alloy composition displayed an inverse dependence on the Co(II) concentration but tended to become independent of concentration as the potential was decreased to 0 V. A rotating ring-disk electrode voltammetry technique was developed to analyze the composition and structure of the Co-Al alloy deposits. This technique takes advantage of the fact that the mass-transport-limited reduction of cobalt(II) occurs at potentials considerably more positive than that at which aluminum codeposition occurs. Scanning electron microscopy and energy dispersive X-ray analysis of bulk electrodeposits revealed that deposit morphology depends strongly upon aluminum content/deposition potential; deposits produced at 0.40 V from 50.0 mmol/liter Co(II) solutions consisted of 10 to 20 {micro}m diam multifaceted nodules of pure hcp cobalt, whereas those obtained at 0.20 V were dense and fine grained, containing about 4 a/o Al. Deposits produced at 0 V had the visual appearance of a loosely adherent black powder. X-ray diffraction measurements revealed a lattice expansion and a decrease in grain size as the hcp cobalt was alloyed with increasing amounts of aluminum.
Mitchell, J.A.; Pitner, W.R.; Hussey, C.L. [Univ. of Mississippi, University, MS (United States). Dept. of Chemistry; Stafford, G.R. [National Inst. of Standards and Technology, Gaithersburg, MD (United States). Materials Science and Engineering Lab.
Electrorefining of Titanium from Bi-Ti Alloys in a NaCl-KCl Molten Salt
A27: Electrochemical Study of Ag Ionization in Molten Lead ...
The concentration of Ag+ in the molten glass significantly increased with ... Electrochemical Deposition of High Purity Silicon in Molten Salts.
Heat transfer and pressure drop measurements in an air/molten salt direct-contact heat exchanger
This paper presents a comparison of experimental data with a recently published model of heat exchange in irrigated packed beds. Heat transfer and pressure drop were measured in a 150 mm (ID) column with a 610-mm bed of metal Pall rings. Molten nitrate salt and preheated air were the working fluids with a salt inlet temperature of approximately 440{degree}C and air inlet temperatures of approximately 230{degree}C. A comparison between the experimental data and the heat transfer model is made on the basis of heat transfer from the salt. For the range of air and salt flow rates tested, 0.3 to 1.2 kg/m{sup 2} s air flow and 6 to 18 kg/m{sup 2} s salt flow, the data agree with the model within 22% standard deviation. In addition, a model for the column pressure drop was validated, agreeing with the experimental data within 18% standard deviation over the range of column pressure drop from 40 to 1250 Pa/m. 25 refs., 7 figs., 2 tabs.
Bohn, M.S.
Interactions Between Fluid-Dynamics and Neutronic Phenomena in the Physics of Molten-Salt Systems
Technical Paper / Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications
S. Dulla; P. Ravetto
Transpiring wall supercritical water oxidation reactor salt deposition studies
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.
Haroldsen, B.L.; Mills, B.E.; Ariizumi, D.Y.; Brown, B.G. [and others
Morphology and photoluminescence of Ba0.5Sr0.5MoO4 powders by a molten salt method
Ba0.5Sr0.5MoO4 powders with scheelite-type tetragonal structure were successfully synthesized by a molten salt method. The structure, morphology, and luminescent property of the as-prepared powders were characterized ...
Ling Wei; Yunfei Liu; Yinong Lu; Tao Wu
Testing thermocline filler materials and molten-salt heat transfer fluids for thermal energy storage systems used in parabolic trough solar power plants.
Parabolic trough power systems that utilize concentrated solar energy to generate electricity are a proven technology. Industry and laboratory research efforts are now focusing on integration of thermal energy storage as a viable means to enhance dispatchability of concentrated solar energy. One option to significantly reduce costs is to use thermocline storage systems, low-cost filler materials as the primary thermal storage medium, and molten nitrate salts as the direct heat transfer fluid. Prior thermocline evaluations and thermal cycling tests at the Sandia National Laboratories' National Solar Thermal Test Facility identified quartzite rock and silica sand as potential filler materials. An expanded series of isothermal and thermal cycling experiments were planned and implemented to extend those studies in order to demonstrate the durability of these filler materials in molten nitrate salts over a range of operating temperatures for extended timeframes. Upon test completion, careful analyses of filler material samples, as well as the molten salt, were conducted to assess long-term durability and degradation mechanisms in these test conditions. Analysis results demonstrate that the quartzite rock and silica sand appear able to withstand the molten salt environment quite well. No significant deterioration that would impact the performance or operability of a thermocline thermal energy storage system was evident. Therefore, additional studies of the thermocline concept can continue armed with confidence that appropriate filler materials have been identified for the intended application.
Kelly, Michael James; Hlava, Paul Frank; Brosseau, Douglas A.
Regenerative mode photo electrochemical cells in molten salt electrolytes. 1st four monthly report (1/31/80)
The most promising photoelectrodes selected for use in the butyl pyridinium chloride-aluminum chloride room temperature molten salt are n-type silicon, gallium arsenide and cadmium telluride. The solubilities of these semiconductors are low, and their conduction and valence band edges are favorably located. Cadmium selenide and sulfide showed significant solubility in the melt, and the conduction band edge for p-type cadmium telluride was too close to the aluminum deposition potential. Several reversible redox couples have been identified, which could potentially be used in a photoelectrochemical cell. These include W/sup 5 +//W/sup 6 +/ and Eu/sup 2 +//Eu/sup 3 +/ as well as ferrocene and its derivatives.
Molten salt oxidation of mixed wastes: Separation of radioactive materials and Resource Conservation and Recovery Act (RCRA) materials
The Oak Ridge National Laboratory (ORNL) is involved in a program to apply a molten salt oxidation (MSO) process to the treatment of mixed wastes at Oak Ridge and other Department of Energy (DOE) sites. Mixed wastes are defined as those wastes that contain both radioactive components, which are regulated by the atomic energy legislation, and hazardous waste components, which are regulated under the Resource Conservation and Recovery Act (RCRA). A major part of our ORNL program involves the development of separation technologies that are necessary for the complete treatment of mixed wastes. The residues from the MSO treatment of the mixed wastes must be processed further to separate the radioactive components, to concentrate and recycle residues, or to convert the residues into forms acceptable for final disposal. This paper is a review of the MSO requirements for separation technologies, the information now available, and the concepts for our development studies.
Bell, J.T.; Haas, P.A.; Rudolph, J.C.
Preparation of Al-Ca Alloys by Molten Salt Electrolysis Method
Content of calcium in alloys was to be higher than 13w%, and tests were carried out with 0.80-1.20A�cm?2 of ... Plutonium Removal from Fluoride Spent Salts.
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 experimentation.
Mathur, Anoop [Terrafore Inc.
Freeze-thaw tests of trough receivers employing a molten salt working fluid.
Several studies predict an economic benefit of using nitrate-based salts instead of the current synthetic oil within a solar parabolic trough field. However, the expected economic benefit can only be realized if the reliability and optical performance of the salt trough system is comparable to today's oil trough. Of primary concern is whether a salt-freeze accident and subsequent thaw will lead to damage of the heat collection elements (HCEs). This topic was investigated by experiments and analytical analysis. Results to date suggest that damage will not occur if the HCEs are not completely filled with salt. However, if the HCE is completely filled at the time of the freeze, the subsequent thaw can lead to plastic deformation and significant bending of the absorber tube.
Moss, Timothy A.; Iverson, Brian D.; Siegel, Nathan Phillip; Kolb, Gregory J.; Ho, Clifford Kuofei
Controlled temperature expansion in oxygen production by molten alkali metal salts
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.
Erickson, Donald C. (Annapolis)
Metals removal from spent salts
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.
Hsu, Peter C. (Pleasanton, CA); Von Holtz, Erica H. (Livermore, CA); Hipple, David L. (Livermore, CA); Summers, Leslie J. (Livermore, CA); Brummond, William A. (Livermore, CA); Adamson, Martyn G. (Danville, CA)
Actinide removal from spent salts
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.
Hsu, Peter C. (Pleasanton, CA); von Holtz, Erica H. (Livermore, CA); Hipple, David L. (Livermore, CA); Summers, Leslie J. (Livermore, CA); Adamson, Martyn G. (Danville, CA)
Partition of actinides and fission products between metal and molten salt phases: Theory, measurement, and application to IFR pyroprocess development
The chemical basis of Integral Fast Reactor fuel reprocessing (pyroprocessing) is partition of fuel, cladding, and fission product elements between molten LiCl-KCl and either a solid metal phase or a liquid cadmium phase. The partition reactions are described herein, and the thermodynamic basis for predicting distributions of actinides and fission products in the pyroprocess is discussed. The critical role of metal-phase activity coefficients, especially those of rare earth and the transuranic elements, is described. Measured separation factors, which are analogous to equilibrium constants but which involve concentrations rather than activities, are presented. The uses of thermodynamic calculations in process development are described, as are computer codes developed for calculating material flows and phase compositions in pyroprocessing.
Ackerman, J.P.; Johnson, T.R.
Molten salt cooling//sup 17/Li-/sup 83/Pb breeding blanket concept
A description of a fusion breeding blanket concept using draw salt coolant and static /sup 17/Li-/sup 83/Pb is presented. /sup 17/Li-/sup 83/Pb has high breeding capability and low tritium solubility. Draw salt operates at low pressure and is inert to water. Corrosion, MHD, and tritium containment problems associated with the MARS design are alleviated because of the use of a static LiPb blanket. Blanket tritium recovery is by permeation toward the plasma. A direct contact steam generator is proposed to eliminate some generic problems associated with a tube shell steam generator.
Sze, D.K.; Cheng, E.T.
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.
Erickson, D.C.
Thermal hydraulic design of a salt-cooled highly efficient environmentally friendly reactor
A 1 OOOMWth liquid-salt cooled thermal spectrum reactor was designed with a long fuel cycle, and high core exit temperature. These features are desirable in a reactor designed to provide process heat applications such as ...
Whitman, Joshua (Joshua J.)
Assessment of feasibility, economics, and market potential for a molten salt system at 1000/sup 0/F reheat steam: feasibility, economics, and market potential
As a result of the Advanced Central Receiver (ACR) Phase I systems studies, Martin-Marietta Corporation (MMC) developed a conceptual design employing 1050/sup 0/F molten salt and a 950/sup 0/F/950/sup 0/F reheat turbine. This concept appears to have the potential for providing higher steam conditions leading to higher performance and wider market application. This report presents the results of a preliminary investigation of the system to determine the feasibility of providing 1000/sup 0/F/1000/sup 0/F steam and the impact of the required design modifications on the system performance, cost, and market potential for solar repowering. Two modified designs are investigated. In one modified design, the temperature of the molten salt is the same as in the MMC baseline design (1050/sup 0/F), but the steam generators have been modified to provide 1000/sup 0/F/1000/sup 0/F steam. In the other modified design, the enhanced steam conditions are obtained using molten salt at a temperature of 1100/sup 0/F.
DeRienzo, P.; Masaki, M.; Mathur, P.
Molten salt as a heat transfer fluid for heating a subsurface formation
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.
Nguyen, Scott Vinh (Houston, TX); Vinegar, Harold J. (Bellaire, TX)
Oxide Skin Strength Measurements on Molten Aluminum
Presentation Title, OXIDE SKIN STRENGTH MEASUREMENTS ON MOLTEN ALUMINUM � MANGANESE ALLOYS WITH AND WITHOUT SALT ON SURFACE
Gas Turbine/Solar Parabolic Trough Hybrid Design Using Molten Salt Heat Transfer Fluid: Preprint
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.
Turchi, C. S.; Ma, Z.
Molten Oxide Electrolysis Application to Steelmaking: A New ...
Abstract Scope, Molten oxide electrolysis (MOE) is a new steelmaking ... Electrochemical Reduction of Tantalum Oxide in a CaCl2 � CaO Molten Salt Electrolyte.
Wetting Properties of Molten Silicon with Graphite Materials
Abstract Scope, The wetting behavior of molten-silicon/refractory-materials system is important in ... Electrorefining of Metallurgical Grade Silicon in Molten Salts.
The Interface Reaction and Transport of Oxygen between the Molten ...
Effect of Silicon on the Viscosity and Solidification Properties of Molten Irons with ... Stibnite in Low Temperature Molten Salt Smelting Process without Reductant.
Tritium production analysis and management strategies for a Fluoride-salt-cooled high-temperature test reactor (FHTR)
The Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a test reactor concept that aims to demonstrate the neutronics, thermal-hydraulics, materials, tritium management, and to address other reactor operational ...
Rodriguez, Judy N
Results of molten salt panel and component experiments for solar central receivers: Cold fill, freeze/thaw, thermal cycling and shock, and instrumentation tests
Experiments have been conducted with a molten salt loop at Sandia National Laboratories in Albuquerque, NM to resolve issues associated with the operation of the 10MW{sub e} Solar Two Central Receiver Power Plant located near Barstow, CA. The salt loop contained two receiver panels, components such as flanges and a check valve, vortex shedding and ultrasonic flow meters, and an impedance pressure transducer. Tests were conducted on procedures for filling and thawing a panel, and assessing components and instrumentation in a molten salt environment. Four categories of experiments were conducted: (1) cold filling procedures, (2) freeze/thaw procedures, (3) component tests, and (4) instrumentation tests. Cold-panel and -piping fill experiments are described, in which the panels and piping were preheated to temperatures below the salt freezing point prior to initiating flow, to determine the feasibility of cold filling the receiver and piping. The transient thermal response was measured, and heat transfer coefficients and transient stresses were calculated from the data. Freeze/thaw experiments were conducted with the panels, in which the salt was intentionally allowed to freeze in the receiver tubes, then thawed with heliostat beams. Slow thermal cycling tests were conducted to measure both how well various designs of flanges (e.g., tapered flanges or clamp type flanges) hold a seal under thermal conditions typical of nightly shut down, and the practicality of using these flanges on high maintenance components. In addition, the flanges were thermally shocked to simulate cold starting the system. Instrumentation such as vortex shedding and ultrasonic flow meters were tested alongside each other, and compared with flow measurements from calibration tanks in the flow loop.
Pacheco, J.E.; Ralph, M.E.; Chavez, J.M.; Dunkin, S.R.; Rush, E.E.; Ghanbari, C.M.; Matthews, M.W.
Design of a 2400MW liquid-salt cooled flexible conversion ratio reactor
A 2400MWth liquid-salt cooled flexible conversion ratio reactor was designed, utilizing the ternary chloride salt NaCl-KCl-MgCI2 (30%-20%-50%) as coolant. The reference design uses a wire-wrapped, hex lattice core, and is ...
Petroski, Robert C
Thermal Transport and Heat Exchanger Design for the Space Molten Salt Reactor Concept.
??Surface power and nuclear electric propulsion in space necessitate the development of high energy density, long term continuous power sources. Research at The Ohio State� (more)
Flanders, Justin M.
Progress is reported in two separate abstracts: MSRE design, engineering analysis, and component development; and material studies. (N.W.R.)
File: AHTR: Japan.AHTR.Nov2004.Paper The submitted manuscript has been authored by a contractor of the U.S. Government under contract DE-AC05-00OR22725. Accordingly, the U.S. Government retains a nonexclusive, royalty-free license to publish or reproduce the published form of this contribution, or allow others to do so, for U.S. Government purposes.
Charles Forsberg; Liquid-metal-reactor Plant; Charles Forsberg
Cost of Processing Fuel from a Molten Salt, Fusion/Fission, Hybrid Reactor Blanket
Blanket and Process Engineering / Proceedings of the Second National Topical Meeting on Tritium Technology in Fission, Fusion and Isotopic Applications (Dayton, Ohio, April 30 to May 2, 1985)
J. S. Watson; W. R. Grimes; D. E. Brashears
Granular Dynamics in Pebble Bed Reactor Cores
and P. S. Pickard, �Molten-salt-cooled advanced high-heat transfer of a molten salt coolant and the passive
Laufer, Michael Robert
A fuel cycle economic study has been made for a 315Mwe graphite- moderated slightly enriched fused-salt reactor. Fuel cycle costs of less than 1.5 mills may be possible for such reactors operating on a ten-year cycle even when the fuel is discarded at the end of the cycle. Recovery of the uranium and plutonium at the end of the cycle reduces the fuel cycle costs to approximates 1 mill/kwh. Changes in the waste storage cost, reprocessing cost or salt inventory have a relatively minor effect on fuel cycle costs. (auth)
Guthrie, C.E.
Molten Metal Safety Approach through a Network
Abstract Scope, Molten Metal explosion or splash is a major risk encountered in the ... In-Line Salt-ACD: A Chlorine�Free Technology for Metal Treatment.
Overview of Component Testing Requirements for a Small Fluoride Salt-Cooled High Tempreature Reactor
This article summarizes the information necessary to provide reasonable assurance that components for a small fluoride salt-cooled high temperature reactor will meet their functional requirements. In support of the analysis of testing requirements, a simplified, conceptual description of the systems, structures, and components specific to this reactor class was developed. These reactor system elements were divided into major categories based on their functions: (1) reactor core systems, (2) heat transport system, (3) reactor auxiliary cooling system, and (4) instrumentation and controls system. An assessment of technical maturity for each element was made, and a gap analysis was performed to identify specific areas that require further testing. A prioritized list of the testing requirements was then developed. The prioritization was based on both the relative importance of the system to reactor viability, and performance and time requirements to perform the testing.
Cetiner, Mustafa Sacit [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL
Considerations of Alloy N for Fluoride Salt-Cooled High-Temperature Reactor Applications
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.
Ren, Weiju [ORNL; Muralidharan, Govindarajan [ORNL; Wilson, Dane F [ORNL; Holcomb, David Eugene [ORNL
Fluoride Salt-Cooled High-Temperature Reactor Technology Development and Demonstration Roadmap
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.
Holcomb, David Eugene [ORNL] [ORNL; Flanagan, George F [ORNL] [ORNL; Mays, Gary T [ORNL] [ORNL; Pointer, William David [ORNL] [ORNL; Robb, Kevin R [ORNL] [ORNL; Yoder Jr, Graydon L [ORNL] [ORNL
Direct Electroreduction of Oxides in Molten Fluorides
However, up to now, the use of chloride salts is still problematic partially because ... Electrochemical Reduction of Tantalum Oxide in a CaCl2 � CaO Molten Salt�...
An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components
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.
Holcomb, David Eugene [ORNL; Cetiner, Mustafa Sacit [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL
of noble gases in molten salts, which also provide a modeln Hexane B2 275�C. Hydrogen-Molten Salt WI (dynes/em) WI PI
Joyce, Peter James
Salt Salt Nature Bulletin No. 340-A April 12, 1969 Forest Preserve District of Cook County George W. Dunne, President Roland F. Eisenbeis, Supt. of Conservation SALT It is fortunate that Salt -- common salt, known to chemists as sodium chloride and to mineralogists as Halite -- is one of the most abundant substances on earth, because most of us crave it and must have it. Eskimos get along without salt because they live mostly on the uncooked flesh of fish and mammals. A few nomad tribes never eat it and do not need it because their diet contains so much milk cheese, and meat eaten raw or roasted. We people who eat boiled meat and many vegetables must have salt. Of the millions of tons produced commercially each year, only about three percent is used as table salt. Large quantities are required for refrigeration meat packing, curing and preserving fish, pickles, sauerkraut, and for other foods prepared in brine. A lot of it is needed for livestock. Salt is spread on sidewalks, streets and highways to melt ice in winter. It is used to glaze pottery, sewer pipe and other ceramics. It is required in many metallurgical processes, chemical industries, and the manufacture of such products as leather, glass, soap, bleaching powder and photographic supplies. It has about 14,000 uses.
Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors
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 results show similar trends as the computational fluid dynamics (CFD) results presented in this report; however, some differences exist that will need to be assessed in future studies. The results of this testing will be used to improve the diode design to be tested in the liquid salt loop system.
Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)
Physical Similitude in Hierarchical Engineered Systems
LS-VHTR), that research in molten salts as reactor primaryfast reactors, and molten salt reactors. Coated particleby the 8-MWth ORNL Molten Salt Reactor Experiment with the
Blandford, Edward David
Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors
H. G. MacPherson The molten salt adventure Nuclear Scienceand P.F. Peterson, Molten-Salt-Cooled Advanced High-Clarno Assessment of candidate molten salt coolants for the
Galvez, Cristhian
Towards Sustainable Metals Production by Molten Oxide Electrolysis
Liquid-metal/molten-salt cells have been shown to operate as rechargeable batteries that have the potential to handle colossal currents thereby enabling us to�...
A New Vacuum Degassing Process for Molten Aluminum
In order to maintain a low hydrogen content in molten aluminum, A porous refractory ... Metallurgical Performance of Salt and Chlorine Fluxing Technologies in�...
Development and Demonstration of a Molten Metal Cooling Trough ...
This paper presents a new technology that allows cooling molten metal directly into ... Metallurgical Performance of Salt and Chlorine Fluxing Technologies in�...
Fast Pyrolysis of Poplar Using a Captive Sample Reactor: Effects of Inorganic Salts on Primary Pyrolysis Products
We have constructed a captive sample reactor (CSR) to study fast pyrolysis of biomass. The reactor uses a stainless steel wire mesh to surround biomass materials with an isothermal environment by independent controlling of heating rates and pyrolysis temperatures. The vapors produced during pyrolysis are immediately entrained and transported in He carrier gas to a molecular beam mass spectrometer (MBMS). Formation of secondary products is minimized by rapidly quenching the sample support with liquid nitrogen. A range of alkali and alkaline earth metal (AAEM) and transition metal salts were tested to study their effect on composition of primary pyrolysis products. Multivariate curve resolution (MCR) analysis of the MBMS data shows that transition metal salts enhance pyrolysis of carbohydrates and AAEM salts enhances pyrolysis of lignin. This was supported by performing similar separate studies on cellulose, hemicellulose and extracted lignin. The effect of salts on char formation is also discussed.
Mukarakate, C.; Robichaud, D.; Donohoe, B.; Jarvis, M.; Mino, K.; Bahng, M. K.; Nimlos, M.
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.
Buyers, A.G.; Rosen, F.D.; Motta, E.E.
thermal energy storage tanks Sandia has developed a heat transfer fluid (HTF) for use at elevated temperatures that has a lower freezing point
October 1977-September 1978, Argonne National Lab Report 78-New York, 1976, p. 473. Argonne National Laboratory, AnnualOctober 1978-March 1979, Argonne National Lab Report No. 70-
Stability of Molten Core Materials
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.
Layne Pincock; Wendell Hintze
Pre-Conceptual Design of a Fluoride-Salt-Cooled Small Modular Advanced High Temperature Reactor (SmAHTR)
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.
Greene, Sherrell R [ORNL; Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Carbajo, Juan J [ORNL; Ilas, Dan [ORNL; Cisneros, Anselmo T [ORNL; Varma, Venugopal Koikal [ORNL; Corwin, William R [ORNL; Wilson, Dane F [ORNL; Yoder Jr, Graydon L [ORNL; Qualls, A L [ORNL; Peretz, Fred J [ORNL; Flanagan, George F [ORNL; Clayton, Dwight A [ORNL; Bradley, Eric Craig [ORNL; Bell, Gary L [ORNL; Hunn, John D [ORNL; Pappano, Peter J [ORNL; Cetiner, Mustafa Sacit [ORNL
Generation -IV Reactor Concepts
Generation-IV Reactor Concepts Generation-IV Reactor Concepts Thomas H. Fanning Argonne National Laboratory 9700 South Cass Avenue Argonne, Illinois 60439, USA The Generation-IV International Forum (GIF) is a multi-national research and development (R&D) collaboration. The GIF pursues the development of advanced, next generation reactor technology with goals to improve: a) sustainability (effective fuel utilization and minimization of waste) b) economics (competitiveness with respect to other energy sources) c) safety and reliability (e.g., no need for offsite emergency response), and d) proliferation resistance and physical protection The GIF Technology Roadmap exercise selected six generic systems for further study: the Gas- cooled Fast Reactor (GFR), the Lead-cooled Fast Reactor (LFR), the Molten Salt Reactor (MSR),
Burnup Analysis of Thorium-Uranium Based Molten Salt Blanket in a Fusion-Fission Hybrid Reactor
Fusion Technologies: Heating and Fueling / Proceedings of the Twentieth Topical Meeting on the Technology of Fusion Energy (TOFE-2012) (Part 2) Nashville, Tennessee, August 27-31, 2012
Jing Zhao; Yongwei Yang; Sicong Xiao; Zhiwei Zhou
Experimental and Analytical Simulation of MFCI (Molten Fuel Coolant Interaction) during CDA (Core Disruptive Accident) in Sodium Cooled Fast Reactor.
??With increasing demand for understanding Severe Accident Scenario in Sodium Cooled Fast Reactors, there is an urgent need of enhancing numerical and experimental simulation techniques.� (more)
Natarajan, Venkataraman
Ions on the Electrodeposition of Titanium in Molten Fluoride-chloride ...
Electrochemical Behavior of Calcium-Lead Alloys in Molten Salt Electrolytes ... on the Corrosion of Ni-Cased Alloys (NiCrW and NiCrMo) in Molten Fluorides.
Optical properties of a solar-absorbing molten salt heat transfer fluid. [Eutectic mixture of KNO3, NaNO2, and NaNO3 with particle suspensions of cobalt oxides or copper oxides
The optical absorption properties of a high temperature molten salt heat transfer fluid were measured from 0.35 ..mu..m to 2.5 ..mu..m using both hemispherical transmission and reflection techniques. This fluid has application as a direct-absorbing working fluid in a high temperature central receiver solar energy facility. The absorption spectrum of the pure molten fluid--a eutectic mixture of KNO/sub 3/, NaNO/sub 2/, and NaNO/sub 3/, known as Hitec (Du Pont trade name)--displays a fundamental absorption edge near 410 nm, which was found to shift to longer wavelength linearly with temperature. Throughout the remainder of the visible spectrum, the fluid is transparent. To enhance its solar absorption, particulate metallic oxides of Co or Cu were introduced into the fluid. Absorption spectra of these oxide particle suspensions in the molten salt were determined as a function of dopant concentration ranging from 0 to 0.1 wt% metal nitrate added to the Hitec. These measurements were carried out at 200/sup 0/C under flow conditions to cause a homogeneous suspension of particles. Special transmission and reflection flow cells were designed and constructed to handle 200/sup 0/C fluids. The suspended particles cause an additional optical absorption throughout the visible spectrum which is characteristic of the particular metallic oxide and closely follows a Beer-Lambert concentration dependence. The solar averaged absorption in a fixed layer thickness was calculated for various concentrations of the fluid-oxide mixtures. The fluid without oxide particles absorbs approximately 8% of the solar spectrum per cm of path length. Addition of 0.1 wt% of Co(NO/sub 3/)/sub 2/.6H/sub 2/O increases this absorption to approximately 90% per cm. Of the oxides studied, Co/sub 3/O/sub 4/ particle suspensions offer better solar absorption characteristics than CuO. Effects of particulate scattering on the measurements are discussed.
Drotning, W.D.
Challenges in the Development of Advanced Reactors
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.
P. Sabharwall; M.C. Teague; S.M. Bragg-Sitton; M.W. Patterson
Waste disposal in underground salt cavities is considered. Theoretical Investigations for spherical and cylindrical cavities included analysis of elastic stress, thermal stress, and stress redistribution due to the development of a plastic zone around the cavity. The problems of temperature distribution and accompanying thermal stress, due to heat emission from the waste, were also studied. The reduction of the cavity volume, the development of the plastic zone, and the resulting stress redistribution around the cavity are presented as functions of cavity depth, internal pressure of cavity, strenzth of salt, and cavity teraperature rise. It is shown that a salt cavity can be designed such that it is structurally stable as a storage container assuming a chemical equilibrium can be established between the liquid waste and salt. (W.D.M.)
Serata, S.; Gloyna, E.F.
Investigation on Corrosion Behaviour of Ni-Based Alloys in Molten ...
In this paper, corrosion processes of Ni-based superalloys including Inconel 600, Hastelloy X and Hastelloy C-276 were investigated in molten fluoride salts�...
TEM Analysis of Incoloy 800H Exposed to Molten LiF-NaF-KF
However, molten salt corrosion is not as well understood as conventional aqueous corrosion. Focused Ion Beam machining was used to prepare site-
It is deemed feasible to store reactor fuel wastes in a salt dome cavity to a depth where the differential in pressure between the soil over-burden pressure and pressure of the fluid inside the cavity does not exceed 3000 psi, and the temperature is less than 400 deg F. Tests at pressure increments of 1000 psi were conducted on a 2" cylindrical cavity contained in a 6-in. long by 6-in. cylindrical salt core. Tests indicate that the cavity exhibited complete stability under pressures to 3000 psi and temperatures to 300 deg F. At temperatures of 100 to 400 deg F and pressures to 5000 psi continuous deformation of the cavity resulted. Initial movement of the salt was observed at all pressures. This was evidenced by vertical deformation and cavity size reduction. It was noted that a point of structural equilibrium was reached at lower temperatures when the pressure did not exceed 5000 psi. A literature study reveals that the most common type of cavity utilized in liquefied petroleum gas storage is either cylindrical or ellipsoidal. A few are pear or inverted cone shaped. There was no indication of leakage for cavities when pressure tested for as long as 72 hr. This indicates that the salt mass is not permeable under conditions of prevailing underground temperature and pressure. Salt specimens tested under atmospheric Pressure and temperature exhibited permeabilities of 0.1 to 0.2 millidarcys. The cost of completing underground storage cavities in salt masses is expected to be approximately 05 per barrel of storage space. (auth)
Brown, K.E.; Jessen, F.W.; Gloyna, E.F.
Molten carbonate fuel cell research at ORNL
The activities at ORNL during the period July 1976 to February 1977 on the molten carbonate fuel cell program, funded by the ERDA Division of Conservation Research and Technology, are summarized. This period marks the initiation of molten carbonate fuel cell research at ORNL, making use of the extensive background of expertise and facilities in molten salt research. The activities described include a literature survey on molten carbonates, design, acquisition and installation of apparatus for experimental studies of molten carbonates, initial experiments on materials compatibility with molten carbonates, electrolysis experiments for the determination of transference numbers, and theoretical studies of transport behavior and the coupling of mass flows in molten carbonate mixtures. Significant accomplishments were the theoretical prediction of a possibly appreciable change in the alkali ion ratio at molten carbonate fuel cell electrodes, operated at high current densities, as a result of mobility differences of the alkali ions; design, construction and assembly of an electrolysis cell, and initiation of measurements of composition profiles in mixed alkali carbonate electrolytes; initiation of differential scanning calorimetry of pure alkali carbonates for quantitative measurement of transition enthalpies, eventually leading to new, more reliable values of the enthalpies and free energies of formation of the pure and mixed carbonates.
Braunstein, J.; Bronstein, H. R.; Cantor, S.; Heatherly, D.; Vallet, C. E.
An Overview of Liquid Fluoride Salt Heat Transport Technology
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.
Cetiner, Mustafa Sacit [ORNL; Holcomb, David Eugene [ORNL
Fluoride-Salt-Cooled High-Temperature Reactor (FHR) with Silicon-Carbide-Matrix Coated-Particle Fuel
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.
Forsberg, C. W. [Massachusetts Institute of Technology (MIT); Terrani, Kurt A [ORNL; Snead, Lance Lewis [ORNL; Katoh, Yutai [ORNL
An Atomistic Study of the Structure and Thermodynamics of Molten ...
The molten salt-mixtures of ionic (NaCl, KCl etc.) and covalent (AlCl3, ZnCl2, etc.) chlorides are proposed as potential candidate materials, which can offer the�...
Aurivillius phases of PbBi{sub 4}Ti{sub 4}O{sub 15} doped with Mn{sup 3+} synthesized by molten salt technique: Structure, dielectric, and magnetic properties
Doping of manganese (Mn{sup 3+}/Mn{sup 4+}) into the Aurivillius phase Pb{sub 1-x}Bi{sub 4+x}Ti{sub 4-x}Mn{sub x}O{sub 15} was carried out using the molten salt technique for various Mn concentrations (x=0, 0.2, 0.4, 0.6, 0.8, and 1). Single phase samples could be obtained in the composition range with x up to 0.6 as confirmed by X-ray and neutron diffraction analysis. Dielectric measurements show a peak at 801, 803, 813 and 850 K for samples with x=0, 0.2, 0.4, and 0.6, respectively, related to the ferroelectric transition temperature (T{sub c}). The main contribution of the in-plane polarization for x{=}0.4 the polarization originates from the dipole moment in the Ti(2)O{sub 6} layer. Mn doping in the Pb{sub 1-x}Bi{sub 4+x}Ti{sub 4-x}Mn{sub x}O{sub 15} does not show any long range magnetic ordering. -- Graphical abstract: The dipole moment of TiO{sub 6} dependence of x in Pb{sub 1-x}Bi{sub 4+x}Ti{sub 4-x}Mn{sub x}O{sub 15} (0{0.2. {yields} Ferromagnetic interactions show the contribution of mixed valence of Mn{sup 3+}/Mn{sup 4+}.
Zulhadjri; Prijamboedi, B. [Inorganic and Physical Chemistry Group, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesha No. 10, Bandung (Indonesia); Nugroho, A.A. [Magnetic and Photonic Physics Research-Group, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesha No. 10, Bandung (Indonesia); Mufti, N. [Physics Department, Universitas Negeri Malang, Jl. Surabaya 6, Malang 65145 (Indonesia); Fajar, A. [Centre for Technology of Nuclear Industry Materials - BATAN Puspiptek Serpong, Tangerang (Indonesia); Palstra, T.T.M. [Solid State Materials Laboratory, Zernike Institute for Advanced Materials, Rijksuniversiteit Groningen, Nijenborgh 4, 9747AG Groningen (Netherlands); Ismunandar, E-mail: email@example.com [Inorganic and Physical Chemistry Group, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesha No. 10, Bandung (Indonesia)
Effects of Coolant Temperature Changes on Reactivity for Various Coolants in a Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)
The purpose of this study is to perform an investigation into the relative merit of various salts and salt compounds being considered for use as coolants in the liquid salt cooled very high temperature reactor platform (LS-VHTR). Most of the non-nuclear properties necessary to evaluate these salts are known, but the neutronic characteristics important to reactor core design are still in need of a more extensive examination. This report provides a two-fold approach to further this investigation. First, a list of qualifying salts is assembled based upon acceptable non-nuclear properties. Second, the effect on system reactivity for a secondary system transient or an off-normal or accident condition is examined for each of these salt choices. The specific incident to be investigated is an increase in primary coolant temperature beyond normal operating parameters. In order to perform the relative merit comparison of each candidate salt, the System Temperature Coefficient of Reactivity is calculated for each candidate salt at various state points throughout the core burn history. (author)
Casino, William A. Jr. [AREVA - Framatome ANP, 3315 Old Forest Road OF-15, P.O. Box 10935, Lynchburg, VA 24506-0935 (United States)
Production of chlorine from chloride salts
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.
Salt Fluxes for Alkali and Alkaline Earth Element Removal from ...
Sep 1, 2001... for Alkali and Alkaline Earth Element Removal from Molten Aluminum ... Solid chloride salts containing MgC2 can be used to remove alkali�...
Separation and Recovery of Uranium Metal from Spent Light Water Reactor Fuel via Electrolytic Reduction and Electrorefining
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.
S. D. Herrmann; S. X. Li
Liquid surface skimmer apparatus for molten lithium and method
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.
Robinson, Samuel C. (Knoxville, TN); Pollard, Roy E. (Maryville, TN); Thompson, William F. (Oak Ridge, TN); Stark, Marshall W. (Gastonia, NC); Currin, Jr., Robert T. (Salisbury, NC)
Environmental impacts of nonfusion power systems. [Data on environmental effects of all power sources that may be competitive with fusion reactor power plants
Data were collected on the environmental effects of power sources that may be competitive with future fusion reactor power plants. Data are included on nuclear power plants using HTGR, LMBR, GCFR, LMFBR, and molten salt reactors; fossil-fuel electric power plants; geothermal power plants; solar energy power plants, including satellite-based solar systems; wind energy power plants; ocean thermal gradient power plants; tidal energy power plants; and power plants using hydrogen and other synthetic fuels as energy sources.
Brouns, R.J.
An Overview of Liquid Fluoride Salt Heat Transport Systems
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 this report also describes more recently developed technologies such as dry gas seals. This report also provides a high-level, parametric evaluation of LSHT loop performance to allow general intercomparisons between heat-transport fluid options as well as provide an overview of the properties and requirements for a representative loop. A compilation of relevant thermophysical properties of useful fluoride salts is also included for salt heat transport systems. Fluoride salts can be highly corrosive depending on the container materials selected, the salt chemistry, and the operating procedures used. The report includes an overview of the state-of-the-art in reduction-oxidation chemistry control methodologies employed to minimize corrosion issues. Salt chemistry control technology, however, remains at too low a level of understanding for widespread industrial usage. Loop operational issues such as start-up procedures and system freeze-up vulnerability are also discussed. 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 over
Holcomb, David Eugene [ORNL; Cetiner, Mustafa Sacit [ORNL
Method for continuously recovering metals using a dual zone chemical reactor
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.
Bronson, M.C.
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.
Bronson, Mark C. (Livermore, CA)
Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR)
A study has been completed to develop a new baseline core design for the liquid-salt-cooled very high-temperature reactor (LS-VHTR) that is better optimized for liquid coolant and that satisfies the top-level operational and safety targets, including strong passive safety performance, acceptable fuel cycle parameters, and favorable core reactivity response to coolant voiding. Three organizations participated in the study: Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), and Argonne National Laboratory (ANL). Although the intent was to generate a new reference LS-VHTR core design, the emphasis was on performing parametric studies of the many variables that constitute a design. The results of the parametric studies not only provide the basis for choosing the optimum balance of design options, they also provide a valuable understanding of the fundamental behavior of the core, which will be the basis of future design trade-off studies. A new 2400-MW(t) baseline design was established that consists of a cylindrical, nonannular core cooled by liquid {sup 7}Li{sub 2}BeF{sub 4} (Flibe) salt. The inlet and outlet coolant temperatures were decreased by 50 C, and the coolant channel diameter was increased to help lower the maximum fuel and vessel temperatures. An 18-month fuel cycle length with 156 GWD/t burnup was achieved with a two-batch shuffling scheme, while maintaining a core power density of 10 MW/m{sup 3} using graphite-coated uranium oxicarbide particle fuel enriched to 15% {sup 235}U and assuming a 25 vol-% packing of the coated particles in the fuel compacts. The revised design appears to have excellent steady-state and transient performance. The previous concern regarding the core's response to coolant voiding has been resolved for the case of Flibe coolant by increasing the coolant channel diameter and the fuel loading. Also, the LSVHTR has a strong decay heat removal performance and appears capable of surviving a loss of forced circulation (LOFC) even with failure to scram. Significant natural convection of the coolant salt occurs, resulting in fuel temperatures below steady-state values and nearly uniform temperature distributions during the transient.
Ingersoll, DT
Electrochromic Salts, Solutions, and Devices
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.
Burrell, Anthony K. (Los Alamos, NM); Warner, Benjamin P. (Los Alamos, NM); McClesky, T. Mark (Los Alamos, NM)
Electrochromic salts, solutions, and devices
Burrell, Anthony K. (Los Alamos, NM); Warner, Benjamin P. (Los Alamos, NM); McClesky,7,064,212 T. Mark (Los Alamos, NM)
(ODS) Ferritic Steel in Molten Fluoride Salts
About this Abstract. Meeting, 2010 TMS Annual Meeting & Exhibition. Symposium , Advanced Materials and Fuels Enabling Future Fusion, Fission and Hybrid�...
Molten salt fuels with high plutonium solubility
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.
Moir, Ralph W; Turchi, Patrice E.A.; Shaw, Henry F; Kaufman, Larry
Molten Glass for Thermal Storage: Advanced Molten Glass for Heat Transfer and Thermal Energy Storage
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.
Fundamental Properties of Salts
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.
Toni Y Gutknecht; Guy L Fredrickson
Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)
A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the equivalent temperature of heat delivered to either the power conversion system or a hydrogen production plant. Using a comparative cost analysis, the construction costs per unit output are projected to be 50-55% of the costs for modular gas-cooled or sodium-cooled reactor systems. This is primarily a consequence of substantially larger power output and higher conversion efficiency for the AHTR. The AHTR has a number of unique technical challenges in meeting the NGNP requirements; however, it appears to offer advantages over high-temperature helium-cooled reactors and provides an alternative development path to achieve the NGNP requirements. Primary challenges include optimizing the core design for improved response to transients, designing an internal blanket to thermally protect the reactor vessel, and engineering solutions to high-temperature refueling and maintenance.
Ingersoll, D.T.
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.
Buyers, A.G.
Electrodeposition of molten silicon
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.
De Mattei, Robert C. (Sunnyvale, CA); Elwell, Dennis (Palo Alto, CA); Feigelson, Robert S. (Saratoga, CA)
J. Plasma Fusion Res. SERIES, Vol. 10 (2013) Flibe-Tritium Research for Fission or Fusion Reactors at Kyushu University
There is increasing interest in using ionic molten-salt Flibe not only as self-cooled tritium(T)-breeding material in a fusion reactor blanket but also as fuel solvent of molten-salt fission reactors. Application of Flibe to T-breeding fluid for a stellarator-type fusion reactor operated at a high magnetic field brings large simplification of its blanket structure, allowing continuous operation under high-beta plasma conditions. Using mixed Flibe-ThF 4+UF 4 fuel in molten salt fission reactors permits stable long-term operation without fuel exchange. When Flibe or Flinak is irradiated by neutrons, however, acid and corrosive TF is generated, and some T permeates through structural walls. In order to solve these problems, chemical conditions of Flibe are changed using the redox-control reaction, Be+2TF=BeF 2+T 2. In addition, permeation of hydrogen isotopes is lowered by enhancing T recovery rates. Part of Flibe-tritium researches are performed at Idaho National Laboratory (INL) under the Japan-US collaboration work of JUPITER-II. Our own contributions to the topics are shortly introduced in this paper.
Satoshi Fukada
Cycle thorium et r�acteurs � sel fondu. Exploration du champ des param�tres et des contraintes d�finissant le "Thorium Molten Salt Reactor".
??Le recours � l'�nergie �lectronucl�aire pour diminuer les �missions anthropiques de CO2 n�cessite des avanc�es technologiques majeures. Les r�acteurs nucl�aires de IVe g�n�ration doivent r�pondre� (more)
Mathieu, Ludovic
The possibility of fuel cycle design for ABC/ATW complex with molten fuel on LiF-BeF2 basis
The experience gained in the field of the development of molten salt reactors (MSR) can be made a basis of chemical processing of the ABC/ATW liquid fuel. The following combination of two processing principles are proposed for the ABC/ATW fuel (LiF-BeF2-PuF3,(4)-MAFn): -continious removal of radioactive gases, volatile impurities and 'noble fission products'; -portion-by-portion electrochemical processing with removal of rare earth elements and some other fission products at an autonomous plant. After processing the fuel salt is brought back to the blanket of the ABC/ATW complex. The analysis of information previously published in different countries allows for a safe assumption that the ABC/ATW fuel cycle with liquid fuel salt is feasible and can be demonstrated experimentally.
Naumov, V. S.; Bychkov, A. V. [Federal Scientific Center of Russia Research Institute of Atomic Reactors (RIAR) Russia, Dimitrovgrad 433510 (Russian Federation)
Configurational Entropy and Structure of the Molten NaCl-KCl-ZnCl2 ...
In this context, we examine NaCl-KCl-ZnCl2 molten salts and pay particular attention to characterizing the thermodynamics and structure of these liquids in order�...
Management of Salt Waste from Electrochemical Processing of Used Nuclear Fuel
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.
Michael F. Simpson; Michael N. Patterson; Joon Lee; Yifeng Wang; Joshua Versey; Ammon Williams; Supathorn Phongikaroon; James Allensworth; Man-Sung Yim
Modeling of Molten Core Concrete Interactions and Fission Product Release
The study of molten core concrete interactions is important in estimating the possible consequences of a severe nuclear reactor accident. CORCON-Mod2 is a computer program that models the thermal, chemical, and physical phenomena associated with molten core concrete interactions. Models have been added to extend the modeling of these phenomena. An ideal solution chemical equilibrium methodology predicts the fission product vaporization release. Additional chemical species have been added, and the calcula...
Oak Ridge Reservation Facilities
processed for shipment to the Nevada Test Site or other appropriate disposal facility. Molten Salt Reactor Experiment Facility The Molten Salt Reactor Experiment (MSRE) operated...
An Account of Oak Ridge National Laboratory's Thirteen Research Reactors
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.
Rosenthal, Murray Wilford [ORNL
An Assessment of Reactor Types for Thermochemical Hydrogen Production
Nuclear energy has been proposed as a heat source for producing hydrogen from water using a sulfur-iodine thermochemical cycle. This document presents an assessment of the suitability of various reactor types for this application. The basic requirement for the reactor is the delivery of 900 C heat to a process interface heat exchanger. Ideally, the reactor heat source should not in itself present any significant design, safety, operational, or economic issues. This study found that Pressurized and Boiling Water Reactors, Organic-Cooled Reactors, and Gas-Core Reactors were unsuitable for the intended application. Although Alkali Metal-Cooled and Liquid-Core Reactors are possible candidates, they present significant development risks for the required conditions. Heavy Metal-Cooled Reactors and Molten Salt-Cooled Reactors have the potential to meet requirements, however, the cost and time required for their development may be appreciable. Gas-Cooled Reactors (GCRs) have been successfully operated in the required 900 C coolant temperature range, and do not present any obvious design, safety, operational, or economic issues. Altogether, the GCRs approach appears to be very well suited as a heat source for the intended application, and no major development work is identified. This study recommends using the Gas-Cooled Reactor as the baseline reactor concept for a sulfur-iodine cycle for hydrogen generation.
ASME Material Challenges for Advanced Reactor Concepts
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.
Piyush Sabharwall; Ali Siahpush
separate effects test steam generators small modular reactorNuclear Generating Station (SONGS) steam generators (SG).January of 2012, a steam generator tube leak was detected,
Nuclear Energy Governance and the Politics of Social Justice: Technology, Public Goods, and Redistribution in Russia and France
Reactors (SWCRs) Thermal-Molten Salt Reactors (MSRs) Fast-water, heavy water, gas, molten salt and liquid metal formreactors using gas, molten salt or liquid metal as their
Grigoriadis, Theocharis N
Monte-Carlo Modeling of Parameters of a Subcritical Cascade Reactor Based on MSBR and LMFBR Technologies
Parameters of a subcritical cascade reactor driven by a proton accelerator and based on a primary lead-bismuth target, main reactor constructed analogously to the molten salt breeder (MSBR) reactor core and a booster-reactor analogous to the core of the BN-350 liquid metal cooled fast breeder reactor (LMFBR). It is shown by means of Monte-Carlo modeling that the reactor under study provides safe operation modes (k_{eff}=0.94-0.98), is apable to transmute effectively radioactive nuclear waste and reduces by an order of magnitude the requirements on the accelerator beam current. Calculations show that the maximal neutron flux in the thermal zone is 10^{14} cm^{12}\\cdot s^_{-1}, in the fast booster zone is 5.12\\cdot10^{15} cm^{12}\\cdot s{-1} at k_{eff}=0.98 and proton beam current I=2.1 mA.
Bznuni, S A; Zhamkochyan, V M; Polanski, A; Sosnin, A N; Khudaverdyan, A H
SALT-ANL. Systems Analysis Process Simulation
SALT (Systems Analysis Language Translator), a systems- analysis and process-simulation program for steady-state and dynamic systems, can also be used for optimization and sensitivity studies. SALT employs state-of-the-art numerical techniques including a hybrid steepest-descent/quasi-Newtonian multidimensional nonlinear equation solver, sequential quadratic programming methods for optimization, and multistep integration methods for both stiff and nonstiff systems of differential equations. Based on a preprocessor concept where a `new` system driver can be written for each application, SALT-ANL contains precompiled component models, several flow types, and a number of thermodynamic and transport property routines, including a gas chemical-equilibrium code. It has been applied to the study of open-cycle and liquid-metal magnetohydrodynamic systems, fuel cells, ocean thermal energy conversion, municipal solid-waste processing, fusion, breeder reactors, and geothermal and solar-energy systems. Models available include: combustor, compressor, deaerator, gas-diffuser, fuel-dryer, feedwater-heater, flash-tank, gas-turbine, heater, heat-exchanger, flow-initiator, fuel-flow-initiator, molten-carbonate fuel-cell, liquid-metal diffuser, magnetohydrodynamic-generator, liquid-metal magnetohydrodynamic-generator, liquid-metal nozzle, liquid-metal pipe, flow-mixer, gas-nozzle, phosphoric acid fuel-cell, pump, pipe-calculator, steam-condenser, steam-drum, liquid-gas separator, stack, solid-oxide fuel-cell, flow-splitter, steam-turbine, two-phase diffuser, two-phase mixer, and two-phase nozzle. Input data to the SALT program describe the system configuration for the specific problem to be analyzed and provide instructions defining system constraints, objective functions, parameter sweeps, etc. to generate a PL/I program representing the system problem and performing the various analytic tasks.
Berry, G.F.; Geyer, H.K. [Argonne National Lab., IL (United States)
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.
Gruen, D.M.
Apparatus for controlling molten core debris
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.
Golden, Martin P. (Trafford, PA); Tilbrook, Roger W. (Monroeville, PA); Heylmun, Neal F. (Pittsburgh, PA)
Apparatus for controlling molten core debris. [LMFBR
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.
Golden, M.P.; Tilbrook, R.W.; Heylmun, N.F.
Cleanup of plutonium oxide reduction black salts
This work describes pyrochemical processes employed to convert direc oxide reduction (DOR) black salts into discardable white salt and plutonium metal. The DOR process utilizes calcium metal as the reductant in a molten calcium chloride solvent salt to convert plutonium oxide to plutonium metal. An insoluble plutonium-rich dispersion called black salt sometimes forms between the metal phase and the salt phase. Black salts accumulated for processing were treated by one of two methods. One method utilized a scrub alloy of 70 wt % magnesium/30 wt % zinc. The other method utilized a pool of plutonium metal to agglomerate the metal phase. The two processes were similar in that calcium metal reductant and calcium chloride solvent salt were used in both cases. Four runs were performed by each method, and each method produced greater than 93% conversion of the black salt.
Giebel, R.E.; Wing, R.O.
Method for Making a Uranium Chloride Salt Product
The subject apparatus provides a means to produce UCl3, in large quantities without incurring corrosion of the containment vessel or associated apparatus. Gaseous Cl is injected into a lower layer of Cd where CdCl2 is formed. Due to is lower density, the CdCl2 rises through the Cd layer into a layer of molten LiCl-KCL salt where a rotatable basket containing uranium ingots is suspended. The CdCl2 reacts with the uranium to form UCl, and Cd. Due to density differences, the Cd sinks down to the liquid Cd layer and is reused. The UCl3 combines with the molten salt. During production the temperature is maintained at about 600 degrees 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% UCl3 is solidified.
Miller, William F.; Tomczuk, Zygmunt
Evaluation of potential for MSRE spent fuel and flush salt storage and treatment at the INEL
The potential for interim storage as well as for treatment of the Molten Salt Reactor Experiment spent fuel at INEL has been evaluated. Provided that some minimal packaging and chemical stabilization prerequisites are satisfied, safe interim storage of the spent fuel at the INEL can be achieved in a number of existing or planned facilities. Treatment by calcination in the New Waste Calcining Facility at the INEL can also be a safe, effective, and economical alternative to treatment that would require the construction of a dedicated facility. If storage at the INEL is chosen for the Molten Salt Reactor Experiment (MSRE) spent fuel salts, their transformation to the more stable calcine solid would still be desirable as it would result in a lowering of risks. Treatment in the proposed INEL Remote-Handled Immobilization Facility (RHIF) would result in a waste form that would probably be acceptable for disposal at one of the proposed national repositories. The cost increment imputable to the treatment of the MSRE salts would be a small fraction of the overall capital and operating costs of the facility or the cost of building and operating a dedicated facility. Institutional and legal issues regarding shipments of fuel and waste to the INEL are summarized. The transfer of MSRE spent fuel for interim storage or treatment at the INEL is allowed under existing agreements between the State of idaho and the Department of energy and other agencies of the Federal Government. In contrast, current agreements preclude the transfer into Idaho of any radioactive wastes for storage or disposal within the State of Idaho. This implies that wastes and residues produced from treating the MSRE spent fuel at locations outside Idaho would not be acceptable for storage in Idaho. Present agreements require that all fuel and high-level wastes stored at the INEL, including MSRE spent fuel if received at the INEL, must be moved to a location outside Idaho by the year 2035.
Ougouag, A.M.; Ostby, P.A.; Nebeker, R.L.
Rapid quenching of molten lithium-aluminum jets in water
Control rods for the K production reactor at Savannah River, are grouped in assemblies of seven rods, called ``septifoils``. A problem area is that overheated cooling rods for these control rods might partially melt, with the resulting molten metal draining into the water at the bottom. Experiments were conducted in which up to 1 kg molten alloy was contacted with water at a time. Conditions were varied in an attempt to include those factors that might trigger a vapor explosion. Results indicate that a steam explosion that would damage the septifoil is unlikely.
Greene, G.A. [Brookhaven National Lab., Upton, NY (United States); Cho, D.H. [Argonne National Lab., IL (United States); Hyder, M.L.; Allison, D.K. [Westinghouse Savannah River Co., Aiken, SC (United States); Ellison, P.G. [Idaho National Engineering Lab., Idaho Falls, ID (United States)
Very High Efficiency Reactor (VHER) Concepts for Electrical Power Generation and Hydrogen Production
The goal of the Very High Efficiency Reactor study was to develop and analyze concepts for the next generation of nuclear power reactors. The next generation power reactor should be cost effective compared to current power generation plant, passively safe, and proliferation-resistant. High-temperature reactor systems allow higher electrical generating efficiencies and high-temperature process heat applications, such as thermo-chemical hydrogen production. The study focused on three concepts; one using molten salt coolant with a prismatic fuel-element geometry, the other two using high-pressure helium coolant with a prismatic fuel-element geometry and a fuel-pebble element design. Peak operating temperatures, passive-safety, decay heat removal, criticality, burnup, reactivity coefficients, and material issues were analyzed to determine the technical feasibility of each concept.
Molten carbonate fuel cell
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.
Kaun, Thomas D. (New Lenox, IL); Smith, James L. (Lemont, IL)
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.
Kaun, T.D.; Smith, J.L.
MOLTEN METAL REACTORS - Energy Innovation Portal
Bingham, Dennis N. (Idaho Falls, ID), Klingler, Kerry M. (Idaho Falls, ID), Turner, Terry D. (Idaho Falls, ID), Wilding, Bruce M. (IdahoFalls, ID) ...
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.
Roman, W.G.
Noncentrosymmetric salt inclusion oxides: Role of salt lattices and counter ions in bulk polarity
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.
West, J. Palmer [Department of Chemistry, Clemson University, Clemson, SC 29634-0973 (United States)] [Department of Chemistry, Clemson University, Clemson, SC 29634-0973 (United States); Hwu, Shiou-Jyh, E-mail: firstname.lastname@example.com [Department of Chemistry, Clemson University, Clemson, SC 29634-0973 (United States)] [Department of Chemistry, Clemson University, Clemson, SC 29634-0973 (United States)
Neutronics and Depletion Methods for Parametric Studies of Fluoride Salt Cooled High Temperature Reactors with Slab Fuel Geometry and Multi-Batch Fuel Management Schemes
The Advanced High Temperature Reactor (AHTR) is a 3400 MWth fluoride salt cooled high temperature reactor (FHR) that uses TRISO particle fuel compacted into slabs rather than spherical fuel pebbles or cylindrical fuel compacts. Simplified methods are required for parametric design studies such that analyzing the entire feasible design space for an AHTR is tractable. These simplifications include fuel homogenization techniques to increase the speed of neutron transport calculations in depletion analysis and equilibrium depletion analysis methods to analyze systems with multi-batch fuel management schemes. This paper presents three elements of significant novelty. First, the reactivity-equivalent physical transformation (RPT) methodology usually applied in systems with coated particle fuel in cylindrical and spherical geometries was extended to slab geometries. Secondly, based on this newly developed RPT method for slab geometries, a methodology that uses Monte Carlo depletion approaches was further developed to search for the maximum discharge burnup in a multi-batch system by iteratively estimating the beginning of equilibrium cycle composition and sampling different discharge burnups. This iterative equilibrium depletion search (IEDS) method fully defines an equilibrium fuel cycle (keff, power, flux and composition evolutions across space and time), but is computationally demanding, although feasible on single-processor workstations. Finally, an analytical method, the non-linear reactivity model, was developed by expanding the linear reactivity model to include an arbitrary number of higher order terms to extrapolate single-batch depletion results to estimate the maximum discharge burnup and BOEC keff in systems with multi-batch fuel management schemes. Results from this method were benchmarked against equilibrium depletion analysis results using the IEDS.
Cisneros, Anselmo T. [University of California, Berkeley; Ilas, Dan [ORNL
The Advanced High-Temperature Reactor (AHTR) is a 3400 MWth fluoride-salt-cooled high-temperature reactor (FHR) that uses TRISO particle fuel compacted into slabs rather than spherical or cylindrical fuel compacts. Simplified methods are required for parametric design studies such that analyzing the entire feasible design space for an AHTR is tractable. These simplifications include fuel homogenization techniques to increase the speed of neutron transport calculations in depletion analysis and equilibrium depletion analysis methods to analyze systems with multi-batch fuel management schemes. This paper presents three elements of significant novelty. First, the Reactivity-Equivalent Physical Transformation (RPT) methodology usually applied in systems with coated-particle fuel in cylindrical and spherical geometries has been extended to slab geometries. Secondly, based on this newly developed RPT method for slab geometries, a methodology that uses Monte Carlo depletion approaches was further developed to search for the maximum discharge burnup in a multi-batch system by iteratively estimating the beginning of equilibrium cycle (BOEC) composition and sampling different discharge burnups. This Iterative Equilibrium Depletion Search (IEDS) method fully defines an equilibrium fuel cycle (keff, power, flux, and composition evolutions) but is computationally demanding, although feasible on single-processor workstations. Finally, an analytical method, the Non-Linear Reactivity Model, was developed by expanding the linear reactivity model to include an arbitrary number of higher order terms so that single-batch depletion results could be extrapolated to estimate the maximum discharge burnup and BOEC keff in systems with multi-batch fuel management schemes. Results from this method were benchmarked against equilibrium depletion analysis results using the IEDS.
Design report on SCDAP/RELAP5 model improvements - debris bed and molten pool behavior
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.
Allison, C.M.; Rempe, J.L.; Chavez, S.A.
A Feasibility Study of Steelmaking by Molten Oxide Electrolysis (TRP9956)
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.
Donald R. Sadoway; Gerbrand Ceder
Status of the Integral Fast Reactor fuel cycle demonstration and waste management practices
Over the past few years, Argonne National Laboratory has been preparing for the demonstration of the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety and operations, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle, which will be demonstrated at Argonne-West in Idaho, employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The required facility modifications and process equipment for the demonstration are nearing completion. Their status and the results from initial fuel fabrication work, including the waste management aspects, are presented. Additionally, estimated compositions of the various process waste streams have been made, and characterization and treatment methods are being developed. The status of advanced waste processing equipment being designed and fabricated is described.
Benedict, R.W.; Goff, K.M.; McFarlane, H.F.
Lithium-ferrate-based cathodes for molten carbonate fuel cells
Argonne National Laboratory is developing advanced cathodes for pressurized operation of the molten carbonate fuel cell (MCFC) at approximately 650 degrees Centigrade. These cathodes are based on lithium ferrate (LiFeO[sub 2]) which is attractive because of its very low solubility in the molten (Li,K)[sub 2]CO[sub 3] electrolyte. Because of its high resistivity, LiFeO[sub 2] cannot be used as a direct substitute for NiO. Cation substitution is, therefore, necessary to decrease resistivity. The effect of cation substitution on the resistivity and deformation of LiFeO[sub 2] was determined. The substitutes were chosen because their respective oxides as well as LiFeO[sub 2] crystallize with the rock-salt structure.
Lanagan, M.T.; Wolfenstine, J. [Argonne National Lab., IL (United States). Energy Technology Div.; Bloom, I.; Kaun, T.D.; Krumpelt, M. [Argonne National Lab., IL (United States). Chemical Technology Div.
Distribution of Calcium and Aluminum between Molten Silicon and ...
Presentation Title, Distribution of Calcium and Aluminum between Molten ... Electrochemical deposition of high purity silicon from molten fluoride electrolytes.
Three-dimensional modeling and simulation of vapor explosions in Light Water Reactors.
??Steam explosions can occur during a severe accident in light water nuclear reactors with the core melting as the consequence of interaction of molten core� (more)
Schr�der, Maxim
Waste removal in pyrochemical fuel processing for the Integral Fast Reactor
Electrorefining in a molten salt electrolyte is used in the Integral Fast Reactor fuel cycle to recover actinides from spent fuel. Processes that are being developed for removing the waste constituents from the electrorefiner and incorporating them into the waste forms are described in this paper. During processing, halogen, chalcogen, alkali, alkaline earth, and rare earth fission products build up in the molten salt as metal halides and anions, and fuel cladding hulls and noble metal fission products remain as metals of various particle sizes. Essentially all transuranic actinides are collected as metals on cathodes, and are converted to new metal fuel. After processing, fission products and other waste are removed to a metal and a mineral waste form. The metal waste form contains the cladding hulls, noble metal fission products, and (optionally) most rare earths in a copper or stainless steel matrix. The mineral waste form contains fission products that have been removed from the salt into a zeolite or zeolite-derived matrix.
Ackerman, J.P.; Johnson, T.R.; Laidler, J.J.
Recovery of protactinium from molten fluoride nuclear fuel compositions
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)
Baes, C.F. Jr.; Bamberger, C.; Ross, R.G.
Cathode-preparation method for molten-carbonate fuel cell
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.
Smith, J.L.; Sim, J.W.; Kucera, E.H.
MOLTEN SALT APPROACHES TO MITIGATE CLIMATE CHANGE Frank H. Shu
·· Catalysis ·· Electrolytes for electrochemical applications, heat transfer, and energy storage September 9, 2008. Inventors Sheng Dai1 and Huimin Luo2 1 Chemical Sciences Division 2 Nuclear Science
Williams, Gary A.
Behavior of Silicon Electrodepositing in Fluoride Molten Salts
About this Abstract. Meeting, 2012 TMS Annual Meeting & Exhibition. Symposium , Electrometallurgy 2012. Presentation Title, Behavior of Silicon�...
Electrochemistry for Nd electrowinning from fluoride-oxide molten salts
About this Abstract. Meeting, 2014 TMS Annual Meeting & Exhibition. Symposium , Rare Metal Extraction & Processing Symposium. Presentation Title�...
Molten Salts: Bath Chemistry and Process Design in Aluminum ...
ABOUT THE PRESENTERS. Donald R. Sadoway is a professor of materials chemistry in the Department of Materials Science and Engineering at the�...
C1 Technology of Molten salt Electrolysis of Magnesium Chloride
D14 Gold Nanoparticles in Red Ruby Glasses Used for Decoration in Thailand � D15 Soft Magnetic Properties of Nanocrystalline Fe-based P/M Cores Mixed�...
Electrochemical Deposition of High Purity Silicon in Molten Salts
The energy consumption was estimated to be less than 3 kWh/kg Si. Such a low energy requirement suggests that electrorefining by using repeated steps may�...
Material Constraints on Accelerator Driven Sub-Critical Molten Salt ...
... machines can be used for neutron spallation sources. Future materials advances in these machines can be expected to improve their operating efficiencies.
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
Reactor (SFR) * Gas-Cooled Fast Reactor (GFR) * Lead-Bismuth-Cooled Fast Reactor (LFR) * Molten Salt Reactor (MSR). While the international community will study all six concepts,...
Modeled Salt Density for Nuclear Material Estimation in the Treatment of Spent Nuclear Fuel
Spent metallic nuclear fuel is being treated in a pyrometallurgical process that includes electrorefining the uranium metal in molten eutectic LiCl-KCl as the supporting electrolyte. We report a model for determining the density of the molten salt. Inventory operations account for the net mass of salt and for the mass of actinides present. It was necessary to know the molten salt density but difficult to measure, and it was decided to model the salt density for the initial treatment operations. The model assumes that volumes are additive for the ideal molten salt solution as a starting point; subsequently a correction factor for the lanthanides and actinides was developed. After applying the correction factor, the percent difference between the net salt mass in the electrorefiner and the resulting modeled salt mass decreased from more than 4.0% to approximately 0.1%. As a result, there is no need to measure the salt density at 500 C for inventory operations; the model for the salt density is found to be accurate.
DeeEarl Vaden; Robert. D. Mariani
Nuclear Reactor Severe Accident Experiments
Nuclear Reactor Severe Accident Experiments Nuclear Reactor Severe Accident Experiments Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Nuclear Reactor Severe Accident Experiments 1 2 3 4 5 6 7 We perform experiments simulating reactor core melt phenomena in which molten core debris ("corium") erodes the concrete floor of a containment building. This occurred during the Fukushima nuclear power plant accident though the extent of concrete damage is yet unknown. This video shows the top view of a churning molten pool of uranium oxide at 2000°C (3600°F) seen during an experiment at Argonne. Corium behaves much like lava.
Dissolution Behavior of Rhodium into Molten Slag
Determination of FeO Containing Liquid Slag Surface Tensions Using the Sessile Drop Method � Dissolution Behavior of Rhodium into Molten Slag.
Technology Development Roadmap for the Advanced High Temperature Reactor Secondary Heat Exchanger
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 research efforts on the near-term qualification, selection, or maturation strategy as detailed in this report. Development of the integration methodology feasibility study, along with research and development (R&D) needs, are ongoing tasks that will be covered in the future reports as work progresses. Section 2 briefly presents the integration of AHTR technology with conventional chemical industrial processes., See Idaho National Laboratory (INL) TEV-1160 (2011) for further details
P. Sabharwall; M. McCllar; A. Siahpush; D. Clark; M. Patterson; J. Collins
Modular Accident Analysis Program, Version 5, Molten Corium�Concrete Interaction and Debris Coolability Model Enhancement Description
This report describes proposed enhancements to the Modular Accident Analysis Program (MAAP) molten corium�concrete interaction (MCCI) model. MAAP is a computer program that simulates the operation of light-water and heavy-water moderated nuclear power plants for both current and advanced light-water reactor designs.Engineers at Fukushima observed that water pumped into the reactor vessel rose to a certain height, but it did not rise further as more water was pumped into the reactor ...
Potential applications of helium-cooled high-temperature reactors to process heat use
High-Temperature Gas-Cooled Reactors (HTRs) permit nuclear energy to be applied to a number of processes presently utilizing fossil fuels. Promising applications of HTRs involve cogeneration, thermal energy transport using molten salt systems, steam reforming of methane for production of chemicals, coal and oil shale liquefaction or gasification, and - in the longer term - energy transport using a chemical heat pipe. Further, HTRs might be used in the more distant future as the energy source for thermochemical hydrogen production from water. Preliminary results of ongoing studies indicate that the potential market for Process Heat HTRs by the year 2020 is about 150 to 250 GW(t) for process heat/cogeneration application, plus approximately 150 to 300 GW(t) for application to fossil conversion processes. HTR cogeneration plants appear attractive in the near term for new industrial plants using large amounts of process heat, possibly for present industrial plants in conjunction with molten-salt energy distribution systems, and also for some fossil conversion processes. HTR reformer systems will take longer to develop, but are applicable to chemicals production, a larger number of fossil conversion processes, and to chemical heat pipes.
Gambill, W.R.; Kasten, P.R.
Measurement of Thermodynamic Properties of Tellurium in Molten ...
... dissolution of tellurium gas (Te2) into molten iron by equilibrating molten iron in ... Quantifying the Export Flow of Used Electronics from the United States: The�...
The Influence of Lewis Acid/Base Chemistry on the Removal of Gallium by Volatility from Weapons-Grade Plutonium Dissolved in Molten Chlorides
It has been proposed that GaCl{sub 3} can be removed by direct volatilization from a Pu-Ga alloy that is dissolved in a molten chloride salt. Although pure GaCl{sub 3} is quite volatile (boiling point: 201 deg. C), the behavior of GaCl{sub 3} dissolved in chloride salts is quite different because of solution effects and is critically dependent upon the composition of the solvent salt (i.e., its Lewis acid/base character). In this technical note, the behavior of gallium in prototypical Lewis acid and Lewis base salts is contrasted. It is found that gallium volatility is suppressed in basic melts and promoted in acidic melts. These results have an important influence on the potential for simple gallium removal in molten salt systems.
Williams, David F.; Cul, Guillermo D. del [Oak Ridge National Laboratory (United States); Toth, Louis M. [Electrochemical Systems (United States); Collins, Emory D. [Oak Ridge National Laboratory (United States)
Method for making a uranium chloride salt product
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.
Miller, William E. (Naperville, IL); Tomczuk, Zygmunt (Lockport, IL)
Selective Reduction of Active Metal Chlorides from Molten LiCl-KCl using Lithium Drawdown
In support of optimizing electrorefining technology for treating spent nuclear fuel, lithium drawdown has been investigated for separating actinides from molten salt electrolyte. Drawdown reaction selectivity is a major issue that needs to be investigated, since the goal is to remove actinides while leaving the fission products in the salt. A series of lithium drawdown tests with surrogate fission product chlorides was run to obtain selectivity data with non-radioactive salts, develop a predictive model, and draw conclusions about the viability of using this process with actinide-loadd salt. Results of tests with CsCl, LaCl3, CeCl3, and NdCl3 are reported here. An equilibrium model has been formulated and fit to the experimental data. Excellent fits to the data were achieved. Based on analysis and results obtained to date, it is concluded that clean separation between minor actinides and lanthanides will be difficult to achieve using lithium drawdown.
Michael F. Simpson; Daniel LaBrier; Michael Lineberry; Tae-Sic Yoo
Iodized Salt
Iodized Salt Iodized Salt Name: Theresa Location: N/A Country: N/A Date: N/A Question: Why do they put iodine in salt? Replies: Iodine was introduced into salt at earlier this century when it was discovered that certain areas of the US had a mark deficiency in iodine in the diet of people, and people developed a neck swelling (goiter). The Great Lakes region is one of these areas where the soil is lacking iodine. Goiter can be caused when the thyroid gland swells because of a lack of iodine in the diet. Most medical advise now states that iodine in salt is no longer necessary due to our food sources arising from all over the world. Steve Sample Hi Theresa...see, there are a variety of elements and compounds that are necessary for the proper maintenance of our life. One of these is iodine, since a small quantity of iodine is needed for the adequate functioning of the thyroid gland. A deficiency of iodine produces dire effects, as goiter, where the thyroid gland swollens due to the lack of iodine traces in the diet. The iodine affects directly the tyrhoid gland secretions, which themselves, to a great extent, control heart action, nerve response to stimuli, rate of body growth and metabolism.
Molten carbonate fuel cell separator
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.
Nickols, Richard C. (East Hartford, CT)
Nickols, R.C.
Coal derived fuel gases for molten carbonate fuel cells
Product streams from state-of-the-art and future coal gasification systems are characterized to guide fuel cell program planners and researchers in establishing performance goals and developing materials for molten carbonate fuel cells that will be compatible with gasifier product gases. Results are presented on: (1) the range of gasifier raw-gas compositions available from the major classes of coal gasifiers; (2) the degree of gas clean-up achievable with state-of-the-art and future gas clean-up systems; and (3) the energy penalties associated with gas clean-up. The study encompasses fixed-bed, fluid-bed, entrained-bed, and molten salt gasifiers operating with Eastern bituminous and Western subbituminous coals. Gasifiers operating with air and oxygen blowing are evaluated, and the coal gasification product streams are characterized with respect to: (1) major gas stream constituents, e.g., CO, H/sub 2/, CO/sub 2/, CH/sub 4/, N/sub 2/, H/sub 2/O; (2) major gas stream contaminants, e.g., H/sub 2/S, COS, particulates, tars, etc.; and (3) trace element contaminants, e.g., Na, K, V, Cl, Hg, etc.
Capillary-Pumped Passive Reactor Concept for Space Nuclear Power
To develop the passively-cooled space reactor concept using the capillary-induced lithium flow, since molten lithium possesses a very favorable surface tension characteristic. In space where the gravitational field is minimal, the gravity-assisted natural convection cooling is not effective nor an option for reactor heat removal, the capillary induced cooling becomes an attractive means of providing reactor cooling.
Dr. Thomas F. Lin; Dr. Thomas G. Hughes; Christopher G. Miller
Cathode for molten carbonate fuel cell
Disclosed are a porous sintered cathode for a molten carbonate fuel cell and method of making same. The cathode includes 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 has 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.
Kaun, T.D.; Mrazek, F.C.
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.
Kaun, Thomas D. (New Lenox, IL); Mrazek, Franklin C. (Hickory Hills, IL)
Stainless steel corrosion by molten nitrates : analysis and lessons learned.
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 pitting occurred near welds; however this was the hottest region in the chamber. Pitting was observed up to two inches above the weld, indicating independence from weld effects.
Kruizenga, Alan Michael
Recirculating Molten Metal Supply System And Method
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).
Kinosz, Michael J. (Apollo, PA); Meyer, Thomas N. (Murrysville, PA)
Rare Earth Extraction by Molten Oxide Electrolysis
Symposium, Production, Refining and Recycling of Rare Earth Metals ... Electrolysis in molten halides is an established method for the reduction but requires ... Recycling of Different Sintered Magnet Grades by Hydrogen Processing Yielding�...
17048 Biochemistry 1998, 37, 17048-17053 Sugar-Induced Molten-Globule Model �
ABSTRACT: Proteins denature at low pH because of intramolecular electrostatic repulsions. The addition of salt partially overcomes this repulsion for some proteins, yielding a collapsed conformation called the A-state. A-states have characteristics expected for the molten globule, a notional kinetic protein folding intermediate. Here we show that the addition of neutral sugars to solutions of acid-denatured equine ferricytochrome c induces formation of the A-state in the absence of added salt. We characterized the structure and stability of the sugar-induced A-state with circular dichroism spectropolarimetry (CD) and NMR-monitored hydrogen-deuterium exchange experiments. We also examined the stability of the sugarinduced A-state as a function of sugar size and concentration. The results are interpreted using several models and we conclude that the stabilizing effect is consistent with increased steric repulsion between the protein and the sugar solutions. Structural and stability studies of molten globule-like states have been instrumental in expanding our knowledge of protein folding. Molten globules, notional kinetic folding intermediates, are compact and possess nativelike secondary structure and some native tertiary contacts (1-3). An
Paula R. Davis-searles; Artemiza S. Morar; Aleister J. Saunders; Dorothy A. Erie; Gary J. Pielak
Neutronic Study of Slightly Modified Water Reactors and Application to Transition Scenarios Richard Chambon
the new generation (GenIV) of nuclear reactors, currently under development, has to face. These reactors conver- ting or full breeding cycles with technologies such as Fast Breeder Reactors (FBRs) or Molten viability of this kind of reactor and to meet current licensing standards, before detailed sa- fety studies
Paris-Sud XI, Université de
A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.
Christy, R.F.
Catalytic Gasification of Coal using Eutectic Salt Mixtures
The objectives of this study are to: identify appropriate eutectic salt mixture catalysts for coal gasification; assess agglomeration tendency of catalyzed coal; evaluate various catalyst impregnation techniques to improve initial catalyst dispersion; evaluate effects of major process variables (such as temperature, system pressure, etc.) on coal gasification; evaluate the recovery, regeneration and recycle of the spent catalysts; and conduct an analysis and modeling of the gasification process to provide better understanding of the fundamental mechanisms and kinetics of the process. A review of the collected literature was carried out. The catalysts which have been used for gasification can be roughly classified under the following five groups: alkali metal salts; alkaline earth metal oxides and salts; mineral substances or ash in coal; transition metals and their oxides and salts; and eutectic salt mixtures. Studies involving the use of gasification catalysts have been conducted. However, most of the studies focused on the application of individual catalysts. Only two publications have reported the study of gasification of coal char in CO2 and steam catalyzed by eutectic salt mixture catalysts. By using the eutectic mixtures of salts that show good activity as individual compounds, the gasification temperature can be reduced possibly with still better activity and gasification rates due to improved dispersion of the molten catalyst on the coal particles. For similar metal/carbon atomic ratios, eutectic catalysts were found to be consistently more active than their respective single salts. But the exact roles that the eutectic salt mixtures play in these are not well understood and details of the mechanisms remain unclear. The effects of the surface property of coals and the application methods of eutectic salt mixture catalysts with coal chars on the reactivity of gasification will be studied. Based on our preliminary evaluation of the literature, a ternary eutectic salt mixture consisting of Li- Na- and K- carbonates has the potential as gasification catalyst. To verify the literature reported, melting points for various compositions consisting of these three salts and the temperature range over which the mixture remained molten were determined in the lab. For mixtures with different concentrations of the three salts, the temperatures at which the mixtures were found to be in complete molten state were recorded. By increasing the amount of Li2CO3, the melting temperature range was reduced significantly. In the literature, the eutectic mixtures of Li- Na- and K-carbonates are claimed to have a lower activation energy than that of K2CO3 alone and they remain molten at a lower temperature than pure K2CO3. The slow increase in the gasification rates with eutectics reported in the literature is believed to be due to a gradual penetration of the coals and coal char particles by the molten and viscous catalyst phase. The even spreading of the salt phase seems to increase the overall carbon conversion rate. In the next reporting period, a number of eutectic salts and methods of their application on the coal will be identified and tested.
Atul Sheth; Pradeep Agrawal; Yaw D. Yeboah
Structure and dynamics in yttrium-based molten rare earth alkali fluorides
The transport properties of molten LiF-YF$_3$ mixtures have been studied by pulsed field gradient nuclear magnetic resonance spectroscopy, potentiometric experiments, and molecular dynamics simulations. The calculated diffusion coefficients and electric conductivities compare very well with the measurements accross a wide composition range. We then extract static (radial distribution functions, coordination numbers distributions) and dynamic (cage correlation functions) quantities from the simulations. Then, we discuss the interplay between the microscopic structure of the molten salts and their dynamic properties. It is often considered that variations in the diffusion coefficient of the anions are mainly driven by the evolution of its coordination with the metallic ion (Y$^{3+}$ here). We compare this system with fluorozirconate melts and demonstrate that the coordination number is a poor indicator of the evolution of the diffusion coefficient. Instead, we propose to use the ionic bonds lifetime. We show th...
Levesque, Maximilien; Salanne, Mathieu; Gobet, Mallory; Groult, Henri; Bessada, Catherine; Madden, Paul A; Rollet, Anne-Laure
Simulated first operating campaign for the Integral Fast Reactor fuel cycle demonstration
This report discusses the Integral Fast Reactor (IFR) which is an innovative liquid-metal-cooled reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid-metal cooling to offer significant improvements in reactor safety, operation, fuel cycle-economics, environmental protection, and safeguards. Over the next few years, the IFR fuel cycle will be demonstrated at Argonne-West in Idaho. Spent fuel from the Experimental Breeder Reactor II (EBR-II) win be processed in its associated Fuel Cycle Facility (FCF) using a pyrochemical method that employs molten salts and liquid metals in an electrorefining operation. As part of the preparation for the fuel cycle demonstration, a computer code, PYRO, was developed at Argonne to model the electrorefining operation using thermodynamic and empirical data. This code has been used extensively to evaluate various operating strategies for the fuel cycle demonstration. The modeled results from the first operating campaign are presented. This campaign is capable of processing more than enough material to refuel completely the EBR-II core.
Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Park, K.H. [Argonne National Lab., Idaho Falls, ID (United States); Ackerman, J.P. [Argonne National Lab., IL (United States)
Method and apparatus for atomization and spraying of molten metals
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.
Hobson, David O. (Oak Ridge, TN); Alexeff, Igor (Oak Ridge, TN); Sikka, Vinod K. (Clinton, TN)
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.
Hobson, D.O.; Alexeff, I.; Sikka, V.K.
Temperature-dependent mechanical property testing of nitrate thermal storage salts.
Three salt compositions for potential use in trough-based solar collectors were tested to determine their mechanical properties as a function of temperature. The mechanical properties determined were unconfined compressive strength, Young's modulus, Poisson's ratio, and indirect tensile strength. Seventeen uniaxial compression and indirect tension tests were completed. It was found that as test temperature increases, unconfined compressive strength and Young's modulus decreased for all salt types. Empirical relationships were developed quantifying the aforementioned behaviors. Poisson's ratio tends to increase with increasing temperature except for one salt type where there is no obvious trend. The variability in measured indirect tensile strength is large, but not atypical for this index test. The average tensile strength for all salt types tested is substantially higher than the upper range of tensile strengths for naturally occurring rock salts. Interest in raising the operating temperature of concentrating solar technologies and the incorporation of thermal storage has motivated studies on the implementation of molten salt as the system working fluid. Recently, salt has been considered for use in trough-based solar collectors and has been shown to offer a reduction in levelized cost of energy as well as increasing availability (Kearney et al., 2003). Concerns regarding the use of molten salt are often related to issues with salt solidification and recovery from freeze events. Differences among salts used for convective heat transfer and storage are typically designated by a comparison of thermal properties. However, the potential for a freeze event necessitates an understanding of salt mechanical properties in order to characterize and mitigate possible detrimental effects. This includes stress imparted by the expanding salt. Samples of solar salt, HITEC salt (Coastal Chemical Co.), and a low melting point quaternary salt were cast for characterization tests to determine unconfined compressive strength, indirect tensile strength, coefficient of thermal expansion (CTE), Young's modulus, and Poisson's ratio. Experiments were conducted at multiple temperatures below the melting point to determine temperature dependence.
Everett, Randy L.; Iverson, Brian D.; Broome, Scott Thomas; Siegel, Nathan Phillip; Bronowski, David R.
Molten carbonate fuel cell programs in the United States
The environmental, performance, and economic aspects of molten carbonate fuel cell power plants are discussed. Design, components, and operation of molten carbonate fuel cells are discussed, and US research is outlined. (WHK)
Ackerman, J.P.
Updated Generation IV Reactors Integrated Materials Technology Program Plan, Revision 2
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.
Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Halsey, William [Lawrence Livermore National Laboratory (LLNL); Hayner, George [Idaho National Laboratory (INL); Katoh, Yutai [ORNL; Klett, James William [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Stoller, Roger E [ORNL; Wilson, Dane F [ORNL
Destruction of organic wastes with molten oxidizers
A process for destruction of biologically hazardous organic chemical wastes by using liquefied strongly oxidizing inorganic salts, such as the alkali metal nitrates, at high temperatures and atmospheric pressure, to yield inorganic salts, carbon dioxide, and water. The oxidizing salts are regenerated and recycled.
Bradshaw, R.W.; Holmes, J.T.; Tyner, C.E.
Y-12's first Open House-September 2-3, 1967, part 2
work is done for the Laboratory. For instance, Y-12 fabricated major components for the Molten Salt Reactor Experiment, the High Flux Isotope Reactor, the DCX fusion experiment...
Advanced High Temperature Reactor Systems and Economic Analysis
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 advanced supercritical-water power cycles. The current design activities build upon a series of small-scale efforts over the past decade to evaluate and describe the features and technology variants of FHRs. Key prior concept evaluation reports include the SmAHTR preconceptual design report,1 the PB-AHTR preconceptual design, and the series of early phase AHTR evaluations performed from 2004 to 2006. This report provides a power plant-focused description of the current state of the AHTR. The report includes descriptions and sizes of the major heat transport and power generation components. Component configuration and sizing are based upon early phase AHTR plant thermal hydraulic models. The report also provides a top-down AHTR comparative economic analysis. A commercially available advanced supercritical water-based power cycle was selected as the baseline AHTR power generation cycle both due to its superior performance and to enable more realistic economic analysis. The AHTR system design, however, has several remaining gaps, and the plant cost estimates consequently have substantial remaining uncertainty. For example, the enriched lithium required for the primary coolant cannot currently be produced on the required scale at reasonable cost, and the necessary core structural ceramics do not currently exist in a nuclear power qualified form. The report begins with an overview of the current, early phase, design of the AHTR plant. Only a limited amount of information is included about the core and vessel as the core design and refueling options are the subject of a companion report. The general layout of an AHTR system and site showing the relationship of the major facilities is then provided. Next is a comparative evaluation of the AHTR anticipated performance and costs. Finally, the major system design efforts necessary to bring the AHTR design to a pre-conceptual level are then presented.
Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Qualls, A L [ORNL
In-Vessel Retention of Molten Corium: Lessons Learned and Outstanding Issues
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.
J.L. Rempe; K.Y. Suh; F. B. Cheung; S. B. Kim
Nuclear reactor melt-retention structure to mitigate direct containment heating
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.
Tutu, Narinder K. (Manorville, NY); Ginsberg, Theodore (East Setauket, NY); Klages, John R. (Mattituck, NY)
The Advanced High-Temperature Reactor (AHTR) for Producing Hydrogen to Manufacture Liquid Fuels
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 and thus reduce three temperature losses in the system associated with (1) heat transfer from the fuel to the reactor coolant, (2) temperature rise across the reactor core, and (3) heat transfer across the heat exchangers between the reactor and H2 production plant. Lowering the peak reactor temperatures and thus reducing the high-temperature materials requirements may make the AHTR the enabling technology for low-cost nuclear hydrogen production.
Forsberg, C.W.; Peterson, P.F.; Ott, L.
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)
Wall, E.M.V.; Bauer, D.T.; Hahn, H.T.
Molten metal holder furnace and casting system incorporating the molten metal holder furnace
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).
Method and apparatus for spraying molten materials
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.
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)
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.
Glovan, R.J.; Tierney, J.C.; McLean, L.L.; Johnson, L.L.; Nelson, G.L.; Lee, Y.M.
Status of Molten Carbonate Fuel Cell Technology
Fuel cell technology development and commercialization continues to be a major thrust in the alternative energy sector of distributed generation (DG). Second generation, molten carbonate fuel cell technology (MCFC) is now entering a critical commercialization phase. Given recent MCFC developments and advances in other distributed generation technologies, an assessment and update on the prospects for MCFC power systems is needed to guide future utility investments.
Molten Air -- A new, highest energy class of rechargeable batteries
This study introduces the principles of a new class of batteries, rechargeable molten air batteries, and several battery chemistry examples are demonstrated. The new battery class uses a molten electrolyte, are quasi reversible, and have amongst the highest intrinsic battery electric energy storage capacities. Three examples of the new batteries are demonstrated. These are the iron, carbon and VB2 molten air batteries with respective intrinsic volumetric energy capacities of 10,000, 19,000 and 27,000 Wh per liter.
Licht, Stuart
Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials
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 materials needed to ensure their safe and reliable operation. The focus of this document will be the overall range of DOE's structural materials research activities being conducted to support VHTR development. By far, the largest portion of material's R&D supporting VHTR development is that being performed directly as part of the Next-Generation Nuclear Plant (NGNP) Project. Supplementary VHTR materials R&D being performed in the DOE program, including university and international research programs and that being performed under direct contracts with the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, will also be described. Specific areas of high-priority materials research that will be needed to deploy the NGNP and provide a basis for subsequent VHTRs are described, including the following: (1) Graphite: (a) Extensive unirradiated materials characterization and assessment of irradiation effects on properties must be performed to qualify new grades of graphite for nuclear service, including thermo-physical and mechanical properties and their changes, statistical variations from billot-to-billot and lot-to-lot, creep, and especially, irradiation creep. (b) Predictive models, as well as codification of the requirements and design methods for graphite core supports, must be developed to provide a basis for licensing. (2) Ceramics: Both fibrous and load-bearing ceramics must be qualified for environmental and radiation service as insulating materials. (3) Ceramic Composites: Carbon-carbon and SiC-SiC composites must be qualified for specialized usage in selected high-temperature components, such as core stabilizers, control rods, and insulating covers and ducting. This will require development of component-specific designs and fabrication processes, materials characterization, assessment of environmental and irradiation effects, and establishment of codes and standards for materials testing and design requirements. (4) Pressure Vessel Steels: (a) Qualification of short-term, high-temperature properties of light water rea
Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL
Preventing Molten Aluminium Water Explosions through the Use of ...
The energy released from one kilogram of molten aluminium reacted with oxygen is equivalent to detonating 3 kilograms of trinitrotoluene (TNT). For over 60�...
Recent Progress in Molten Oxide Electrolysis for Iron Production
Presentation Title, Recent Progress in Molten Oxide Electrolysis for Iron Production ... Concentrated Solar Power for Producing Liquid Fuels from CO2 and H2O.
Process of making electrolyte structure for molten carbonate fuel cells
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.
Arendt, R.H.; Curran, M.J.
Arendt, Ronald H. (Schenectady, NY); Curran, Matthew J. (Schenectady, NY)
Electrochemical cell utilizing molten alkali metal electrode-reactant
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.
Virkar, Anil V. (Sandy, UT); Miller, Gerald R. (Salt Lake City, UT)
Thermal Barrier Coatings for Resistance Against Attack by Molten ...
Presentation Title, Thermal Barrier Coatings for Resistance Against Attack by Molten Silicate Deposits from CMAS Sand, Volcanic Ash, or Coal Fly Ash Ingested�...
Hybrid Molten Bed Gasifier for High Hydrogen Syngas Production
Hybrid Molten Bed Gasifier for High Hydrogen (H2) Syngas Production Gas Technology Institute (GTI) Project Number: FE0012122 Project Description The research team will evaluate and...
SunShot Initiative: Advanced Nitrate Salt Central Receiver Power Plant
Advanced Nitrate Salt Central Receiver Power Plant Advanced Nitrate Salt Central Receiver Power Plant Abengoa logo Photo of two lit towers surrounded by much smaller blue flat plates that are mounted on the ground. Commercial central receiver plant designs Abengoa, under the Baseload CSP FOA, will demonstrate a 100-megawatt electrical (MWe) central receiver plant using nitrate salt as the receiver coolant, thermal storage medium, and heat transport fluid in the steam generator. Approach The plan is to operate the plant at full load for 6,400 hours each year using only solar energy. Abengoa is working to create a team of suppliers capable of deploying a commercially ready nitrate salt central receiver technology that can be competitive in the current power marketplace. Innovation Abengoa is developing a new molten-salt power tower technology with a surround heliostat field. Key components include:
Detection of frozen salt in pipes using gamma-ray spectrometry of potassium self-activity
Solar plants that use molten salts as heat transfer fluid need careful control to avoid the freezing of the salt in the pipes; if such a problem occurs, a diagnostic instrument to localize where is the frozen salt plug and to determine its length is useful. If the salt contains potassium (as is the case of the most common mixture used in solar plants, NaNO{sub 3}/KNO{sub 3} 60/40% by weight), the gamma decay of the natural unstable isotope {sup 40}K can be exploited to detect the frozen salt in a non-invasive way. Simulations and experimental results regarding the detectability of such plugs with different masses/lengths are presented. (author)
Grena, Roberto; Scafe, Raffaele; Pisacane, Fabrizio; Pilato, Renzo; Crescenzi, Tommaso; Mazzei, Domenico [ENEA, Casaccia Research Centre, via Anguillarese 301, 00123 S. Maria di Galeria, Rome (Italy)
Plutonium recovery from spent reactor fuel by uranium displacement
This report discusses a process for separating uranium values and transuranic values from fission products containing rare earth values when the values which 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 re-established.
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.
Ackerman, John P. (Downers Grove, IL)
Flexible Conversion Ratio Fast Reactor Systems Evaluation
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.
Neil Todreas; Pavel Hejzlar
Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.
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 both the PBMR and prismatic design. The main focus of this report is the RPV for both design concepts with emphasis on material selection.
Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division
Method for removing semiconductor layers from salt substrates
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.
Shuskus, Alexander J. (West Hartford, CT); Cowher, Melvyn E. (East Brookfield, MA)
Pump for molten metal or other fluid
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.
Horton, James A. (Livermore, CA); Brown, Donald L. (Livermore, CA)
Energy Department Completes Salt Coolant Material Transfer to Czech
Completes Salt Coolant Material Transfer to Czech Completes Salt Coolant Material Transfer to Czech Republic for Advanced Reactor Research Energy Department Completes Salt Coolant Material Transfer to Czech Republic for Advanced Reactor Research May 20, 2013 - 12:52pm Addthis News Media Contact (202) 586-4940 PRAGUE, CZECH REPUBLIC - 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 (ORNL) to the Czech Nuclear Research Institute Řež for experiments at Řež's critical test facility. This partnership builds on a strong history of U.S.-Czech energy collaboration and follows President Obama's speech in Prague in April 2009, where he laid out the importance of international
Amine salts of nitroazoles
Compositions of matter, a method of providing chemical energy by burning said compositions, and methods of making said compositions. These compositions are amine salts of nitroazoles.
Lee, Kien-yin (Los Alamos, NM); Stinecipher, Mary M. (Los Alamos, NM)
The transport properties of molten LiF-YF3 mixtures have been studied by pulsed field gradient nuclear magnetic resonance spectroscopy, potentiometric experiments, and molecular dynamics simulations. The calculated diffusion coefficients and electric conductivities compare very well with the measurements accross a wide composition range. We then extract static (radial distribution functions, coordination numbers distributions) and dynamic (cage correlation functions) quantities from the simulations. Then, we discuss the interplay between the microscopic structure of the molten salts and their dynamic properties. It is often considered that variations in the diffusion coefficient of the anions are mainly driven by the evolution of its coordination with the metallic ion (Y3+ here). We compare this system with fluorozirconate melts and demonstrate that the coordination number is a poor indicator of the evolution of the diffusion coefficient. Instead, we propose to use the ionic bonds lifetime. We show that the weak Y-F ionic bonds in LiF-YF3 do not induce the expected tendency of the fluoride diffusion coefficient to converge toward the one of yttrium cation when the content in YF3 increases. Implications on the validity of the Nernst-Einstein relation for estimating the electrical conductivity are discussed.
Maximilien Levesque; Vincent Sarou-Kanian; Mathieu Salanne; Mallory Gobet; Henri Groult; Catherine Bessada; Paul A. Madden; Anne-Laure Rollet
Electrolyte paste for molten carbonate fuel cells
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.
Bregoli, Lawrance J. (Southwick, MA); Pearson, Mark L. (New London, CT)
Method for the regeneration of spent molten zinc chloride
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.
Zielke, Clyde W. (McMurray, PA); Rosenhoover, William A. (Pittsburgh, PA)
Molten carbonate fuel cell technology improvement
This report summarizes the work performed under Department of Energy Contract DEAC21-87MC23270, Molten Carbonate Fuel Cell Technology Improvement.'' This work was conducted over a three year period and consisted of three major efforts. The first major effort was the power plant system study which reviewed the competitive requirements for a coal gasifier/molten carbonate fuel cell power plant, produced a conceptual design of a CG/MCFC, and defined the technology development requirements. This effort is discussed in Section 1 of the report. The second major effort involved the design and development of a new MCFC cell configuration which reduced the material content of the cell to a level competitive with competing power plants, simplified the cell configuration to make the components more manufacturable and adaptable to continuous low cost processing techniques, and introduced new-low-pressure drop flow fields for both reactant gases. The new flow fields permitted the incorporation of recirculation systems in both reactant gas systems, permitting simplified cooling techniques and the ability to operate on both natural gas and a wide variety of gasifier fuels. This cell technology improvement is discussed in Section 2. The third major effort involved the scaleup of the new cell configuration to the full-area, 8-sq-ft size and resulted in components used for a 25-kW, 20-cell stack verification test. The verification test was completed with a run of 2200 hours, exceeding the goal of 2000 hours and verifying the new cell design. TWs test, in turn, provided the confidence to proceed to a 100-kW demonstration which is the goal of the subsequent DOE program. The scaleup and stack verification tests are discussed in Sections 3, 4, 5, and 6 of this report.
Le cycle Thorium en r�acteurs � sels fondus peut-il �tre une solution au probl�me �nerg�tique du XXI�me si�cle ? Le concept de TMSR-NM.
??Un concept innovant de r�acteurs nucl�aires � sels fondus, le Thorium Molten Salt Reactor (TMSR), a �t� d�fini au LPSC Grenoble. Le pr�sent m�moire porte� (more)
Merle-Lucotte, Elsa
Facility Representative Program: 2008 Facility Representative...
Sherman Chao, LSO Conduct of Operations Improvements at K Basins Dennis Humphreys, RL Molten Salt Reactor Experiment (MSRE) facility lessons learned Charlie Wright, ORO...
Electrochemistry and Materials Properties
Mar 6, 2013 ... In the frame of the development of a Molten Salt Fast Reactor concept, the corrosion behavior of structural metallic materials in contact with the�...
Processing of Superalloys II
... superalloy C276 alloy has been selected as the primary structure materials of the simulation experimental loop in thorium molten salt reactor (TMSR) program.
Effect of Long Time Thermal Exposure on Microstructure and ...
A BP neural network predictor model for desulfurizing molten iron
Desulfurization of molten iron is one of the stages of steel production process. A back-propagation (BP) artificial neural network (ANN) model is developed to predict the operation parameters for desulfurization process in this paper. The primary objective ...
Zhijun Rong; Binbin Dan; Jiangang Yi
Molten Mold Flux Technology for Continuous Casting of the ULC ...
Heat flux from the molten steel to the cupper plate of the casting mold was .... of Conventional and High Niobium API 5L X80 Line Pipe Steel Using EBSD.
Candidate anode materials for iron production by molten oxide electrolysis
Molten oxide electrolysis (MOE) has been identified by the American Iron and Steel Institute (AISI) as one of four possible breakthrough technologies to alleviate the environmental impact of iron and steel production. This ...
Paramore, James D
Towards Sustainable Metal Production by Molten Oxide Electrolysis
Current Efficiency for Aluminium Deposition from Molten Cryolite ...
Electrical Conductivity of the KF-NaF- AlF3 Molten System at Low Cryolite Ratio ... Experimental Investigation of Single Bubble Characteristics in a Cold Model of a ... Impact of Amperage Creep on Potroom Busbars and Electrical Insulation:�...
Recovery boiler superheater corrosion - solubility of metal oxides in molten salt .
??The recovery boiler in a pulp and paper mill plays a dual role of recovering pulping chemicals and generating steam for either chemical processes or� (more)
Meyer, Joseph Freeman
Molten Salt Electrorefining of Zr-Hf Impure Metal for Nuclear ...
A21: First-Principles Molecular Dynamics Simulation of Chemical ... A3: Investigation on Co-combustion Kinetics of Anthracite Coal and Biomass Char by ... Lithium Redox Process for Thermochemical Water-Splitting as Energy Conversion.
Molten Salt Oxidation: A Thermal Technology for Waste Treatment and Demilitarization
MSO is a good alternative to incineration for the treatment of a variety of organic wastes including obsolete explosives, low-level mixed waste streams, PCB contaminated oils, spent resins and carbon. The Lawrence Livermore National Laboratory (LLNL) has demonstrated the MSO process for the effective destruction of explosives, explosives-contaminated materials, and other wastes on a 1.5 kg/hr bench-scale unit and in an integrated MSO facility capable of treating 8 kg/hr of low-level radioactive mixed wastes. LLNL, under the direction and support of the Joint Demilitarization Technology (JDT) program, is currently building an integrated MSO plant for destroying explosives, explosives-contaminated sludge and explosives-contaminated activated charcoal. In a parallel effort, LLNL also provides technical support to DOE for the implementation of the MSO technology at industrial scale at Richland, Washington. Over 30 waste streams have been demonstrated with LLNL-built MSO systems. In this paper we will present our latest experimental data, our operational experience with MSO and also discuss its process capabilities.
Hsu, P C; Watkins, B; Pruneda, C; Kwak, S
Optical and chemical properties of molten salt mixtures for use in high temperature power systems
A future, robust energy portfolio will include, together with fossil fuel technologies and nuclear systems, a mix of renewable energy systems. Within each type of system there will also be variants used to strengthen a ...
Passerini, Stefano
Bulk Vitrification Performance Enhancement: Refractory Lining Protection Against Molten Salt Penetration
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.
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
Laser Fusion-Fission Hybrid / Eighteenth Topical Meeting on the Technology of Fusion Energy (Part 2)
R. W. Moir; H. F. Shaw; A. Caro; Larry Kaufman; J. F. Latkowski; J. Powers; P. E. A. Turchi
From molten salts to room temperature ionic liquids: Simulation studies on chloroaluminate systems
An interaction potential including chloride anion polarization effects, constructed from first-principles calculations, is used to examine the structure and transport properties of a series of chloroaluminate melts. A particular emphasis was given to the study of the equimolar mixture of aluminium chloride with 1-ethyl-3-methylimidazolium chloride, which forms a room temperature ionic liquid EMI-AlCl 4. The structure yielded by the classical simulations performed within the framework of the polarizable ion model is compared to the results obtained from entirely electronic structure-based simulations: An excellent agreement between the two flavors of molecular dynamics is observed. When changing the organic cation EMI+ by an inorganic cation with a smaller ionic radius (Li+, Na+, K+), the chloroaluminate speciation becomes more complex, with the formation of Al2Cl 7- in small amounts. The calculated transport properties (diffusion coefficients, electrical conductivity and viscosity) of EMI-AlCl4 are in good ag...
Salanne, Mathieu; Seitsonen, Ari P; Madden, Paul A; Kirchner, Barbara; 10.1039/C1FD00053E
Molten-Salt Synthesis of Pure and Doped LaAlO3 Nanoparticles
About this Abstract. Meeting, Materials Science & Technology 2011. Symposium, Innovative Processing and Synthesis of Ceramics, Glasses and Composites.
C48 TiFe Alloy Prepared by Molten Salt Electrolysis Ilmenite
High thermal energy storage density molten salts for parabolic trough solar power generation.
??New alkali nitrate-nitrite systems were developed by using thermodynamic modeling and the eutectic points were predicted based on the change of Gibbs energy of fusion.� (more)
Wang, Tao
Abstract Scope, Uranium can be recovered from uranium oxide (UO2) spent fuel through the ... Cathodic Behavior of Silicon (?) in BaF2-CaF2 �SiO2 Melts.
Molten Salt Electrolysis of MgCl2 in a Cell with Rapid Chlorine ...
Energy consumption of 7.0 kWh kg-1 Mg that is slightly above the theoretical minimum, 6.2 kWh kg-1 Mg, at 0.68 A cm-2 anodic current density was achieved for�...
Mixed Conducting Molten Salt Electrolyte for Na/NiCl 2 Cell
Catalytic Properties of Ni-Al Intermetallic Nanoparticle Catalysts for Hydrogen Production from Methanol and Methane � Ca, Li and Mg Based Lightweight�...
In-Vessel Retention of Molten Core Debris in the Westinghouse AP1000 Advanced Passive PWR
In-vessel retention (IVR) of molten core debris via external reactor vessel cooling is the hallmark of the severe accident management strategies in the AP600 passive PWR. The vessel is submerged in water to cool its external surface via nucleate boiling heat transfer. An engineered flow path through the reactor vessel insulation provides cooling water to the vessel surface and vents steam to promote IVR. For the 600 MWe passive plant, the predicted heat load from molten debris to the lower head wall has a large margin to the critical heat flux on the external surface of the vessel, which is the upper limit of the cooling capability. Up-rating the power of the passive plant from 600 to 1000 MWe (AP1000) significantly increases the heat loading from the molten debris to the reactor vessel lower head in the postulated bounding severe accident sequence. To maintain a large margin to the coolability limit for the AP1000, design features and severe accident management (SAM) strategies to increase the critical heat flux on the external surface of the vessel wall need to be implemented. A test program at the ULPU facility at University of California Santa Barbara (UCSB) has been initiated to investigate design features and SAM strategies that can enhance the critical heat flux. Results from ULPU Configuration IV demonstrate that with small changes to the ex-vessel design and SAM strategies, the peak critical heat flux in the AP1000 can be increased at least 30% over the peak critical heat flux predicted for the AP600 configuration. The design and SAM strategy changes investigated in ULPU Configuration IV can be implemented in the AP1000 design and will allow the passive plant to maintain the margin to critical heat flux for IVR, even at the higher power level. Continued testing for IVR phenomena is being performed at UCSB to optimize the AP1000 design and to ensure that vessel failure in a severe accident is physically unreasonable. (authors)
Scobel, James H.; Conway, L.E. [Westinghouse Electric Company LLC, P.O. Box 355, Pittsburgh, PA 15230-0355 (United States); Theofanous, T.G. [Center for Risk Studies and Safety, University of California Santa Barbara (United States)
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.
King, L.D.P.
Solid tags for identifying failed reactor components
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.
Bunch, Wilbur L. (Richland, WA); Schenter, Robert E. (Richland, WA)
Salt Waste Processing Initiatives
Patricia Suggs Patricia Suggs Salt Processing Team Lead Assistant Manager for Waste Disposition Project Office of Environmental Management Savannah River Site Salt Waste Processing Initiatives 2 Overview * Current SRS Liquid Waste System status * Opportunity to accelerate salt processing - transformational technologies - Rotary Microfiltration (RMF) and Small Column Ion Exchange (SCIX) - Actinide Removal Process/Modular Caustic Side Solvent Extraction (ARP/MCU) extension with next generation extractant - Salt Waste Processing Facility (SWPF) performance enhancement - Saltstone enhancements * Life-cycle impacts and benefits 3 SRS Liquid Waste Total Volume >37 Million Gallons (Mgal) Total Curies 183 MCi (51% ) 175 MCi (49% ) >358 Million Curies (MCi) Sludge 34.3 Mgal (92% ) 3.0 Mgal (8%)
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.
Kienyin Lee; Stinecipher, M.M.
Salt Selected (FINAL)
WHY SALT WAS SELECTED AS A DISPOSAL MEDIUM 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
Thermal Interaction Between Molten Metal Jet and Sodium Pool: Effect of Principal Factors Governing Fragmentation of the Jet
To clarify the effects of the principal factors that govern the thermal fragmentation of a molten metallic fuel jet in the course of fuel-coolant interaction, which is important in evaluating the sequence of core disruptive accidents (CDAs) for metallic fuel fast reactors, basic experiments were carried out using molten metallic fuel simulants (copper and silver) and a sodium pool.Fragmentation of a molten metal jet with a solid crust was caused by internal pressure produced by the boiling of sodium, which is locally entrapped inside the jet due to hydrodynamic motion between the jet and the coolant. The superheating and the latent heat of fusion of the jet are the principal factors governing this type of thermal fragmentation. On the other hand, the effect of the initial sodium temperature is regarded as negligible in the case of thermal conditions expected to result in CDAs for practical metallic fuel cores. Based on the fragmentation data for several kinds of jets (Cu, Ag, SUS, U, and U-5 wt% Zr alloy), an empirical correlation is proposed that is applicable to the calculation of a mass median diameter of fragments produced by the thermal fragmentation of the jet with a solid crust under low ambient Weber number conditions.
Nishimura, Satoshi [Central Research Institute of Electric Power Industry (CRIEPI) (Japan); Kinoshita, Izumi [Central Research Institute of Electric Power Industry (CRIEPI) (Japan); Sugiyama, Ken-Ichiro [Hokkaido University (Japan); Ueda, Nobuyuki [Central Research Institute of Electric Power Industry (CRIEPI) (Japan)
Process Heat Exchanger Options for the Advanced High Temperature Reactor
In-situ Analysis of Zinc Electrodeposition within an Ionic Liquid Electrolyte
chloride low temperature molten salt. Electrochimica Acta,room temperature molten salt. Journal of the Electrochemical
Keist, Jayme
Flexographically Printed Rechargeable Zinc-based Battery for Grid Energy Storage
Rechargeable, Lithium-ion Molten Salt Battery for Highare room- temperature molten salts, which are typically
Wang, Zuoqian
The performances of aqueous-homogeneous (AHBR), molten-salt (MSBR), liquid-bismuth (LBBR), gas-cooled graphite-moderated (GGBR), and deuterium- moderated gas-cooled (DGBR) breeder reactors were evaluated in respect to fuel yield, fuel cycle costs, and development status. A net electrical plant capability of 1000 Mwe was selected, and the fuel and fertile streams were processed continuously on-site. The maximum annual fuel yields were 1.5 mills/ kwhr. The minimum estimated fuel cycle costs were 0.9, 0.6, 1.0, 1.2, and 1.3 mills/kwhr at fuel yields of were 0.9, 0.9, 1.5, 1.5, and 1.3 mills/kwhr. Only the AHBR and the MSBR are capable of achieving fuel yields substantially in excess of 4%/yr, and therefore, in view of the uncertainties in nuclear data and efficiencies of processing methods, only these two can be listed with confidence as being able to satisfy the main criterion of the AEC longrange thorium breeder program, viz. a doubling time of 25 years or less. The development effort required to bring the various concepts to the stage where a prototype station could be designed was estimated to be least for the AHBR, somewhat more for the MSBR, and several times as much for the other systems. The AHBR was judged to rank first in regard to nuclear capability, fuel cycle potential, and status of development. (auth)
Alexander, L.G.; Carter, W.L.; Chapman, R.H.; Kinyon, R.W.; Miller, J.W.; Van Winkle, R.
Ketone Production from the Thermal Decomposition of Carboxylate Salts
The MixAlco process uses an anaerobic, mixed-culture fermentation to convert lignocellulosic biomass to carboxylate salts. The fermentation broth must be clarified so that only carboxylate salts, water, and minimal impurities remain. Carboxylate salts are concentrated by evaporation and thermally decomposed into ketones. The ketones can then be chemically converted to a wide variety of chemicals and fuels. The presence of excess lime in the thermal decomposition step reduced product yield. Mixtures of calcium carboxylate salts were thermally decomposed at 450 degrees C. Low lime-to-salt ratios (g Ca(OH)2/g salt) of 0.00134 and less had a negligible effect on ketone yield. In contrast, salts with higher lime-to-salt ratios of 0.00461, 0.0190, and 0.272 showed 3.5, 4.6, and 9.4% loss in ketone yield, respectively. These losses were caused primarily by increases in tars and heavy oils; however, a three-fold increase in hydrocarbon production occurred as well. To predict ketone product distribution, a random-pairing and a Gibbs free energy minimization model were applied to thermal decompositions of mixed calcium and sodium carboxylate salts. Random pairing appears to better predict ketone product composition. For sodium and calcium acetate, two types of mixed sodium carboxylate salts, and two types of mixed calcium carboxylate salts, activation energy (EA) was determined using three isoconversional methods. For each salt type, EA varied significantly with conversion. The average EA for sodium and calcium acetate was 226.65 and 556.75 kJ/mol, respectively. The average EA for the two mixed sodium carboxylate salts were 195.61, and 218.18 kJ/mol. The average EA for the two mixed calcium carboxylate salts were 232.78, and 176.55 kJ/mol. In addition, three functions of conversion were employed to see which one best modeled the experimental data. The Sestak-Berggren model was the best overall. Possible reactor designs and configurations that address the challenges associated with the continuous thermal decomposition of carboxylate salts are also presented and discussed. Methods of fermentation broth clarification were tested. Flocculation showed little improvement in broth purity. Coagulation yielded broth of 93.23% purity. Filtration using pore sizes from 1 micrometer to 240 Daltons increased broth purity (90.79 to 98.33%) with decreasing pore size.
Landoll, Michael 1984-
Nuclear Reactors
Reactors Nuclear reactors created not only large amounts of plutonium needed for the weapons programs, but a variety of other interesting and useful radioisotopes. They produced...
Molten Carbonate Fuel Cell Product Design Improvement
This annual report provides results of Energy Research Corporation`s technical approach to performing the program `Molten Carbonate Fuel Cell (MCFC) Product Design Improvement` covered under the DOE-ERC Cooperative Agreement DE-FC21-95MC31184. This work is supported by DOE/METC and DOD/DARPA as well as ERC Team funds. The objective of the DOE-sponsored program is to advance the direct carbonate fuel cell technology to a level suitable for commercial entry for civilian applications. The overall objective of the DOD/DARPA initiative is to adapt the civilian 2 MW-Class fuel cell power plant for dual fuel DOD applications. This program is designed to advance the carbonate fuel cell technology from the power plant demonstration status to the commercial entry early production unit design stage. The specific objectives which will allow attainment of these overall program goals are: (1) Provide environmental information to support DOE evaluation with respect to the National Environmental Policy Act (NEPA), (2) Define market-responsive power plant requirements and specifications, (3) Establish design for multifuel, low-cost, modular, market-responsive power plant, (4) Resolve power plant manufacturing issues and define the design for the commercial manufacturing facility, (5) Acquire capabilities to support developmental testing of 0370 stacks and BOP equipment as required to prepare for commercial design, and (6) Resolve stack and BOP equipment technology issues and design, build, and field test a modular commercial prototype power plant to demonstrate readiness of the power plant for commercial entry.
Reactor technology: power conversion systems and reactor operation and maintenance
The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He/sup 3/ reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored.
Powell, J.R.
Salt Creek Student Homepage
Salt Creek Investigation Salt Creek Investigation</2> "Whales Dying in the Pacific Ocean" "Fish Dying in Lake Michigan" Recent headlines remind us of environmental problems near and far away. Scientists have been wondering if these problems could be due to the warmer temperatures this past spring and summer or could there be other reasons? Lack of rain and near drought conditions have forced many areas to restrict water use. We know from past history that pollution affects our drinking water and marine life. Remember what we read about Lake Erie and from reading A River Ran Wild by Lynne Cherry. There are many factors affecting the environment around us . . . even in Salt Creek which runs through our area. We may not be able to investigate the Pacific Ocean and Lake Michigan