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Sample records for salt fuel version

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

    International Nuclear Information System (INIS)

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

    2008-01-01

    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 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 4 , whose melting point is 490 C. The use of 232 Th as a fuel is also being studied. ( 232 Th does not produce Pu under neutron irradiation.) The temperature of the molten salt would be ∼550 C at the inlet (60 C above the solidus temperature) and ∼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 (∼1 mol%) of UF 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 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 233 U production rate is ∼0.6 atoms per 14.1 MeV neutron

  2. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aufiero, M.; Cammi, A.; Fiorina, C. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Leppänen, J. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Luzzi, L., E-mail: lelio.luzzi@polimi.it [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Ricotti, M.E. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy)

    2013-10-15

    In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

  3. Molten fluoride fuel salt chemistry

    International Nuclear Information System (INIS)

    Toth, L.M.; Del Cul, G.D.; Dai, S.; Metcalf, D.H.

    1995-01-01

    The chemistry of molten fluorides is traced from their development as fuels in the Molten Salt Reactor Experiment with important factors in their selection being discussed. Key chemical characteristics such as solubility, redox behavior, and chemical activity are explained as they relate to the behavior of molten fluoride fuel systems. Development requirements for fitting the current state of the chemistry to modern nuclear fuel system are described. It is concluded that while much is known about molten fluoride behavior which can be used effectively to reduce the amount of development required for future systems, some significant molten salt chemical questions must still be addressed. copyright American Institute of Physics 1995

  4. Fuel processing for molten-salt reactors

    International Nuclear Information System (INIS)

    Hightower, J.R. Jr.

    1976-01-01

    Research devoted to development of processes for the isolation of protactinium and for the removal of fission products from molten-salt breeder reactors is reported. During this report period, engineering development progressed on continuous fluorinators for uranium removal, the metal transfer process for rare-earth removal, the fuel reconstitution step, and molten salt--bismuth contactors to be used in reductive extraction processes. The metal transfer experiment MTE-3B was started. In this experiment all parts of the metal transfer process for rare-earth removal are demonstrated using salt flow rates which are about 1 percent of those required to process the fuel salt in a 1000-MW(e) MSBR. During this report period the salt and bismuth phases were transferred to the experimental vessels, and two runs with agitator speeds of 5 rps were made to measure the rate of transfer of neodymium from the fluoride salt to the Bi--Li stripper solution. The uranium removed from the fuel salt by fluorination must be returned to the processed salt in the fuel reconstitution step before the fuel salt is returned to the reactor. An engineering experiment to demonstrate the fuel reconstitution step is being installed. In this experiment gold-lined equipment will be used to avoid introducing products of corrosion by UF 6 and UF 5 . Alternative methods for providing the gold lining include electroplating and mechanical fabrication

  5. LIFE Materails: Molten-Salt Fuels Volume 8

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-12-11

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

  6. LIFE Materails: Molten-Salt Fuels Volume 8

    International Nuclear Information System (INIS)

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

    2008-01-01

    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.

  7. Static fuel molten salt reactors - simpler, cheaper and safer

    International Nuclear Information System (INIS)

    Scott, Ian

    2015-01-01

    refuelling. These reactors are relatively straightforward simplifications of conventional solid fuelled reactors. The fuel assemblies are similar both in design and in construction materials. Replacement of water as coolant with a (fissile free) molten salt removes explosion risks from the reactor containment. There are many possible designs of reactors utilising this form of fuel. One design, a fast spectrum actinide burning reactor called the Stable Salt Reactor has been developed to the stage where realistic capital cost estimates can be made. This was done independently of Moltex Energy by Atkins Ltd. The capital cost (UK prices) for a 1GWe nuclear island was estimated (rough order of magnitude, reflecting the early stage of the design) as £718 per kW, a small fraction of the cost for any conventional nuclear island. Of particular interest to this conference may be the potential for a thorium breeding version of the reactor. Simply replacing the coolant salt with one based on ThF 4 turns the reactor into an efficient 233 U breeder. The basic principles of this version will be described during the talk. (author)

  8. Molten salt burner fuel behaviour and treatment

    International Nuclear Information System (INIS)

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

    2001-01-01

    The objective of this paper is to discuss the feasibility of molten salt reactor technology for treatment of Pu, minor actinides and fission products, when the reactor and fission product clean-up unit are planned as an integral system. This contribution summarises the available R and D which led to selection of the fuel compositions for the molten salt reactor of the TRU burner type (MSB). Special characteristics of behaviour of TRUs and fission products during power operation of MSB concepts are presented. The present paper briefly reviews the processing developments underlying the prior molten salt reactor programmes and relates them to the separation requirements of the MSB concept, including the permissible range of processing cycle times and removal times. Status and development needs in the thermodynamic properties of fluorides, fission product clean-up methods and container materials compatibility with the working fluids for the fission product clean-up unit are discussed. (authors)

  9. Disposition of the fluoride fuel and flush salts from the Molten Salt Reactor experiment at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Peretz, F.J.

    1996-01-01

    The Molten Salt Reactor Experiment (MSRE) is an 8 MW reactor that was operated at Oak Ridge National Laboratory (ORNL) from 1965 through 1969. The reactor used a unique liquid salt fuel, composed of a mixture of LIF, BeF 2 , ZrF 4 , and UF 4 , and operated at temperatures above 600 degrees C. The primary fuel salt circulation system consisted of the reactor vessel, a single fuel salt pump, and a single primary heat exchanger. Heat was transferred from the fuel salt to a coolant salt circuit in the primary heat exchanger. The coolant salt was similar to the fuel salt, except that it contains only LiF (66%) and BeF, (34%). The coolant salt passed from the primary heat exchanger to an air-cooled radiator and a coolant salt pump, and then returned to the primary heat exchanger. Each of the salt loops was provided with drain tanks, located such that the salt could be drained out of either circuit by gravity. A single drain tank was provided for the non-radioactive coolant salt. Two drain tanks were provided for the fuel salt. Since the fuel salt contained radioactive fuel, fission products, and activation products, and since the reactor was designed such that the fuel salt could be drained immediately into the drain tanks in the event of a problem in the fuel salt loop, the fuel salt drain tanks were provided with a system to remove the heat generated by radioactive decay. A third drain tank connected to the fuel salt loop was provided for a batch of flush salt. This batch of salt, similar in composition to the coolant salt, was used to condition the fuel salt loop after it had been exposed to air and to flush the fuel salt loop of residual fuel salt prior to accessing the reactor circuit for maintenance or experimental activities. This report discusses the disposition of the fluoride fuel and flush salt

  10. Fuel processing for molten-salt reactors

    International Nuclear Information System (INIS)

    Hightower, J.R. Jr.

    1975-01-01

    Progress is reported on the development of processes for the isolation of protactinium and for the removal of fission products from molten-salt breeder reactors. The metal transfer experiment MTE-3 (for removing rare earths from MSRE fuel salt) was completed and the equipment used in that experiment was examined. The examination showed that no serious corrosion had occurred on the internal surfaces of the vessels, but that serious air oxidation occurred on the external surfaces of the vessels. Analyses of the bismuth phases indicated that the surfaces in contact with the salts were enriched in thorium and iron. Mass transfer coefficients in the mechanically agitated nondispersing contactors were measured in the Salt/Bismuth Flow-through Facility. The measured mass transfer coefficients are about 30 to 40 percent of those predicted by the preferred literature correlation, but were not as low as those seen in some of the runs in MTE-3. Additional studies using water--mercury systems to simulate molten salt-bismuth systems indicated that the model used to interpret results from previous measurements in the water--mercury system has significant deficiencies. Autoresistance heating studies were continued to develop a means of internal heat generation for frozen-wall fluorinators. Equipment was built to test a design of a side arm for the heating electrode. Results of experiments with this equipment indicate that for proper operation the wall temperature must be held much lower than that for which the equipment was designed. Studies with an electrical analog of the equipment indicate that no regions of abnormally high current density exist in the side arm. (JGB)

  11. Parametric studies on the fuel salt composition in thermal molten salt breeder reactors

    International Nuclear Information System (INIS)

    Nagy, K.; Kloosterman, J.L.; Lathouwers, D.; Van der Hagen, T.H.J.J.

    2008-01-01

    In this paper the salt composition and the fuel cycle of a graphite moderated molten salt self-breeder reactor operating on the thorium cycle is investigated. A breeder molten salt reactor is always coupled to a fuel processing plant which removes the fission products and actinides from the core. The efficiency of the removal process(es) has a large influence on the breeding capacity of the reactor. The aim is to investigate the effect on the breeding ratio of several parameters such as the composition of the molten salt, moderation ratio, power density and chemical processing. Several fuel processing strategies are studied. (authors)

  12. Molten salt fueled reactors with a fast salt draining

    International Nuclear Information System (INIS)

    Ventre, Edmond; Blum, J.M.

    1976-01-01

    This invention relates to a molten salt nuclear reactor which comprises a new arrangement for shutting it down in complete safety. This nuclear reactor has a molten salt primary circuit comprising, in particular, the core of this reactor. It includes a leak tight vessel the capacity of which is appreciably greater than that of the molten salt volume of the circuit and placed so that the level of the molten salt, when all the molten salt of the circuit is contained in this vessel, is less than that of the base of the core. There are facilities for establishing and maintaining an inert gas pressure in the vessel above the molten salt, for releasing the compressed gas and for connecting the vessel to the primary circuit entering this vessel at a lower level than that of the molten salt and enabling molten salt to enter or leave the vessel according to the pressure of the inert gas. The particular advantage of this reactor is that it can be shut down safely since the draining of the primary circuit no longer results from a 'positive action' but from the suppression of an arrangement essential for the operation of the reactor consisting of the build-up of the said inert gas pressure in the said vessel [fr

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  14. Advancing Molten Salts and Fuels at Sandia National Laboratories

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-26

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

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

    NARCIS (Netherlands)

    Capelli, E.

    2016-01-01

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

  16. Thermodynamic characterization of the molten salt reactor fuel - 5233

    International Nuclear Information System (INIS)

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

    2015-01-01

    The Molten Salt Reactor (MSR) has been selected as one of the Generation IV nuclear systems. The very unique feature of this reactor concept is the liquid nature of the fuel which offers numerous advantages concerning the reactor safety. Nowadays, the research in Europe is focused on an innovative concept, the MSFR (Molten Salt Fast Reactor), that combines the generic assets of molten salt as liquid fuel with those related to fast neutron reactors and the thorium fuel cycle. For the design and safety assessment of the MSFR concept, it is extremely important to have a thorough knowledge of the physico-chemical properties of fluorides salts, which is the class of materials that is the best suited for nuclear applications. Potential chemical systems have been critically reviewed and an extensive thermodynamic database describing the most relevant systems has been created at the Institute for Transuranium Elements of the Joint Research Centre (JRC). Thermochemical equilibrium calculations are a very important tool that allows the evaluation of the performance of several salt mixtures predicting their properties and thus the optimization of the fuel composition. The work combines the experimental determination of different salt properties with the modelling of the thermodynamic functions, using the Calphad method. An overview of the experimental work and the thermodynamic assessments will be given in this paper and different fuel options for the MSFR will be discussed. (authors)

  17. Accelerator molten-salt breeding and thorium fuel cycle

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  18. Fuel cycle cost analysis on molten-salt reactors

    International Nuclear Information System (INIS)

    Shimazu, Yoichiro

    1976-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V.; Feynberg, O.; Merzlyakov, A.; Surenkov, A.; Zagnitko, A. [National Research Center, Kurchatov Institute, Moscow (Russian Federation); Afonichkin, V.; Bovet, A.; Khokhlov, V. [Institute of High Temperature Electrochemisty, Ekaterinburg (Russian Federation); Subbotin, V.; Gordeev, M.; Panov, A.; Toropov, A. [Institute of Technical Physics, Snezhinsk (Russian Federation)

    2013-07-01

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

  20. Salt Matters: Preserving Choice, Protecting Health (Short Version)

    Centers for Disease Control (CDC) Podcasts

    2011-02-01

    This video presents information about salt as a major contributor to high blood pressure, heart disease, and stroke.  Created: 2/1/2011 by National Center for Chronic Disease Prevention and Health Promotion (NCCDPHP), Division for Heart Disease and Stroke Prevention (DHDSP).   Date Released: 2/1/2011.

  1. Salt Matters: Preserving Choice, Protecting Health (Long Version)

    Centers for Disease Control (CDC) Podcasts

    2011-02-01

    This video presents information about salt as a major contributor to high blood pressure, heart disease, and stroke.  Created: 2/1/2011 by National Center for Chronic Disease Prevention and Health Promotion (NCCDPHP), Division for Heart Disease and Stroke Prevention (DHDSP).   Date Released: 4/20/2011.

  2. Molten Salt Fuel Cycle Requirements for ADTT Applications

    International Nuclear Information System (INIS)

    Del Cul, G.D.; Toth, L.M.; Williams, D.F.

    1999-01-01

    The operation of an ADT system with the associated nuclear reactions has a profound effect upon the chemistry of the fuel - especially with regards to container compatibility and the chemical separations that may be required. The container can be protected by maintaining the redox chemistry within a relatively narrow, non-corrosive window. Neutron economy as well as other factors require a sophisticated regime of fission product separations. Neither of these control requirements has been demonstrated on the scale or degree of sophistication necessary to support an ADT device. We review the present situation with respect to fluoride salts, and focus on the critical issues in these areas which must be addressed. One requirement for advancement in this area - a supply of suitable materials - will soon be fulfilled by the remediation of ORNLs Molten Salt Reactor Experiment, and the removal of a total of 11,000 kg of enriched (Li-7 > 99.9%) coolant, flush, and fuel salts

  3. Tests of prototype salt stripper system for IFR fuel cycle

    International Nuclear Information System (INIS)

    Carls, E.L.; Blaskovitz, R.J.; Johnson, T.R.; Ogata, T.

    1993-01-01

    One of the waste treatment steps for the on-site reprocessing of spent fuel from the Integral Fast Reactor fuel cycles is stripping of the electrolyte salt used in the electrorefining process. This involves the chemical reduction of the actinides and rare earth chlorides forming metals which then dissolve in a cadmium pool. To develop the equipment for this step, a prototype salt stripper system has been installed in an engineering scale argon-filled glovebox. Pumping trails were successful in transferring 90 kg of LiCl-KCl salt containing uranium and rare earth metal chlorides at 500 degree C from an electrorefiner to the stripper vessel at a pumping rate of about 5 L/min. The freeze seal solder connectors which were used to join sections of the pump and transfer line performed well. Stripping tests have commenced employing an inverted cup charging device to introduce a Cd-15 wt % Li alloy reductant to the stripper vessel

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

    International Nuclear Information System (INIS)

    Choi, In-Kyu; Yeon, Jei-Won; Kim, Won-Ho

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Lelek, Vladimir; Vocka, Radim

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-29

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

  8. Fuel cycle costs for molten-salt reactors

    International Nuclear Information System (INIS)

    Nagashima, Kikusaburo

    1983-01-01

    This report describes FCC (fuel cycle cost) estimates for MSCR (molten-salt converter reactor) and MSBR (molten-salt breeder reactor) compared with those for LWRs (PWR and BWR). The calculation is based on the present worth technique with a given discount rate for each cost item, which enables us to make comparison between FCC's for MSCR, MSBR and LWRs. As far as the computational results obtained here are concerned, shown that the FCC's for MSCR and MSBR are 70 -- 60 % lower than the values for LWRs. And it could be said that the FCC for MSCR (Pu-converter) is about 10 % lower than that for MSBR, because of the smaller amount of fissile inventory of MSCR than the inventory of MSBR. (author)

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

    International Nuclear Information System (INIS)

    Leefe, S.; Jackson-Laver, P.; Scott, I.R.

    2015-01-01

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

  10. Fuel behaviour calculations with version 2.0 of the code FUROM

    International Nuclear Information System (INIS)

    Kulacsy, K.

    2011-01-01

    The fuel modelling code FUROM (FUel ROd Model), suitable for calculating the normal operation condition behaviour of PWR and WWER fuels, has been developed at AEKI for several years. In 2010 the new version of the code, FUROM-2.0 was released. Calculations performed with this version and results are presented. (author)

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

    International Nuclear Information System (INIS)

    Zhang Dalin; Qiu Suizheng; Su Guanghui; Liu Changliang

    2008-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2008-01-01

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

  13. Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Yajuan, E-mail: yajuan.zhong@gmail.com [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences, Taiyuan 030001 (China); Zhang, Junpeng [CAS Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences, Taiyuan 030001 (China); Lin, Jun, E-mail: linjun@sinap.ac.cn [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Xu, Liujun [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Guo, Quangui [CAS Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences, Taiyuan 030001 (China)

    2017-07-15

    Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10{sup −6} K{sup −1} (α{sub ∥}) and 6.15 × 10{sup −6} K{sup −1} (α{sub ⊥}) at the temperature range of 25–700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

  14. Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

    International Nuclear Information System (INIS)

    Zhong, Yajuan; Zhang, Junpeng; Lin, Jun; Xu, Liujun; Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong; Guo, Quangui

    2017-01-01

    Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10 −6 K −1 (α ∥ ) and 6.15 × 10 −6 K −1 (α ⊥ ) at the temperature range of 25–700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

  15. Residual salt separation from simulated spent nuclear fuel reduced in a LiCl-Li2O salt

    International Nuclear Information System (INIS)

    Hur, Jin-Mok; Hong, Sun-Seok; Seo, Chung-Seok

    2006-01-01

    The electrochemical reduction of spent nuclear fuel in LiCl-Li 2 O molten salt for the conditioning of spent nuclear fuel requires the separation of the residual salts from a reduced metal product after the reduction process. Considering the behavior of spent nuclear fuel during the electrochemical reduction process, a surrogate material matrix was constructed and inactive tests on a salt separation were carried out to produce the data required for active tests. Fresh uranium metal prepared from the electrochemical reduction of U 3 O 8 powder was used as the surrogates of the spent nuclear fuel Atomic Energy Society of Japan, Tokyo, Japan, All rights reservedopyriprocess. LiCl, Li 2 O, Y 2 O 3 and SrCl 2 were selected as the components of the residual salts. Interactions between the salts and their influence on the separation of the residual salts were analyzed by differential scanning calorimetry (DSC) and thermogravimetry (TG). Eutectic melting of LiCl-Li 2 O and LiCl-SrCl 2 led to a melting point which was lower than that of the LiCl molten salt was observed. Residual salts were separated by a vaporization method. Co-vaporization of LiCl-Li 2 O and LiCl-SrCl 2 was achieved below the temperatures which could make the uranium metal oxidation by Li 2 O possible. The salt vaporization rates at 950degC were measured as follows: LiCl-8 wt% Li 2 O>LiCl>LiCl-8 wt% SrCl 2 >SrCl 2 . (author)

  16. The concept of fuel cycle integrated molten salt reactor for transmuting Pu+MA from spent LWR fuels

    International Nuclear Information System (INIS)

    Hirose, Y.; Takashima, Y.

    2001-01-01

    Japan should need a new fuel cycle, not to save spent fuels indefinitely as the reusable resources but to consume plutonium and miner actinides orderly without conventional reprocessing. The key component is a molten salt reactor fueled with the Pu+MA (PMA) separated from LWR spent fuels using fluoride volatility method. A double-tiered once-through reactor system can burn PMA down to 5% remnant ratio, and can make PMA virtually free from the HAW to be disposed geometrically. A key issue to be demonstrated is the first of all solubility behavior of trifluoride species in the molten fuel salt of 7 LiF-BeF 2 mixture. (author)

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  18. Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

    2017-11-01

    This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability. (C) 2017 Published by Elsevier B.V.

  19. Evaluation of Fluorine-Trapping Agents for Use During Storage of the MSRE Fuel Salt

    Energy Technology Data Exchange (ETDEWEB)

    Brynestad, J.; Williams, D.F.

    1999-05-01

    A fundamental characteristic of the room temperature Molten Salt Reactor Experiment (MSRE) fuel is that the radiation from the retained fission products and actinides interacts with this fluoride salt to produce fluorine gas. The purpose of this investigation was to identify fluorine-trapping materials for the MSRE fuel salt that can meet both the requirement of interim storage in a sealed (gastight) container and the vented condition required for disposal at the Waste Isolation Pilot Plant (WIPP). Sealed containers will be needed for interim storage because of the large radon source that remains even in fuel salt stripped of its uranium content. An experimental program was undertaken to identify the most promising candidates for efficient trapping of the radiolytic fluorine generated by the MSRE fuel salt. Because of the desire to avoid pressurizing the closed storage containers, an agent that traps fluorine without the generation of gaseous products was sought.

  20. Evaluation of Fluorine-Trapping Agents for Use During Storage of the MSRE Fuel Salt

    International Nuclear Information System (INIS)

    Brynestad, J.; Williams, D.F.

    1999-01-01

    A fundamental characteristic of the room temperature Molten Salt Reactor Experiment (MSRE) fuel is that the radiation from the retained fission products and actinides interacts with this fluoride salt to produce fluorine gas. The purpose of this investigation was to identify fluorine-trapping materials for the MSRE fuel salt that can meet both the requirement of interim storage in a sealed (gastight) container and the vented condition required for disposal at the Waste Isolation Pilot Plant (WIPP). Sealed containers will be needed for interim storage because of the large radon source that remains even in fuel salt stripped of its uranium content. An experimental program was undertaken to identify the most promising candidates for efficient trapping of the radiolytic fluorine generated by the MSRE fuel salt. Because of the desire to avoid pressurizing the closed storage containers, an agent that traps fluorine without the generation of gaseous products was sought

  1. Numerical study on heat transfer characteristics of liquid-fueled molten salt using OpenFOAM

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Bang, In Cheol

    2017-01-01

    To pursue sustainability and safety enhancement of nuclear energy, molten salt reactor is regarded as a promising candidate among various types of gen-IV reactors. Besides, pyroprocessing, which treats molten salt containing fission products, should consider safety related to decay heat from fuel material. For design of molten salt-related nuclear system, it is required to consider both thermal-hydraulic characteristics and neutronic behaviors for demonstration. However, fundamental heat transfer study of molten salt in operation condition is not easy to be experimentally studied due to its large scale, high temperature condition as well as difficulties of treating fuel material. >From that reason, numerical study can have benefit to investigate behaviors of liquid-fueled molten salt in real condition. In this study, open source CFD package OpenFOAM was used to analyze liquid-fueled molten salt loop having internal heat source as a first step of research. Among various molten salts considered as a candidate of liquid fueled molten salt reactors, in this study, FLiBe was chosen as liquid salt. For simulating heat generation from fuel material within fluid flow, volumetric heat source was set for fluid domain and OpenFOAM solver was modified as fvOptions as customized. To investigate thermal-hydraulic behavior of molten salt, CFD model was developed and validated by comparing experimental results in terms of heat transfer and pressure drop. As preliminary stage, 2D cavity simulations were performed to validate the modeling capacity of modified solver of OpenFOAM by comparison with those of ANSYS-CFX. In addition, cases of external heat flux and internal heat source were compared to configure the effect of heat source setting in various operation condition. As a result, modified solver of OpenFOAM considering internal heat source have sufficient modeling capacity to simulate liquid-fueled molten salt systems including heat generation cases. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  3. Development of fuel cycle technology for molten-salt reactor systems

    International Nuclear Information System (INIS)

    Uhlir, J.

    2006-01-01

    Full text: Full text: The Molten-Salt Reactor (MSR) represents one of promising advanced reactor type assigned to the GEN IV reactor systems. It can be operated either as thorium breeder within the Th -133U fuel cycle or as actinide transmuter incinerating transuranium fuel. Essentially the main advantage of MSR comes out from the prerequisite, that this reactor type should be directly connected with the 'on-line' reprocessing of circulating liquid (molten-salt) fuel. This principle should allow very effective extraction of freshly constituted fissile material (233U). Besides, the on-line fuel salt clean up is necessary within a long run to keep the reactor in operation. As a matter of principle, it permits to clear away typical reactor poisons like xenon, krypton, lanthanides etc. and possibly also other products of burned plutonium and transmuted minor actinides. The fuel salt clean up technology should be linked with the fresh MSR fuel processing to continuously refill the new fuel (thorium or transuranics) into the reactor system. On the other hand, the technologies of fresh transuranium molten-salt fuel processing from the current LWR spent fuel and of the on-line reprocessing of MSR fuel represent two killing points of the whole MSR technology, which have to be successfully solved before MSR deployment in the future. There are three main pyrochemical partitioning techniques proposed for processing and/or reprocessing of MSR fuel: Fluoride volatilization processes, Molten salt / liquid metal extraction processes and Electrochemical separation processes. Two of them - Fluoride Volatility Method and Electrochemical separation process from fluoride media are under development in the Nuclear Research Institute Rez pic. R and D in the field of Fluoride Volatility Method is concentrated to the development and verification of experimental semi-pilot technology for LWR spent fuel reprocessing, which may result in a product the form and composition of which might be

  4. SALT 2010 Bilingual S/E Version: A Tool for Assessing the Language Production of Bilingual (Spanish/English) Children

    Science.gov (United States)

    Miller, Jon F.; Iglesias, Aquiles; Rojas, Raul

    2010-01-01

    Assessing the language development of bilingual children can be a challenge--too often, children in the complex process of learning both Spanish and English are under- or over-diagnosed with language disorders. SLPs can change that with "SALT 2010 Bilingual S/E Version" for grades K-3, the first tool to comprehensively assess children's language…

  5. Treatment of waste salt from the advanced spent fuel conditioning process (I): characterization of Zeolite A in Molten LiCl Salt

    International Nuclear Information System (INIS)

    Kim, Jeong Guk; Lee, Jae Hee; Yoo, Jae Hyung; Kim, Joon Hyung

    2004-01-01

    The oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) and the long-lived radioactive nuclides partitioning process based on electro-refining process, which are being developed ay the Korea Atomic Energy Research Institute (KAERI), are to generate two types of molten salt wastes such as LiCl salt and LiCl-KCl eutectic salt, respectively. These waste salts must meet some criteria for disposal. A conditioning process for LiCl salt waste from ACP has been developed using zeolite A. This treatment process of waste salt using zeolite A was first developed by US ANL (Argonne National Laboratory) for LiCl-KCl eutectic salt waste from an electro-refining process of EBR (Experimental Breeder Reactor)-II spent fuel. This process has been developed recently, and a ceramic waste form (CWF) is produced in demonstration-scale V-mixer (50 kg/batch). However, ANL process is different from KAERI treatment process in waste salt, the former is LiCl-KCl eutectic salt and the latter is LiCl salt. Because of melting point, the immobilization of eutectic salt is carried out at about 770 K, whereas LiCl salt at around 920 K. Such difference has an effect on properties of immobilization media, zeolite A. Here, zeolite A in high-temperature (923 K) molten LiCl salt was characterized by XRD, Ion-exchange, etc., and evaluated if a promising media or not

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  7. Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

    2017-11-01

    This paper provides an overview of research evaluating the use of tellurite glass as a waste form for salt wastes from electrochemical processing. The capacities to immobilize different salts were evaluated including: a LiCl-Li2O oxide reduction salt (for oxide fuel) containing fission products, a LiCl-KCl eutectic salt (for metallic fuel) containing fission products, and SrCl2. Physical and chemical properties of the glasses were characterized by using X-ray diffraction, bulk density measurements, chemical durability tests, scanning electron microscopy, and energy dispersive X-ray emission spectroscopy. These glasses were found to accommodate high concentrations of halide salts and have high densities. However, improvements are needed to meet chemical durability requirements.

  8. Molten-salt reactor strategies viewed from fuel conservation effect, (1)

    International Nuclear Information System (INIS)

    Furuhashi, Akira

    1976-01-01

    Saving of material requirements in the long-term fuel cycle is studied by introducing molten-salt reactors with good neutron economy into a projection of nuclear generating capacity in Japan. In this first report an examination is made on the effects brought by the introduction of molten-salt converter reactors starting with Pu which are followed by 233 U breeders of the same type. It is shown that the sharing of some Pu in the light water- and fast breeder-reactor system with molten-salt reactors provides a more rapid transition to the self-supporting, breeding cycle than the simple fast breeding system, thus leading to an appreciable fuel conservation. Considerations are presented on the strategic repartition of generating capacity among reactor types and it is shown that all of the converted 233 U should be promptly invested to molten-salt breeders to quickly establish the dual breeding system, instead of recycling to converters themselves. (auth.)

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

    International Nuclear Information System (INIS)

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

    1996-12-01

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

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

    International Nuclear Information System (INIS)

    Ananich, P.I.; Bunin, E.N.; Kazazyan, V.T.; Nemtsev, V.A.; Sikorin, S.N.

    2001-01-01

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

  11. Sulfate Salts in Gasoline and Ethanol Fuels -- Historical Perspective and Analysis of Available Data

    Energy Technology Data Exchange (ETDEWEB)

    McCormick, Robert L. [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Alleman, Teresa [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Yanowitz, Janet [Ecoengineering, Inc., Sharonville, OH (United States)

    2017-09-21

    This report reviews the chemistry of sulfate salts dissolved in ethanol and gasoline, potential sources of sulfate salts in ethanol and gasoline, the history of consumer vehicle issues with sulfate salt deposits in the early 2000s, and the corresponding changes to the denatured fuel ethanol specification. Recommendations for future research are provided. During a period of rapid market expansion in 2004-05, issues were reported with vehicles running on E10 provided by certain suppliers in some markets. It was commonly believed that these vehicle problems were caused by sulfate salts precipitating from the fuel. Investigators identified sodium sulfate, and in one case also ammonium sulfate, as the predominate salts found in the engines. Several stakeholders believed the issue was excess sulfate ions in the ethanol portion of the E10, and in 2005 the ASTM specification for ethanol (D4806) was modified to include a 4-part per million (ppm) limit on sulfate ions. While there have been no further reports of consumer vehicle issues, the recently approved increase of ethanol in gasoline from 10 to 15 volume percent has resulted in renewed interest in the sulfate ion concentration in fuel ethanol. This report reviews published data on the solubility of sulfate salts in ethanol. The possible sources of sulfate anions and charge balancing cations (such as sodium) in fuel ethanol and petroleum derived blendstocks are discussed. Examination of historical information on the consumer vehicle issues that occurred in 2004-2005 reveals that a source of sodium or ammonium ions, required for the formation of the observed insoluble salts, was never identified. Recommendations for research to better understand sulfate salt solubility issues in ethanol, hydrocarbon blendstocks, and ethanol-gasoline blends are presented.

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

    International Nuclear Information System (INIS)

    Dudnikov, A. A.; Alekseev, P. N.; Subbotin, S. A.

    2007-01-01

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

  13. Romanian experience with rock salt characterisation methods and the implications for disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Diaconu, Daniela; Balan, Valeriu; Mirion, Ilie

    2001-01-01

    The disposal in deep geological formations as rock salt, granite or clay seems to be now the most appropriate solution for final storage of the spent fuel. At this moment, rock salt is one of the Romanian options for spent fuel disposal, but the final decision will be made only after a performance assessment of this geological formation, having as input data the specific characteristics of the salt rock. In order to provide the data requested by the safety assessment programs, the Institute for Nuclear Research - Pitesti developed complex and modern methodologies for thermodynamic parameter determination as well as studies on salt convergence and radionuclide migration. The methodologies pursued to determine those thermal properties specific for spent fuel disposal as dilatation coefficient, heat conductivity and specific heat. The convergence and migration studies pursued a better understanding of these processes, very important in the disposal safety. The paper is a review of those studies and presents the methodologies and the main results obtained on salt samples from Slanic Prahova Salt Mine. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-04-01

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

  15. Waste management analysis for the nuclear fuel cycle. I. Actinide recovery from aqueous salt wastes

    International Nuclear Information System (INIS)

    Martella, L.L.; Navratil, J.D.

    1979-01-01

    A preliminary feasibility study of solvent extraction methods has been completed for removing actinides from selected salt wastes likely to be produced during reactor fuel fabrication and reprocessing. The use of a two-step solvent extraction system, tributyl phosphate (TBP) followed by a bidentate organophosphorus extractant (DHDECMP), appears most efficient for removing actinides from salt waste. The TBP step would remove most of the plutonium and >99.99% of the uranium. The second step, using DHDECMP, would remove >99.91% of the americium, the remaining plutonium (>99.98%), and other actinides from the acidified salt waste

  16. Treatment of waste salt from the advanced spent fuel conditioning process (II) : optimum immobilization condition

    International Nuclear Information System (INIS)

    Kim, Jeong Guk; Lee, Jae Hee; Yoo, Jae Hyung; Kim, Joon Hyung

    2004-01-01

    Since zeolite is known to be stable at a high temperature, it has been reported as a promising immobilization matrix for waste salt. The crystal structure of dehydrated zeolite A breaks down above 1060 K, resulting in the formation of an amorphous solid and re-crystallization to beta-Cristobalite. This structural degradation depends on the existence of chlorides. When contacted to HCl, zeolite 4A is not stable even at 473 K. The optimum consolidation condition for LiCl salt waste from the oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) has been studied using zeolite A since 2001. Actually the constituents of waste salt are water-soluble. And, alkali halides are known to be readily radiolyzed to yield interstitial halogens and metal colloids. For disposal in a geological repository, the waste salt must meet the acceptance criteria. For a waste form containing chloride salt, two of the more important criteria are leach resistance and waste form durability. In this work, we prepared some samples with different mixing ratios of LiCl salt to zeolite A, and then compared some characteristics such as thermal stability, salt occlusion, free chloride content, leach resistance, mixing effect, etc

  17. U-Mo fuels handbook. Version 1.0

    International Nuclear Information System (INIS)

    Rest, Jeffrey; Kim, Yeon Soo; Hofman, Gerard L.; Meyer, Mitchell K.; Hayes, Steven L.

    2006-01-01

    This handbook provides an overview of property data and fuel performance topics with an emphasis on data available for U-Mo alloys. These data often exist only in report format and have not been widely disseminated in the journal literature. For some topics there is more than one source of data, which are sometimes inconsistent. In this situation, the authors have attempted to select the best dataset to provide a standard for fuel designers and reactor operators. Following the section on unirradiated and irradiated materials properties for the monolithic U-Mo alloy, property data for cladding and matrix aluminum are presented. Property data for cladding aluminum are more widely available, and are not presented in great depth. Finally, some properties of (U-Mo)/Al dispersions are also included in this document. Where no data are available, best estimate correlations are provided. Best fits to the data are presented in order to facilitate use by fuel designers and reactor operators.

  18. Fuel rod computations. The COMETHE code in its CEA version

    International Nuclear Information System (INIS)

    Lenepveu, Dominique.

    1976-01-01

    The COMETHE code (COde d'evolution MEcanique et THermique) is intended for computing the irradiation behavior of water reactor fuel pins. It is concerned with steadily operated cylindrical pins, containing fuel pellet stacks (UO 2 or PuO 2 ). The pin consists in five different axial zones: two expansion chambers, two blankets, and a central core that may be divided into several stacks parted by plugs. As far as computation is concerned, the pin is divided into slices (maximum 15) in turn divided into rings (maximum 50). Information are obtained for each slice: the radial temperature distribution, heat transfer coefficients, thermal flux at the pin surface, changes in geometry according to temperature conditions, and specific burn-up. The physical models involved take account for: heat transfer, fission gas release, fuel expansion, and creep of the can. Results computed with COMETHE are compared with those from ELP and EPEL irradiation experiments [fr

  19. Maw and spent HTR Fuel Element Test storage in Boreholes in rock salt

    International Nuclear Information System (INIS)

    Barnert, E.; Brucher, P.H.; Kroth, K.; Merz, E.; Niephaus, D.

    1986-01-01

    The Budesminister fur Forschung und Technolgie (BMFT, Federal Ministry for Research and Technology) is sponsoring a project at the Kernforschungsanlage Julich (KFA, Juelich Nuclear Research Centre) entitled ''MAW and HTR Fuel Element Test disposal in Boreholes.'' The aim of this project is to develop a technique for the final disposal of (1) dissolver sludge, (2) cladding hulls/structural components and (3) spent HTR fuels elements in salt, and to test this technique in the abandoned Asse salt mine, including safety calculations and safety engineering demonstrations. The project is divided into the sub-projects I ''Disposal/sealing technique'' and II ''Retrievable disposal test.''

  20. Evaluation of potential for MSRE spent fuel and flush salt storage and treatment at the INEL

    International Nuclear Information System (INIS)

    Ougouag, A.M.; Ostby, P.A.; Nebeker, R.L.

    1996-09-01

    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

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

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Dudnikov, A.A.; Ignatiev, V.V.; Prusakov, V.N.; Ponomarev-Stepnoy, N.N.; Subbotin, S.A.

    2000-01-01

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

  2. Thermoelastic analyses of spent fuel repositories in bedded and dome salt. Technical memorandum report RSI-0054

    International Nuclear Information System (INIS)

    Callahan, G.D.; Ratigan, J.L.

    1978-01-01

    Global thermoelastic analyses of bedded and dome salt models showed a slight preference for the bedded salt model through the range of thermal loading conditions. Spent fuel thermal loadings should be less than 75 kW/acre of the repository pending more accurate material modeling. One should first limit the study to one or two spent fuel thermal loading (i.e. 75 kW/acre and/or 50 kW/acre) analyses up to a maximum time of approximately 2000 years. Parametric thermoelastic type analyses could then be readily obtained to determine the influence of the thermomechanical properties. Recommendations for further study include parametric analyses, plasticity analyses, consideration of the material interfaces as joints, and possibly consideration of a global joint pattern (i.e. jointed at the same orientation everywhere) for the non-salt materials. Subsequently, the viscoelastic analyses could be performed

  3. Corrosion behavior of spent MTR fuel elements in a drowned salt mine repository

    International Nuclear Information System (INIS)

    Brodda, B.G.; Fachinger, J.

    1995-01-01

    Spent MTR fuel from German Material Test Reactors will not be reprocessed, but stored in a final salt repository in the deep geologic underground. Fuel elements will be placed in POLLUX containers, which are assumed to resist the corrosive attack of an accidentally formed concentrated salt brine for about 500 years. After a container failure the brine would contact the fuel element, corrode the aluminum plating and possibly leach radionuclides from the fuel. A source term for the calculation of radionuclide mobilization results from the investigation of the behavior of MTR fuel in this scenario, which has to be considered for the long-term safety analysis of a deep mined rock salt repository. Experiments with the different plating materials show that the considered aluminum alloys will not resist the corrosive attack of a brine solution, especially in the presence of iron, under the conditions in a drowned salt mine repository. Although differences in the corrosion rates of about two orders of magnitude were observed when applying different parameter sets, the deterioration must be considered to be almost instantaneous in geological terms. Radionuclides are mobilized from irradiated MTR fuel, when the meat of the fuel element becomes accessible to the brine solution. It seems, however, that the radionuclides are effectively trapped by the aluminum hydroxide formed, as the activity concentrations in the brine solution soon reach a constant level with the progressing corrosion of the cladding aluminum. In the presence of iron a more significant initial release was observed, but also in this case an equilibrium activity seems to be reached as a consequence of radionuclide trapping

  4. Investigation and Taguchi Optimization of Microbial Fuel Cell Salt Bridge Dimensional Parameters

    Science.gov (United States)

    Sarma, Dhrupad; Barua, Parimal Bakul; Dey, Nabendu; Nath, Sumitro; Thakuria, Mrinmay; Mallick, Synthia

    2018-01-01

    One major problem of two chamber salt bridge microbial fuel cells (MFCs) is the high resistance offered by the salt bridge to anion flow. Many researchers who have studied and optimized various parameters related to salt bridge MFC, have not shed much light on the effect of salt bridge dimensional parameters on the MFC performance. Therefore, the main objective of this research is to investigate the effect of length and cross sectional area of salt bridge and the effect of solar radiation and atmospheric temperature on MFC current output. An experiment has been designed using Taguchi L9 orthogonal array, taking length and cross sectional area of salt bridge as factors having three levels. Nine MFCs were fabricated as per the nine trial conditions. Trials were conducted for 3 days and output current of each of the MFCs along with solar insolation and atmospheric temperature were recorded. Analysis of variance shows that salt bridge length has significant effect both on mean (with 53.90% contribution at 95% CL) and variance (with 56.46% contribution at 87% CL), whereas the effect of cross sectional area of the salt bridge and the interaction of these two factors is significant on mean only (with 95% CL). Optimum combination was found at 260 mm salt bridge length and 506.7 mm2 cross sectional area with 4.75 mA of mean output current. The temperature and solar insolation data when correlated with each of the MFCs average output current, revealed that both external factors have significant impact on MFC current output but the correlation coefficient varies from MFC to MFC depending on salt bridge dimensional parameters.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-30

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

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

    International Nuclear Information System (INIS)

    Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik

    2015-01-01

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

  7. The light-water-reactor version of the Uranus integral fuel-rod code

    International Nuclear Information System (INIS)

    Moreno, A.; Lassmann, K.

    1977-01-01

    The LWR of the Uranus code, a digital computer programme for the thermal and mechanical analysis of fuel rods, is presented. Material properties are discussed and their effect on integral fuel rod behaviour elaborated via Uranus results for some carefully selected reference experiments. The numerical results do not represent post-irradiation analysis of in-pile experiments, they illustrate rather typical and diverse Uranus capabilities. The performance test shows that Uranus is reliable and efficient, thus the code is a most valuable tool in fuel fod analysis work. K. Lassmann developed the LWR version of the Uranus code, material properties were reviewed and supplied by A. Moreno. (author)

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  9. Disposal Of Spent Fuel In Salt Using Borehole Technology: BSK 3 Concept

    Energy Technology Data Exchange (ETDEWEB)

    Fopp, Stefan; Graf, Reinhold [GNS Gesellschaft fuer Nuklear-Service mbH, Hollestrasse 7A, D-45127 Essen (Germany); Filbert, Wolfgang [DBE TECHNOLOGY GmbH, Eschenstrasse 55, D-31224 Peine (Germany)

    2008-07-01

    The BSK 3 concept was developed for the direct disposal of spent fuel in rock salt. It is based on the conditioning of fuel assemblies and inserting fuel rods into a steel canister which can be placed in vertical boreholes. The BSK 3 canister is suitable for spent fuel rods from 3 PWR or 9 BWR fuel assemblies. The emplacement system developed for the handling and disposal of BSK 3 canisters comprises a transfer cask which provides appropriate shielding during the transport and emplacement process, a transport cart, and an emplacement device. Using the emplacement device the transfer cask will be positioned onto the top of the borehole lock. The presentation describes the development and the design of the transfer cask and the borehole lock. A technically feasible and safe design for the transfer cask and the borehole lock was found regarding the existing safety requirements for radiation shielding, heat dissipation and handling procedure. (authors)

  10. The Light-Water-Reactor Version of the URANUS Integral fuel-rod code

    Energy Technology Data Exchange (ETDEWEB)

    Labmann, K; Moreno, A

    1977-07-01

    The LWR version of the URANUS code, a digital computer programme for the thermal and mechanical analysis of fuel rods, is presented. Material properties are discussed and their effect on integral fuel rod behaviour elaborated via URANUS results for some carefully selected reference experiments. The numerical results do not represent post-irradiation analyses of in-pile experiments, they illustrate rather typical and diverse URANUS capabilities. The performance test shows that URANUS is reliable and efficient, thus the code is a most valuable tool in fuel rod analysis work. K. LaBmann developed the LWR version of the URANUS code, material properties were reviewed and supplied by A. Moreno. (Author) 41 refs.

  11. Reoxidation of uranium in electrolytically reduced simulated oxide fuel during residual salt distillation

    International Nuclear Information System (INIS)

    Eun-Young Choi; Jin-Mok Hur; Min Ku Jeon; University of Science and Technology, Yuseong-gu, Daejeon

    2017-01-01

    We report that residual salt removal by high-temperature distillation causes partial reoxidation of uranium metal to uranium oxide in electrolytically reduced simulated oxide fuel. Specifically, the content of uranium metal in the above product decreases with increasing distillation temperatures, which can be attributed to reoxidation by Li 2 O contained in residual salt (LiCl). Additionally, we estimate the fractions of Li 2 O reacted with uranium metal under these conditions, showing that they decrease with decreasing temperature, and calculate some thermodynamic parameters of the above reoxidation. (author)

  12. Methods of characterization of salt formations in view of spent fuel final disposal

    International Nuclear Information System (INIS)

    Diaconu, Daniela; Balan, Valeriu; Mirion, Ilie

    2002-01-01

    Deep disposal in geological formations of salt, granite and clay seems to be at present the most proper and commonly adopted solution for final disposal of high-level radioactive wastes and spent fuel. Disposing such wastes represents the top-priority issue of the European research community in the field of nuclear power. Although seemingly premature for Romanian power system, the interest for final disposal of spent fuel is justified by the long duration implied by the studies targeting this objective. At the same time these studies represent the Romanian nuclear research contribution in the frame of the efforts of integration within the European research field. Although Romania has not made so far a decision favoring a given geological formation for the final disposal of spent fuel resulting from Cernavoda NPP, the most generally taken into consideration appears the salt formation. The final decision will be made following the evaluation of its performances to spent fuel disposal based on the values of the specific parameters of the geological formation. In order to supply the data required as input parameters in the codes of evaluation of the geological formation performances, the INR Pitesti initiated a package of modern and complex methodologies for such determinations. The studies developed so far followed up the special phenomenon of salt convergence, a phenomenon characteristic for only this kind of rock, as well as the radionuclide migration. These studies allow a better understanding of these processes of upmost importance for disposal's safety. The methods and the experimental installation designed and realized at INR Pitesti aimed at determination of thermal expansion coefficient, thermal conductivity, specific heat, which are all parameters of high specific interest for high level radioactive waste or spent fuel disposal. The paper presents the results of these studies as well as the methodologies, the experimental installations and the findings

  13. Thermodynamic analysis and kinetic modelling of dioxin formation and emissions from power boilers firing salt-laden hog fuel.

    Science.gov (United States)

    Duo, Wenli; Leclerc, Denys

    2007-04-01

    Both organic chlorine (e.g. PVC) and inorganic chlorides (e.g. NaCl) can be significant chlorine sources for dioxin and furan (PCDD/F) formation in combustion processes. This paper presents a thermodynamic analysis of high temperature salt chemistry. Its influence on PCDD/F formation in power boilers burning salt-laden wood waste is examined through the relationships between Cl2, HCl, NaCl(g) and NaCl(c). These analyses show that while HCl is a product of combustion of PVC-laden municipal solid waste, NaCl can be converted to HCl in hog fuel boilers by reactions with SO2 or alumino-silicate materials. Cl2 is a strong chlorinating agent for PCDD/F formation. HCl can be oxidized to Cl2 by O2, and Cl2 can be reduced back to HCl by SO2. The presence of sulphur at low concentrations thus enhances PCDD/F formation by increasing HCl concentrations. At high concentrations, sulphur inhibits de novo formation of PCDD/Fs through Cl2 reduction by excess SO2. The effect of NH3, CO and NOx on PCDD/F formation is also discussed. A semi-empirical kinetic model is proposed. This model considers both precursor and de novo formation mechanisms. A simplified version is used as a stack emission model. The kinetic model indicates that stack dioxin emissions will increase linearly with decreasing electrostatic precipitator (ESP) efficiency and exponentially with increasing ESP temperature.

  14. DART-TM: A thermomechanical version of DART for LEU VHD dispersed and monolithic fuel analysis

    International Nuclear Information System (INIS)

    Saliba, Roberto; Taboada, Horacio; Moscarda, Ma.Virginia; Rest, Jeff

    2003-01-01

    A collaboration agreement between ANL/USDOE and CNEA Argentina, in the area of Low Enriched Uranium Advanced Fuels has been in place since October 16, 1997 under the 'Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy'. An annex concerning DART code optimization has been operative since February 8, 1999. Previously, as a part of this annex a visual thermal FASTDART version was developed that includes mechanistic models for the calculation of the fission-gas-bubble and fuel particle size distribution, reaction layer thickness, and meat thermal conductivity. FASTDART was presented at the last RERTR Meeting that included validation against RERTR 3 irradiation data. The thermal FASTDART version was assessed as an adequate tool for modeling the behavior of LEU U-Mo dispersed fuels under irradiation against PIE RERTR irradiation data. During this past year the development of a 3-D thermo-mechanical version of the code for modeling the irradiation behavior of LEU U-Mo monolithic and dispersion fuel was initiated. Some preliminary results of this work will be shown during RERTR-2003 meeting. (author)

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

    Directory of Open Access Journals (Sweden)

    Rodrigues Davide

    2015-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

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

    International Nuclear Information System (INIS)

    Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui

    2016-01-01

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

  18. Small molten-salt reactors with a rational thorium fuel-cycle

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Mitachi, Kohshi; Kato, Yoshio

    1992-01-01

    In the fission-energy utilization for solving global social and environmental problems including the 'Greenhouse Effect' in the next century, a new strategy should be introduced considering high safety and economy, simplicity, size-flexibility, anti-nuclear proliferation and terrorism, high temperature heat supply, etc., aiming to establish a rational breeding fuelcycle. Thorium Molten-Salt Nuclear Energy Synergetics based on [I] Th utilization, [II] fluid-fuel concept and [III] separation of fissile breeding and power generation functions would be one of the most promising approach. A design study of a standard Molten-Salt Reactor: FUJI-II (350 MWth, 155-161 MWe) ensuring fuel self-sustaining nature (conversion-ratio ∝ 1.0) in spite of small-size, and pilot-plant miniFUJI-II has been proceeded. (orig.)

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

    International Nuclear Information System (INIS)

    Jaskierowicz, S.

    2012-01-01

    The fuel reprocessing of the molten salt reactor (Gen IV concept) is a multi-steps process in which actinides and lanthanides extraction is performed by a reductive extraction technique. The development of an analytic model has showed that the contact between the liquid fuel LiF-ThF 4 and a metallic phase constituted of Bi-Li provide firstly a selective and quantitative extraction of actinides and secondly a quantitative extraction of lanthanides. The control of this process implies the knowledge of saline phase properties. Studies of the physico-chemical properties of fluoride salts lead to develop a technique based on potentiometric measurements to evaluate the fluoro-acidity of the salts. An acidity scale was established in order to classify the different fluoride salts considered. Another electrochemical method was also developed in order to determine the solvation properties of solutes in fluoride F- environment (and particularly ThF 4 by F-) in reductive extraction technique, a metallic phase is also involved. A method to prepare this phase was developed by electro-reduction of lithium on a bismuth liquid cathode in LiCl-LiF melt. This technique allows to accurately control the molar fraction of lithium introduced into the liquid bismuth, which is a main parameter to obtain an efficient extraction. (author)

  20. Achieving salt-cooled reactor goals: economics, variable electricity, no major fuel failures - 15118

    International Nuclear Information System (INIS)

    Forsberg, C.

    2015-01-01

    The Fluoride-salt-cooled High-temperature Reactor (FHR) with a Nuclear air-Brayton Combined Cycle (NACC) and Firebrick Resistance-Heated Energy Storage (FIRES) is a new reactor concept. The FHR uses High-Temperature Gas-cooled Reactor (HTGR) coated-particle fuel and liquid-salt coolants originally developed for molten salt reactors (MSRs) where the fuel was dissolved in the coolant. The FIRES system consists of high-temperature firebrick heated to high temperatures with electricity at times of low electric prices. For a modular FHR operating with a base-load 100 MWe output, the station output can vary from -242 MWe to +242 MWe. The FHR can be built in different sizes. The reactor concept was developed using a top-down approach: markets, requirements, reactor design. The goals are: (1) increase plant revenue by 50 to 100% relative to base-load nuclear plants with capital costs similar to light-water reactors, (2) enable a zero-carbon nuclear renewable electricity grid, and (3) no potential for major fuel failure and thus no potential for major radionuclide offsite releases in a beyond-design-basis accident (BDBA). The basis for the goals and how they may be achieved is described

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-01

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

  2. Validation of the metal fuel version of the SAS4A accident analysis code

    International Nuclear Information System (INIS)

    Tentner, A.M.

    1991-01-01

    This paper describes recent work directed towards the validation of the metal fuel version of the SAS4A accident analysis code. The SAS4A code system has been developed at Argonne National Laboratory for the simulation of hypothetical severe accidents in Liquid Metal-Cooled Reactors (LMR), designed to operate in a fast neutron spectrum. SAS4A was initially developed for the analysis of oxide-fueled liquid metal-cooled reactors and has played an important role in the simulation and assessment of the energetics potential for postulated severe accidents in these reactors. Due to the current interest in the metal-fueled liquid metal-cooled reactors, a metal fuel version of the SAS4A accident analysis code is being developed in the Integral Fast Reactor program at Argonne. During such postulated accident scenarios as the unprotected (i.e. without scram) loss-of-flow and transient overpower events, a large number of interrelated physical phenomena occur during a relatively short time. These phenomena include transient heat transfer and hydrodynamic events, coolant boiling, and fuel and cladding melting and relocation. Due to strong neutronic feedbacks these events can significantly influence the reactor power history in the accident progression. The paper presents the results of a recent SAS4A simulation of the M7 TREAT experiment. 6 refs., 5 figs

  3. Effects of gaseous radioactive nuclides on the design and operation of repositories for spent LWR fuel in rock salt

    International Nuclear Information System (INIS)

    Jenks, G.H.

    1979-12-01

    Information relating to the identities and amounts of gaseous radionuclides present in spent LWR fuel and to their release from canistered spent fuel under plausible storage and disposal conditions was assembled, reviewed, and analyzed. Information was also reviewed and analyzed on several other subjects that relate to the integrity of the carbon steel canister in which the spent fuel is to be encapsulated and to the expected rates of transfer of gaseous radionuclides through crushed salt backfill within a disposal room in a reference repository in rock salt. The advantages and disadvantages were considered for several different canister-backfill materials, and recommendations were made regarding preferred materials. Other recommendations relate to encapsulation procedures and specifications and to needs for additional experimental studies. The objective of this work was to provide reference information, conclusions, and recommendations that could be used to establish design and operating conditions and procedures for a bedded salt repository for spent LWR fuel and that could also be used to help evaluate the safety of the repository. The results of this work will also generally apply to spent fuel repositories in domal salt. However, because the domal salt may have little or no brine inclusions within it, there may be little or no possibility that brine will migrate into open spaces around an emplaced canister. Addordingly, some of the concerns that result from the possible occurrence of brine migration in bedded salt may be of no importance in domal salt

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  5. Sources of variation in δ13C of fossil fuel emissions in Salt Lake City, USA

    International Nuclear Information System (INIS)

    Bush, S.E.; Pataki, D.E.; Ehleringer, J.R.

    2007-01-01

    The isotopic composition of fossil fuels is an important component of many studies of C sources and sinks based on atmospheric measurements of CO 2 . In C budget studies, the isotopic composition of crude petroleum and CH 4 are often used as a proxy for the isotopic composition of CO 2 emissions from combustion. In this study, the C isotope composition (δ 13 C) of exhaust from the major fossil fuel emission sources in Salt Lake City, USA, was characterized with 159 measurements of vehicle exhaust of various types and eight measurements of residential furnace exhaust. These two sources were found to be isotopically distinct, and differed from global-scale estimates based on average values for crude petroleum and CH 4 . Vehicle-specific factors such as engine load and operation time had no effect on δ 13 C of vehicle exhaust. A small difference was found between the mean δ 13 C of vehicle exhaust collected randomly from different vehicles and the mean δ 13 C of gasoline collected from multiple fueling stations representing major gasoline distributors in Salt Lake City and the surrounding area. However, a paired comparison of δ 13 C of exhaust and gasoline for six different vehicles did not show any consistent C isotope fractionation during vehicle combustion. The mean δ 13 C of crude petroleum processed for local distribution differed slightly from refined gasoline collected at multiple fueling stations, but time lags between processing and transportation cannot be ruled out as an uncontrollable contributing factor. Measured isotope ratios were then combined with fuel consumption statistics to predict the annual cycle of δ 13 C of fossil fuel emissions for the Salt Lake City metropolitan area. The results showed that the isotopic composition of CO 2 emissions from fossil fuel combustion varied by almost 3 per mille over the course of the 2002 calendar year. This study illustrates that on a regional scale, the isotopic composition of fossil fuel emissions shows

  6. Preliminary analysis on in-core fuel management optimization of molten salt pebble-bed reactor

    International Nuclear Information System (INIS)

    Xia Bing; Jing Xingqing; Xu Xiaolin; Lv Yingzhong

    2013-01-01

    The Nuclear Hot Spring (NHS) is a molten salt pebble-bed reactor featured by full power natural circulation. The unique horizontal coolant flow of the NHS demands the fuel recycling schemes based on radial zoning refueling and the corresponding method of fuel management optimization. The local searching algorithm (LSA) and the simulated annealing algorithm (SAA), the stochastic optimization methods widely used in the refueling optimization problems in LWRs, were applied to the analysis of refueling optimization of the NHS. The analysis results indicate that, compared with the LSA, the SAA can survive the traps of local optimized solutions and reach the global optimized solution, and the quality of optimization of the SAA is independent of the choice of the initial solution. The optimization result gives excellent effects on the in-core power flattening and the suppression of fuel center temperature. For the one-dimensional zoning refueling schemes of the NHS, the SAA is an appropriate optimization method. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-01

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

  8. Proposals on the organization of a fuel cycle of the cascade sub-critical molten salt reactor (CSMSR)

    International Nuclear Information System (INIS)

    Bychkov, A.V.; Kormilitsyn, M.V.; Melnik, M.I.; Babikov, L.G.; Ponomarev, L.I.

    2002-01-01

    At present the approach of burning out long-lived radioactive waste (RW) in the reactor core neutron flux is the most feasible one. Currently the way of closing nuclear fuel cycle (NFC) on the basis of the nuclear chemical concept of the cascade sub-critical molten salt reactor (CSMSR) is considered as the most promising one. It is characterised by a number of advantages. CSMSR controlled by a beam of protons or electrons is the optimal reactor for closing the NFC using non-aqueous fluoride methods of fuel reprocessing. They, in comparison with aqueous methods, are characterised by a small waste quantity and are less laborious because of the absence of severe requirements to the product purity. A high productivity of high-temperature electrochemical processes allows the implementation of the fuel recycling process as part of the CSMSR total technological cycle. It can be conducted in the 'on-line' mode in the bypass molten salt circuit that brings the transportation volume of high-activity materials to a minimum. In order to reprocess the CSMSR irradiated molten salt fuel on the basis of salt composition LiF-NaF-(BeF 2 ) an option, based on the following three main operations of the melt treatment, was proposed at SSC RF RIAR: (i) On-line argon treatment of molten salt fuel for removal of gaseous fission products (FP) and also FP that form volatile fluorides and aerosols; (ii) Organisation of the fuel-active metal (probably with a fine-dispersed plutonium alloy) interaction in the on-line mode for removal of 'noble' and 'semi-noble' FP and corrosion products such as Ni, Fe, Cr (when using Pu alloy it allows to regenerate at the same time of the burned-out plutonium component); (iii) Portion-by-portion (fuel composition partially being removed from the CSMSR molten salt circuit) pyroelectrochemical reprocessing of the molten salt composition aimed at the removal of lanthanides - FP followed by a return of actinides to the CSMSR fuel cycle. This technology will allow

  9. Estimation of the development possibility of the ABC/ATW fuel cycle based on LiF-BeF2 fuel salt. Part 2

    International Nuclear Information System (INIS)

    Bychkov, A.V.; Naumov, V.S.

    1994-01-01

    The aim of the first chapter was generalization of data on solubility and equilibrium states of fission product and actinide fluorides in fluoride salt melts-solvents and fuel composition melts based on LiF-BeF 2 mixture which was proposed as fuel basis for ABC/ATW facility. The second chapter is devoted to description of processes proposed for the chemical-technological complex of the ABC/ATW facility and their physico-chemical peculiarities. The complex is responsible for the removal of fission products and actinides from irradiated fuel salt

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  11. Investigation of an Alternative Fuel Form for the Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

    International Nuclear Information System (INIS)

    Casino, William A. Jr.

    2006-01-01

    Much of the recent studies investigating the use of liquid salts as reactor coolants have utilized a core configuration of graphite prismatic fuel block assemblies with TRISO particles embedded into cylindrical fuel compacts arranged in a triangular pitch lattice. Although many calculations have been performed for this fuel form in gas cooled reactors, it would be instructive to investigate whether an alternative fuel form may yield improved performance for the liquid salt-cooled Very High Temperature Reactor (LS-VHTR). This study investigates how variations in the fuel form will impact the performance of the LS-VHTR during normal and accident conditions and compares the results with a similar analysis that was recently completed for a LS-VHTR core made up of prismatic block fuel. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering; Greenspan, Ehud [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering

    2015-02-09

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

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

    International Nuclear Information System (INIS)

    Peterson, Per; Greenspan, Ehud

    2015-01-01

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3 . This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel

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

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

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

  15. The risk-rewards structure of using spent nuclear fuel in molten salt reactor - 5513

    International Nuclear Information System (INIS)

    He, X.; Du, Z.; Macian-Juan, R.; Seidl, M.

    2015-01-01

    The molten salt reactor concept naturally lends itself to a re-use of fuel either by online reprocessing or by using spent nuclear fuel as part of the driver fuel. Moreover some well-known safety advantages over traditional LWR designs are promised: the primary loop can be operated at atmospheric pressure, refueling can be done online, only a minimum amount of excess reactivity needs to be stored inside the core and the continuous circulation and inter-mixing of the fuel results in a more homogenous redistribution of fission products. In this paper the feasibility of running a molten salt reactor on spent LWR fuel is discussed in a number of scenarios in order to make the various trade-offs transparent: using SNF in a classic graphite moderated MSR and doing the same for a lead-cooled dual-fluid MSR. From a commercial company's point of view the MSR concept faces already substantial risks even without the use of SNF: licensing concerns due to an enrichment of fissile nuclides typically above 5% of heavy metal mass, limited practical experience with the reliability of proposed MSR materials and almost no experience with online reprocessing. For one thing one could therefore aim for the most conservative design which would rely on the design of ORNL's graphite moderated MSR operated in the sixties. While appearing realistic from a technical perspective, the potential for SNF re-use in the sense of actinide destruction appears limited. On the other hand one can maximize the risk and the potential payoff by concentrating on the most speculative design, i.e. a dual fluid reactor with an ultra-hard neutron spectrum in order to most efficiently burn higher actinides. In this paper the neutronic design calculations for the above described MSR concepts are presented in order to maximize SNF's contribution for the reactors' energy generation and their potential for actinide destruction. Among the optimization parameters are the lattice constants, the type

  16. Identification and evaluation of alternatives for the disposition of fluoride fuel and flush salts from the molten salt reactor experiment at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1996-01-01

    This document presents an initial identification and evaluation of the alternatives for disposition of the fluoride fuel and flush salts stored in the drain tanks at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). It will serve as a resource for the U.S. Department of Energy contractor preparing the feasibility study for this activity under the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA). This document will also facilitate further discussion on the range of credible alternatives, and the relative merits of alternatives, throughout the time that a final alternative is selected under the CERCLA process

  17. A Thermodynamic Model for the Fuel of a Molten Salt Actinide Burner

    Energy Technology Data Exchange (ETDEWEB)

    Benes, Ondrej [Institute of Chemical Technology, Technicka 5, 16603 Prague (Czech Republic); European Commission - Joint Research Centre - Institute for Transuranium Elements, P.O. BOX 2340, 76125 Karlsruhe (Germany)

    2008-07-01

    In this study the importance of the thermodynamic description of a multi-component system when optimizing the fuel choice for a molten salt reactor is demonstrated. It is shown on the MF-PuF{sub 3} (M=Li,Na,K,Rb) system, one of the fuel alternatives, how properties such as vapour pressure or the solubility of the actinides in the alkali halide matrix can be obtained. Moreover it is shown that much bigger PuF{sub 3} solubility is achieved in the matrix containing only alkali halides than in a matrix that contains some concentrations of BeF{sub 2}. In order to obtain full thermodynamic description of the MF-PuF{sub 3} (M=Li,Na,K,Rb,Cs) system all the binary phase diagrams must be assessed. This is done according to the CALPHAD method including the critical review of all available data followed by an interactive optimization of the phase diagram to achieve the best possible agreement between the measurement and the calculation. A novel approach of obtaining the excess enthalpies of the (Rb,Cs)F solid solution by Ab initio has been used and the results are compared to the experimentally determined phase diagram measured in this study as well. For the measurement of the phase diagrams of the volatile fluoride salts special encapsulation technique has been developed. (authors)

  18. Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

    International Nuclear Information System (INIS)

    Shi, Chengbin; Cheng, Maosong; Liu, Guimin

    2016-01-01

    Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

  19. Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Chengbin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Cheng, Maosong, E-mail: mscheng@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Liu, Guimin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

    2016-08-15

    Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-02-15

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

  1. Dynamic tritium inventory of a NET/ITER fuel cycle with lithium salt solution blanket

    International Nuclear Information System (INIS)

    Spannagel, G.; Gierszewski, P.

    1991-01-01

    At the Karlsruhe Nuclear Research Center (KfK) a flexible tool is being developed to simulate the dynamics of tritium inventories. This tool can be applied to any tritium handling system, especially to the fuel cycle components of future nuclear fusion devices. This instrument of simulation will be validated in equipment to be operated at the Karlsruhe Tritium Laboratory. In this study tritium inventories in a NET/ITER type fuel cycle involving a lithium salt solution blanket are investigated. The salt solution blanket serves as an example because it offers technological properties which are attractive in modeling the process; the example does not impair the general validity of the tool. Usually, the operation strategy of complex structures will deteriorate due to failures of the subsystems involved. These failures together with the reduced availability ensuing from them will be simulated. The example of this study is restricted to reduced availabilities of two subsystems, namely the reactor and the tritium recovery system. For these subsystems the influence of statistically varying intervals of operation is considered. Strategies we selected which are representative of expected modes of operation. In the design of a fuel cycle, care will be taken that prescribed availabilities of the subsystems can be achieved; however, the description of reactor operation is a complex task since operation breaks down into several campaigns for which rules have been specified which enable determination of whether a campaign has been successful and can be stopped. Thus, it is difficult to predict the overall behavior prior to a simulation which includes stochastic elements. Using the example mentioned above the capabilities of the tool will be illustrated; besides the presentation of results of inventory simulation, the applicability of these data will be discussed. (orig.)

  2. Modeling CANDU type fuel behaviour during extended burnup irradiations using a revised version of the ELESIM code

    International Nuclear Information System (INIS)

    Arimescu, V.I.; Richmond, W.R.

    1992-05-01

    The high-burnup database for CANDU fuel, with a variety of cases, offers a good opportunity to check models of fuel behaviour, and to identify areas for improvement. Good agreement of calculated values of fission-gas release, and sheath hoop strain, with experimental data indicates that the global behaviour of the fuel element is adequately simulated by a computer code. Using, the ELESIM computer code, the fission-gas release, swelling, and fuel pellet expansion models were analysed, and changes made for gaseous swelling, and diffusional release of fission-gas atoms to the grain boundaries. Using this revised version of ELESIM, satisfactory agreement between measured values of fission-gas release was found for most of the high-burnup database cases. It is concluded that the revised version of the ELESIM code is able to simulate with reasonable accuracy high-burnup as well as low-burnup CANDU fuel

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

    International Nuclear Information System (INIS)

    Forsberg, C. W.; Snead, Lance Lewis; Katoh, Yutai

    2012-01-01

    The FHR is a new reactor concept that uses coated-particle fuel and a low-pressure liquid-salt coolant. Its neutronics are similar to a high-temperature gas-cooled reactor (HTGR). The power density is 5 to 10 times higher because of the superior cooling properties of liquids versus gases. The leading candidate coolant salt is a mixture of 7 LiF and BeF 2 (FLiBe) possessing a boiling point above 1300 C and the figure of merit ρC 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.

  4. Reuse of spent fuel cladding Zr by molten salt toward advanced recycle society

    International Nuclear Information System (INIS)

    Amano, Osamu; Kobayashi, Hiroaki; Suzuki, Kazunori; Yasuike, Y.; Sato, Nobuaki

    2003-01-01

    Cladding tubes of zircaloy 95% generated from reprocessing process for spent nuclear fuels are to be chopped in about 3 cm length, compacted and solidified with cements. This paper reports the summary of investigation of the present possible techniques for zirconium recovery: (1) electrolysis of molten salts (Zr-chlorides and/or fluorides) and (2) separation as volatile zirconium chlorides (ZrCl 4 ) (chloride volatility process) followed by reaction with metallic magnesium at 900degC to produce sponged Zr (Kroll method). The feasibility are discussed from the point of view of reduction of secondary radioactive wastes, accumulation of such nuclides as Co-60 and Ni-63 in electrolytic basin, radioactivity estimation in the products, and also problems of cleaning and reducing chemicals. (S. Ohno)

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

    International Nuclear Information System (INIS)

    Kim Davies; Shelly X Li

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim Davies; Shelly X Li

    2007-09-01

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

  7. Thermal-Hydraulics Phenomena Important in Modeling and Simulation of Liquid-Fuel Molten Salt Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bajorek, Stephen; Diamond, David J.

    2018-11-11

    This paper discusses liquid-fuel molten salt reactors, how they will operate under normal, transient, and accident conditions, and the results of an expert elicitation to determine the corresponding thermalhydraulic phenomena important to understanding their behavior. Identifying these phenomena will enable the U.S. Nuclear Regulatory Commission (NRC) to develop or identify modeling functionalities and tools required to carry out confirmatory analyses that examine the validity and accuracy of an applicant’s calculations and help determine the margin of safety in plant design. NRC frequently does an expert elicitation using a Phenomena Identification and Ranking Table (PIRT) to identify and evaluate the state of knowledge of important modeling phenomena. However, few details about the design of these reactors and the sequence of events during accidents are known, so the process used was considered a preliminary PIRT. A panel met to define phenomena that would need to be modeled and considered the impact/importance of each phenomenon with respect to specific figures-of-merit (FoMs) (e.g., salt temperature, velocity, and composition). Each FoM reflected a potential impact on radionuclide release or loss of a barrier to release. The panel considered what the path forward might be with respect to being able to model the phenomenon in a simulation code. Results are explained for both thermal and fast spectrum designs.

  8. Development of cesium phosphotungstate salt and chitosan composite membrane for direct methanol fuel cells.

    Science.gov (United States)

    Xiao, Yanxin; Xiang, Yan; Xiu, Ruijie; Lu, Shanfu

    2013-10-15

    A novel composite membrane has been developed by doping cesium phosphotungstate salt (CsxH3-xPW12O40 (0≤x≤3), Csx-PTA) into chitosan (CTS/Csx-PTA) for application in direct methanol fuel cells (DMFCs). Uniform distribution of Csx-PTA nanoparticles has been achieved in the chitosan matrix. The proton conductivity of the composite membrane is significantly affected by the Csx-PTA content in the composite membrane as well as the Cs substitution in PTA. The highest proton conductivity for the CTS/Csx-PTA membranes was obtained with x=2 and Cs2-PTA content of 5 wt%. The value is 6×10(-3) S cm(-1) and 1.75×10(-2) S cm(-1) at 298 K and 353 K, respectively. The methanol permeability of CTS/Cs2-PTA membrane is about 5.6×10(-7), 90% lower than that of Nafion-212 membrane. The highest selectivity factor (φ) was obtained on CTS/Cs2-PTA-5 wt% composite membrane, 1.1×10(4)/Scm(-3)s. The present study indicates the promising potential of CTS/Csx-PTA composite membrane as alternative proton exchange membranes in direct methanol fuel cells. Copyright © 2013 Elsevier Ltd. All rights reserved.

  9. Fuel Cell Power Model Version 2: Startup Guide, System Designs, and Case Studies. Modeling Electricity, Heat, and Hydrogen Generation from Fuel Cell-Based Distributed Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Steward, D.; Penev, M.; Saur, G.; Becker, W.; Zuboy, J.

    2013-06-01

    This guide helps users get started with the U.S. Department of Energy/National Renewable Energy Laboratory Fuel Cell Power (FCPower) Model Version 2, which is a Microsoft Excel workbook that analyzes the technical and economic aspects of high-temperature fuel cell-based distributed energy systems with the aim of providing consistent, transparent, comparable results. This type of energy system would provide onsite-generated heat and electricity to large end users such as hospitals and office complexes. The hydrogen produced could be used for fueling vehicles or stored for later conversion to electricity.

  10. Estimates of relative areas for the disposal in bedded salt of LWR wastes from alternative fuel cycles

    International Nuclear Information System (INIS)

    Lincoln, R.C.; Larson, D.W.; Sisson, C.E.

    1978-01-01

    The relative mine-level areas (land use requirements) which would be required for the disposal of light-water reactor (LWR) radioactive wastes in a hypothetical bedded-salt formation have been estimated. Five waste types from alternative fuel cycles have been considered. The relative thermal response of each of five different site conditions to each waste type has been determined. The fuel cycles considered are the once-through (no recycle), the uranium-only recycle, and the uranium and plutonium recycle. The waste types which were considered include (1) unreprocessed spent reactor fuel, (2) solidified waste derived from reprocessing uranium oxide fuel, (3) plutonium recovered from reprocessing spent reactor fuel and doped with 1.5% of the accompanying waste from reprocessing uranium oxide fuel, (4) waste derived from reprocessing mixed uranium/plutonium oxide fuel in the third recycle, and (5) unreprocessed spent fuel after three recycles of mixed uranium/plutonium oxide fuels. The relative waste-disposal areas were determined from a calculated value of maximum thermal energy (MTE) content of the geologic formations. Results are presented for each geologic site condition in terms of area ratios. Disposal area requirements for each waste type are expressed as ratios relative to the smallest area requirement (for waste type No. 2 above). For the reference geologic site condition, the estimated mine-level disposal area ratios are 4.9 for waste type No. 1, 4.3 for No. 3, 2.6 for No. 4, and 11 for No. 5

  11. Ultimate storage of spent fuel elements of the AVR test power station in the Asse salt mine

    International Nuclear Information System (INIS)

    Wolf, J.

    1975-02-01

    With regard to the ultimate storage of irradiated AVR pebble-bed reactor carbide fuel elements in the saline of Asse, a number of tests and calculations are presented to demonstrate that there is no credible possibility of the MCA (maximum credible accident) defined for the saline. The safety of persons is not threatened during the operation of spent fuel storage nor at any later time (extrapolation up to approx. 1,000 years after storage). 1,000 fuel elements at a time are packed up in gas-tight containers which are stacked in boreholes. The boreholes are then sealed with concrete. Lay-out and functions of the special airlock and transportation systems - from the packing of the containers in a hot cell to the final storage in the borehole - are described with special reference to aspects of the safety of the overall procedure. The possible accidents in the mine are discussed in detail. 85 Kr and T release rates are determined in laboratory tests by heating of the spherical fuel elements. Tests with fuel elements embedded in salt or stagnant brine were carried out at varies temperatures to investigate their behaviour in final storage. Kr and T release, extraction of fission products, mechanical resistance and corrosion were examined in these tests. Finally, the permeability of salt and salt concrete to radioactive gases were investigated in a special experimental arrangement. The diffusion and permeability coefficients obtained for 85 Kr, HT and HTO allow an estimation of the gas discharge of the stored fuel element. (RB/AK) [de

  12. Neutronics Phenomena Important in Modeling and Simulation of Liquid-Fuel Molten Salt Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, David J.

    2018-11-11

    This paper discusses liquid-fuel molten salt reactors, how they will operate under normal, transient, and accident conditions, and the results of an expert elicitation to determine the corresponding neutronic phenomena important to understanding their behavior. Identifying these phenomena will enable the U.S. Nuclear Regulatory Commission (NRC) to develop or identify modeling functionalities and tools required to carry out confirmatory analyses that examine the validity and accuracy of applicants’ calculations and help determine the margin of safety in plant design. NRC frequently does an expert elicitation using a Phenomena Identification and Ranking Table (PIRT) to identify and evaluate the state of knowledge of important modeling phenomena. However, few details about the design of these reactors and the sequence of events during accidents are known, so the process used was considered a preliminary PIRT. A panel met to define phenomena that would need to be modeled and considered the impact/importance of each phenomenon with respect to specific figures-of-merit (FoMs) (e.g., power distribution, fluence, kinetics parameters and reactivity). Each FoM reflected a potential impact on radionuclide release or loss of a barrier to release. The panel considered what the path forward might be with respect to being able to model the phenomenon in a simulation code. Results are explained for both thermal and fast spectrum designs.

  13. Design study on advanced nuclear fuel recycle system. Conceptual design study of recycle system using molten salt

    International Nuclear Information System (INIS)

    Kasai, Y.; Kakehi, I.; Moro, T.; Higashi, T.; Tobe, K.; Kawamura, F.; Yonezawa, S.; Yoshiuji, T.

    1998-10-01

    Advanced recycle system engineering group of OEC (Oarai Engineering Center) has being carried out a design study of the advanced nuclear fuel recycle system using molten salt (electro-metallurgical process). This system is aiming for improvements of fuel cycle economy and reduction of environmental burden (MA recycles, Minimum of radioactive waste disposal), and also improvement of safety and nuclear non-proliferation. This report describes results of the design study that has been continued since December 1996. (1) A design concept of the advanced nuclear fuel recycle system, that is a module type recycles system of pyrochemical reprocessing and fuel re-fabrication was studied. The module system has advantage in balance of Pu recycle where modules are constructed in coincidence with the construction plan of nuclear power plants, and also has flexibility for technology progress. A demonstration system, minimum size of the above module, was studies. This system has capacity of 10 tHM/y and is able to demonstrate recycle technology of MOX fuel, metal fuel and nitride fuel. (2) Each process of the system, which are pyrochemical electrorefining system, cathode processor, de-cladding system, waste disposal system, etc., were studied. In this study, capacity of an electrorefiner was discussed, and vitrification experiment of molten salt using lead-boric acid glass was conducted. (3) A hot cell system and material handling system of the demonstration system was studied. A robot driven by linear motor was studied for the handling system, and an arrangement plan of the cell system was made. Criticality analysis in the cell system and investigation of material accountancy system of the recycle plant were also made. This design study will be continued in coincidence with design study of reactor and fuel, aiming to establish the concept of FBR recycle system. (author)

  14. Matpro--version 10: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    International Nuclear Information System (INIS)

    Reymann, G.A.

    1978-02-01

    The materials properties correlations and computer subcodes (MATPRO--Version 10) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory are described. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures

  15. MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    International Nuclear Information System (INIS)

    Hagrman, D.L.; Reymann, G.A.

    1979-02-01

    This handbook describes the materials properties correlations and computer subcodes (MATPRO-Version 11) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures

  16. MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    Energy Technology Data Exchange (ETDEWEB)

    Hagrman, D.L.; Reymann, G.A. (comps.)

    1979-02-01

    This handbook describes the materials properties correlations and computer subcodes (MATPRO-Version 11) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures.

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

    International Nuclear Information System (INIS)

    Boussier, H.; Heuer, D.

    2010-01-01

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

  18. Fabrication drawings of fuel pins for FUJI project among PSI, JNC and NRG. Revised version

    International Nuclear Information System (INIS)

    Ozawa, Takayuki; Nakazawa, Hiroaki; Abe, Tomoyuki; Nagayama, Masahiro

    2002-02-01

    Irradiation tests and post-irradiation examinations in the framework of JNC-PSI-NRG collaboration project will be performed in 2003-2005. Irradiation fuel pins will be fabricated by the middle of 2003. The fabrication procedure for irradiation fuel pins has been started in 2001. Several fabrication tests and qualification tests in JNC and PSI (Paul Scherrer Institute, Switzerland) have been performed before the fuel pin fabrication. According to the design assignment between PSI and JNC in the frame of this project, PSI should make a specification document for the fuel pellet, the sphere-pac fuel particles, the vipac fuel particles, and the fuel pin. JNC should make a fabrication drawing for irradiation pins. JNC has been performed the fuel design in cooperation with PSI and NRG (Nuclear Research and Consultancy Group, Netherlands). In this project, the pelletized fuel, the sphere-pac fuel, and the vipac fuel will be simultaneously irradiated on HFR (High Flux Reactor, Netherlands). This fabrication drawing has been made under the design assignment with PSI, and consists of the drawing of MOX pellet, thermal insulator pellet, pin components, fuel segments, and the constructed pin. The fabrication drawings were approved in October 2001, but after that, the optimization of specifications has been discussed and agreed among all partners. In this report, the revised fabrication drawings will be shown. Based on the commission of Plutonium Fuel Technology Group, Advanced Fuel Recycle Technology Division, this design work has been performed in Fuel Design and Evaluation Group, Plutonium Fuel Fabrication Division, Plutonium Fuel Center. (author)

  19. Fabrication drawings of fuel pins for FUJI project among PSI, JNC and NRG. Revised version 2

    International Nuclear Information System (INIS)

    Ozawa, Takayuki; Nakazawa, Hiroaki; Abe, Tomoyuki; Nagayama, Masahiro

    2002-10-01

    Irradiation tests and post-irradiation examinations in the framework of JNC-PSI-NRG collaboration project will be performed in 2003-2005. Irradiation fuel pins will be fabricated by the middle of 2003. The fabrication procedure for irradiation fuel pins has been started in 2001. Several fabrication tests and qualification tests in JNC and PSI (Paul Scherrer Institut, Switzerland) have been performed before the fuel pin fabrication. According to the design assignment between PSI and JNC in the frame of this project, PSI should make specification documents for the fuel pellet, the sphere-pac fuel particles, the vipac fuel fragments, and the fuel segment fabrication. JNC should make the fabrication drawings for irradiation pins. JNC has been performed the fuel design in cooperation with PSI and NRG (Nuclear Research and Consultancy Group, Holland). In this project, the pelletized fuel, the sphere-pac fuel, and the vipac fuel will be simultaneously irradiated on HFR (High Flux Reactor, Holland). The fabrication drawings have been made under the design assignment with PSI, and consist of the drawings of MOX pellet, thermal insulator pellet, pin components, fuel segments, and the constructed pin. The fabrication drawings were approved in October 2001, but after that, the optimization of specifications was discussed and agreed among all partners. According to this agreement, the fabrication drawings were revised in January 2002. After the earlier revision, the shape of particle retainer to be made by PSI was modified from its drawing beforehand delivered. In this report, the fabrication drawings revised again will be shown, and the fabrication procedure (welding Qualification Tests) will be modified in accordance with the result of discussion on the 3rd technical meeting held in September 2002. These design works have been performed in Fuel Design and Evaluation Group, Plutonium Fuel Fabrication Division, Plutonium Fuel Center under the commission of Plutonium Fuel

  20. V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

    International Nuclear Information System (INIS)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

    2010-07-01

    V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

  1. V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

    Energy Technology Data Exchange (ETDEWEB)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

    2010-07-15

    V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

  2. Safety studies dedicated to molten salt reactors with a fast neutron spectrum and operated in the Thorium fuel cycle - Innovative concept of Molten Salt Fast Reactor

    International Nuclear Information System (INIS)

    Brovchenko, Mariya

    2013-01-01

    The nuclear reactors of the 4. generation must allow an optimized use of natural resources, while performing at a high safety level. The framework of this thesis is the deployment study of one of such a system, an innovative and still little studied Molten Salt Fast Reactor. An excellent safety is an ultimate requirement of the nuclear energy deployment, so it is important to raise this question at the current early stage of the MSFR concept development. This concept was the subject of a neutronic tool benchmark within a European project EVOL. Definition, calculations and results analyses were performed during this thesis. Comparisons of static neutronic and burn-up calculations, performed by the project participants, concluded to a good agreement between the different codes and methods used and pointed out the sensibility of the nuclear database choice on the results. With the aim of safety analysis of the MSFR, the decay heat was studied in detail. The tool used for the decay heat calculation was developed and validated, to finally evaluate the decay heat in the reactor. The decay heat source presented in different zones was quantified, concluding to a high importance of the cooling of the fuel salt and the bubbling system enclosing a part of the fission products. The safety analysis methodology was also studied in this thesis. Even if the safety principles are directly transposable to the MSFR, the precise recommendations are not. This is due to the specificity of the design that relies on the liquid state of the fuel, on the reprocessing systems located in the reactor and the embryonic stage of the design. First, a preliminary transposition work of some criteria to the MSFR design was realized, resulting amongst other things in a list of accidental scenarios particular for MSFR. Finally, a preliminary physical study of some types of accidental scenarios was performed, that can be used as a basis for further analyses with more sophisticated tools. (author) [fr

  3. Molten salt breeder reactor

    International Nuclear Information System (INIS)

    1977-01-01

    MSBR Study Group formed in October 1974 has studied molten salt breeder reactor and its various aspects. Usage of a molten salt fuel, extremely interesting as reactor chemistry, is a great feature to MSBR; there is no need for separate fuel making, reprocessing, waste storage facilities. The group studied the following, and these results are presented: molten salt technology, molten salt fuel chemistry and reprocessing, reactor characteristics, economy, reactor structural materials, etc. (Mori, K.)

  4. Handbook on process and chemistry of nuclear fuel reprocessing version 2

    International Nuclear Information System (INIS)

    2008-10-01

    Aqueous nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of aqueous reprocessing, because it contributes to establish and develop fuel reprocessing technology and nuclear fuel cycle treating high burn-up UO 2 fuel and spent MOX fuel, and to utilize aqueous reprocessing technology much widely. This handbook is the second edition of the first report, which summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing' from FY 1993 until FY 2000. (author)

  5. Fuel salt reprocessing influence on the MSFR behavior and on its associated reprocessing unit

    International Nuclear Information System (INIS)

    Doligez, X.

    2010-10-01

    In order to face with the growing of the energy demand, the nuclear industry has to reach the fourth generation technology. Among those concept, molten salt reactor, and especially the fast neutron spectrum configuration, seems very promising: indeed breeding is achievable while the feedback coefficient are still negative. However, the reprocessing salt scheme is not totally set down yet. A lot of uncertainties remain on chemical properties of the salt. Thanks to numerical simulation we studied the behavior of the molten Salt Fast Reactor coupled to a nominal reprocessing unit. We are now able to determine heat transfer and radiation in each elementary step of the unit and, by this way determine those that need special study for radioprotection. We also studied which elements are fundamental to extract for the reactor operation. Finally, we present a sensibility analysis of the chemical uncertainties to few relevant properties of the reactor behavior. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, R.; Takahashi, Y.; Nakamura, H.; Mizuguchi, K. [Power and Industrial Research and Development Center, Toshiba Corporation Power Systems Company, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862 (Japan); Oomori, T. [Chemical System Design and Engineering Department, Toshiba Corporation Power Systems Company, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan)

    2013-07-01

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

  7. Prediction of temperature increases in a salt repository expected from the storage of spent fuel or high-level waste

    International Nuclear Information System (INIS)

    Llewellyn, G.H.

    1978-04-01

    Comparisons in temperature increases incurred from hypothetical storage of 133 MW of 10-year-old spent fuel (SF) or high-level waste (HLW) in underground salt formations have been made using the HEATING5 computer code. The comparisons are based on far-field homogenized models that cover areas of 65 and 25 sq miles for SF and HLW, respectively, and near-field unit-cell models covering respective areas of 610 ft 2 and 400 ft 2 . Preliminary comparisons based on heat loads of 150 kW/acre and 3.5 kW/canister indicated near-field temperature increases about 20% higher for the storage of the spent fuel than for the high-level waste. In these comparisons, it was also found that the thermal energy deposited in the salt after 500 years is about twice the energy deposited by the high-level waste. The thermal load in a repository containing 10-year-old spent fuel was thus limited to 60 kW/acre to obtain comparable far-field thermal effects as obtained in a repository containing 10-year-old high-level waste loaded at 150 kW/acre. Detailed far-field and unit-cell comparisons of transient temperature increases have been made based on these loadings. Unit-cell comparisons were made between a canister containing high-level waste with an initial heat production rate of 2.1 kW and a canister containing a PWR spent fuel assembly producing 0.55 kW. Using a three-dimensional unit-cell model, a maximum salt temperature increase of 260 0 F was calculated for the high-level waste prior to back-filling (5 years after burial), whereas a maximum temperature increase of 110 0 F was calculated for the spent fuel prior to backfilling (25 years after burial). Comparisons were also made between various configurational models for the high-level waste showing the applicability of each model

  8. Computer code SICHTA-85/MOD 1 for thermohydraulic and mechanical modelling of WWER fuel channel behaviour during LOCA and comparison with original version of the SICHTA code

    International Nuclear Information System (INIS)

    Bujan, A.; Adamik, V.; Misak, J.

    1986-01-01

    A brief description is presented of the expansion of the SICHTA-83 computer code for the analysis of the thermal history of the fuel channel for large LOCAs by modelling the mechanical behaviour of fuel element cladding. The new version of the code has a more detailed treatment of heat transfer in the fuel-cladding gap because it also respects the mechanical (plastic) deformations of the cladding and the fuel-cladding interaction (magnitude of contact pressure). Also respected is the change in pressure of the gas filling of the fuel element, the mechanical criterion is considered of a failure of the cladding and the degree is considered of the blockage of the through-flow cross section for coolant flow in the fuel channel. The LOCA WWER-440 model computation provides a comparison of the new SICHTA-85/MOD 1 code with the results of the original 83 version of SICHTA. (author)

  9. Graphite anode surface modification with controlled reduction of specific aryl diazonium salts for improved microbial fuel cells power output.

    Science.gov (United States)

    Picot, Matthieu; Lapinsonnière, Laure; Rothballer, Michael; Barrière, Frédéric

    2011-10-15

    Graphite electrodes were modified with reduction of aryl diazonium salts and implemented as anodes in microbial fuel cells. First, reduction of 4-aminophenyl diazonium is considered using increased coulombic charge density from 16.5 to 200 mC/cm(2). This procedure introduced aryl amine functionalities at the surface which are neutral at neutral pH. These electrodes were implemented as anodes in "H" type microbial fuel cells inoculated with waste water, acetate as the substrate and using ferricyanide reduction at the cathode and a 1000 Ω external resistance. When the microbial anode had developed, the performances of the microbial fuel cells were measured under acetate saturation conditions and compared with those of control microbial fuel cells having an unmodified graphite anode. We found that the maximum power density of microbial fuel cell first increased as a function of the extent of modification, reaching an optimum after which it decreased for higher degree of surface modification, becoming even less performing than the control microbial fuel cell. Then, the effect of the introduction of charged groups at the surface was investigated at a low degree of surface modification. It was found that negatively charged groups at the surface (carboxylate) decreased microbial fuel cell power output while the introduction of positively charged groups doubled the power output. Scanning electron microscopy revealed that the microbial anode modified with positively charged groups was covered by a dense and homogeneous biofilm. Fluorescence in situ hybridization analyses showed that this biofilm consisted to a large extent of bacteria from the known electroactive Geobacter genus. In summary, the extent of modification of the anode was found to be critical for the microbial fuel cell performance. The nature of the chemical group introduced at the electrode surface was also found to significantly affect the performance of the microbial fuel cells. The method used for

  10. National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel. Volume I. Conceptual design report

    International Nuclear Information System (INIS)

    1978-12-01

    In February 1976, the Energy Research and Development Administration (ERDA), now the Department of Energy (DOE), established a National Waste Terminal Storage (NWTS) program. As a part of this program, two parallel conceptual design efforts were initiated in January 1977. One was for deep geologic storage, in domed salt, of high level waste resulting from the reprocessing of spent fuel. The other was for deep geologic storage of unreprocessed spent fuel in bedded salt. These two concepts are identified as NWTS Repository 1 and Repository 2, respectively. Repository 2 (NWTSR2) is the concept which is covered by this Conceptual Design Report. Volume I of the conceptual design report contains the following information: physical description of the report; project purpose and justification; principal safety, fire, and health hazards; environmental impact considerations; quality assurance considerations; assessment of operational interfaces; assessment of research and development interfaces; project schedule; proposed method of accomplishment; summary cost estimate; and outline specifications. The conceptual design for Repository 2 was developed in sufficient detail to permit determination of scope, engineering feasibility, schedule, and cost estimates, all of which are necessary for planning and budgeting the project

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  12. Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    2002-01-01

    The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [∼1000 Mw(e)] with passive safety systems to provide the potential for improved economics

  13. Experimental base for experiments with molten salt fuel compositions at Chelyabinsk-70

    International Nuclear Information System (INIS)

    Subbotin, V.; Avrorin, E.; Grebyonkin, K.; Zouev, Yu.; Panov, A.

    1997-01-01

    Now some conceptual projects of Molten-Salts Based Nuclear Reactors (MSBNR) exists and problem of creating of full-scale demonstration installation of such type is working up seriously enough. Wide researches, confirming reality of solving of the problem of MSBNR building, have already been carried out. At the same time engineer realization of the project needs tests of a whole number of technical and technological solutions, and obtaining of additional data in physics and chemistry of salts and compatibility of materials. Possessing powerful scientific and technical potential and developed experimental base RFNC-VNIITF would have a possibility to bring in adequate contribution to the problem of creating MSBNR

  14. Reprocessing method of ceramic nuclear fuels in low-melting nitrate molten salts

    International Nuclear Information System (INIS)

    Brambilla, G.; Caporali, G.; Zambianchi, M.

    1976-01-01

    Ceramic nuclear fuel is reprocessed through a method wherein the fuel is dispersed in a molten eutectic mixture of at least two alkali metal nitrates and heated to a temperature in the range between 200 and 300 0 C. That heated mixture is then subjected to the action of a gaseous stream containing nitric acid vapors, preferably in the presence of a catalyst such as sodium fluoride. Dissolved fuel can then be precipitated out of solution in crystalline form by cooling the solution to a temperature only slightly above the melting point of the bath

  15. Mock-up facilities for the development of an advanced spent fuel management process using molten salt technology

    International Nuclear Information System (INIS)

    Young-Joon Shin; Ik-Soo Kim; Seung-Chul Oh; Soo-Haeng Cho; Yo-Taik Song; Hyun-Soo Park

    2000-01-01

    The Korea Atomic Energy Research Institute (KAERI) has investigated a new approach to spent fuel storage technology that would reduce the total storage volume and the amount of decay heat. The technology utilizes the reduction of oxide fuel to a metal to reduce the volume and preferentially removing the fission products to reduce the decay heat. The uranium oxide is reduced to uranium metal by Li metal in a molten LiCl salt bath. During the reduction process, fission products are dissolved into the LiCl bath and some of the highly radioactive elements, such as Sr and Cs, are preferentially removed from the bath. The reduced uranium metal is cast into an ingot, put into a storage capsule, and stored using conventional storage methods. The fission products are treated as high level radioactive wastes. Each process of the technology has been studied and analyzed for technical feasibility, and has come to the point for designing and constructing of the mock-up for a demonstration of the technology. This paper presents the detailed design of the mock-up of the system and operational characteristics, along with all the details of the equipment for the system. KAERI plans to use the mock-up for the demonstration using an in-active spent fuel specimen. (authors)

  16. Retrieval system for emplaced spent unreprocessed fuel (SURF) in salt bed depository. Baseline concept criteria specifications and mechanical failure probabilities

    International Nuclear Information System (INIS)

    Hudson, E.E.; McCleery, J.E.

    1979-05-01

    One of the integral elements of the Nuclear Waste Management Program is the material handling task of retrieving Canisters containing spent unreprocessed fuel from their emplacement in a deep geologic salt bed Depository. A study of the retrieval concept data base predicated this report. In this report, alternative concepts for the tasks are illustrated and critiqued, a baseline concept in scenario form is derived and basic retrieval subsystem specifications are presented with cyclic failure probabilities predicted. The report is based on the following assumptions: (a) during retrieval, a temporary radiation seal is placed over each Canister emplacement; (b) a sleeve, surrounding the Canister, was initially installed during the original emplacement; (c) the emplacement room's physical and environmental conditions established in this report are maintained while the task is performed

  17. Risk analysis methodology for unreprocessed spent fuel disposal in bedded salt

    International Nuclear Information System (INIS)

    Pepping, R.E.; Chu, M.S.Y.; Cranwell, R.M.

    1982-01-01

    In accordance with the decision to defer the reprocessing of commercially generated spent fuel, we are investigating the implications on risk of direct disposal of spent fuel assemblies. To the extent possible, we are using the methodology developed at Sandia for the NRC to evaluate risks from the disposal of wastes from reprocessing of spent fuel. This allows direct comparison of the risks calculated for the two waste forms. A number of differences between the two waste forms with implications on risk have been identified and investigation of their effects has begun. Among these are the presence of gases and additional plutonium and uranium isotopes, the potential for differing leach behavior, and the difference in the decay heat source which determines the overall thermomechanical response of the host media. We have analyzed a number of scenarios for a hypothetical geologic repository that have been identified as important contributors to risk from the disposal of both reprocessed and unreprocessed spent fuel. For each scenario, we employ the Groundwater Transport, Pathways to Man, and Dosimetry and Health Effects models of the High Level Waste Methodology. Risks are compared for the reprocessed and unreprocessed spent fuel wastes and the effects of uncertainty in the parameters of the various models are compared

  18. New version of NPP Krsko Decommissioning program and LILW and spent fuel management

    International Nuclear Information System (INIS)

    Zeleznik, N.; Mele, I.; Jenko, T.; Lokner, V.; Levanat, I.; Rapic, A.

    2004-01-01

    According to the requirements of the bilateral agreement between Republic of Slovenia and Republic of Croatia on the legal and other obligations for Nuclear power plant (NPP) Krsko the Decommissioning program was prepared. The main purpose of the program was to estimate the overall expenses of the future decommissioning, radioactive waste and spent fuel management of the NPP Krsko in order to establish separate fund in Croatia and to correct the rate per kWh collected in the existing decommissioning fund in Slovenia. The program looked at all possible scenarios of dismantling, radioactive waste and spent fuel management and proposed the most plausible two scenarios which are technically possible and financially feasible. (author)

  19. CASKETSS-2: a computer code system for thermal and structural analysis of nuclear fuel shipping casks (version 2)

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1991-08-01

    A computer program CASKETSS-2 has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS-2 means a modular code system for CASK Evaluation code system Thermal and Structural Safety (Version 2). Main features of CASKETSS-2 are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) There are simplified computer programs and a detailed one in the structural analysis part in the code system. (3) Input data generator is provided in the code system. (4) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  20. Investigation of solar parabolic trough power plants with and without integrated TES (thermal energy storage) and FBS (fuel backup system) using thermic oil and solar salt

    International Nuclear Information System (INIS)

    Boukelia, T.E.; Mecibah, M.S.; Kumar, B.N.; Reddy, K.S.

    2015-01-01

    Thermodynamic, economic and environmental analyses of concentrating solar power plants assist in identifying an effective and viable configuration. In this paper, a 4E (energy-exergy-environmental-economic) comparative study of 8 different configurations of parabolic trough solar thermal power plants with two different working fluids (Therminol VP-1 -oil and molten solar salt), with and without integrated thermal energy storage or/and backup fuel system is presented. The results of the comparative study indicate relevant differences among the 8 configurations. The molten solar salt configuration with integrated thermal energy storage and fossil fuel backup system exhibits the highest overall energy efficiency (18.48%) compared to other configurations. Whereas, the highest overall exergy efficiency (21.77%), capacity factor (38.20%) and annual energy generation (114 GWh) are found for the oil based configuration with integrated thermal energy storage and fossil fuel backup system. The results indicate that the configurations based on molten salt are better in terms of environmental and economical parameters. The configurations with integrated thermal energy storage and fossil fuel backup system are found to be techno-economical, but on the other hand are less environment friendly. A detailed comparison of these plants after optimization must be performed before drawing a final conclusion about the best configuration to be adopted in parabolic trough solar thermal power plant. - Highlights: • 4E comparative study of 8 configurations of PTSTPP with two different fluids. • Comparison of the configurations with and without integrated TES (thermal energy storage) and FBS (fuel backup system). • The overall energy efficiency of the salt plant with TES and FBS is the highest. • The overall exergy efficiency of the oil plant with TES and FBS is the highest. • The salt plants are the best configurations in terms of environ–eco parameters

  1. Molten salt actinide recycler and transforming system without and with Th–U support: Fuel cycle flexibility and key material properties

    International Nuclear Information System (INIS)

    Ignatiev, V.; Feynberg, O.; Gnidoi, I.; Merzlyakov, A.; Surenkov, A.; Uglov, V.; Zagnitko, A.; Subbotin, V.; Sannikov, I.; Toropov, A.; Afonichkin, V.; Bovet, A.; Khokhlov, V.; Shishkin, V.; Kormilitsyn, M.; Lizin, A.; Osipenko, A.

    2014-01-01

    Highlights: • We examine feasibility of MOSART system without and with U–Th support. • We experimentally studied key material properties to prove MOSART flowsheet. • MOSART potential as the system with flexible fuel cycle scenarios is emphasized. • MOSART can operate with different TRU loadings in transmuter or even breeder modes. - Abstract: A study is under progress to examine the feasibility of MOlten Salt Actinide Recycler and Transforming (MOSART) system without and with U–Th support fuelled with different compositions of transuranic elements (TRU) trifluorides from spent LWR fuel. New design options with homogeneous core and fuel salt with high enough solubility for transuranic elements trifluorides are being examined because of new goals. The paper has the main objective of presenting the fuel cycle flexibility of the MOSART system while accounting technical constrains and experimental data received in this study. A brief description is given of the experimental results on key physical and chemical properties of fuel salt and combined materials compatibility to satisfy MOSART system requirements

  2. Molten salt fuels for treatment of plutonium and radwastes in ADS critical systems

    International Nuclear Information System (INIS)

    Ignatiev, Victor V.

    2000-01-01

    Introduction of the innovative reactor concept of the incinerator type in the future nuclear power system should provide the following: Low Plutonium and Minor Actinides Total Inventory in the Nuclear Fuel Cycle (M) Reduced Actinides Total Losses to Waste (W) Minimal Uranium-235 SupportMinimal Neutron Captures Outside Actinides (Coolant and Structural Material Activation Products). Estimations have shown strong dependence of the first two parameters (M and W), which are responsible for incinerator efficiency, from the burnup (c) reached in the core of an incinerator and the actinides mass flow rate in the fuel cycle (A(t)=G(t)/Q(t), where G(t)=amount of TRU fed to the process during t, and Q(t)=electricity produced during (t)

  3. Assay of uranium in fused salt cake generated at the natural uranium metal fuel fabrication plants by gamma-ray spectrometry

    International Nuclear Information System (INIS)

    Kalsi, P.C.; Bhanu, A.U.; Sahoo, S.; Iyer, R.H.

    1986-01-01

    A passive gamma-ray spectroscopic method is employed for the assay of uranium in fused salt cake, a scrap produced at the natural uranium metal fuel fabrication plants. The method makes use of NaI(TI) detector coupled with a multichannel analyser. The 1 MeV gamma-ray of 238 U was used for the calibration. The calibration curve was made by counting synthetic mixtures made of U 3 O 8 powder, the heat treatment salt and iron in the form of fine powder. The uranium content in these synthetic mixtures was kept in the range of 1-11 per cent. 23 lots of the fused salt cake taken from three different batches of the salt cake were then analysed by this method. The uranium content of fused salt cake was found to be in the range of 1.70-11.43 per cent. To compare the gamma spectrometric results with a completely independent method, chemical analysis of all the fused salt cakes were also carried out. The NDA results were found to agree within ± 17 per cent with the chemical analysis results. (author)

  4. Measures of stress corrosion cracking in the canister storage facility of spent nuclear fuel. Vol.3. Development of salt particle collection device

    International Nuclear Information System (INIS)

    Takeda, Hirofumi; Saegusa, Toshiari

    2009-01-01

    A natural ventilation system is generally adopted for storage facilities of spent nuclear fuel. At the storage facilities of concrete casks built near the seashore, the air including the sea salt particles comes into the concrete casks and could cause SCC to the canister made of stainless steel. In this study, we proposed a salt particle collection device with a low flow resistance which does not block the air flow into the building. The effect of the device was evaluated quantitatively in laboratory experiments and in field tests. Obtained results are as follows: (1) The pressure loss of the device is smaller than one-sevenths of pressure loss of a filter used in a forced ventilation system and the efficiency of salt particle collection is more than 80% in both laboratory experiments and field tests. However, the efficiency of salt particle collection depends on the diameter of a salt particle. (2) It was clarified the diameter of the particle which can be collected by the device under the condition of the size of the device, the density and velocity of the particle. And the pressure loss of the device was evaluated. In the case of setting the device in the air inlet of a concrete cask, salt particles lager than 27μm in diameter can be collected by the device under the condition of the same pressure loss of a bard screen which opening ratio is 80%. (author)

  5. Uranium in selected endorheic basins as partial analogue for spent fuel behavior in salt

    International Nuclear Information System (INIS)

    Van Luik, A.E.

    1987-01-01

    If uranium (U) behavior with respect to the components of certain endorheic (closed) basin subsurface, playa, or terminal lake brines were quantitatively understood, the ability to predict the long-term redistribution of emplaced U among analogous components of salt formations may be enhanced. Tests that determine the nature of U interactions with pure mineral and organic matter surfaces are important, but studying the natural systems available could give indications of long-term stabilities of processes, and of preferential processes. For example, some metals present in trace quantities, such as U, may be coprecipitated in the oxidized zone with an evaporite mineral that may afterward undergo diagenesis, especially if conditions become more reducing. During diagenesis, the trace metal may be remobilized, but scavenged by sulfides or organic particulates, leaving the evaporite mineral depleted of its trace metal content. A survey of the literature shows trace metal behavior in closed basins has been studied. However, information on U consists of only a few abundance determinations for some evaporite systems. Obtaining and interpreting natural analogue data for the U and Th decay series in selected endorheic basin environments is suggested. 44 refs., 3 figs

  6. Uranium in selected endorheic basins as partial analogue for spent fuel behavior in salt

    International Nuclear Information System (INIS)

    Luik, A.E. van

    1987-01-01

    If uranium (U) behavior with respect to the components of certain endorheic (closed) basin subsurface, playa, or terminal lake brines were quantitatively understood, the ability to predict the long-term redistribution of emplaced U among analogous components of salt formations may be enhanced. Tests that determine the nature of U interactions with pure mineral and organic matter surfaces are important, but studying the natural systems available could give indications of long-term stabilities of processes, and of preferential processes. For example, some metals present in trace quantities, such as U, may be coprecipitated in the oxidized zone with an evaporite mineral that may afterwards undergo diagenesis, especially if conditions become more reducing. During diagenesis, the trace metal may be remobilized, but scavenged by sulfides or organic particulates, leaving the evaporite mineral depleted of its trace metal content. A survey of the literature shows some trace metal behavior in closed basins has been studied. However, information on U consists of only a few abundance determinations for some evaporite systems. Obtaining and interpreting natural analogue data for the U and Th decay series in selected endorheic basin environments is suggested. (author)

  7. DIONISIO 2.0: New version of the code for simulating a whole nuclear fuel rod under extended irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro, E-mail: soba@cnea.gov.ar; Denis, Alicia

    2015-10-15

    Highlights: • A new version of the DIONISIO code is developed. • DIONISIO is devoted to simulating the behavior of a nuclear fuel rod in operation. • The formerly two-dimensional simulation of a pellet-cladding segment is now extended to the whole rod length. • An acceptable and more realistic agreement with experimental data is obtained. • The prediction range of our code is extended up to average burnup of 60 MWd/kgU. - Abstract: The version 2.0 of the DIONISIO code, that incorporates diverse new aspects, has been recently developed. One of them is referred to the code architecture that allows taking into account the axial variation of the conditions external to the rod. With this purpose, the rod is divided into a number of axial segments. In each one the program considers the system formed by a pellet and the corresponding cladding portion and solves the numerous phenomena that take place under the local conditions of linear power and coolant temperature, which are given as input parameters. To do this a bi-dimensional domain in the r–z plane is considered where cylindrical symmetry and also symmetry with respect to the pellet mid-plane are assumed. The results obtained for this representative system are assumed valid for the complete segment. The program thus produces in each rod section the values of the temperature, stress, strain, among others as outputs, as functions of the local coordinates r and z. Then, the general rod parameters (internal rod pressure, amount of fission gas released, pellet stack elongation, etc.) are evaluated. Moreover, new calculation tools designed to extend the application range of the code to high burnup, which were reported elsewhere, have also been incorporated to DIONISIO 2.0 in recent times. With these improvements, the code results are compared with some 33 experiments compiled in the IFPE data base, that cover more than 380 fuel rods irradiated up to average burnup levels of 40–60 MWd/kgU. The results of these

  8. DIONISIO 2.0: New version of the code for simulating a whole nuclear fuel rod under extended irradiation

    International Nuclear Information System (INIS)

    Soba, Alejandro; Denis, Alicia

    2015-01-01

    Highlights: • A new version of the DIONISIO code is developed. • DIONISIO is devoted to simulating the behavior of a nuclear fuel rod in operation. • The formerly two-dimensional simulation of a pellet-cladding segment is now extended to the whole rod length. • An acceptable and more realistic agreement with experimental data is obtained. • The prediction range of our code is extended up to average burnup of 60 MWd/kgU. - Abstract: The version 2.0 of the DIONISIO code, that incorporates diverse new aspects, has been recently developed. One of them is referred to the code architecture that allows taking into account the axial variation of the conditions external to the rod. With this purpose, the rod is divided into a number of axial segments. In each one the program considers the system formed by a pellet and the corresponding cladding portion and solves the numerous phenomena that take place under the local conditions of linear power and coolant temperature, which are given as input parameters. To do this a bi-dimensional domain in the r–z plane is considered where cylindrical symmetry and also symmetry with respect to the pellet mid-plane are assumed. The results obtained for this representative system are assumed valid for the complete segment. The program thus produces in each rod section the values of the temperature, stress, strain, among others as outputs, as functions of the local coordinates r and z. Then, the general rod parameters (internal rod pressure, amount of fission gas released, pellet stack elongation, etc.) are evaluated. Moreover, new calculation tools designed to extend the application range of the code to high burnup, which were reported elsewhere, have also been incorporated to DIONISIO 2.0 in recent times. With these improvements, the code results are compared with some 33 experiments compiled in the IFPE data base, that cover more than 380 fuel rods irradiated up to average burnup levels of 40–60 MWd/kgU. The results of these

  9. National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel. Quality assurance program for licensing

    International Nuclear Information System (INIS)

    1978-12-01

    A National Waste Terminal Storage Repository, in bedded salt, for spent unreprocessed fuel is the subject of a conceptual design project which began in January 1977. This volume presents a preliminary quality assurance program to guide the license applicant in developing a detailed program that will be compatible with anticipated National Waste Terminal Storage (NWTSR2) contracting arrangements and provide the documentation required by regulatory bodies. This QA program is designed to provide confidence that the quality-related activities pertaining to safety-related structures, systems, and components will be identified and controlled. Specific responsibilities for quality-related activities are documented and assigned to personnel and organizations for the major phases of facility design and construction. These responsibilities encompass a broad range of activities and are addressed in this preliminary program. The quality assurance program elements are organized and discussed herein as follows: (1) quality assurance during design and construction; (2) the applicant (DOE); (3) siting contractor; (4) architect/engineer; (5) project field management; and (6) operations contractor

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V; Merzlyakov, A [Kurchatov Institute - KI (Russian Federation); Afonichkin, V; Khokhlov, V; Salyulev, A [Institute of High Temperature Electrochemisty (IHTE), RF Yuri Golovatov, Konstantin Grebenkine, Vladimir Subbotin Institute of Technical Physics (VNIITF) (Russian Federation)

    2003-07-01

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

  12. Alternative Fuel and Advanced Technology Commercial Lawn Equipment (Spanish version); Clean Cities, Energy Efficiency & Renewable Energy (EERE)

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, Erik

    2015-06-01

    Powering commercial lawn equipment with alternative fuels or advanced engine technology is an effective way to reduce U.S. dependence on petroleum, reduce harmful emissions, and lessen the environmental impacts of commercial lawn mowing. Numerous alternative fuel and fuel-efficient advanced technology mowers are available. Owners turn to these mowers because they may save on fuel and maintenance costs, extend mower life, reduce fuel spillage and fuel theft, and demonstrate their commitment to sustainability.

  13. Analysis on the use of engineered barriers for geologic isolation of spent fuel in a reference salt site repository

    International Nuclear Information System (INIS)

    Cloninger, M.O.; Cole, C.R.; Washburn, J.F.

    1980-12-01

    A perspective on the potential durability and effectiveness requirements for the waste form, container and other engineered barriers for geologic disposal of spent nuclear fuel has been developed. This perspective is based on calculated potential doses to individuals who may be exposed to radioactivity released from a repository via a groundwater transport pathway. These potential dose commitments were calculated with an integrated geosphere transport and bioshpere transport model. A sensitivity analysis was accomplished by varying four important system parameters, namely the waste radionuclide release rate from the repository, the delay prior to groundwater contact with the waste (leach initiation), aquifer flow velocity and flow path length. The nuclide retarding capacity of the geologic media, a major determinant of the isolation effectiveness, was not varied as a parameter but was held constant for a particular reference site. This analysis is limited to looking only at engineered barriers whose net effect is either to delay groundwater contact with the waste form or to limit the rate of release of radionuclides into the groundwater once contact has occurred. The analysis considers only leach incident scenarios, including a water well intrusion into the groundwater near a repository, but does not consider other human intrusion events or catastrophic events. The analysis has so far been applied to a reference salt site repository system and conclusions are presented.Basically, in nearly all cases, the regional geology is the most effective barrier to release of radionuclides to the biosphere; however, for long-lived isotopes of carbon, technetium and iodine, which were poorly sorbed on the geologic media, the geology is not very effective once a leach incident is initiated

  14. Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Maldonado, Ivan [Univ. of Tennessee, Knoxville, TN (United States)

    2016-04-14

    The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate (“plank”) fuel. Proposal to FY12 NEUP entitled “Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors” was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project’s success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

  15. Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Maldonado, Ivan

    2016-01-01

    The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate ('plank') fuel. Proposal to FY12 NEUP entitled 'Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors' was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project's success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

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

    International Nuclear Information System (INIS)

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

    2002-12-01

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

  17. Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1. 11, September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

    1982-01-01

    This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts.

  18. Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1.11, September 1981

    International Nuclear Information System (INIS)

    Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

    1982-01-01

    This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts

  19. Molten salt reactors: chemistry

    International Nuclear Information System (INIS)

    1983-01-01

    This work is a critical analysis of the 1000 MW MSBR project. Behavior of rare gases in the primary coolant circuit, their extraction from helium. Coating of graphite by molybdenum, chemistry of protactinium and niobium produced in the molten salt, continuous reprocessing of the fuel salt and use of stainless steel instead of hastelloy are reviewed [fr

  20. Retrieval system for emplaced spent unreprocessed fuel (SURF) in salt bed depository: accident event analysis and mechanical failure probabilities. Final report

    International Nuclear Information System (INIS)

    Bhaskaran, G.; McCleery, J.E.

    1979-10-01

    This report provides support in developing an accident prediction event tree diagram, with an analysis of the baseline design concept for the retrieval of emplaced spent unreprocessed fuel (SURF) contained in a degraded Canister. The report contains an evaluation check list, accident logic diagrams, accident event tables, fault trees/event trees and discussions of failure probabilities for the following subsystems as potential contributors to a failure: (a) Canister extraction, including the core and ram units; (b) Canister transfer at the hoist area; and (c) Canister hoisting. This report is the second volume of a series. It continues and expands upon the report Retrieval System for Emplaced Spent Unreprocessed Fuel (SURF) in Salt Bed Depository: Baseline Concept Criteria Specifications and Mechanical Failure Probabilities. This report draws upon the baseline conceptual specifications contained in the first report

  1. Electrolytic reduction runs of 0.6 kg scale-simulated oxide fuel in a Li2O-LiCl molten salt using metal anode shrouds

    Science.gov (United States)

    Choi, Eun-Young; Lee, Jeong; Heo, Dong Hyun; Lee, Sang Kwon; Jeon, Min Ku; Hong, Sun Seok; Kim, Sung-Wook; Kang, Hyun Woo; Jeon, Sang-Chae; Hur, Jin-Mok

    2017-06-01

    Ten electrolytic reduction or oxide reduction (OR) runs of a 0.6 kg scale-simulated oxide fuel in a Li2O-LiCl molten salt at 650 °C were conducted using metal anode shrouds. During this procedure, an anode shroud surrounds a platinum anode and discharges hot oxygen gas from the salt to outside of the OR apparatus, thereby preventing corrosion of the apparatus. In this study, a number of anode shrouds made of various metals were tested. Each metallic anode shroud consisted of a lower porous shroud for the salt phase and an upper nonporous shroud for the gas phase. A stainless steel (STS) wire mesh with five-ply layer was a material commonly used for the lower porous shroud for the OR runs. The metals tested for the upper nonporous shroud in the different OR runs are STS, nickel, and platinum- or silver-lined nickel. The lower porous shroud showed no significant damage during two consecutive OR runs, but exhibited signs of damage from three or more runs due to thermal stress. The upper nonporous shrouds made up of either platinum- or silver-lined nickel showed excellent corrosion resistance to hot oxygen gas while STS or nickel without any platinum or silver lining exhibited poor corrosion resistance.

  2. Molten salt reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Simon, N.; Renault, C.

    2014-01-01

    Molten salt reactors are one of the 6 concepts retained for the 4. generation of nuclear reactors. The principle of this reactor is very innovative: the nuclear fuel is dissolved in the coolant which allows the online reprocessing of the fuel and the online recovery of the fission products. A small prototype: the Molten Salt Reactor Experiment (MSRE - 8 MWt) was operating a few years in the sixties in the USA. The passage towards a fast reactor by the suppression of the graphite moderator leads to the concept of Molten Salt Fast Reactor (MSFR) which is presently studied through different European projects such as MOST, ALISIA and EVOL. Worldwide the main topics of research are: the adequate materials resisting to the high level of corrosiveness of the molten salts, fuel salt reprocessing, the 3-side coupling between neutron transport, thermohydraulics and thermo-chemistry, the management of the changing chemical composition of the salt, the enrichment of lithium with Li 7 in the case of the use of lithium fluoride salt and the use of MSFR using U 233 fuel (thorium cycle). The last part of the article presents a preliminary safety analysis of the MSFR. (A.C.)

  3. Risk analysis methodology for spent fuel repositories in bedded salt: methodlogy summary and differences between spent fuel and high level wastes

    International Nuclear Information System (INIS)

    Pepping, R.E.; Chu, M.S.

    1981-06-01

    In the absence of spent fuel reprocessing plans, unreprocessed spent fuel has become a candidate waste form for geologic disposal. In order to understand the public health risks from such disposal and to gain insights into the factors that influence them, a methodology is needed to combine the effects of site geology and hydrology, physical and chemical properties of the waste form, and the details of the engineering design. This report outlines such a methodology which the authors currently are applying to the analysis of unreprocessed spent fuel disposal. The methodology is the same methodology as was developed to describe the risks from geologic disposal of wastes from reprocessed spent fuel. The difference between spent fuel wastes and wastes from reprocessing that may affect the applicability of the methodology are highlighted

  4. Effect of the type of ammonium salt on the extractive desulfurization of fuels using deep eutectic solvents

    NARCIS (Netherlands)

    Warrag, Samah E.E.; Adeyemi, Idowu; Rodriguez, Nerea R.; Nashef, Inas M.; van Sint Annaland, Martin; Kroon, Maaike C.; Peters, Cor J.

    2018-01-01

    In a previous work, we proved that the deep eutectic solvents (DESs) consisting of mixtures of tetraalkylammonium salts with polyols are promising candidates for oil desulfurization based on the obtained liquid-liquid equilibrium (LLE) data. In this study, the capability of DESs containing other

  5. Development of a mechanistic model for release of radionuclides from spent fuel in brines: Salt Repository Project

    International Nuclear Information System (INIS)

    Reimus, P.W.; Windisch, C.F.

    1988-03-01

    At present there are no comprehensive mechanistic models describing the release of radionuclides from spent fuel in brine environments. This report provides a comprehensive review of the various factors that can affect radionuclide release from spent fuel, suggests a modeling approach, and discusses proposed experiments for obtaining a better mechanistic understanding of the radionuclide release processes. Factors affecting radionuclide release include the amount, location, and disposition of radionuclides in the fuel and environmental factors such as redox potential, pH, the presence of complexing anions, temperature, and radiolysis. It is concluded that a model describing the release of radionuclides from spent fuel should contain separate terms for release from the gap, grain boundaries, and grains of the fuel. Possible functional forms for these terms are discussed in the report. Experiments for assessing their validity and obtaining key model parameters are proposed. 71 refs., 4 figs., 6 tabs

  6. The local implementation of clean(er) fuels policies in Europe. A Handbook with guidelines. Final version

    International Nuclear Information System (INIS)

    Mulder, F.; Amara, Sliman Abu; Uustal, M.; Pelkmans, L.; Devriendt, N.; Rogulska, M.; Defranceschi, P.

    2009-05-01

    This handbook aims to guide the local/regional governments all over Europe who are involved in implementing clean(er) fuel policies in transport. The general challenge these governments are facing is how local policies on clean(er) fuels and vehicles can be made operational. Hence, how can the step be made from a vision on the strategic policy level, to a vision on the implementation of these policies. A local/regional policy on clean(er) fuels and vehicles is commonly part of the larger category 'sustainable transport policy', which in itself is part of a broader local environmental policy. The encompassing local/regional sustainable mobility policy will in most cases be based on the three well known main policy aims in this area: CO2 reduction; Improving the local air quality; and Improving the security of supply (locally often less stressed). This handbook will focus on the actual implementation of a clean(er) fuels and vehicles policy. It will describe the main challenges and how these can be overcome. It will describe how the market conditions for clean(er) fuels and vehicles can be created by establishing the vital market elements and which process is required to do so. And it will show how local enterprises can be involved and what the role of the local governments in this process can be. In order to identify the local success factors in overcoming the main challenges for implementation, case studies have been carried out in three European cities, namely Stockholm (Sweden), Graz (Austria) and Lille (France). The choice of these cities was based on their successes in implementing clean(er) fuel policies (although they followed different paths) and the fact that they managed to achieve ambitious clean(er) fuel/ clean(er) vehicle targets. These cities may thus be considered as ?good practice examples?. The case studies are based on existing literature, on multiple stakeholders? interviews in all three cities, and on two small surveys. The objectives of this

  7. Application of modified version of SPPS-1 - HEXAB-2DB computer code package for operational analyses of fuel behaviour in WWER-440 reactors at Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kharalampieva, Ts; Stoyanova, I; Antonov, A; Simeonov, T [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Petkov, P [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

    1994-12-31

    The modified version of SPPS-1 code called SPPS-1-HEXAB-2DB was applied for the purposes of the operational analysis and power peaking factors and reactor core critical parameters predictions of WWER-440s. The results of the calculations performed by the use of SPPS-1-HEXAB-2DB code and the corresponding parameters obtained from experiments at Kozloduy NPP WWER-440s as well as the results of fuel rod power distribution are presented. The method of operation simulation of reactor core with 349 assemblies (Unit 4) and with 313 fuel assemblies and 36 dummy fuel assemblies (Unit 1) is outlined. The modified code calculates not only fuel burnup and Pm-149 and Sm-149 concentrations distributions but also the space distribution of I-135 and Xe-135 concentrations. In this way it makes possible to perform the reactor operation simulation during the immediate periods after the reactor start-up or shut-down and to predict the critical reactor core parameters during transients. The results obtained show that SPPS-1-HEXAB-2DB code describes adequately the reactor core status. The new SPPS-1 code algorithm for estimation of assembly-wise power peaking factors distribution in reactor core is also described. The new code provides an option for checking the correctness of reactor core symmetry. The experience from the use of the modified SPPS-1-HEXAB-2DB code system confirms the provision of improved availability of operational analysis, prediction of Kozloduy NPP WWER-440s safe operations and fuel behaviour estimation. 14 tabs., 4 figs., 5 refs.

  8. Thermoelastic analysis of spent fuel and high level radioactive waste repositories in salt. A semi-analytical solution

    International Nuclear Information System (INIS)

    St John, C.M.

    1977-04-01

    An underground repository containing heat generating, High Level Waste or Spent Unreprocessed Fuel may be approximated as a finite number of heat sources distributed across the plane of the repository. The resulting temperature, displacement and stress changes may be calculated using analytical solutions, providing linear thermoelasticity is assumed. This report documents a computer program based on this approach and gives results that form the basis for a comparison between the effects of disposing of High Level Waste and Spent Unreprocessed Fuel

  9. New Method for Super Hydrophobic Treatment of Gas Diffusion Layers for Proton Exchange Membrane Fuel Cells Using Electrochemical Reduction of Diazonium Salts.

    Science.gov (United States)

    Thomas, Yohann R J; Benayad, Anass; Schroder, Maxime; Morin, Arnaud; Pauchet, Joël

    2015-07-15

    The purpose of this article is to report a new method for the surface functionalization of commercially available gas diffusion layers (GDLs) by the electrochemical reduction of diazonium salt containing hydrophobic functional groups. The method results in superhydrophobic GDLs, over a large area, without pore blocking. An X-ray photoelectron spectroscopy study based on core level spectra and chemical mapping has demonstrated the successful grafting route, resulting in a homogeneous distribution of the covalently bonded hydrophobic molecules on the surface of the GDL fibers. The result was corroborated by contact angle measurement, showing similar hydrophobicity between the grafted and PTFE-modified GDLs. The electrochemically modified GDLs were tested in proton exchange membrane fuel cells under automotive, wet, and dry conditions and demonstrated improved performance over traditional GDLs.

  10. Sigma-R: an improved version of the delayed neutron counting device for THTR fuel element verification

    International Nuclear Information System (INIS)

    Galantucci, R.; Haas, R.; Janssens, A.; Agostini, P.; Becker, L.; Bernede, M.; Caldon, L.; Cordani, G.; Guardini, S.

    1989-01-01

    SIGMA is a device designed for the Safeguards verification measurements of the 235 U content of THTR (or AVR) pebble fuel elements; it has been in operation for more than 12 years at the HOBEG fuel fabrication plant in HANAU (FRG). The THTR fuel element is a 60 mm diameter sphere composed of particles of a U-Th mixture in a graphite matrix. The device is schematically composed of an irradiation facility where the pebble is irradiated by 252 Cf and a counting facility where the delayed fission neutrons are monitored. The system has recently been thoroughly updated, the modified instrument being called SIGMA-R. SIGMA-R automatically governs the whole charge-discharge, irradiation and counting sequence, collects the measured counts in the microprocessor memory and performs the relevant calculations, including the statistical evaluation. The instructions are given interactively to the inspector through an IRIS terminal. SIGMA-R provides extensive self diagnosis features. The inspector is also informed explicity of malfunctions during routine operation. The instrument is equipped with a permanent memory which facilitates the inspector's work considerably. SIGMA is complementary to the so-called fuel pebble sampling devices (FPSD) which are designed to take a random sample of 10 pebbles out of 1000 directly from the production line. With SIGMA-R it is now possible to measure the 235 U content of THTR pebbles very accurately. The uncertainty essentially is limited to Poisson statistics and allows us to evaluate the 235 U content in routine operation, with a random error down to 0.15% and a systematic uncertainty estimated at 0.25%, the latter coming from calibration and normalization procedures

  11. Chemistry of Uranium in brines related to the spent fuel disposal in a salt repository. Part I

    International Nuclear Information System (INIS)

    Diaz Arocas, P.; Grambow, B.

    1993-01-01

    This report describes the work performed from september 1991 to december 1992. Our work is focused on the chemistry of uranium in highly saline solution. Experiments were performed to study the formation process and the stability of solid phases of U(VI) in NaCl solution at different ionic strength. The characterization of solid phases and of uranium concentration in solution are reported as a function of time. Experiments in NaClO 4 at low concentration have been performed for comparison. A method is proposed for uranium analyses in highly concentrated salt solution. The work has been carried out in KfK (INE), Germany. (Author) 10 figs

  12. DIONISIO 2.0: new version of the code for simulating the behavior of a power fuel rod under irradiation

    International Nuclear Information System (INIS)

    Soba, A; Denis, A; Lemes, M; Gonzalez, M E

    2012-01-01

    During the latest ten years the Codes and Models Section of the Nuclear Fuel Cycle Department has been developing the DIONISIO code, which simulates most of the main phenomena that take place within a fuel rod during the normal operation of a nuclear reactor: temperature distribution, thermal expansion, elastic and plastic strain, creep, irradiation growth, pellet-cladding mechanical interaction, fission gas release, swelling and densification. Axial symmetry is assumed and cylindrical finite elements are used to discretized the domain. The code has a modular structure and contains more than 40 interconnected models. A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO 2 fuels in LWR conditions, predict the radial distribution of power density, burnup and concentration of diverse nuclides within the pellet. New models of porosity and fission gas release in the rim, as well as the influence of the microstructure of this zone on the thermal conductivity of the pellet, are presently under development. A considerable computational challenge was the inclusion of the option of simulating the whole bar, by dividing it in a number of axial segments, at the user's choice, and solving in each segment the complete problem. All the general rod parameters (pressure, fission gas release, volume, etc.) are evaluated at the end of every time step. This modification allows taking into account the axial variation of the linear power and, consequently, evaluating the dependence of all the significant rod parameters with that coordinate. DIONISIO was elected for participating in the FUMEX III project of codes intercomparison, organized by IAEA, from 2008 to 2011. The results of the simulations performed within this project were compared with more than 30 experiments that involve more than 150 irradiated rods. The high number

  13. A history of salt.

    Science.gov (United States)

    Cirillo, M; Capasso, G; Di Leo, V A; De Santo, N G

    1994-01-01

    The medical history of salt begins in ancient times and is closely related to different aspects of human history. Salt may be extracted from sea water, mineral deposits, surface encrustations, saline lakes and brine springs. In many inland areas, wood was used as a fuel source for evaporation of brine and this practice led to major deafforestation in central Europe. Salt played a central role in the economies of many regions, and is often reflected in place names. Salt was also used as a basis for population censuses and taxation, and salt monopolies were practised in many states. Salt was sometimes implicated in the outbreak of conflict, e.g. the French Revolution and the Indian War of Independence. Salt has also been invested with many cultural and religious meanings, from the ancient Egyptians to the Middle Ages. Man's innate appetite for salt may be related to his evolution from predominantly vegetarian anthropoids, and it is noteworthy that those people who live mainly on protein and milk or who drink salty water do not generally salt their food, whereas those who live mainly on vegetables, rice and cereals use much more salt. Medicinal use tended to emphasize the positive aspects of salt, e.g. prevention of putrefaction, reduction of tissue swelling, treatment of diarrhea. Evidence was also available to ancient peoples of its relationship to fertility, particularly in domestic animals. The history of salt thus represents a unique example for studying the impact of a widely used dietary substance on different important aspects of man's life, including medical philosophy.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  15. Encyclopaedia of Features, Events and Processes (FEPs) for the Swedish SFR and Spent Fuel Repositories. Preliminary Version

    International Nuclear Information System (INIS)

    Miller, Bill; Savage, Dave; McEwen, Tim; White, Matt

    2002-08-01

    This is an 'Encyclopaedia' providing descriptions of Features, Events and Processes (FEPs) that are relevant to the Swedish repository for low and intermediate-level wastes (the SFR) and the proposed Swedish repository for spent fuel. Although the FEPs and their descriptions found in this encyclopaedia are specific to these two repository concepts, many of the descriptions will also be relevant to other repository designs and concepts, although they have not been written to be inclusive of the features of other repositories. As such, this encyclopaedia may be of interest to a wide range of individuals and organisations involved in repository safety assessment around the world. The purpose of this encyclopaedia is to describe, in qualitative terms, the various FEPs which have been identified as being relevant to the two Swedish repository designs. These descriptions may be used in a variety of ways. One important role will be to support quantitative performance assessments (PAs) by describing the conceptual understanding of the various components of the repository (e.g. the barriers, the rock and the groundwater) and their evolution: this conceptual understanding is crucial because it is the foundation upon which the mathematical analysis is based. The descriptions have been written at a level of detail appropriate for a scientifically literate reader without specialist knowledge of radioactive waste disposal technology or assessment procedures. As such, the descriptions avoid the use of specialist terms, acronyms and equations. Many of the FEP descriptions relate to issues which are the focus of ongoing research and, thus, they reflect the current state of knowledge and may require updating at regular intervals, either to include more recent technical information or procedures for treating the FEP in mathematical safety assessment. When attempting to predict the future behaviour of a repository for radioactive wastes, it is sensible to use a systematic approach to

  16. Encyclopaedia of Features, Events and Processes (FEPs) for the Swedish SFR and Spent Fuel Repositories. Preliminary Version

    Energy Technology Data Exchange (ETDEWEB)

    Miller, Bill; Savage, Dave; McEwen, Tim; White, Matt [QuantiSci Ltd, Melton Mowbray (United Kingdom)

    2002-08-01

    This is an 'Encyclopaedia' providing descriptions of Features, Events and Processes (FEPs) that are relevant to the Swedish repository for low and intermediate-level wastes (the SFR) and the proposed Swedish repository for spent fuel. Although the FEPs and their descriptions found in this encyclopaedia are specific to these two repository concepts, many of the descriptions will also be relevant to other repository designs and concepts, although they have not been written to be inclusive of the features of other repositories. As such, this encyclopaedia may be of interest to a wide range of individuals and organisations involved in repository safety assessment around the world. The purpose of this encyclopaedia is to describe, in qualitative terms, the various FEPs which have been identified as being relevant to the two Swedish repository designs. These descriptions may be used in a variety of ways. One important role will be to support quantitative performance assessments (PAs) by describing the conceptual understanding of the various components of the repository (e.g. the barriers, the rock and the groundwater) and their evolution: this conceptual understanding is crucial because it is the foundation upon which the mathematical analysis is based. The descriptions have been written at a level of detail appropriate for a scientifically literate reader without specialist knowledge of radioactive waste disposal technology or assessment procedures. As such, the descriptions avoid the use of specialist terms, acronyms and equations. Many of the FEP descriptions relate to issues which are the focus of ongoing research and, thus, they reflect the current state of knowledge and may require updating at regular intervals, either to include more recent technical information or procedures for treating the FEP in mathematical safety assessment. When attempting to predict the future behaviour of a repository for radioactive wastes, it is sensible to use a systematic

  17. Ultimate storage of spent fuel elements from the AVR experimental nuclear power plant in the Asse Salt Mine

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, J.

    1975-02-15

    The present paper is intended to serve as the basis for the licensing procedures both in respect of the transport and storage techniques and also for the ultimate storage itself. In regard to the technique it will be shown on the basis of design drawings and calculations that the handling, transport and storage of the fuel elements can be safely carried out in accordance with the regulations in force. In regard to the ultimate storage itself, since no highly radioactive wastes with a long-lived actinide content have yet been stored, it will be necessary to show that an unacceptable contamination of the biosphere will be avoided even in the long term under all anticipated conditions. It will further be necessary to show by calculations and suitable tests, in view of the radioactive gas and fissile material content, that no danger due to gas release from the fuel elements will arise during the operating life of the mine and that a nuclear criticality risk can be excluded for all time.

  18. Fundamental Properties of Salts

    Energy Technology Data Exchange (ETDEWEB)

    Toni Y Gutknecht; Guy L Fredrickson

    2012-11-01

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

  19. National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel. Special study No. 1. 25-year retrievability

    International Nuclear Information System (INIS)

    1978-12-01

    National Waste Terminal Storage Repository 2 (NWTSR2), in bedded salt, for spent unreprocessed fuel is the subject of a conceptual design project which began in January 1977. In the base conceptual design the spent fuel is to be stored in a retrievable mode for 5 years. This report presents the results of a separate study on 25-year retrievability with open storage rooms. Drawings prepared for 25-year retrievability are only those which differ from the companion 5-year retrievability drawings which comprise Volume II of the Conceptual Design Report (CDR). NWTSR2 capital facilities originally were to be constructed in phases. Phase I was to include those facilities required for the first 5 years of operation, which were hypothesized to be from 1985 through 1990; Phase II was to include facilities required from 1991 until the capacity of the 2000-acre mine was reached. These dates are used as the basis for conceptual design and for Special Study No. 1 but do not represent actual schedule dates for the repository. The terms Phase I and Phase II are used throughout the drawings and test, so they need to be understood. However, it is not now intended to phase the capital construction because of the relatively small effort represented by Phase II surface construction. An optional future expansion of the mine, Phase III, would add 1000 acres on the end of the mine opposite the shaft pillar. The optional Phase III expansion would commence operation upon completion of Phase II operations. The drawings and design comparisons for Phase III mine operations appear only in this report

  20. Molten salt reactor concept

    International Nuclear Information System (INIS)

    Sood, D.D.

    1980-01-01

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

  1. National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel. Special study No. 1. Twenty-five-year retrievability, decommissioning cost estimate

    International Nuclear Information System (INIS)

    1978-11-01

    This estimate covers decommissioning costs of facilities for the receipt, handling, and storage in bedded salt of canistered spent fuel assemblies from both BWR and PWR commercial power plants. The estimate includes all decommissioning costs for a repository, based on 25-year retrievability, constructed in accordance with the design shown in Conceptual Design Report (CDR), as modified by Special Study No. 1 (KE Report No. 78-60-RE) and decommissioned in accordance with the program outlined in the Conceptual Design Description Report, KE Report No. 78-58-R. Costs for Operating Contractor personnel on the site at this time are included in this report and not in the Operating Cost Estimate (KE Report 78-63-RE). The operating cost estimates end with the completion of storage room backfilling. The three major elements of decommissioning are: demolition of surface facilities, backfilling of main entries and airways, and shaft liner removal and shaft plugging. EDIT, ECON, and DELOX computer programs and a chart of accounts were furnished by UCC-ND under direction of the Government and the line item-capital-cost estimate was prepared according to the prescribed format. The decommissioning cost estimate referenced herein is in the same format as its companion line item-capital-cost estimate KE Report 78-62-RE

  2. Nuclear data covariances and sensitivity analysis, validation of a methodology based on the perturbation theory; application to an innovative concept: the molten thorium salt fueled reactor

    International Nuclear Information System (INIS)

    Bidaud, A.

    2005-10-01

    Neutron transport simulation of nuclear reactors is based on the knowledge of the neutron-nucleus interaction (cross-sections, fission neutron yields and spectra...) for the dozens of nuclei present in the core over a very large energy range (fractions of eV to several MeV). To obtain the goal of the sustainable development of nuclear power, future reactors must have new and more strict constraints to their design: optimization of ore materials will necessitate breeding (generation of fissile material from fertile material), and waste management will require transmutation. Innovative reactors that could achieve such objectives - generation IV or ADS (accelerator driven system) - are loaded with new fuels (thorium, heavy actinides) and function with neutron spectra for which nuclear data do not benefit from 50 years of industrial experience, and thus present particular challenges. After validation on an experimental reactor using an international benchmark, we take classical reactor physics tools along with available nuclear data uncertainties to calculate the sensitivities and uncertainties of the criticality and temperature coefficient of a thorium molten salt reactor. In addition, a study based on the important reaction rates for the calculation of cycle's equilibrium allows us to estimate the efficiency of different reprocessing strategies and the contribution of these reaction rates on the uncertainty of the breeding and then on the uncertainty of the size of the reprocessing plant. Finally, we use this work to propose an improvement of the high priority experimental request list. (author)

  3. Radiocarbon Records of Fossil Fuel Emissions From Urban Trees in the Greater Salt Lake Valley From Mid-Century to Present.

    Science.gov (United States)

    Chritz, K.; Buchert, M.; Walker, J. C.; Mendoza, D.; Pataki, D. E.; Xu, X.; Lin, J. C.

    2017-12-01

    Generating long term records of fossil fuel emissions of urban environments is complicated by the fact that direct observations of emissions and urban atmospheric CO2 concentrations were only collected in the recent past. Radiocarbon (14C) in tree rings from urban trees can provide archives of fossil fuel emissions that may track population growth over time, as higher population density is typically correlated with increased vehicular traffic and associated CO2 emissions, which are radiocarbon dead. We present radiocarbon measurements (n=125) from five roadside green ash trees (Fraxinus pennsylvanica) located in three cities of northern Utah - Salt Lake City (urban, 2016 population: 193,744), Logan City (agricultural, 2016 population: 49,110) and Heber (rural, 2016 population: 14,969). Urban trees were cored in four cardinal directions and ring widths were measured and counted to establish a chronology. One ring from every third year in a single core from each tree was removed and holocellulose was extracted from bulk wood of individual rings for 14C analysis. Fraction CO2 from fossil fuel burning (CO2-ff) was calculated using a simple mass-balance calculation from measured 14C values and remote background atmospheric 14CO2 values for NH Zone 2. The data from all three cities indicate a general trend of increasing CO2-ff uptake by the trees from 1980s to present, as expected with increased population growth and vehicular traffic. However, records in all three cities show unique elevated CO2-ff prior to the 1980s, assuming similar climate patterns through time, diverging from historic population size. We employed atmospheric simulations from the STILT (Stochastic Time-Inverted Lagrangian Transport) models for each of these trees to create footprints to determine source areas for CO2. These footprints reveal that atmospheric sampling areas can be large for certain trees, and other sources of 14C dead carbon, such as coal and natural gas from industrial emissions

  4. Assessment of effectiveness of geologic isolation systems. Test case release consequence analysis for a spent fuel repository in bedded salt

    International Nuclear Information System (INIS)

    Raymond, J.R.; Bond, F.W.; Cole, C.R.; Nelson, R.W.; Reisenauer, A.E.; Washburn, J.F.; Norman, N.A.; Mote, P.A.; Segol, G.

    1980-01-01

    Geologic and geohydrologic data for the Paradox Basin have been used to simulate movement of ground water and radioacrtive contaminants from a hypothetical nuclear reactor spent fuel repository after an assumed accidental release. The pathlines, travel times and velocity of the ground water from the repository to the discharge locale (river) were determined after the disruptive event by use of a two-dimensional finite difference hydrologic model. The concentration of radioactive contaminants in the ground water was calculated along a series of flow tubes by use of a one-dimensional mass transport model which takes into account convection, dispersion, contaminant/media interactions and radioactive decay. For the hypothetical site location and specific parameters used in this demonstration, it is found that Iodine-129 (I-129) is tthe only isotope reaching the Colorado River in significant concentration. This concentration occurs about 8.0 x 10 5 years after the repository has been breached. This I-129 ground-water concentration is about 0.3 of the drinking water standard for uncontrolled use. The groundwater concentration would then be diluted by the Colorado River. None of the actinide elements reach more than half the distance from the repository to the Colorado River in the two-million year model run time. This exercise demonstrates that the WISAP model system is applicable for analysis of contaminant transport. The results presented in this report, however, are valid only for one particular set of parameters. A complete sensitivity analysis must be performed to evaluate the range of effects from the release of contaminants from a breached repository

  5. Moltex Energy's stable salt reactors

    International Nuclear Information System (INIS)

    O'Sullivan, R.; Laurie, J.

    2016-01-01

    A stable salt reactor is a molten salt reactor in which the molten fuel salt is contained in fuel rods. This concept was invented in 1951 and re-discovered and improved recently by Moltex Energy Company. The main advantage of using molten salt fuel is that the 2 problematic fission products cesium and iodine do not exist in gaseous form but rather in a form of a salt that present no danger in case of accident. Another advantage is the strongly negative temperature coefficient for reactivity which means the reactor self-regulates. The feasibility studies have been performed on a molten salt fuel composed of sodium chloride and plutonium/uranium/lanthanide/actinide trichloride. The coolant fluid is a mix of sodium and zirconium fluoride salts that will need low flow rates. The addition of 1 mol% of metal zirconium to the coolant fluid reduces the risk of corrosion with standard steels and the addition of 2% of hafnium reduces the neutron dose. The temperature of the coolant is expected to reach 650 Celsius degrees at the exit of the core. This reactor is designed to be modular and it will be able to burn actinides. (A.C.)

  6. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. This part 'CIRCUITS' regroups under a condensed form - in French and using international units - the essential information contained in both basic documents of the American project for a molten-salt breeder power plant. This part is only dealing with things relating to the CEA-EDF workshop 'CIRCUITS'. It is not concerned with information on: the reactor and the moderator replacement, the primary and secondary salts, and the fuel salt reprocessing, that are dealt with in parts 'CORE' and 'CHEMISTRY' respectively. The possible evolutions in the data - and solutions - taken by the American designers for their successive projects (1970 to 1972) are shown. The MSBR power plant comprises three successive heat transfer circuits. The primary circuit (Hastelloy N), radioactive and polluted, containing the fuel salt, includes the reactor, pumps and exchangers. The secondary circuit (pipings made of modified Hastelloy N) contaminated in the exchanger, ensures the separation between the fuel and the fluid operating the turbo-alternator. The water-steam circuit feeds the turbine with steam. This steam is produced in the steam generator flowed by the secondary fluid. Some subsidiary circuits (discharge and storage of the primary and secondary salts, ventilation of the primary circuit ...) complete the three principal circuits which are briefly described. All circuits are enclosed inside the controlled-atmosphere building of the nuclear boiler. This building also ensures the biological protection and the mechanical protection against outer aggressions [fr

  7. Electrolytic reduction runs of 0.6 kg scale-simulated oxide fuel in a Li{sub 2}O-LiCl molten salt using metal anode shrouds

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Eun-Young, E-mail: eychoi@kaeri.re.kr [Korea Atomic Energy Research Institute, Daedoek-daero 989-111, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Lee, Jeong; Heo, Dong Hyun; Lee, Sang Kwon [Korea Atomic Energy Research Institute, Daedoek-daero 989-111, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Jeon, Min Ku [Korea Atomic Energy Research Institute, Daedoek-daero 989-111, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Department of Quantum Energy Chemical Engineering, University of Science and Technology, Gajeong-ro 217, Yuseong-gu, Daejeon 34113 (Korea, Republic of); Hong, Sun Seok [Korea Atomic Energy Research Institute, Daedoek-daero 989-111, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Kim, Sung-Wook [Korea Atomic Energy Research Institute, Daedoek-daero 989-111, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Department of Quantum Energy Chemical Engineering, University of Science and Technology, Gajeong-ro 217, Yuseong-gu, Daejeon 34113 (Korea, Republic of); Kang, Hyun Woo; Jeon, Sang-Chae; Hur, Jin-Mok [Korea Atomic Energy Research Institute, Daedoek-daero 989-111, Yuseong-gu, Daejeon 34057 (Korea, Republic of)

    2017-06-15

    Ten electrolytic reduction or oxide reduction (OR) runs of a 0.6 kg scale-simulated oxide fuel in a Li{sub 2}O-LiCl molten salt at 650 °C were conducted using metal anode shrouds. During this procedure, an anode shroud surrounds a platinum anode and discharges hot oxygen gas from the salt to outside of the OR apparatus, thereby preventing corrosion of the apparatus. In this study, a number of anode shrouds made of various metals were tested. Each metallic anode shroud consisted of a lower porous shroud for the salt phase and an upper nonporous shroud for the gas phase. A stainless steel (STS) wire mesh with five-ply layer was a material commonly used for the lower porous shroud for the OR runs. The metals tested for the upper nonporous shroud in the different OR runs are STS, nickel, and platinum- or silver-lined nickel. The lower porous shroud showed no significant damage during two consecutive OR runs, but exhibited signs of damage from three or more runs due to thermal stress. The upper nonporous shrouds made up of either platinum- or silver-lined nickel showed excellent corrosion resistance to hot oxygen gas while STS or nickel without any platinum or silver lining exhibited poor corrosion resistance. - Highlights: •Electrolytic reduction runs of a 0.6 kg scale-simulated oxide fuel in a Li{sub 2}O-LiCl molten salt at 650 °C were conducted using metal anode shrouds. •Each metallic anode shroud consisted of a lower porous shroud for the salt phase and an upper nonporous shroud for the gas phase. •The upper nonporous shrouds made up of noble metal-lined nickel showed excellent corrosion resistance to hot oxygen gas.

  8. Effect of cesium salt of tungstophosphoric acid (Cs-TPA) on the properties of sulfonated polyether ether ketone (SPEEK) composite membranes for fuel cell applications

    Energy Technology Data Exchange (ETDEWEB)

    Dogan, Hacer; Inan, Tuelay Y.; Unveren, Elif [The Scientific and Technological Research Council of Turkey (TUeBiTAK), Marmara Research Center, Chemistry Institute, P.K. 21, 41470 Gebze-Kocaeli (Turkey); Kaya, Metin [DEMIRDOeKUeM A.S. 4 Eyluel Mah, ismet inoenue Cad. No:245 Bozueyuek/Bilecik (Turkey)

    2010-08-15

    We have prepared composite membranes for fuel cell applications. Cesium salt of tungstophosphoric acid (Cs-TPA) particles was synthesized by aqueous solutions of tungstophosphoric acid and cesium hydroxide and, Cs-TPA particles and sulfonated (polyether ether ketone) (SPEEK) with two sulfonation degrees (DS), 60 and 70%have been used. We examined both the effects of Cs-TPA in SPEEK membranes as functions of sulfonation degrees of SPEEK and the content of Cs-TPA. The performance of the composite membranes was evaluated in terms of water uptake, ion exchange capacity, proton conductivity, chemical stability, hydrolytic stability, thermal stability and methanol permeability. The morphology of the membranes was investigated with SEM micrographs. Increasing sulfonation degree of SPEEK from 60 to 70 caused agglomeration of the Cs-TPA particles. The methanol permeability was reduced to 4.7 x 10{sup -7} cm{sup 2}/s for SPEEK (DS: 60%)/Cs-TPA membrane with 10 wt.% Cs-TPA concentration, and acceptable proton conductivity of 1.3 x 10{sup -1} S/cm was achieved at 80 C under 100% RH. The weight loss at 900 C increased with the addition of inorganic particles, as expected. The hydrolytic stability of the SPEEK/Cs-TPA based composite membranes was improved with the incorporation of the Cs-TPA particles into the matrix. We also noted that SPEEK60/Cs-TPA composite membranes were hydrolytically more stable than SPEEK70/Cs-TPA composite membranes. On the other hand, Methanol, water vapor, and hydrogen permeability values of SPEEK60 composite membranes were found to be lower than that of Nafion {sup registered}. (author)

  9. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. Emphasize is put essentially on the fuel salt of the primary circuit inside which fission reactions occur. The reasons why the (LiF-BeF 2 -ThF 4 -UF 4 ) salt was chosen for the M.S.B.R. concept are examined; the physical, physicochemical and chemical properties of this salt are discussed with its interactions with the structural materials and its evolution in time. An important part of this volume is devoted to the continuous reprocessing of the active salt, the project designers having deemed advisable to take advantage at best from the availability of a continuous purification, in a thermal breeding. The problem of tritium formation and distribution inside the reactor is also envisaged and the fundamentals of the chemistry of the secondary coolant salt are given. The solutions proposed are: the hydrogen scavenging of the primary circuit, a reduction in metal permeability by an oxyde layer deposition on the side in contact with the vapor, and tritium absorption through an isotope exchange with the hydroxifluoroborate [fr

  10. Residual salts separation from metal reduced electrolytically in a LiCl-Li2O molten salt

    International Nuclear Information System (INIS)

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

    2005-01-01

    The PWR spent oxide fuel can be reduced electrolytically in a hot molten salt for the conditioning and the preparation of a metallic fuel. Then the metal product is smelted into an ingot to be treated in the post process. Incidentally, the residual salt which originated from the molten salt and spent fuel elements should be separated from the metal product during the smelting. In this work, we constructed a surrogate material system to simulate the salt separation from the reduced spent fuel and studied the vaporization behaviors of the salts

  11. Molten-salt converter reactors

    International Nuclear Information System (INIS)

    Perry, A.M.

    1975-01-01

    Molten-salt reactors appear to have substantial promise as advanced converters. Conversion ratios of 0.85 to 0.9 should be attainable with favourable fuel cycle costs, with 235 U valued at $12/g. An increase in 235 U value by a factor of two or three ($10 to $30/lb. U 3 O 8 , $75/SWU) would be expected to increase the optimum conversion ratio, but this has not been analyzed in detail. The processing necessary to recover uranium from the fuel salt has been partially demonstrated in the MSRE. The equipment for doing this would be located at the reactor, and there would be no reliance on an established recycle industry. Processing costs are expected to be quite low, and fuel cycle optimization depends primarily on inventory and burnup or replacement costs for the fuel and for the carrier salt. Significant development problems remain to be resolved for molten-salt reactors, notably the control of tritium and the elimination of intergranular cracking of Hastelloy-N in contact with tellurium. However, these problems appear to be amenable to solution. It is appropriate to consider separating the development schedule for molten-salt reactors from that for the processing technology required for breeding. The Molten-Salt Converter Reactor should be a useful reactor in its own right and would be an advance towards the achievement of true breeding in thermal reactors. (author)

  12. Salt Tolerance

    OpenAIRE

    Xiong, Liming; Zhu, Jian-Kang

    2002-01-01

    Studying salt stress is an important means to the understanding of plant ion homeostasis and osmo-balance. Salt stress research also benefits agriculture because soil salinity significantly limits plant productivity on agricultural lands. Decades of physiological and molecular studies have generated a large body of literature regarding potential salt tolerance determinants. Recent advances in applying molecular genetic analysis and genomics tools in the model plant Arabidopsis thaliana are sh...

  13. Sea salt

    OpenAIRE

    Galvis-Sánchez, Andrea C.; Lopes, João Almeida; Delgadillo, Ivone; Rangel, António O. S. S.

    2013-01-01

    The geographical indication (GI) status links a product with the territory and with the biodiversity involved. Besides, the specific knowledge and cultural practices of a human group that permit transforming a resource into a useful good is protected under a GI designation. Traditional sea salt is a hand-harvested product originating exclusively from salt marshes from specific geographical regions. Once salt is harvested, no washing, artificial drying or addition of anti-caking agents are all...

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

    Energy Technology Data Exchange (ETDEWEB)

    Azizov, E. A.; Gladush, G. G., E-mail: gladush@triniti.ru; Dokuka, V. N.; Khayrutdinov, R. R. [State Research Center of the Russian Federation, Troitsk Institute for Innovation and Fusion Research (Russian Federation)

    2015-12-15

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

  15. Molten salt electrorefining method

    International Nuclear Information System (INIS)

    Tanaka, Hiroshi; Nakamura, Hitoshi; Shoji, Yuichi; Matsumaru, Ken-ichi.

    1994-01-01

    A molten cadmium phase (lower side) and a molten salt phase (upper side) are filled in an electrolytic bath. A basket incorporating spent nuclear fuels is inserted/disposed in the molten cadmium phase. A rotatable solid cathode is inserted/disposed in the molten salt phase. The spent fuels, for example, natural uranium, incorporated in the basket is dissolved in the molten cadmium phase. In this case, the uranium concentration in the molten salt phase is determined as from 0.5 to 20wt%. Then, electrolysis is conducted while setting a stirring power for stirring at least the molten salt phase of from 2.5 x 10 2 to 1 x 10 4 based on a reynolds number. Crystalline nuclei of uranium are precipitated uniformly on the surface of the solid cathode, and they grow into fine dendrites. With such procedures, since short-circuit between the cathode precipitates and the molten cadmium phase (anode) is scarcely caused, to improve the recovering rate of uranium. (I.N.)

  16. The Open-source Data Inventory for Anthropogenic CO2, version 2016 (ODIAC2016: a global monthly fossil fuel CO2 gridded emissions data product for tracer transport simulations and surface flux inversions

    Directory of Open Access Journals (Sweden)

    T. Oda

    2018-01-01

    Full Text Available The Open-source Data Inventory for Anthropogenic CO2 (ODIAC is a global high-spatial-resolution gridded emissions data product that distributes carbon dioxide (CO2 emissions from fossil fuel combustion. The emissions spatial distributions are estimated at a 1  ×  1 km spatial resolution over land using power plant profiles (emissions intensity and geographical location and satellite-observed nighttime lights. This paper describes the year 2016 version of the ODIAC emissions data product (ODIAC2016 and presents analyses that help guide data users, especially for atmospheric CO2 tracer transport simulations and flux inversion analysis. Since the original publication in 2011, we have made modifications to our emissions modeling framework in order to deliver a comprehensive global gridded emissions data product. Major changes from the 2011 publication are (1 the use of emissions estimates made by the Carbon Dioxide Information Analysis Center (CDIAC at the Oak Ridge National Laboratory (ORNL by fuel type (solid, liquid, gas, cement manufacturing, gas flaring, and international aviation and marine bunkers; (2 the use of multiple spatial emissions proxies by fuel type such as (a nighttime light data specific to gas flaring and (b ship/aircraft fleet tracks; and (3 the inclusion of emissions temporal variations. Using global fuel consumption data, we extrapolated the CDIAC emissions estimates for the recent years and produced the ODIAC2016 emissions data product that covers 2000–2015. Our emissions data can be viewed as an extended version of CDIAC gridded emissions data product, which should allow data users to impose global fossil fuel emissions in a more comprehensive manner than the original CDIAC product. Our new emissions modeling framework allows us to produce future versions of the ODIAC emissions data product with a timely update. Such capability has become more significant given the CDIAC/ORNL's shutdown. The ODIAC data

  17. The Open-source Data Inventory for Anthropogenic CO2, version 2016 (ODIAC2016): a global monthly fossil fuel CO2 gridded emissions data product for tracer transport simulations and surface flux inversions

    Science.gov (United States)

    Oda, Tomohiro; Maksyutov, Shamil; Andres, Robert J.

    2018-01-01

    The Open-source Data Inventory for Anthropogenic CO2 (ODIAC) is a global high-spatial-resolution gridded emissions data product that distributes carbon dioxide (CO2) emissions from fossil fuel combustion. The emissions spatial distributions are estimated at a 1 × 1 km spatial resolution over land using power plant profiles (emissions intensity and geographical location) and satellite-observed nighttime lights. This paper describes the year 2016 version of the ODIAC emissions data product (ODIAC2016) and presents analyses that help guide data users, especially for atmospheric CO2 tracer transport simulations and flux inversion analysis. Since the original publication in 2011, we have made modifications to our emissions modeling framework in order to deliver a comprehensive global gridded emissions data product. Major changes from the 2011 publication are (1) the use of emissions estimates made by the Carbon Dioxide Information Analysis Center (CDIAC) at the Oak Ridge National Laboratory (ORNL) by fuel type (solid, liquid, gas, cement manufacturing, gas flaring, and international aviation and marine bunkers); (2) the use of multiple spatial emissions proxies by fuel type such as (a) nighttime light data specific to gas flaring and (b) ship/aircraft fleet tracks; and (3) the inclusion of emissions temporal variations. Using global fuel consumption data, we extrapolated the CDIAC emissions estimates for the recent years and produced the ODIAC2016 emissions data product that covers 2000-2015. Our emissions data can be viewed as an extended version of CDIAC gridded emissions data product, which should allow data users to impose global fossil fuel emissions in a more comprehensive manner than the original CDIAC product. Our new emissions modeling framework allows us to produce future versions of the ODIAC emissions data product with a timely update. Such capability has become more significant given the CDIAC/ORNL's shutdown. The ODIAC data product could play an important

  18. Fuel model studies. Comparison of our present version of GAPCON-THERMAL-2 with results from the EPRI code comparison study. Partial report

    International Nuclear Information System (INIS)

    Malen, K.; Jansson, L.

    1978-08-01

    Runs with our present version of GAPCON-THERMAL-2 have been compared to results from the EPRI code comparison study. Usually also our version of GAPCON predicts high temperatures, 100-300 K or 10-15% higher than average code predictions and experimental results. The well-known temperaturegas release instablility is found also with GAPCON. In this case one identifies the gas release limits 1400 deg C and 1700 deg C as instablility points. (author)

  19. Calculation of β-effective of a molten salt reactor

    International Nuclear Information System (INIS)

    Hirakawa, N.; Sakaba, H.

    1987-01-01

    A method to calculate the β eff of a molten salt reactor was developed taking the effect of the flow of the molten salt into account. The method was applied to the 1000MW MSR design made by ORNL. The change in β eff due to the change in the residence time outside of the core of the fuel salt and to the change in the flow velocity when the total amount of the fuel salt is kept constant were investigated. It was found that β eff was reduced to 47.9% of the value when the fuel salt is at rest for the present design. (author)

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

    International Nuclear Information System (INIS)

    Uhlir, Jan

    2007-01-01

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

  1. Salt cookbook

    CERN Document Server

    Saha, Anirban

    2015-01-01

    If you are a professional associated with system and infrastructure management, looking at automated infrastructure and deployments, then this book is for you. No prior experience of Salt is required.

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

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J. J.; Harrison, T. J.; Gehin, J. C. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6172 (United States)

    2013-07-01

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

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

    International Nuclear Information System (INIS)

    Powers, J. J.; Harrison, T. J.; Gehin, J. C.

    2013-01-01

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

  4. National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel. Special study No. 3. Waste retrieval from backfilled regions

    International Nuclear Information System (INIS)

    1978-09-01

    Methods and costs were studied for delayed canister retrieval from rooms that had been backfilled immediately after canister storage. The effects of this method of storage on mine geometry, thermal and rock mechanics environments, mine development and operations, mine ventilation, time schedule, retrieval machinery and safety were investigated. Salt and air temperatures were determined. Pillar width, number of rooms, extraction ratio, tonnages of mined salt, and salt handling and hoisting requirements were calculated. The required changes in mining equipment were established. Salt handling and elapsed time schedules were developed. Ventilation requirements - size and number of shafts, size the arrangement of airways, number of stacks, and size and number of fans were then calculated. The development sequence of these facilities was established. Canister retrieval problems were analyzed for canisters stuck in the hole as well as free. Retrieval methods and machinery were studied and are described. Safety with respect to both radiation and room collapse was studied and compared with CDR safety conditions. The effects of a reduced themal loading of 30 KW/acre on temperatures, room closure, mine layout, ventilation and ground control were studied and reported. A cost estimate was prepared, giving cost differentials between the base CDR costs and Special Study No. 3. Two appendices are included. The first contains nine Heat Transfer memoranda that state the thermal basis of this study. The second appendix provides a detailed operating time analysis of the retrieval machinery

  5. The molten salt reactor adventure

    International Nuclear Information System (INIS)

    MacPherson, H.G.

    1985-01-01

    A personal history of the development of molten salt reactors in the United States is presented. The initial goal was an aircraft propulsion reactor, and a molten fluoride-fueled Aircraft Reactor Experiment was operated at Oak Ridge National Laboratory in 1954. In 1956, the objective shifted to civilian nuclear power, and reactor concepts were developed using a circulating UF 4 -ThF 4 fuel, graphite moderator, and Hastelloy N pressure boundary. The program culminated in the successful operation of the Molten Salt Reactor Experiment in 1965 to 1969. By then the Atomic Energy Commission's goals had shifted to breeder development; the molten salt program supported on-site reprocessing development and study of various reactor arrangements that had potential to breed. Some commercial and foreign interest contributed to the program which, however, was terminated by the government in 1976. The current status of the technology and prospects for revived interest are summarized

  6. Treat upgrade fuel fabrication

    International Nuclear Information System (INIS)

    Davidson, K.V.; Schell, D.H.

    1979-01-01

    An extrusion and thermal treatment process was developed to produce graphite fuel rods containing a dispersion of enriched UO 2 . These rods will be used in an upgraded version of the Transient Reactor Test Facility (TREAT). The improved fuel provides a higher graphite matrix density, better fuel dispersion and higher thermal capabilities than the existing fuel

  7. Factsheets for road transportation fuels. Specifications and perspectives. Revised version 2012; Factsheets Brandstoffen voor het wegverkeer. Kenmerken en perspectief. Herziene versie 2012

    Energy Technology Data Exchange (ETDEWEB)

    Verbeek, R. [TNO, Delft (Netherlands); Kampman, B. [CE Delft, Delft (Netherlands)

    2012-07-15

    Vehicles powered by green gas, electric transport, hydrogen as fuel; a whole new series of applications has become available. These applications have been listed, offering policy makers (government, province and local authority) as well as car fleet managers an overview of the current state of affairs. The factsheets show per fuel what the main specifications of alternative fuels and engines are. Subjects discussed include applicability of a fuel in a vehicle; air pollution compared to current fuel; greenhouse gas emissions and energy use; infrastructure (distribution and refueling/charging infrastructure; cost and fiscal aspects; development stage [Dutch] Auto's op groen gas, elektrisch vervoer, waterstof als brandstof, er is inmiddels een reeks van nieuwe toepassingen beschikbaar. Deze zijn op een rijtje gezet, zodat voor beleidsmakers (Rijk, provincie en gemeente) maar ook voor wagenparkbeheerders, overzichtelijk wordt wat de stand van zaken is. In de factsheets worden per brandstof de belangrijkste eigenschappen van alternatieve brandstoffen en motoren op een rijtje gezet. Onderwerpen die aan bod komen zijn: toepasbaarheid van een brandstof in een voertuig; luchtverontreiniging t.o.v. bestaande brandstof; uitstoot broeikasgassen en energieverbruik; infrastructuur (distributie en tank/laadinfrastructuur); kosten en fiscale aspecten; ontwikkelingsstadium.

  8. The problems of calculation of heat transfer crisis in fuel assemblies of PW reactors based on modern versions of thermohydraulic codes

    International Nuclear Information System (INIS)

    Fialko, N.M.; Sharaevskij, G.I.; Sharaevskaya, E.I.; Babak, E.I.

    2014-01-01

    The article gives an analysis of the adequacy of computer software systems FASCICLE BM-DF and COBRA, which are designed to calculate the main parameters of the safety of water-cooled nuclear reactors. This calculation is based on determining the local thermal-hydraulic parameters of the flow of coolant in the fuel rod assembled elements. In this article introduced the results of the comparison of experiments performed to determine the distribution of the main thermal-hydraulic flow parameters characteristic of subchannels of fuel rod assembled elements with the data for calculating these parameters on the basis of declared computer codes. Particular attention is paid to the analysis of experimental and calculation data, by definition, burnout in rod fuel assembled elements

  9. CFTSIM-ITER dynamic fuel cycle model

    International Nuclear Information System (INIS)

    Busigin, A.; Gierszewski, P.

    1998-01-01

    Dynamic system models have been developed for specific tritium systems with considerable detail and for integrated fuel cycles with lesser detail (e.g. D. Holland, B. Merrill, Analysis of tritium migration and deposition in fusion reactor systems, Proceedings of the Ninth Symposium Eng. Problems of Fusion Research (1981); M.A. Abdou, E. Vold, C. Gung, M. Youssef, K. Shin, DT fuel self-sufficiency in fusion reactors, Fusion Technol. (1986); G. Spannagel, P. Gierszewski, Dynamic tritium inventory of a NET/ITER fuel cycle with lithium salt solution blanket, Fusion Eng. Des. (1991); W. Kuan, M.A. Abdou, R.S. Willms, Dynamic simulation of a proposed ITER tritium processing system, Fusion Technol. (1995)). In order to provide a tool to understand and optimize the behavior of the ITER fuel cycle, a dynamic fuel cycle model called CFTSIM is under development. The CFTSIM code incorporates more detailed ITER models, specifically for the important isotope separation system, and also has an easier-to-use graphical interface. This paper provides an overview of CFTSIM Version 1.0. The models included are those with significant and varying tritium inventories over a test campaign: fueling, plasma and first wall, pumping, fuel cleanup, isotope separation and storage. An illustration of the results is shown. (orig.)

  10. Molten salt reactors. The AMSTER concept

    International Nuclear Information System (INIS)

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

    2001-01-01

    This article presents the concept of actinide molten salt transmuter (AMSTER). This reactor is graphite-moderated and is dedicated to the burning of actinides. The main difference with a molten salt reactor is that its liquid fuel undergoes an on-line partial reprocessing in which fission products are extracted and heavy nuclei are reintroduced into the fuel. In order to maintain the reactivity regular injections of 235 U-salt are made. In classical reactors, fuel burn-up is limited by the swelling of the cladding and the radiation fuel pellets resistance, in AMSTER there is no limitation to the irradiation time of the fuel, so all the actinides can be burnt or transmuted. (A.C.)

  11. Bath Salts

    Science.gov (United States)

    ... deaths and been blamed for a handful of suicides and murders. Two of the chemicals in bath salts (mephedrone and MDPV) are Schedule I class drugs. That means they have a high potential for abuse and no accepted medical use . People who are ...

  12. Low temperature chemical processing of graphite-clad nuclear fuels

    Science.gov (United States)

    Pierce, Robert A.

    2017-10-17

    A reduced-temperature method for treatment of a fuel element is described. The method includes molten salt treatment of a fuel element with a nitrate salt. The nitrate salt can oxidize the outer graphite matrix of a fuel element. The method can also include reduced temperature degradation of the carbide layer of a fuel element and low temperature solubilization of the fuel in a kernel of a fuel element.

  13. Study of the moderating effect of salts on the sodium-water reaction on the cleaning of irradiated fuel assemblies from fast neutron reactors, using fluid sodium heat transfer

    International Nuclear Information System (INIS)

    Lacroix, Marie

    2014-01-01

    Within the framework of the development of generation IV reactors one of the research tracks is related to the development of fast neutron reactors using fluid sodium heat transfer. The CEA (French Alternative Energies and Atomic Energy Commission) plans to build a prototype of reactor of this type called 'ASTRID'. To address development requirements for this prototype, research is in progress on the reactor's availability and in particular on the reduction of the washing duration for residual sodium fuel assemblies during their discharge. In fact, because sodium is very reactive with water (presently the only available process), the washing is done, for example, by very gradual addition. A solution currently being studied at the CEA and which is the subject of this thesis report consists of the addition of an aqueous salts solutions to the washing water in order to slow down the kinetic reaction. This doctoral dissertation describes the various salts, which have been evaluated and aims to explain their action mode. (author) [fr

  14. Innovative nuclear system based on liquid fuel

    International Nuclear Information System (INIS)

    Delpech, S.; Jaskierowicz, S.; Picard, G.; Merle-Lucotte, E.; Heuer, D.; Doligez, X.

    2009-01-01

    The aim of this paper is to present the physical properties and characteristics of the innovative concept of Molten Salt Fast Reactor (MSFR) developed by CNRS (France) and the corresponding fuel salt reprocessing proposed to clean up the fuel salt based on an analytical approach of lanthanides and actinides extraction. (author)

  15. Evaluation of light LPG-fueled vehicles and comparison with their diesel-fueled and gasoline-fueled versions. Measurements of regulated and non-regulated pollutant emissions. Measurement of CO{sub 2} emissions and fuel consumption; Evaluation de vehicules legers fonctionnant au GPL et comparatif avec leurs versions essence et diesel. Mesures des emissions polluantes reglementees et non reglementees. Mesure des emissions de CO{sub 2} et de la consommation de carburant

    Energy Technology Data Exchange (ETDEWEB)

    Gagnepain, L.

    2004-04-01

    During the end of the 1990's the number of light LPG-fueled vehicles has increased thanks to the environmental advantages of this automotive fuel and to its tax depreciation advantage. A European emission test program (EETP) has been initiated by LPG companies (BP LPG, French committee of butane propane, liquefied petroleum gas association, Shell global autogas, SHV gas, Totalgas, Vereniging Vloeibaar Gas) and by environment agencies (Ademe, energy saving trust) in order to compare the environmental performances of LPG-fueled vehicles with their equivalent diesel-fueled and gasoline-fueled models. Four laboratories have participated to this evaluation: TUV (Germany), IFP (France), TNO (Netherlands) and Millbrook (UK). The comparative results are presented in tables and graphs (CO, HC, NO{sub x} and CO{sub 2} emissions, fuel consumption, CH{sub 4}, N{sub 2}O, benzene, 1,3-butadiene, formaldehyde, acetaldehyde, polycyclic aromatic hydrocarbons (PAH) emissions, 'well to wheel' greenhouse impact, ozone formation potential, carcinogen risk). The results show important differences among the different models and differences in the environmental performances depending on the vehicle utilization (highway, urban area use). In general the new generation (Euro 3) of LPG-fueled vehicles is significantly better in terms of environmental impact than the previous generation (Euro 2) of vehicles. (J.S.)

  16. Immobilization of IFR salt wastes in mortar

    International Nuclear Information System (INIS)

    Fisher, D.F.; Johnson, T.R.

    1988-01-01

    Portland cement-base mortars are being considered for immobilizing chloride salt wastes from the fuel cycle of an integral fast reactor (IFR). The IFR is a sodium-cooled fast reactor with metal fuel. It has a close-coupled fuel cycle in which fission products are separated from the actinides in an electrochemical cell operating at 500 degrees C. This cell has a cadmium anode and a liquid salt electrolyte. The salt will be a low-melting mixture of alkaline and alkaline earth chlorides. This paper discusses one method being considered for immobilizing this treated salt, to disperse it in a portland cement-base motar, which would then be sealed in corrosion-resistant containers. For this application, the grout must be sufficiently fluid that it can be pumped into canisters where it will solidify into a strong, leach-resistant material

  17. The thorium fuel cycle in molten salt reactors as a solution for the energetic problem of the 21. century? The TMSR-NM concept

    International Nuclear Information System (INIS)

    Merle-Lucotte, E.

    2008-06-01

    Within the frame of development of sustainable nuclear programs, this report does not only deal with the development of nuclear systems, but with the general context in which such a development will occur. While describing and commenting her professional career in different nuclear research institutions and on various research programs, the author describes the assets and challenges of the electro-nuclear sector, and then focuses on the research structures and contexts for future possible nuclear concepts, and more particularly like melted salt reactors for which she highlights scientific and technical problems which are still to be solved. She describes French, European and world programs which were to start by 2009

  18. Challenges Related to the Use of Liquid Metal and Molten Salt Coolants in Advanced Reactors. Report of the collaborative project COOL of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2013-05-01

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was launched in 2000, based on a resolution by the IAEA General Conference (GC(44)/RES/21). INPRO aims at helping to ensure that nuclear energy is available in the twenty-first century in a sustainable manner, and seeks to bring together all interested Member States, both technology holders and technology users, to jointly consider actions to achieve desired innovations. INPRO is taking care of the specific needs of developing countries. One of the aims of INPRO is to develop options for enhanced sustainability through promotion of technical and institutional innovations in nuclear energy technology through collaborative projects among IAEA Member States. Collaboration among INPRO members is fostered on selected innovative nuclear technologies to bridge technology gaps. Collaborative projects have been selected so that they complement other national and international R and D activities. The INPRO Collaborative Project COOL on Investigation of Technological Challenges Related to the Removal of Heat by Liquid Metal and Molten Salt Coolants from Reactor Cores Operating at High Temperatures investigated the technological challenges of cooling reactor cores that operate at high temperatures in advanced fast reactors, high temperature reactors and accelerator driven systems by using liquid metals and molten salts as coolants. The project was initiated in 2008 and was led by India; experts from Brazil, China, Germany, India, Italy and the Republic of Korea participated and provided chapters of this report. The INPRO Collaborative Project COOL addressed the following fields of research regarding liquid metal and molten salt coolants: (i) survey of thermophysical properties; (ii) experimental investigations and computational fluid dynamics studies on thermohydraulics, specifically pressure drop and heat transfer under different operating conditions; (iii) monitoring and control of coolant

  19. High throughput salt separation from uranium deposits

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, S.W.; Park, K.M.; Kim, J.G.; Kim, I.T.; Park, S.B., E-mail: swkwon@kaeri.re.kr [Korea Atomic Energy Research Inst. (Korea, Republic of)

    2014-07-01

    It is very important to increase the throughput of the salt separation system owing to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites in pyroprocessing. Multilayer porous crucible system was proposed to increase a throughput of the salt distiller in this study. An integrated sieve-crucible assembly was also investigated for the practical use of the porous crucible system. The salt evaporation behaviors were compared between the conventional nonporous crucible and the porous crucible. Two step weight reductions took place in the porous crucible, whereas the salt weight reduced only at high temperature by distillation in a nonporous crucible. The first weight reduction in the porous crucible was caused by the liquid salt penetrated out through the perforated crucible during the temperature elevation until the distillation temperature. Multilayer porous crucibles have a benefit to expand the evaporation surface area. (author)

  20. Thorium Molten-Salt Nuclear Energy Synergetics

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  1. Electrochemical ion separation in molten salts

    Science.gov (United States)

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

    2017-12-19

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

  2. Immobilization of IFR salt wastes in mortar

    International Nuclear Information System (INIS)

    Fischer, D.F.; Johnson, T.R.

    1988-01-01

    Portland cement-base mortars are being considered for immobilizing chloride salt wastes produced by the fuel cycles of Integral Fast Reactors (IFR). The IFR is a sodium-cooled fast reactor with metal alloy fuels. It has a close-coupled fuel cycle in which fission products are separated from the actinides in an electrochemical cell operating at 500/degree/C. This cell has a liquid cadmium anode in which the fuels are dissolved and a liquid salt electrolyte. The salt will be a mixture of either lithium, potassium, and sodium chlorides or lithium, calcium, barium, and sodium chlorides. One method being considered for immobilizing the treated nontransuranic salt waste is to disperse the salt in a portland cement-base mortar that will be sealed in corrosion-resistant containers. For this application, the grout must be sufficiently fluid that it can be pumped into canister-molds where it will solidify into a strong, leach-resistant material. The set times must be longer than a few hours to allow sufficient time for processing, and the mortar must reach a reasonable compressive strength (/approximately/7 MPa) within three days to permit handling. Because fission product heating will be high, about 0.6 W/kg for a mortar containing 10% waste salt, the effects of elevated temperatures during curing and storage on mortar properties must be considered

  3. Compatibility of molten salt and structural materials

    International Nuclear Information System (INIS)

    Kawakami, Masahiro

    1994-01-01

    As the important factors for considering the compatibility of fuel salt and coolant salt with structural materials in molten salt reactors, there are the moisture remaining in molten salt and the fluorine potential in molten salt. In this study, as for the metals which are the main components of corrosion resistant alloys, the corrosion by the moisture remaining in molten salt and the dependence of the corrosion on fluorine potential were examined. As the molten salts, an eutectic molten salt LiF-BeF 2 was mainly used, and LiF-KF was used in combination. As the metallic materials, Cr, Ni and Cu which are the main components of corrosion resistant and heat resistant alloys, Hastelloy and Monel, were used. In the experiment, the metal pieces were immersed in the molten salt, and by sampling the molten salt, the change with time lapse of the concentration of the dissolved metals was examined. Besides, the electrochemical measurement was carried out for Cr, of which the corrosion was remarkable, and the change with time lapse of the dissolved ions was examined. The experimental setup, the experimental method, and the results of the immersion test and the electrochemical test are reported. The experiment on the corrosion of metals depending on fluorine potential is also reported. (K.I.)

  4. Compatibility of molten salt and structural materials

    Energy Technology Data Exchange (ETDEWEB)

    Kawakami, Masahiro [Toyohashi Univ. of Technology, Aichi (Japan)

    1994-12-01

    As the important factors for considering the compatibility of fuel salt and coolant salt with structural materials in molten salt reactors, there are the moisture remaining in molten salt and the fluorine potential in molten salt. In this study, as for the metals which are the main components of corrosion resistant alloys, the corrosion by the moisture remaining in molten salt and the dependence of the corrosion on fluorine potential were examined. As the molten salts, an eutectic molten salt LiF-BeF{sub 2} was mainly used, and LiF-KF was used in combination. As the metallic materials, Cr, Ni and Cu which are the main components of corrosion resistant and heat resistant alloys, Hastelloy and Monel, were used. In the experiment, the metal pieces were immersed in the molten salt, and by sampling the molten salt, the change with time lapse of the concentration of the dissolved metals was examined. Besides, the electrochemical measurement was carried out for Cr, of which the corrosion was remarkable, and the change with time lapse of the dissolved ions was examined. The experimental setup, the experimental method, and the results of the immersion test and the electrochemical test are reported. The experiment on the corrosion of metals depending on fluorine potential is also reported. (K.I.).

  5. Neutronics of a liquid salt cooled - very high temperature reactor

    International Nuclear Information System (INIS)

    Zakova, J.

    2007-01-01

    matrix. The external radius of a TRISO particle has been set to 410μm, the radius of the fuel kernel to 150μm, in case of plutonium fueled core, and 215μm in case of uranium fueled core. The core was modelled in stochastic, three-dimensional code MCNP, version 4c3, in the finest detail. First, an under moderated core setup was found for both types of fuel by modifying the fuel to moderator ratio; then, the void and the thermal coefficients of reactivity were investigated. Few single - component molten salts were involved in the study of the void effect, in order to estimate worth of these components; NaF, BeF 2 , LiF, ZrF 4 . As a reference multi-component salt, Li 2 Be 4 F, referred to as FLiBe, was investigated. Results: It can be seen that removing BeF 2 from the core brings a negative reactivity contribution, while other three components, NaF, LiF and ZrF 4 would in a mixture contribute to the reactivity positively. Voiding FLiBe, which is a mixture of 66% of LiF and 34% BeF 2 , is equivalent to a negative reactivity insertion. Both the moderator and the fuel temperature coefficients of reactivity are large and negative for both plutonium and uranium fueled core. In the operational temperature interval (1200 K for graphite and 1500 K for fuel), the total temperature feedback is - 7.82 pcm/K for the plutonium fueled core and -2.47 pcm/K for the uranium fueled core. This results show, that the LS-VHTR core has a potential to meet the basic safety requirements as both uranium, and spent LWR fuel burner. References: [1] D. T. Ingersoll, L. J. Ott, J. P. Renier, S. J. Ball, W. R. Corwin, C. W. Forsberg, D. F. Williams, D. F. Wilson, L. Reid, G. D. Del Cul, P. F. Peterson, H. Zhao, P. S. Pickard, E. J. Parma. Status of Preconceptual Design of the Advanced High-Temperature Reactor. (ORNL, The United States of America, Tennessee 2004). [2] A. Talamo, W. Gudowski, F. Venneri, Annals of Nuclear Energy, 31, 173-196 (2004), The burnup capabilities of the Deep Burn Modular

  6. Fundamental study on the salt distillation from the mixtures of rare earth precipitates and LiCl-KCl eutectic salt

    International Nuclear Information System (INIS)

    Yang, H. C.; Eun, H. C.; Cho, Y. Z.; Lee, H. S.; Kim, I. T.

    2008-01-01

    An electrorefining process of spent nuclear fuel generates waste salt containing some radioactive metal chlorides. The most effective method to reduce salt waste volume is to separate radioactive metals from non-radioactive salts. A promising approach is to change radioactive metal chlorides into salt-insoluble oxides by an oxygen sparging. Following this, salt distillation process is available to effectively separate the precipitated particulate metal oxides from salt. This study investigated the distillation rates of LiCl-KCl eutectic salt under different vacuums at elevated temperatures. The first part study investigated distillation rates of eutectic salt under different vacuums at high temperatures by using thermo-gravimetric furnace system. In the second part, we tested the removal of eutectic salt from the RE precipitates by using the laboratory vacuum distillation furnace system. Investigated variables were the temperature of mixture, the degree of vacuum and the time

  7. Molten salt reactors: reactor cores

    International Nuclear Information System (INIS)

    1983-01-01

    In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr

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

    International Nuclear Information System (INIS)

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

    1995-03-01

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

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

    International Nuclear Information System (INIS)

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

    1995-02-01

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

  10. Cooking without salt

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/patientinstructions/000760.htm Cooking without salt To use the sharing features on ... other dishes to add zest. Try Salt-free Cooking Explore cooking with salt substitutes. Add a splash ...

  11. Studies on components for a molten salt reactor

    International Nuclear Information System (INIS)

    Nejedly, M.; Matal, O.

    2003-01-01

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

  12. Molten salts processes and generic simulation

    International Nuclear Information System (INIS)

    Ogawa, Toru; Minato, Kazuo

    2001-01-01

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

  13. Molten salts processes and generic simulation

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-12-01

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

  14. Salt Reductions in Some Foods in The Netherlands: Monitoring of Food Composition and Salt Intake.

    Science.gov (United States)

    Temme, Elisabeth H M; Hendriksen, Marieke A H; Milder, Ivon E J; Toxopeus, Ido B; Westenbrink, Susanne; Brants, Henny A M; van der A, Daphne L

    2017-07-22

    High salt intake increases blood pressure and thereby the risk of chronic diseases. Food reformulation (or food product improvement) may lower the dietary intake of salt. This study describes the changes in salt contents of foods in the Dutch market over a five-year period (2011-2016) and differences in estimated salt intake over a 10-year period (2006-2015). To assess the salt contents of foods; we obtained recent data from chemical analyses and from food labels. Salt content of these foods in 2016 was compared to salt contents in the 2011 version Dutch Food Composition Database (NEVO, version 2011), and statistically tested with General Linear Models. To estimate the daily dietary salt intake in 2006, 2010, and 2015, men and women aged 19 to 70 years were recruited through random population sampling in Doetinchem, a small town located in a rural area in the eastern part of the Netherlands. The characteristics of the study population were in 2006: n = 317, mean age 49 years, 43% men, in 2010: n = 342, mean age 46 years, 45% men, and in 2015: n = 289, mean age 46 years, 47% men. Sodium and potassium excretion was measured in a single 24-h urine sample. All estimates were converted to a common metric: salt intake in grams per day by multiplication of sodium with a factor of 2.54. In 2016 compared to 2011, the salt content in certain types of bread was on average 19 percent lower and certain types of sauce, soup, canned vegetables and legumes, and crisps had a 12 to 26 percent lower salt content. Salt content in other types of foods had not changed significantly. Between 2006, 2010 and 2015 the estimated salt intake among adults in Doetinchem remained unchanged. In 2015, the median estimated salt intake was 9.7 g per day for men and 7.4 g per day for women. As in 2006 and 2010, the estimated salt intake in 2015 exceeded the recommended maximum intake of 6 g per day set by the Dutch Health Council. In the Netherlands, the salt content of bread, certain sauces, soups

  15. Salt Reductions in Some Foods in The Netherlands: Monitoring of Food Composition and Salt Intake

    Directory of Open Access Journals (Sweden)

    Elisabeth H. M. Temme

    2017-07-01

    Full Text Available Background and objectives. High salt intake increases blood pressure and thereby the risk of chronic diseases. Food reformulation (or food product improvement may lower the dietary intake of salt. This study describes the changes in salt contents of foods in the Dutch market over a five-year period (2011–2016 and differences in estimated salt intake over a 10-year period (2006–2015. Methods. To assess the salt contents of foods; we obtained recent data from chemical analyses and from food labels. Salt content of these foods in 2016 was compared to salt contents in the 2011 version Dutch Food Composition Database (NEVO, version 2011, and statistically tested with General Linear Models. To estimate the daily dietary salt intake in 2006, 2010, and 2015, men and women aged 19 to 70 years were recruited through random population sampling in Doetinchem, a small town located in a rural area in the eastern part of the Netherlands. The characteristics of the study population were in 2006: n = 317, mean age 49 years, 43% men, in 2010: n = 342, mean age 46 years, 45% men, and in 2015: n = 289, mean age 46 years, 47% men. Sodium and potassium excretion was measured in a single 24-h urine sample. All estimates were converted to a common metric: salt intake in grams per day by multiplication of sodium with a factor of 2.54. Results. In 2016 compared to 2011, the salt content in certain types of bread was on average 19 percent lower and certain types of sauce, soup, canned vegetables and legumes, and crisps had a 12 to 26 percent lower salt content. Salt content in other types of foods had not changed significantly. Between 2006, 2010 and 2015 the estimated salt intake among adults in Doetinchem remained unchanged. In 2015, the median estimated salt intake was 9.7 g per day for men and 7.4 g per day for women. As in 2006 and 2010, the estimated salt intake in 2015 exceeded the recommended maximum intake of 6 g per day set by the Dutch Health Council

  16. Molten salts and nuclear energy production

    International Nuclear Information System (INIS)

    Le Brun, Christian

    2007-01-01

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

  17. Modeling report of DYMOND code (DUPIC version)

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Yacout, Abdellatif M.

    2003-04-01

    The DYMOND code employs the ITHINK dynamic modeling platform to assess the 100-year dynamic evolution scenarios for postulated global nuclear energy parks. Firstly, DYMOND code has been developed by ANL(Argonne National Laboratory) to perform the fuel cycle analysis of LWR once-through and LWR-FBR mixed plant. Since the extensive application of DYMOND code has been requested, the first version of DYMOND has been modified to adapt the DUPIC, MSR and RTF fuel cycle. DYMOND code is composed of three parts; the source language platform, input supply and output. But those platforms are not clearly distinguished. This report described all the equations which were modeled in the modified DYMOND code (which is called as DYMOND-DUPIC version). It divided into five parts;Part A deals model in reactor history which is included amount of the requested fuels and spent fuels. Part B aims to describe model of fuel cycle about fuel flow from the beginning to the end of fuel cycle. Part C is for model in re-processing which is included recovery of burned uranium, plutonium, minor actinide and fission product as well as the amount of spent fuels in storage and disposal. Part D is for model in other fuel cycle which is considered the thorium fuel cycle for MSR and RTF reactor. Part E is for model in economics. This part gives all the information of cost such as uranium mining cost, reactor operating cost, fuel cost etc

  18. Modeling report of DYMOND code (DUPIC version)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan [KAERI, Taejon (Korea, Republic of); Yacout, Abdellatif M [Argonne National Laboratory, Ilinois (United States)

    2003-04-01

    The DYMOND code employs the ITHINK dynamic modeling platform to assess the 100-year dynamic evolution scenarios for postulated global nuclear energy parks. Firstly, DYMOND code has been developed by ANL(Argonne National Laboratory) to perform the fuel cycle analysis of LWR once-through and LWR-FBR mixed plant. Since the extensive application of DYMOND code has been requested, the first version of DYMOND has been modified to adapt the DUPIC, MSR and RTF fuel cycle. DYMOND code is composed of three parts; the source language platform, input supply and output. But those platforms are not clearly distinguished. This report described all the equations which were modeled in the modified DYMOND code (which is called as DYMOND-DUPIC version). It divided into five parts;Part A deals model in reactor history which is included amount of the requested fuels and spent fuels. Part B aims to describe model of fuel cycle about fuel flow from the beginning to the end of fuel cycle. Part C is for model in re-processing which is included recovery of burned uranium, plutonium, minor actinide and fission product as well as the amount of spent fuels in storage and disposal. Part D is for model in other fuel cycle which is considered the thorium fuel cycle for MSR and RTF reactor. Part E is for model in economics. This part gives all the information of cost such as uranium mining cost, reactor operating cost, fuel cost etc.

  19. Subcritical enhanced safety molten-salt reactor concept

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Ignatiev, V.V.; Men'shikov, L.I.; Prusakov, V.N.; Ponomarev-Stepnoy, N.N.; Subbotin, S.A.; Krasnykh, A.K.; Rudenko, V.T.; Somov, L.N.

    1995-01-01

    The nuclear power and its fuel cycle safety requirements can be met in the main by providing nuclear power with subcritical molten salt reactors (SMSR) - 'burner' with an external neutron source. The utilized molten salt fuel is the decisive advantage of the SMSR over other burners. Fissile and fertile nuclides in the burner are solved in a liquid salt in the form of fluorides. This composition acts simultaneously as: a) fuel, b) coolant, c) medium for chemical partitioning and reprocessing. The effective way of reducing the external source power consists in the cascade neutron multiplication in the system of coupled reactors with suppressed feedback between them. (author)

  20. Indian programme on molten salt cooled nuclear reactors

    International Nuclear Information System (INIS)

    DuIera, I.V.; Vijayan, P.K.; Sinha, R.K.

    2013-01-01

    Bhabha Atomic Research Centre (BARC) is developing a 600 MWth pebble bed high temperature reactor, cooled by natural circulation of molten fluoride salts and is capable of supplying process heat at 1000 ℃ to facilitate hydrogen production by splitting water. BARC has also initiated studies for a reactor concept in which salts of molten fluoride fuel and coolant in fluid form, flows through the reactor core of graphite moderator, resulting in nuclear fission within the molten salt. For thorium fuel cycle, this concept is very attractive, since the fuel can be re-processed on-line, enabling it to be an efficient neutron breeder. (author)

  1. Alternative Fuels Data Center: Biodiesel

    Science.gov (United States)

    Biodiesel Printable Version Share this resource Send a link to Alternative Fuels Data Center : Biodiesel to someone by E-mail Share Alternative Fuels Data Center: Biodiesel on Facebook Tweet about Alternative Fuels Data Center: Biodiesel on Twitter Bookmark Alternative Fuels Data Center: Biodiesel on

  2. Experiments and Modeling in Support of Generic Salt Repository Science

    International Nuclear Information System (INIS)

    Bourret, Suzanne Michelle; Stauffer, Philip H.; Weaver, Douglas James; Caporuscio, Florie Andre; Otto, Shawn; Boukhalfa, Hakim; Jordan, Amy B.; Chu, Shaoping; Zyvoloski, George Anthony; Johnson, Peter Jacob

    2017-01-01

    Salt is an attractive material for the disposition of heat generating nuclear waste (HGNW) because of its self-sealing, viscoplastic, and reconsolidation properties (Hansen and Leigh, 2012). The rate at which salt consolidates and the properties of the consolidated salt depend on the composition of the salt, including its content in accessory minerals and moisture, and the temperature under which consolidation occurs. Physicochemical processes, such as mineral hydration/dehydration salt dissolution and precipitation play a significant role in defining the rate of salt structure changes. Understanding the behavior of these complex processes is paramount when considering safe design for disposal of heat-generating nuclear waste (HGNW) in salt formations, so experimentation and modeling is underway to characterize these processes. This report presents experiments and simulations in support of the DOE-NE Used Fuel Disposition Campaign (UFDC) for development of drift-scale, in-situ field testing of HGNW in salt formations.

  3. Experiments and Modeling in Support of Generic Salt Repository Science

    Energy Technology Data Exchange (ETDEWEB)

    Bourret, Suzanne Michelle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stauffer, Philip H. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Weaver, Douglas James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Caporuscio, Florie Andre [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Otto, Shawn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Boukhalfa, Hakim [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Jordan, Amy B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chu, Shaoping [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zyvoloski, George Anthony [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Johnson, Peter Jacob [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-19

    Salt is an attractive material for the disposition of heat generating nuclear waste (HGNW) because of its self-sealing, viscoplastic, and reconsolidation properties (Hansen and Leigh, 2012). The rate at which salt consolidates and the properties of the consolidated salt depend on the composition of the salt, including its content in accessory minerals and moisture, and the temperature under which consolidation occurs. Physicochemical processes, such as mineral hydration/dehydration salt dissolution and precipitation play a significant role in defining the rate of salt structure changes. Understanding the behavior of these complex processes is paramount when considering safe design for disposal of heat-generating nuclear waste (HGNW) in salt formations, so experimentation and modeling is underway to characterize these processes. This report presents experiments and simulations in support of the DOE-NE Used Fuel Disposition Campaign (UFDC) for development of drift-scale, in-situ field testing of HGNW in salt formations.

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

    International Nuclear Information System (INIS)

    Yoshioka, Ritsuo; Mitachi, Koshi

    2013-01-01

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

  5. Electrolytic reduction of a simulated oxide spent fuel and the fates of representative elements in a Li{sub 2}O-LiCl molten salt

    Energy Technology Data Exchange (ETDEWEB)

    Park, Wooshin, E-mail: wooshin@kaeri.re.kr; Choi, Eun-Young; Kim, Sung-Wook; Jeon, Sang-Chae; Cho, Young-Hwan; Hur, Jin-Mok

    2016-08-15

    A series of electrolytic reduction experiments were carried out using a simulated oxide spent fuel to investigate the reduction behavior of elements in a mixed oxide condition and the fates of elements in the reduction process with 1.0 wt% Li{sub 2}O-LiCl. It was found out that 155% of the theoretical charge was enough to reduce the simulated. Te and Eu were expected to possibly exist in the precipitate and on the anode surface, whereas Ba and Sr showed apparent dissolution behaviors. Rare earths showed relatively low metal fractions from 28.2 to 34.0% except for Y. And the solubility of rare earths was observed to be low due to the low concentration of Li{sub 2}O. The reduction of U was successful as expected showing 99.8% of a metal fraction. Also it was shown that the reduction of ZrO{sub 2} would be effective when a relatively small amount was included in a metal oxide mixture.

  6. Supplemental technical information in support of Y/OWI/TM--44. Volume 9. Drawings for repository preconceptual design studies: salt

    International Nuclear Information System (INIS)

    1978-04-01

    Volume 9 contains drawings for a preconceptual design for a nuclear waste storage facility in salt. Drawings are included for three fuel cycles: fuel recycle, throwaway fuel cycle, and uranium recycle with plutonium in the high-level waste

  7. Investigation of molten salt fast reactor

    International Nuclear Information System (INIS)

    Kubota, Kenichi; Konomura, Mamoru

    2002-01-01

    On survey research for practicability strategy of fast reactor (FR) (phase 1), to extract future practicability image candidates of FR from wide options, in addition to their survey and investigation objects of not only solid fuel reactors of conventional research object but also molten salt reactor as a flowing fuel reactor, investigation on concept of molten salt FR plant was carried out. As a part of the first step of the survey research for practicability strategy, a basic concept on plant centered at nuclear reactor facility using chloride molten salt reactor capable of carrying out U-Pu cycle was examined, to perform a base construction to evaluate economical potential for a practical FBR. As a result, a result could be obtained that because of inferior fuel inventory and heat transmission to those in Na cooling reactor in present knowledge, mass of reactor vessel and intermediate heat exchanger were to widely increased to expect reduction of power generation unit price even on considering cheapness of its fuel cycle cost. Therefore, at present step further investigation on concept design of the chloride molten salt reactor plant system is too early in time, and it is at a condition where basic and elementary researches aiming at upgrading of economical efficiency such as wide reduction of fuel inventory, a measure expectable for remarkable rationalization effect of reprocessing system integrating a reactor to a processing facility, and so on. (G.K.)

  8. Research and development of thorium fuel cycle

    International Nuclear Information System (INIS)

    Oishi, Jun.

    1994-01-01

    Nuclear properties of thorium are summarized and present status of research and development of the use of thorium as nuclear fuel is reviewed. Thorium may be used for nuclear fuel in forms of metal, oxide, carbide and nitride independently, alloy with uranium or plutonium or mixture of the compound. Their use in reactors is described. The reprocessing of the spent oxide fuel in thorium fuel cycle is called the thorex process and similar to the purex process. A concept of a molten salt fuel reactor and chemical processing of the molten salt fuel are explained. The required future research on thorium fuel cycle is commented briefly. (T.H.)

  9. Nuclear criticality safety handbook. Version 2

    International Nuclear Information System (INIS)

    1999-03-01

    The Nuclear Criticality Safety Handbook, Version 2 essentially includes the description of the Supplement Report to the Nuclear Criticality Safety Handbook, released in 1995, into the first version of Nuclear Criticality Safety Handbook, published in 1988. The following two points are new: (1) exemplifying safety margins related to modelled dissolution and extraction processes, (2) describing evaluation methods and alarm system for criticality accidents. Revision is made based on previous studies for the chapter that treats modelling the fuel system: e.g., the fuel grain size that the system can be regarded as homogeneous, non-uniformity effect of fuel solution, and burnup credit. This revision solves the inconsistencies found in the first version between the evaluation of errors found in JACS code system and criticality condition data that were calculated based on the evaluation. (author)

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

    Science.gov (United States)

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

    2018-01-01

    A little-known type of nuclear reactor called the "molten salt reactor" (MSR), in which nuclear fuel is dissolved in a liquid carrier salt, was proposed in the 1940s and developed at the Oak Ridge National Laboratory in the 1960s. Recently, the MSR has generated renewed interest as a remedy for the drawbacks associated with conventional…

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

    Energy Technology Data Exchange (ETDEWEB)

    Boussier, H.; Heuer, D.

    2010-07-01

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

  12. Comparison of disposal concepts for rock salt and hard rock

    International Nuclear Information System (INIS)

    Papp, R.

    1998-01-01

    The study was carried out in the period 1994-1996. The goals were to prepare a draft on spent fuel disposal in hard rock and additionally a comparison with existing disposal concepts for rock salt. A cask for direct disposal of spent fuel and a repository for hard rock including a safeguards concept were conceptually designed. The results of the study confirm, that the early German decision to employ rock salt was reasonable. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Sooby, E., E-mail: soobyes@tamu.edu [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States); Baty, A. [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States); Beneš, O. [European Commission, DG Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); McIntyre, P.; Pogue, N. [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States); Salanne, M. [Université Pierre et Marie Curie, CNRS, Laboratoire PECSA, F-75005 Paris (France); Sattarov, A. [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States)

    2013-09-15

    Highlights: • Developing accelerator driven subcritical fission to destroy transuranics in SNF. • The core is a vessel containing a molten mixture of NaCl and transuranic chlorides. • Molecular dynamics used to calculate the thermophysical properties of the salt. • Density and molecular structure for actinide salts reported here. • The neutronics of ADS fission in molten salt are presented. -- Abstract: We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated.

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

    International Nuclear Information System (INIS)

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

    2013-01-01

    Highlights: • Developing accelerator driven subcritical fission to destroy transuranics in SNF. • The core is a vessel containing a molten mixture of NaCl and transuranic chlorides. • Molecular dynamics used to calculate the thermophysical properties of the salt. • Density and molecular structure for actinide salts reported here. • The neutronics of ADS fission in molten salt are presented. -- Abstract: We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated

  15. SHUFFLE. Windows 95/98/2000 version

    International Nuclear Information System (INIS)

    Slavic, S.; Zefran, B.

    2000-01-01

    Program package SHUFFLE was developed to help the user during fuel loading and unloading operations at a nuclear power plant. The first version, developed in 1992, has been written in the CLIPPER program language and run under the DOS operating system. Since the DOS environment exhibits several drawbacks regarding code portability and flexibility, the recent SHUFFLE version has been transformed to run under the MS Windows operating system. (author)

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

    International Nuclear Information System (INIS)

    Bennett, M.R.; Bamberge, C.E.; Kelmers, A.D.

    1980-01-01

    The subject relates to fuel preparation for molten salt breeder reactors, and more particularly to the reconstitution of spent molten fuel salt after fission product removal. During the course of reactor operation, fission products including rare earths and bred-in protactinium build up in the fuel salt and adversely affect the nuclear properties of the fuel. In order to more efficiently operate the reactor, the level of neutron poison fission products must be kept at a minimum. This is accomplished by continuously removing spent fuel from the primary circuit, processing it to remove fission products, and returning the reprocessed molten salt to the primary circuit. It is desirable for safety and economy that the fuel processing plant be a component of the reactor itself and that the salt be kept in the molten state throughout the processing system. (auth)

  17. Preliminary design studies of the draining tanks for the Molten Salt Fast Reactor

    International Nuclear Information System (INIS)

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

    2014-01-01

    reactor called the Molten Salt Fast Reactor (MSFR). The reference MSFR design is a 3000 MWth reactor with a total fuel salt volume of 18 m3, operated at a mean fuel temperature of 750 deg. C. The first confinement barrier of the reactor includes a salt draining system. In case of a planned reactor shut down or in case of accidents leading to an excessive increase of the temperature in the fuel circuit, the fuel configuration may be changed passively by gravitational draining of the fuel salt in dedicated draining tank located under the reactor and designed to provide adequate reactivity margins while insuring a passive cooling of the fuel salt to extract the residual heat from the short to the long term. The present preliminary assessment of this sub-critical draining system has been performed to identify the physical constraints and to give some orders of magnitude of characteristic time periods (authors)

  18. [Historical roles of salt].

    Science.gov (United States)

    Ritz, E; Ritz, C

    2004-12-17

    Recently increasing evidence has been provided pointing to a close relation of salt consumption to hypertension as well as to target organ damage. It is interesting to note that the discussion concerning salt is unusually emotional. This may be explained, at least in part, by the fact that since ancient times salt had deep symbolic significance, as exemplified, mostly subconsciously, by many customs and expressions still in current use. In the past salt was essential to preserve food. The past importance of salt as a commodity can well be compared with that of oil today. These and further historical aspects of the role of salt are briefly dealt with in this article.

  19. Salt Separation from Uranium Deposits in Integrated Crucible

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, S. W.; Park, K. M.; Chang, J. H.; Kim, J. G.; Park, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. A physical separation process, such as distillation separation, is more attractive than a chemical or dissolution process because physical processes generate much less secondary process. Distillation process was employed for the cathode processsing due to the advantages of minimal generation of secondary waste, compact unit process, simple and low cost equipment. The basis for vacuum distillation separation is the difference in vapor pressures between salt and uranium. A solid cathode deposit is heated in a heating region and salt vaporizes, while non-volatile uranium remains behind. It is very important to increase the throughput of the salt separation system due to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. The evaporation rate of the LiCl-KCl eutectic salt in vacuum distiller is not so high to come up with the generation capacity of uranium dendrites in electro-refiner. Therefore, wide evaporation area or high distillation temperature is necessary for the successful salt separation. The adhered salt in the uranium deposits was removed successfully. The salt content in the deposits was below 0.1 wt% after the sequential operation of the liquid salt separation - salt distillation.

  20. Salt Selection for the LS-VHTR

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  1. Sequestration of CO2 in salt caverns

    International Nuclear Information System (INIS)

    Dusseault, M.B.; Rothenburg, L.; Bachu, S.

    2002-01-01

    The greenhouse effect is thought to be greatly affected by anthropogenic and naturally generated gases, such as carbon dioxide. The reduction of greenhouse gas emissions in the atmosphere could be effected through the permanent storage of carbon dioxide in dissolved salt caverns. A large number of suitable salt deposits are located in Alberta, especially the Lotsberg Salt of east-central Alberta. A major advantage of this deposit is its proximity to present and future point sources of carbon dioxide associated with fossil fuel development projects. Using the perspective of the long term fate of the stored carbon dioxide, the authors presented the characteristics of the Lotsberg Salt and the overlying strata. A high level of security against leakage and migration of the gas back to the biosphere is ensured by several features discussed in the paper. The authors propose a procedure that would be applicable for the creation, testing, and filling of a salt cavern. Achieving a long term prediction of the behavior of the cavern during slow closure, coupled to the pressure and volume behavior of the gas within the cavern represents the critical factor. The authors came up with an acceptable prediction by using a semi-analytical model. The use of salt caverns for the permanent sequestration of carbon dioxide has not yet faced technical obstacles that would prevent it. The authors argue that sequestration of carbon dioxide in salt caverns represents an environmentally acceptable option in Alberta. 11 refs., 3 figs

  2. Sampling device for radioactive molten salt

    International Nuclear Information System (INIS)

    Shindo, Masato

    1998-01-01

    The present invention provides a device for accurately sampling molten salts to which various kinds of metals in a molten salt storage tank are mixed for analyzing them during a spent fuel dry type reprocessing. Namely, the device comprises a sampling tube having an opened lower end to be inserted into the radioactive molten salts stored in a tank and keeps reduced pressure from the upper end, and a pressure reducing pipeline having one end connected to the sampling tube and other end connected to an evacuating pump. In this device, the top end of the sampling tube is inserted to a position for sampling the radioactive molten salts (molten salts). The pressure inside the evacuating pipeline connected to the upper portion of the sampling tube is reduced for a while. In this case, the inside of the pressure reducing pipeline is previously evacuated by the evacuating pump so as to keep a predetermined pressure. Since the pressure in the sampling tube is lowered, molten salts are inserted into the sampling tube, the sampling tube is withdrawn, and the molten salts flown in the sampling tube are analyzed. (I.S.)

  3. Sea Salt vs. Table Salt: What's the Difference?

    Science.gov (United States)

    ... and healthy eating What's the difference between sea salt and table salt? Answers from Katherine Zeratsky, R.D., L.D. The main differences between sea salt and table salt are in their taste, texture ...

  4. Low-salt diet

    Science.gov (United States)

    Low-sodium diet; Salt restriction ... control many functions. Too much sodium in your diet can be bad for you. For most people, ... you limit salt. Try to eat a balanced diet. Buy fresh vegetables and fruits whenever possible. They ...

  5. Molten salt/metal extractions for recovery of transuranic elements

    International Nuclear Information System (INIS)

    Chow, L.S.; Basco, J.K.; Ackerman, J.P.; Johnson, T.R.

    1992-01-01

    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

  6. Thermochemical investigation of molten fluoride salts for Generation IV nuclear applications - an equilibrium exercise

    NARCIS (Netherlands)

    van der Meer, J.P.M.

    2006-01-01

    The concept of the Molten Salt Reactor, one of the so-called Generation IV future reactors, is that the fuel, a fissile material, which is dissolved in a molten fluoride salt, circulates through a closed circuit. The heat of fission is transferred to a second molten salt coolant loop, the heat of

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

    International Nuclear Information System (INIS)

    1983-03-01

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

  8. Engineering development studies for molten-salt breeder reactor processing No. 21

    International Nuclear Information System (INIS)

    Hightower, J.R. Jr.

    1976-03-01

    The status of the following programs is reported: (1) continuous fluorinator development: autoresistance heating test AHT-4; (2) development of the metal transfer process; (3) salt-metal contactor development: experiments with a mechanically agitated, nondispersing contactor using water and mercury and in the salt-bismuth flowthrough facility; and (4) fuel reconstitution development: installation of equipment for a fuel reconstitution engineering experiment

  9. Commercialization of fuel-cells

    Energy Technology Data Exchange (ETDEWEB)

    Penner, S.S.; Appleby, A.J.; Baker, B.S.; Bates, J.L.; Buss, L.B.; Dollard, W.J.; Farris, P.J.; Gillis, E.A.; Gunsher, J.A.; Khandkar, A.; Krumpelt, M.; O' Sullivan, J.B.; Runte, G.; Savinell, R.F.; Selman, J.R.; Shores, D.A.; Tarman, P.

    1995-03-01

    This report is an abbreviated version of the ''Report of the DOE Advanced Fuel Cell Commercialization Working Group (AFC2WG),'' released January 1995. We describe fuel-cell commercialization for stationary power applications of phosphoric acid, molten carbonate, solid oxide, and polymer electrolyte membrane fuel cells.

  10. Fission product removal from molten salt using zeolite

    International Nuclear Information System (INIS)

    Pereira, C.; Babcock, B.D.

    1996-01-01

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

  11. Advances in molten salt electrochemistry towards future energy systems

    International Nuclear Information System (INIS)

    Ito, Yasuhiko

    2005-01-01

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

  12. A radioactive tracer dilution method to determine the mass of molten salt

    International Nuclear Information System (INIS)

    Lei Cao; Jarrell, Josh; Hardtmayer, D.E.; White, Susan; Herminghuysen, Kevin; Kauffman, Andrew; Sanders, Jeff; Li, Shelly

    2017-01-01

    A new technique for molten salt mass determination, termed radioactive tracer dilution, that uses 22 Na as a tracer was validated at bench scale. It has been a challenging problem to determine the mass of molten salt in irregularly shaped containers, where a highly radioactive, high-temperature molten salt was used to process nuclear spent/used fuel during electrochemical recycling (pyro-processing) or for coolant/fuel salt from molten salt reactors. A radioactive source with known activity is dissolved into the salt. After a complete mixture, a small amount of the salt is sampled and measured in terms of its mass and radioactivity. By finding the ratio of the mass to radioactivity, the unknown salt mass in the original container can be precisely determined. (author)

  13. Molten salt small modular reactors (MSSMRs): from DMSR to SmAHTR

    International Nuclear Information System (INIS)

    LeBlanc, D.

    2013-01-01

    Molten salt reactors were developed extensively from the 1950s to 1970s as a thermal breeder alternative on the Thorium-U233 cycle. Simplified designs running as fluid fuel convertors without salt processing as well as TRISO fueled, salt cooled reactors both hold much promise as potential small modular reactors. A background will be presented along with the most likely routes forward for a Canadian development program. (author)

  14. The introduction of the safety of molten salt reactor

    International Nuclear Information System (INIS)

    Zuo Jiaxu; Zhang Chunming

    2011-01-01

    This paper introduces the generation TV Nuclear Energy Systems and molten salt reactor which is the only fluid fuel reactor in the Gen-TV. Safety features and attributes of MSR are described. The supply of fuel and the minimum of waste are described. The clean molten salt in the secondary heat transport system transfers the heat from the primary heat exchanger to a high-temperature Brayton cycle that converts the heat to electricity. With the Brayton cycle, the thermal efficiency of the system will be improved. Base on the MSR, the thorium-uranium fuel cycle is also introduced. (authors)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  16. Electrometallurgical treatment of aluminum-matrix fuels

    International Nuclear Information System (INIS)

    Willit, J.L.; Gay, E.C.; Miller, W.E.; McPheeters, C.C.; Laidler, J.J.

    1996-01-01

    The electrometallurgical treatment process described in this paper builds on our experience in treating spent fuel from the Experimental Breeder Reactor (EBR-II). The work is also to some degree, a spin-off from applying electrometallurgical treatment to spent fuel from the Hanford single pass reactors (SPRs) and fuel and flush salt from the Molten Salt Reactor Experiment (MSRE) in treating EBR-II fuel, we recover the actinides from a uranium-zirconium fuel by electrorefining the uranium out of the chopped fuel. With SPR fuel, uranium is electrorefined out of the aluminum cladding. Both of these processes are conducted in a LiCl-KCl molten-salt electrolyte. In the case of the MSRE, which used a fluoride salt-based fuel, uranium in this salt is recovered through a series of electrochemical reductions. Recovering high-purity uranium from an aluminum-matrix fuel is more challenging than treating SPR or EBR-II fuel because the aluminum- matrix fuel is typically -90% (volume basis) aluminum

  17. Conceptual design of Indian molten salt breeder reactor

    International Nuclear Information System (INIS)

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

    2014-01-01

    The fuel in a molten salt breeder reactor is in the form of a continuously circulating molten salt. Fluoride based salts have been almost universally proposed. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. This constitutes a major technological challenge for this type of reactors. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian Molten Salt Breeder Reactor (IMSBR). Presently various design options and possibilities are being studied from the point of view of reactor physics and thermal hydraulic design. In parallel fundamental studies as regards various molten salts have also been initiated. This paper would discuss conceptual design of these reactors, as well as associated issues and technologies

  18. Thermal Characterization of Molten Salt Systems

    Energy Technology Data Exchange (ETDEWEB)

    Toni Y. Gutknecht; Guy L. Fredrickson

    2011-09-01

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

  19. Comparison of the intermediate storage periods and areas required for final storage of high-level radioactive waste and spent fuel in various types of host rock

    International Nuclear Information System (INIS)

    Mueller-Hoeppe, N.; Lerch, C.; Jobmann, M.; Filbert, W.

    2005-01-01

    The present new version of the German concept for radioactive waste and spent fuel management is based on the assumption that a repository for high-level waste and spent fuel will not be required until 2030. One reason frequently given for this date is the intermediate storage period of at least forty years to allow the very high initial heat generation to decay. However, calculations performed by the authors have shown that the minimum intermediate storage period for a repository in rock salt is only between four and nineteen years, depending on the final storage concept and the load of the waste package. In clay as a host rock, the minimum intermediate storage times were calculated to be between 31 and 142 years; the same time spans are expected to apply to final storage in magmatic rock, such as granite. The maximum permissible loads of a container holding spent fuel in salt are many times those in clay and granite, respectively. It was also seen that the area requirement for final storage of the same waste structures is roughly a factor of ten higher in clay than in salt. The differences between granite and salt are similar. The reasons for these grave differences, on the one hand, are the better thermal conductivity of salt and, on the other hand, the better heat tolerance of the crushed salt used as backfill material compared to that of bentonite used in the clay and granite concepts. While salt will allow temperatures of up to 200 C, the maximum temperature in bentonite is limited to 100 C. (orig.)

  20. Development of an integrated crucible for the salt separation

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, S. W.; Park, K. M.; Jeong, J. H.; Lee, H. S.; Kim, J. G. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Pyroprocessing has been developed for the recovery of actinide elements from spent fuel due to its advantages. Electrorefining is a key step in pyroprocessing. The electrorefining process is generally composed of two recovery steps. The deposit of uranium onto a solid cathode and the recovery of the remaining uranium and TRU elements simultaneously by a liquid cadmium cathode. The solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. A physical separation process, such as distillation separation, is more attractive than a chemical or dissolution process because physical processes generate much less secondary process. Distillation process was employed for the cathode process sing due to the advantages of minimal generation of secondary waste, compact unit process, simple and low cost equipment. The basis for vacuum distillation separation is the difference in vapor pressures between salt and uranium. A solid cathode deposit is heated in a heating region and salt vaporizes, while non volatile uranium remains behind. It is very important to increase the throughput of the salt separation system due to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. The evaporation rate of the LiCl-KCl eutectic salt in vacuum distiller is not so high to come up with the generation capacity of uranium dendrites in electro-refiner. Therefore, wide evaporation area or high distillation temperature is necessary for the successful salt separation. In this study, the integrated salt separation system was developed to increase the throughput of the salt removal process by the separation of the liquid salt prior to the distillation of the LiCl-KCl eutectic salt from the uranium deposits

  1. Combined system of accelerator molten-salt breeder (AMSB) apd molten-salt converter reactor (MSCR)

    International Nuclear Information System (INIS)

    Furukawa, K.; Kato, Y.; Ohmichi, T.; Ohno, H.

    1983-01-01

    A design and research program is discUssed of the development of accelerator molten-salt breeder (AMSB) consisting of a proton accelerator and a molten fluoride target. The target simultaneously serves as a blanket for fissionable material prodUction. An addition of some amoUnt of fissile nuclides to a melt expands the AMSB potentialities as the fissionable material production increases and the energy generation also grows up to the level of self-provision. Besides the blanket salts may be used as nuclear fuel for molten-salt converter reactor (MSCR). The combined AM SB+MSCR system has better parameters as compared to other breeder reactors, molten-salt breeder reactors (MSBR) included

  2. Alternative Fuels Data Center: Natural Gas

    Science.gov (United States)

    Natural Gas Printable Version Share this resource Send a link to Alternative Fuels Data Center : Natural Gas to someone by E-mail Share Alternative Fuels Data Center: Natural Gas on Facebook Tweet about Alternative Fuels Data Center: Natural Gas on Twitter Bookmark Alternative Fuels Data Center: Natural Gas on

  3. Submarine Salt Karst Terrains

    Directory of Open Access Journals (Sweden)

    Nico Augustin

    2016-06-01

    Full Text Available Karst terrains that develop in bodies of rock salt (taken as mainly of halite, NaCl are special not only for developing in one of the most soluble of all rocks, but also for developing in one of the weakest rocks. Salt is so weak that many surface-piercing salt diapirs extrude slow fountains of salt that that gravity spread downslope over deserts on land and over sea floors. Salt fountains in the deserts of Iran are usually so dry that they flow at only a few cm/yr but the few rain storms a decade so soak and weaken them that they surge at dm/day for a few days. We illustrate the only case where the rates at which different parts of one of the many tens of subaerial salt karst terrains in Iran flows downslope constrains the rates at which its subaerial salt karst terrains form. Normal seawater is only 10% saturated in NaCl. It should therefore be sufficiently aggressive to erode karst terrains into exposures of salt on the thousands of known submarine salt extrusions that have flowed or are still flowing over the floors of hundreds of submarine basins worldwide. However, we know of no attempt to constrain the processes that form submarine salt karst terrains on any of these of submarine salt extrusions. As on land, many potential submarine karst terrains are cloaked by clastic and pelagic sediments that are often hundreds of m thick. Nevertheless, detailed geophysical and bathymetric surveys have already mapped likely submarine salt karst terrains in at least the Gulf of Mexico, and the Red Sea. New images of these two areas are offered as clear evidence of submarine salt dissolution due to sinking or rising aggressive fluids. We suggest that repeated 3D surveys of distinctive features (± fixed seismic reflectors of such terrains could measure any downslope salt flow and thus offer an exceptional opportunity to constrain the rates at which submarine salt karst terrains develop. Such rates are of interest to all salt tectonicians and the many

  4. Fusion-Fission Hybrid for Fissile Fuel Production without Processing

    Energy Technology Data Exchange (ETDEWEB)

    Fratoni, M; Moir, R W; Kramer, K J; Latkowski, J F; Meier, W R; Powers, J J

    2012-01-02

    Two scenarios are typically envisioned for thorium fuel cycles: 'open' cycles based on irradiation of {sup 232}Th and fission of {sup 233}U in situ without reprocessing or 'closed' cycles based on irradiation of {sup 232}Th followed by reprocessing, and recycling of {sup 233}U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile material in a fusion-fission hybrid reactor and burning the same fuel in a critical reactor without any reprocessing or reconditioning. This fuel cycle requires the hybrid and the critical reactor to use the same fuel form. TRISO particles embedded in carbon pebbles were selected as the preferred form of fuel and an inertial laser fusion system featuring a subcritical blanket was combined with critical pebble bed reactors, either gas-cooled or liquid-salt-cooled. The hybrid reactor was modeled based on the earlier, hybrid version of the LLNL Laser Inertial Fusion Energy (LIFE1) system, whereas the critical reactors were modeled according to the Pebble Bed Modular Reactor (PBMR) and the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) design. An extensive neutronic analysis was carried out for both the hybrid and the fission reactors in order to track the fuel composition at each stage of the fuel cycle and ultimately determine the plant support ratio, which has been defined as the ratio between the thermal power generated in fission reactors and the fusion power required to breed the fissile fuel burnt in these fission reactors. It was found that the maximum attainable plant support ratio for a thorium fuel cycle that employs neither enrichment nor reprocessing is about 2. This requires tuning the neutron energy towards high energy for breeding and towards thermal energy for burning. A high fuel loading in the pebbles allows a faster spectrum in the hybrid blanket; mixing dummy carbon pebbles with fuel pebbles enables a softer spectrum in

  5. Spent fuel reprocessing method

    International Nuclear Information System (INIS)

    Shoji, Hirokazu; Mizuguchi, Koji; Kobayashi, Tsuguyuki.

    1996-01-01

    Spent oxide fuels containing oxides of uranium and transuranium elements are dismantled and sheared, then oxide fuels are reduced into metals of uranium and transuranium elements in a molten salt with or without mechanical removal of coatings. The reduced metals of uranium and transuranium elements and the molten salts are subjected to phase separation. From the metals of uranium and transuranium elements subjected to phase separation, uranium is separated to a solid cathode and transuranium elements are separated to a cadmium cathode by an electrolytic method. Molten salts deposited together with uranium to the solid cathode, and uranium and transuranium elements deposited to the cadmium cathode are distilled to remove deposited molten salts and cadmium. As a result, TRU oxides (solid) such as UO 2 , Pu 2 in spent fuels can be reduced to U and TRU by a high temperature metallurgical method not using an aqueous solution to separate them in the form of metal from other ingredients, and further, metal fuels can be obtained through an injection molding step depending on the purpose. (N.H.)

  6. Projected Salt Waste Production from a Commercial Pyroprocessing Facility

    Directory of Open Access Journals (Sweden)

    Michael F. Simpson

    2013-01-01

    Full Text Available Pyroprocessing of used nuclear fuel inevitably produces salt waste from electrorefining and/or oxide reduction unit operations. Various process design characteristics can affect the actual mass of such waste produced. This paper examines both oxide and metal fuel treatment, estimates the amount of salt waste generated, and assesses potential benefit of process options to mitigate the generation of salt waste. For reference purposes, a facility is considered in which 100 MT/year of fuel is processed. Salt waste estimates range from 8 to 20 MT/year from considering numerous scenarios. It appears that some benefit may be derived from advanced processes for separating fission products from molten salt waste, but the degree of improvement is limited. Waste form production is also considered but appears to be economically unfavorable. Direct disposal of salt into a salt basin type repository is found to be the most promising with respect to minimizing the impact of waste generation on the economic feasibility and sustainability of pyroprocessing.

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

    Science.gov (United States)

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

    2013-09-01

    We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated. A special thanks is due to Prof. Paul Madden for introducing the ADSMS group to the concept of using the molten salt as the spallation target, rather than a conventional heavy metal spallation target. This feature helps to optimize this core as a Pu/TRU burner.

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

    International Nuclear Information System (INIS)

    Lainetti, Paulo E.O.

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Ignatiev, Victor; Feynberg, Olga; Merzlyakov, Aleksandr; Surenkov, Aleksandr; Subbotin, Vladimir; Zakirov, Raul; Toropov, Andrey; Panov, Aleksandr; Afonichkin, Valery

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  11. Hopper Growth of Salt Crystals.

    Science.gov (United States)

    Desarnaud, Julie; Derluyn, Hannelore; Carmeliet, Jan; Bonn, Daniel; Shahidzadeh, Noushine

    2018-06-07

    The growth of hopper crystals is observed for many substances, but the mechanism of their formation remains ill understood. Here we investigate their growth by performing evaporation experiments on small volumes of salt solutions. We show that sodium chloride crystals that grow very fast from a highly supersaturated solution form a peculiar form of hopper crystal consisting of a series of connected miniature versions of the original cubic crystal. The transition between cubic and such hopper growth happens at a well-defined supersaturation where the growth rate of the cubic crystal reaches a maximum (∼6.5 ± 1.8 μm/s). Above this threshold, the growth rate varies as the third power of supersaturation, showing that a new mechanism, controlled by the maximum speed of surface integration of new molecules, induces the hopper growth of cubic crystals in cascade.

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

    International Nuclear Information System (INIS)

    He, Xun

    2016-01-01

    MSR concept using the mathematic tools. In particular, the aim of the first part is to demonstrate the suitability of the TRACE code for the similar MSR designs by using a modified version of the TRACE code to implement the simulations for the steady-state, transient and accidental conditions. The basic approach of this part is to couple the thermal-hydraulic model and the modified point-kinetic model. The equivalent thermal-hydraulic model of the MSRE was built in 1D with three loops including all the critical main components. The point-kinetic model was improved through considering the precursor drift in order to produce more practical results in terms of the delayed neutron behavior. Additionally, new working fluids, namely the molten salts, were embedded into the source code of TRACE. Most results of the simulations show good agreements with the ORNL's reports and with another recent study and the errors were predictable and in an acceptable range. Therefore, the necessary code modification of TRACE appears to be successful and the model will be refined and its functions will be extended further in order to investigate new MSR design. Another part of this thesis is to implement a preliminary study on a new concept of molten salt reactor, namely the Dual Fluid Reactor (DFR). The DFR belongs to the group of the molten salt fast reactors (MSFR) and it is recently considered to be an option of minimum-waste and inherently safe operation of the nuclear reactors in the future. The DFR is using two separately circulating fluids in the reactor core. One is the fuel salt based on the mixture of tri-chlorides of uranium and plutonium (UCl_3-PuCl_3), while another is the coolant composed of the pure lead (Pb). The current work focuses on the basic dynamic behavior of a scaled-down DFR with 500 MW thermal output (DFR-500) instead of its reference design with 3000 MW thermal output (DFR-3000). For this purpose 10 parallel single fuel channels, as the representative samples

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

    Energy Technology Data Exchange (ETDEWEB)

    He, Xun

    2016-06-14

    one is about the demonstration of a new MSR concept using the mathematic tools. In particular, the aim of the first part is to demonstrate the suitability of the TRACE code for the similar MSR designs by using a modified version of the TRACE code to implement the simulations for the steady-state, transient and accidental conditions. The basic approach of this part is to couple the thermal-hydraulic model and the modified point-kinetic model. The equivalent thermal-hydraulic model of the MSRE was built in 1D with three loops including all the critical main components. The point-kinetic model was improved through considering the precursor drift in order to produce more practical results in terms of the delayed neutron behavior. Additionally, new working fluids, namely the molten salts, were embedded into the source code of TRACE. Most results of the simulations show good agreements with the ORNL's reports and with another recent study and the errors were predictable and in an acceptable range. Therefore, the necessary code modification of TRACE appears to be successful and the model will be refined and its functions will be extended further in order to investigate new MSR design. Another part of this thesis is to implement a preliminary study on a new concept of molten salt reactor, namely the Dual Fluid Reactor (DFR). The DFR belongs to the group of the molten salt fast reactors (MSFR) and it is recently considered to be an option of minimum-waste and inherently safe operation of the nuclear reactors in the future. The DFR is using two separately circulating fluids in the reactor core. One is the fuel salt based on the mixture of tri-chlorides of uranium and plutonium (UCl{sub 3}-PuCl{sub 3}), while another is the coolant composed of the pure lead (Pb). The current work focuses on the basic dynamic behavior of a scaled-down DFR with 500 MW thermal output (DFR-500) instead of its reference design with 3000 MW thermal output (DFR-3000). For this purpose 10 parallel

  14. Reactor chemical considerations of the accelerator molten-salt breeders

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Kato, Yoshio; Ohno, Hideo; Ohmichi, Toshihiko

    1982-01-01

    A single phase of the molten fluoride mixture is simultaneously functionable as a nuclear reaction medium, a heat medium and a chemical processing medium. Applying this characteristics of molten salts, the single-fluid type accelerator molten-salt breeder (AMSB) concept was proposed, in which 7 LiF-BeF 2 -ThF 4 was served as a target-and-blanket salt (Fig. 1 and Table 1), and the detailed discussion on the chemical aspects of AMSB are presented (Tables 2 -- 4 and Fig.2). Owing to the small total amount of radiowaste and the low concentrations of each element in target salt, AMSB would be chemically managable. The performance of the standard-type AMSB is improved by adding 0.3 -- 0.8 m/o 233 UF 4 as follows(Tables 1 and 4, and Figs. 2 and 3): (a) this ''high-gain'' type AMSB is feasible to design chemically, in which still only small amount of radiowaste is included ; (b) the fissile material production rate will be increased significantly; (c) this target salt is straightly fed as an 233 U additive to the fuel of molten-salt converter reactor (MSCR) ; (d) the dirty fuel salt suctioned from MSCR is batch-reprocessed in the safeguarded regional center, in which many AMSB are facilitated ; (e) the isolated 233 UF 4 is blended in the target salt sent to many MSCRs, and the cleaned residual fertile salt is used as a diluent of AMSB salt ; (f) this simple and rational thorium fuel breeding cycle system is also suitable for the nuclear nonproliferation and for the fabrication of smaller size power-stations. (author)

  15. Worth its salt?

    Science.gov (United States)

    The idea that all underground salt deposits can serve as storage sites for toxic and nuclear waste does not always hold water—literally. According to Daniel Ronen and Brian Berkowitz of Israel's Weizmann Institute of Science and Yoseph Yechieli of the Geological Survey of Israel, some buried salt layers are in fact highly conductive of liquids, suggesting that wastes buried in their confines could easily leech into groundwater and nearby soil.When drilling three wells into a 10,000-year-old salt layer near the Dead Sea, the researchers found that groundwater had seeped into the layer and had absorbed some of its salt.

  16. Hydrological methods preferentially recover cesium from nuclear waste salt cake

    International Nuclear Information System (INIS)

    Brooke, J.N.; Hamm, L.L.

    1997-01-01

    The Savannah River Site is treating high level radioactive waste in the form of insoluble solids (sludge), crystallized salt (salt cake), and salt solutions. High costs and operational concerns have prompted DOE to look for ways to improve the salt cake treatment process. A numerical model was developed to evaluate the feasibility of pump and treat technology for extracting cesium from salt cake. A modified version of the VAM3DCG code was used to first establish a steady-state flow field, then to simulate 30 days of operation. Simulation results suggest that efficient cesium extraction can be obtained with low displacement volumes. The actual extraction process will probably be less impressive because of nonuniform properties. 2 refs., 2 figs

  17. Carbon fuel particles used in direct carbon conversion fuel cells

    Science.gov (United States)

    Cooper, John F.; Cherepy, Nerine

    2012-10-09

    A system for preparing particulate carbon fuel and using the particulate carbon fuel in a fuel cell. Carbon particles are finely divided. The finely dividing carbon particles are introduced into the fuel cell. A gas containing oxygen is introduced into the fuel cell. The finely divided carbon particles are exposed to carbonate salts, or to molten NaOH or KOH or LiOH or mixtures of NaOH or KOH or LiOH, or to mixed hydroxides, or to alkali and alkaline earth nitrates.

  18. Carbon Fuel Particles Used in Direct Carbon Conversion Fuel Cells

    Science.gov (United States)

    Cooper, John F.; Cherepy, Nerine

    2008-10-21

    A system for preparing particulate carbon fuel and using the particulate carbon fuel in a fuel cell. Carbon particles are finely divided. The finely dividing carbon particles are introduced into the fuel cell. A gas containing oxygen is introduced into the fuel cell. The finely divided carbon particles are exposed to carbonate salts, or to molten NaOH or KOH or LiOH or mixtures of NaOH or KOH or LiOH, or to mixed hydroxides, or to alkali and alkaline earth nitrates.

  19. Residual Salt Separation from the Metal Products Reduced in a LiCl-Li2O Molten Salt

    International Nuclear Information System (INIS)

    Hur, Jin Mok; Hong, Sun Seok; Kang, Dae Seung; Jeong, Meong Soo; Seo, Chung Seok

    2006-02-01

    The electrochemical reduction of spent nuclear fuel in a LiCl-Li 2 O molten salt for the conditioning of spent nuclear fuel requires the separation of the residual salts from a reduced metal product after the reduction process. Considering the behavior of spent nuclear fuel during the electrochemical reduction process, a surrogate material matrix was constructed and inactive tests on a salt separation were carried out to produce the data required for the active tests. Fresh uranium metal prepared from the electrochemical reduction of U 3 O 8 powder was used as the surrogates of the spent nuclear fuel components which might be metallized by the electrochemical reduction process. LiCl, Li 2 O, Y 2 O 3 and SrCl 2 were selected as the components of the residual salts. Interactions between the salts and their influence on the separation of the residual salts were analyzed by differential scanning calorimetry (DSC) and thermogravimetry (TG). Eutectic melting of LiCl-Li 2 O and LiCl-SrCl 2 led to a melting point which was lower than that of a LiCl molten salt was observed. Residual salts were separated by a vaporization method. Co-vaporization of LiCl-Li 2 O and LiCl-SrCl 2 was achieved below temperatures which could make the uranium metal oxidation by Li 2 O possible. The salt vaporization rates at 950 .deg. C were measured as follows: LiCl-8 wt% Li 2 O > LiCl > LiCl-8 wt% SrCl 2 > SrCl 2

  20. Residual Salt Separation from the Metal Products Reduced in a LiCl-Li{sub 2}O Molten Salt

    Energy Technology Data Exchange (ETDEWEB)

    Hur, Jin Mok; Hong, Sun Seok; Kang, Dae Seung; Jeong, Meong Soo; Seo, Chung Seok

    2006-02-15

    The electrochemical reduction of spent nuclear fuel in a LiCl-Li{sub 2}O molten salt for the conditioning of spent nuclear fuel requires the separation of the residual salts from a reduced metal product after the reduction process. Considering the behavior of spent nuclear fuel during the electrochemical reduction process, a surrogate material matrix was constructed and inactive tests on a salt separation were carried out to produce the data required for the active tests. Fresh uranium metal prepared from the electrochemical reduction of U{sub 3}O{sub 8} powder was used as the surrogates of the spent nuclear fuel components which might be metallized by the electrochemical reduction process. LiCl, Li{sub 2}O, Y{sub 2}O{sub 3} and SrCl{sub 2} were selected as the components of the residual salts. Interactions between the salts and their influence on the separation of the residual salts were analyzed by differential scanning calorimetry (DSC) and thermogravimetry (TG). Eutectic melting of LiCl-Li{sub 2}O and LiCl-SrCl{sub 2} led to a melting point which was lower than that of a LiCl molten salt was observed. Residual salts were separated by a vaporization method. Co-vaporization of LiCl-Li{sub 2}O and LiCl-SrCl{sub 2} was achieved below temperatures which could make the uranium metal oxidation by Li{sub 2}O possible. The salt vaporization rates at 950 .deg. C were measured as follows: LiCl-8 wt% Li{sub 2}O > LiCl > LiCl-8 wt% SrCl{sub 2} > SrCl{sub 2}.

  1. Preliminary Study on the High Temperature Transport System for Molten Salt

    International Nuclear Information System (INIS)

    Lee, S. H.; Lee, H. S.; Kim, J. G.

    2012-01-01

    Pyroprocessing technology is one of the the most promising technologies for the advanced fuel cycle with favorable economic potential and intrinsic proliferation-resistance. The electrorefining process, one of main processes is compos- ed of pyroprocess to recover the useful elements from spent fuel, is under development at the Korea Atomic Energy Research Institute as a sub process of the pyrochemical treatment of spent PWR fuel. High-temperature molten salt transport technologies are required because a molten salt should be transported from the electrorefiner to electrowiner after the electrorefining process. Therefore, in pyroprocessing technology, the development of high-temperature transport technologies for molten salt is a crucial prerequisite. However, there have been a few transport studies on high-temperature molten salt. In this study, an apparatus for suction transport experiments was designed and constructed for the development of high temperature molten salt transport technology. Suction transport experiments were performed using LiC-KCl eutectic salt

  2. Molten salt reactors - safety options galore

    International Nuclear Information System (INIS)

    Gat, U.; Dodds, H.L.

    1997-01-01

    Safety features and attributes of molten salt reactors (MSR) are described. The unique features of fluid fuel reactors of on-line continuous processing and the ability for so-called external cooling result in simple and safe designs with low excess reactivity, low fission product inventory, and small source term. These, in turn, make a criticality accident unlikely and reduce the severity of a loss of coolant to where they are no longer severe accidents. A melt down is not an accident for a reactor that uses molten fuel. The molten salts are stable, non-reactive and efficient heat transfer media that operate at high temperatures at low pressures and are highly compatible with selected structural materials. All these features reduce the accident plethora. Freeze valves can be used for added safety. An ultimate safe reactor (U.S.R) is described with safety features that are passive, inherent and non-tamperable (PINT)

  3. Fast Spectrum Molten Salt Reactor Options

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  4. Molten salt reactor related research in Switzerland

    International Nuclear Information System (INIS)

    Krepel, Jiri; Hombourger, Boris; Fiorina, Carlo

    2015-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Eve Mullen

    2017-12-01

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

  6. [External cephalic version].

    Science.gov (United States)

    Navarro-Santana, B; Duarez-Coronado, M; Plaza-Arranz, J

    2016-08-01

    To analyze the rate of successful external cephalic versions in our center and caesarean sections that would be avoided with the use of external cephalic versions. From January 2012 to March 2016 external cephalic versions carried out at our center, which were a total of 52. We collected data about female age, gestational age at the time of the external cephalic version, maternal body mass index (BMI), fetal variety and situation, fetal weight, parity, location of the placenta, amniotic fluid index (ILA), tocolysis, analgesia, and newborn weight at birth, minor adverse effects (dizziness, hypotension and maternal pain) and major adverse effects (tachycardia, bradycardia, decelerations and emergency cesarean section). 45% of the versions were unsuccessful and 55% were successful. The percentage of successful vaginal delivery in versions was 84% (4% were instrumental) and 15% of caesarean sections. With respect to the variables studied, only significant differences in birth weight were found; suggesting that birth weight it is related to the outcome of external cephalic version. Probably we did not find significant differences due to the number of patients studied. For women with breech presentation, we recommend external cephalic version before the expectant management or performing a cesarean section. The external cephalic version increases the proportion of fetuses in cephalic presentation and also decreases the rate of caesarean sections.

  7. Salt bridge as a gatekeeper against partial unfolding.

    Science.gov (United States)

    Hinzman, Mark W; Essex, Morgan E; Park, Chiwook

    2016-05-01

    Salt bridges are frequently observed in protein structures. Because the energetic contribution of salt bridges is strongly dependent on the environmental context, salt bridges are believed to contribute to the structural specificity rather than the stability. To test the role of salt bridges in enhancing structural specificity, we investigated the contribution of a salt bridge to the energetics of native-state partial unfolding in a cysteine-free version of Escherichia coli ribonuclease H (RNase H*). Thermolysin cleaves a protruding loop of RNase H(*) through transient partial unfolding under native conditions. Lys86 and Asp108 in RNase H(*) form a partially buried salt bridge that tethers the protruding loop. Investigation of the global stability of K86Q/D108N RNase H(*) showed that the salt bridge does not significantly contribute to the global stability. However, K86Q/D108N RNase H(*) is greatly more susceptible to proteolysis by thermolysin than wild-type RNase H(*) is. The free energy for partial unfolding determined by native-state proteolysis indicates that the salt bridge significantly increases the energy for partial unfolding by destabilizing the partially unfolded form. Double mutant cycles with single and double mutations of the salt bridge suggest that the partially unfolded form is destabilized due to a significant decrease in the interaction energy between Lys86 and Asp108 upon partial unfolding. This study demonstrates that, even in the case that a salt bridge does not contribute to the global stability, the salt bridge may function as a gatekeeper against partial unfolding that disturbs the optimal geometry of the salt bridge. © 2016 The Protein Society.

  8. The FARC fuel archive of WWER

    International Nuclear Information System (INIS)

    Zizin, M.N.; Parfenova, N.A.; Proselkov, V.N.; Shishkov, L.K.

    1998-01-01

    The principles of organisation are explained and the structure of the FARC fuel archive for WWER reactors is circumscribed. The objective of archive is accumulation of fuel data, data storage and obtaining of fuel using characteristics. The working version of fuel archive on 01 July 1998 is realised, in which the data tables for fuel assemblies for 169 WWER-440 cycles and 35 WWER-1000 cycles are stored. There are two different versions of fuel archive - for WWER-440 (FARC) and for WWER-1000 (FARC1000). A structure of some tables and the texts of programs for them differ. (Authors)

  9. Expected brine movement at potential nuclear waste repository salt sites

    International Nuclear Information System (INIS)

    McCauley, V.S.; Raines, G.E.

    1987-08-01

    The BRINEMIG brine migration code predicts rates and quantities of brine migration to a waste package emplaced in a high-level nuclear waste repository in salt. The BRINEMIG code is an explicit time-marching finite-difference code that solves a mass balance equation and uses the Jenks equation to predict velocities of brine migration. Predictions were made for the seven potentially acceptable salt sites under consideration as locations for the first US high-level nuclear waste repository. Predicted total quantities of accumulated brine were on the order of 1 m 3 brine per waste package or less. Less brine accumulation is expected at domal salt sites because of the lower initial moisture contents relative to bedded salt sites. Less total accumulation of brine is predicted for spent fuel than for commercial high-level waste because of the lower temperatures generated by spent fuel. 11 refs., 36 figs., 29 tabs

  10. Thorium molten-salt nuclear energy synergetics

    International Nuclear Information System (INIS)

    Furukawa, Kazuo

    1989-01-01

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

  11. High Temperature Fluoride Salt Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Aaron, Adam M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cunningham, Richard Burns [Univ. of Tennessee, Knoxville, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kisner, Roger A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peretz, Fred J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yoder, Jr, Graydon L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

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

  12. The effect of salt composition on reductive extraction of some typical elements from molten LiF-BeF2 salt into liquid bismuth

    International Nuclear Information System (INIS)

    Hirotake, M.; Jun, O.; Kimikazu, M.; Kunimitsu, Y.; Yasunobu, T.

    1983-01-01

    The distribution coefficients of thorium and radium between molten LiF-BeF 2 and liquid bismuth solutions were measured at 600 0 C in support of the processing of the molten-salt breeder reactor (MSBR) fuel. The increasing mole fraction of LiF in the salt phase from 40 to 70 mol% resulted in the rapid decrease of the distribution coefficient of thorium and in the slow decrease of that of radium. A comprehensive correlation of distribution behavior with salt composition is given by taking into account the formation of complex ions. The equilibrium distribution data affirm that thorium and radium exist mainly as Li 2 ThF 6 and RaF 2 , respectively, in the salt phase. It is suggested that the lower mole fraction of LiF in the fuel salt is effective in the MSBR fuel processing

  13. Salt Tolerance in Soybean

    Institute of Scientific and Technical Information of China (English)

    Tsui-Hung Phang; Guihua Shao; Hon-Ming Lam

    2008-01-01

    Soybean is an Important cash crop and its productivity is significantly hampered by salt stress. High salt Imposes negative impacts on growth, nodulation, agronomy traits, seed quality and quantity, and thus reduces the yield of soybean. To cope with salt stress, soybean has developed several tolerance mechanisms, including: (I) maintenance of ion homeostasis; (ii) adjustment in response to osmotic stress; (iii) restoration of osmotic balance; and (iv) other metabolic and structural adaptations. The regulatory network for abiotic stress responses in higher plants has been studied extensively in model plants such as Arabidopsis thaliana. Some homologous components involved in salt stress responses have been identified in soybean. In this review, we tried to integrate the relevant works on soybean and proposes a working model to descdbe Its salt stress responses at the molecular level.

  14. Geomechanics of bedded salt

    International Nuclear Information System (INIS)

    Serata, S.; Milnor, S.W.

    1979-01-01

    Creep data from the literature search is reinterpreted by SGI, resulting in a better understanding of the temperature and stress state dependence of the octahedral creep rate and the octahedral shear strength. The concept of a transition strength between the elastic and the plastic states is in agreement with the data. The elastic and rheological properties of salt are described, and a set of constitutive equations is presented. The dependence of material properties on parameters such as temperature is considered. Findings on the permeability of salt are summarized, and the in-situ behavior of openings in bedded salt is described based on extensive engineering experience. A stress measuring system utilizing a finite element computer code is discussed. Geological factors affecting the stability of salt openings are considered, and the Stress Control Technique for designing stable openings in bedded salt formations is explained

  15. Salt og forbrugervalg

    DEFF Research Database (Denmark)

    Mørk, Trine; Grunert, Klaus G

    af saltreducerede fødevarer og deres købsintention af disse. Dette blev undersøgt ved at måle forbrugerens viden om salt, anvendelse af salt, ønske om reduktion af salt og købsintention af saltreducerede fødevarer i en web-baseret undersøgelse. Efter den web-baserede undersøgelse, blev de samme mål...... undersøgt, men i et supermarked, hvor deltagerne blev inddelt i fire grupper for at undersøge effekten af priming og saltmærkning. Desuden blev der foretaget 15 kvalitative interviews, for at studere hvem og hvad der karakteriserer de deltagere i eksperimentet, som enten ender med ingen salt......-reducerede produkter at købe eller som ender med at købe alle de salt-reducerede produkter....

  16. Nuclear Criticality Safety Handbook, Version 2. English translation

    International Nuclear Information System (INIS)

    2001-08-01

    The Nuclear Criticality Safety Handbook, Version 2 essentially includes the description of the Supplement Report to the Nuclear Criticality Safety Handbook, released in 1995, into the first version of the Nuclear Criticality Safety Handbook, published in 1988. The following two points are new: (1) exemplifying safety margins related to modeled dissolution and extraction processes, (2) describing evaluation methods and alarm system for criticality accidents. Revision has been made based on previous studies for the chapter that treats modeling the fuel system: e.g., the fuel grain size that the system can be regarded as homogeneous, non-uniformity effect of fuel solution, an burnup credit. This revision has solved the inconsistencies found in the first version between the evaluation of errors found in JACS code system and the criticality condition data that were calculated based on the evaluation. This report is an English translation of the Nuclear Criticality Safety Handbook, Version 2, originally published in Japanese as JAERI 1340 in 1999. (author)

  17. Fuel related risks; Braenslerisker

    Energy Technology Data Exchange (ETDEWEB)

    Englund, Jessica; Sernhed, Kerstin; Nystroem, Olle; Graveus, Frank (Grontmij AB, (Sweden))

    2012-02-15

    The project, within which this work report was prepared, aimed to complement the Vaermeforsk publication 'Handbook of fuels' on fuel related risks and measures to reduce the risks. The fuels examined in this project where the fuels included in the first version of the handbook from 2005 plus four additional fuels that will be included in the second and next edition of the handbook. Following fuels were included: woodfuels (sawdust, wood chips, powder, briquettes), slash, recycled wood, salix, bark, hardwood, stumps, straw, reed canary grass, hemp, cereal, cereal waste, olive waste, cocoa beans, citrus waste, shea, sludge, forest industrial sludge, manure, Paper Wood Plastic, tyre, leather waste, cardboard rejects, meat and bone meal, liquid animal and vegetable wastes, tall oil pitch, peat, residues from food industry, biomal (including slaughterhouse waste) and lignin. The report includes two main chapters; a general risk chapter and a chapter of fuel specific risks. The first one deals with the general concept of risk, it highlights laws and rules relevant for risk management and it discuss general risks that are related to the different steps of fuel handling, i.e. unloading, storing, processing the fuel, transportation within the facility, combustion and handling of ashes. The information that was used to produce this chapter was gathered through a literature review, site visits, and the project group's experience from risk management. The other main chapter deals with fuel-specific risks and the measures to reduce the risks for the steps of unloading, storing, processing the fuel, internal transportation, combustion and handling of the ashes. Risks and measures were considered for all the biofuels included in the second version in the handbook of fuels. Information about the risks and risk management was gathered through interviews with people working with different kinds of fuels in electricity and heat plants in Sweden. The information from

  18. Synthesis and Characterization of Processable Polyaniline Salts

    International Nuclear Information System (INIS)

    Gul, Salma; Bilal, Salma; Shah, Anwar-ul-Haq Ali

    2013-01-01

    Polyaniline (PANI) is one of the most promising candidates for possible technological applications. PANI has potential applications in batteries, anion exchanger, tissue engineering, inhibition of steel corrosion, fuel cell, sensors and so on. However, its insolubility in common organic solvents limits its range of applications. In the present study an attempt has been made to synthesize soluble polyaniline salt via inverse polymerization pathway using benzoyl peroxide as oxidant and dodecylbenzenesulfonic acid (DBSA) as dopant as well as a surfactant. A mixture of chloroform and 2-butanol was used as dispersion medium for the first time. The influence of synthesis parameters such as concentration of aniline, benzoyl peroxide and DBSA on the yield and other properties of the resulting PANI salt was studied. The synthesized PANI salt was found to be completely soluble in DMSO, DMF, chloroform and in a mixture of toluene and 2-propanol. The synthesized polymer salt was also characterized with cyclic voltam-metry, SEM, XRD, UV-Vis spectroscopy and viscosity measurements. TGA was used to analyze the thermal properties of synthesized polymer. The extent of doping of the PANI salt was determined from UV-Vis spectra and TGA analysis. The activation energy for the degradation of the polymer was calculated with the help of TGA.

  19. Lowering Salt in Your Diet

    Science.gov (United States)

    ... For Consumers Home For Consumers Consumer Updates Lowering Salt in Your Diet Share Tweet Linkedin Pin it ... Subscribe: FDA Consumer Health Information Everyone needs some salt to function. Also known as sodium chloride, salt ...

  20. Water-bearing explosive containing nitrogen-base salt

    Energy Technology Data Exchange (ETDEWEB)

    Dunglinson, C.; Lyerly, W.M.

    1968-10-21

    A water-bearing explosive composition consists of an oxidizing salt component, a fuel component, and water. A sensitizer is included having an oxygen balance more positive than -150%, and consisting of a salt of an inorganic oxidizing acid and of an acyclic nitrogen base having no more than 2 hydrogen atoms bonded to the basic nitrogen and up to 3 carbons per basic nitrogen, and/or of a phenyl amine. 41 claims.

  1. Long term storage of finished gasolines in large salt caverns

    Energy Technology Data Exchange (ETDEWEB)

    Koenig, J.W.J. [German Strategic Petroleum Reserve, Hamburg (Germany)

    1995-05-01

    Strategic oil stocking requires large low cost storage facilities. Crude oil has been held in very large salt mines and/or artificially made salt caverns for many years, notably in Europe and the USA. Following crude oil, gasoils and refinery light feed stocks have been tried also. Military organisations tried jet fuel and early cases of underground aviation gasoline storage in steel tanks have been reported.

  2. Demand driven salt clean-up in a molten salt fast reactor - Defining a priority list.

    Science.gov (United States)

    Merk, B; Litskevich, D; Gregg, R; Mount, A R

    2018-01-01

    The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified.

  3. Effects of heating on salt-occluded zeolite

    International Nuclear Information System (INIS)

    Lewis, M.A.; Hash, M.C.; Pereira, C.; Ackerman, J.P.

    1996-01-01

    The electrometallurgical treatment of spent nuclear fuel generates a waste stream of fission products in the electrolyte, LiCl-KCl eutectic salt. Argonne National Laboratory is developing a mineral waste form for this waste stream. The waste form consists of a composite formed by hot pressing salt-occluded zeolite and a glass binder. Pressing conditions must be judiciously chosen. For a given pressure, increasing temperatures and hold times give denser products but the zeolite is frequently converted to sodalite. Reducing the temperature or hold time leads to a porous zeolite composite. Therefore, conditions that affect the thermal stability of salt-occluded zeolite both with and without glass are being investigated in an ongoing study. The parameters varied in this stage of the work were heating time, temperature, salt loading, and glass content. The heat-treated samples were examined primarily by X-ray diffraction. Large variations were found in the rate at which salt-occluded zeolite converted to other phases such as nepheline, salt, and sodalite. The products depended on the initial salt loading. Heating times required for these transitions depended on the procedure and temperature used to prepare the salt-occluded zeolite. Mixtures of glass and zeolite reacted much faster than the pure salt-occluded zeolite and were almost always converted to sodalite

  4. Kinetics study of thermal decomposition of calcium carboxylate salts

    International Nuclear Information System (INIS)

    Landoll, Michael P.; Holtzapple, Mark T.

    2013-01-01

    The MixAlco™ process ferments lignocellulosic biomass to carboxylate salts that are thermally decomposed into ketones, which are then chemically converted to a wide variety of chemicals and fuels. To perform these decompositions, suitable reaction models are necessary to properly design, scale, and optimize commercial reactors. For three salt types (calcium acetate, and two types of mixed calcium carboxylate salts), activation energy was determined using three isoconversional methods that employed TGA curves at different heating rates. For all three salt types, activation energy varied significantly with conversion. The average activation energy for calcium acetate was 556.75 kJ mol −1 , and the activation energies for the two mixed calcium carboxylate salts were 232.87, and 176.55 kJ mol −1 . In addition, three functions of conversion were employed to see which one best modeled the experimental data. The Sestak–Berggren model provides the best universal fit for all three salt types. -- Highlights: •Calcium carboxylate salts from fermentation broth thermally decompose to ketones. •Activation energy varies with conversion for all three salt types. •Sestak–Berggren model provides best fit overall for all three salt types

  5. Kinetics study of thermal decomposition of sodium carboxylate salts

    International Nuclear Information System (INIS)

    Landoll, Michael P.; Holtzapple, Mark T.

    2012-01-01

    The MixAlco™ process ferments lignocellulosic biomass to carboxylate salts that are thermally decomposed into ketones, which are then chemically converted to a wide variety of chemicals and fuels. To perform these decompositions, suitable reaction models are necessary to properly design, scale, and optimize commercial reactors. For three salt types (sodium acetate, and two types of mixed sodium carboxylate salts), activation energy was determined using three isoconversional methods that employed TGA curves at different heating rates. For all three salt types, activation energy varied significantly with conversion. The average activation energy for sodium acetate was 226.65 kJ/mol, and the activation energies for the two mixed sodium carboxylate salts were 195.61, and 218.18 kJ/mol. In addition, three functions of conversion were employed to see which one best modeled the experimental data. The Sestak-Berggren model fits all three salt types best. -- Highlights: ► Sodium carboxylate salts from fermentation broth thermally decompose to ketones. ► Activation energy varies with conversion for all three salt types. ► Sestak-Berggren model provides best fit for all three salt types.

  6. Water purification using organic salts

    Science.gov (United States)

    Currier, Robert P.

    2004-11-23

    Water purification using organic salts. Feed water is mixed with at least one organic salt at a temperature sufficiently low to form organic salt hydrate crystals and brine. The crystals are separated from the brine, rinsed, and melted to form an aqueous solution of organic salt. Some of the water is removed from the aqueous organic salt solution. The purified water is collected, and the remaining more concentrated aqueous organic salt solution is reused.

  7. FASTDART - A fast, accurate and friendly version of DART code

    International Nuclear Information System (INIS)

    Rest, Jeffrey; Taboada, Horacio

    2000-01-01

    A new enhanced, visual version of DART code is presented. DART is a mechanistic model based code, developed for the performance calculation and assessment of aluminum dispersion fuel. Major issues of this new version are the development of a new, time saving calculation routine, able to be run on PC, a friendly visual input interface and a plotting facility. This version, available for silicide and U-Mo fuels, adds to the classical accuracy of DART models for fuel performance prediction, a faster execution and visual interfaces. It is part of a collaboration agreement between ANL and CNEA in the area of Low Enriched Uranium Advanced Fuels, held by the Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy. (author)

  8. Versioning Complex Data

    Energy Technology Data Exchange (ETDEWEB)

    Macduff, Matt C.; Lee, Benno; Beus, Sherman J.

    2014-06-29

    Using the history of ARM data files, we designed and demonstrated a data versioning paradigm that is feasible. Assigning versions to sets of files that are modified with some special assumptions and domain specific rules was effective in the case of ARM data, which has more than 5000 datastreams and 500TB of data.

  9. Evaluation of the salt deposition on the canister surface of concrete cask. Part 2. Measurement test of the salt concentration in air and salt deposition in the field

    International Nuclear Information System (INIS)

    Wataru, Masumi

    2012-01-01

    Concerning the storage facility of spent nuclear fuel using the concrete cask, there is an issue of stress corrosion cracking(SCC). The cooling air goes up along the canister surface in the concrete cask. To evaluate the initiation of SCC or rusting, it is important to verify the estimation method of the sea salt deposition on the metal canister surface transported by cooling air including sea salt particles. To measure the deposition rate, field tests were performed in Choushi test center. In the field test, it was found that the amount of sea salt deposition was very low because the density of the atmospheric sea salt concentration was very low compared with the laboratory test. Using relation between laboratory data and filed data, it is possible to evaluate the salt deposition rate on the canister surface. We also measured atmospheric sea salt concentration in Choushi test center to make the environment condition clear and compared the measurement data with the calculation data to verify the evaluation model. We are developing the automatic measuring device for atmospheric sea salt concentration. To check its performance, we are measuring atmospheric sea salt concentration in Yokosuka Area of CRIEPI and it was confirmed that the device works for one month automatically and fulfills its specifications. (author)

  10. Molten salt engineering for thorium cycle. Electrochemical studies as examples

    International Nuclear Information System (INIS)

    Ito, Yasuhiko

    1998-01-01

    A Th-U nuclear energy system utilizing accelerator driven subcritical molten salt breeder reactor has several advantages compared to conventional U-Pu nuclear system. In order to obtain fundamental data on molten salt engineering of Th-U system, electrochemical study was conducted. As the most primitive simulated study of beam irradiation of molten salt, discharge electrolysis was investigated in molten LiCl-KCl-AgCl system. Stationary discharge was generated under atmospheric argon gas and fine Ag particles were obtained. Hydride ion (H - ) behavior in molten salts was also studied to predict the behavior of tritide ion (T - ) in molten salt fuel. Finally, hydrogen behavior in metals at high temperature was investigated by electrochemical method, which is considered to be important to confine and control tritium. (author)

  11. Continuous extraction of molten chloride salts with liquid cadmium alloys

    International Nuclear Information System (INIS)

    Chow, L.S.; Basco, J.K.; Ackerman, J.P.; Johnson, T.R.

    1993-01-01

    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

  12. Method of detecting a fuel element failure

    International Nuclear Information System (INIS)

    Cohen, P.

    1975-01-01

    A method is described for detecting a fuel element failure in a liquid-sodium-cooled fast breeder reactor consisting of equilibrating a sample of the coolant with a molten salt consisting of a mixture of barium iodide and strontium iodide (or other iodides) whereby a large fraction of any radioactive iodine present in the liquid sodium coolant exchanges with the iodine present in the salt; separating the molten salt and sodium; if necessary, equilibrating the molten salt with nonradioactive sodium and separating the molten salt and sodium; and monitoring the molten salt for the presence of iodine, the presence of iodine indicating that the cladding of a fuel element has failed. (U.S.)

  13. Development of High Temperature Transport System for Molten Salt

    International Nuclear Information System (INIS)

    Lee, S. H.; Lee, H. S.; Kim, J. G.

    2011-01-01

    Pyroprocessing technology is one of the the most promising technologies for the advanced fuel cycle with favorable economic potential and intrinsic proliferation-resistance. The electrorefining process, one of main processes which is composed of pyroprocess to recover the useful elements from spent fuel, is under development at the Korea Atomic Energy Research Institute as a sub process of the pyrochemical treatment of spent PWR fuel. High-temperature molten salt transport technologies are required because a molten salt should be transported from the electrorefiner to electrowiner after the electrorefining process. Therefore, in pyrometallurgical processing, the development of high-temperature molten salt transport technologies is a crucial prerequisite. However, there have been a few transport studies on high-temperature molten salt. In this study, an apparatus for suction transport experiments was designed and constructed for the development of high temperature transport technology for molten salt, and the performance test of the apparatus was performed. And also, predissolution test of the salt was carried out using the reactor with furnace in experimental apparatus

  14. Studies on the molten salt reactor. Code development and neutronics analysis of MSRE-type design

    International Nuclear Information System (INIS)

    Zhuang Kun; Cao Liangzhi; Zheng Youqi; Wu Hongchun

    2015-01-01

    The molten salt reactor is characterized by its use of the fluid-fuel, which serves both as a fuel and as a coolant simultaneously. The position of delayed neutron precursors continuously changes both in the core and in the external loop due to the fuel circulation, and the fission products are extracted by an online fuel reprocessing unit, which all lead to the modeling methods for the conventional reactors using solid fuel not applicable. This study establishes suitable calculation models for the neutronics analysis of the molten salt reactor and develops a new code named MOREL based on the three-dimensional diffusion steady and transient calculations. Some numerical tests are chosen to verify the code and the numerical results indicate that MOREL can be used for the analysis of the molten salt reactor. After verification, it is applied to analyze the characteristics of a typical molten salt reactor, including the steady characteristics, the influence of fuel circulation on the kinetic behaviors. Besides, the influence of online fuel reprocessing simulation is also examined. The results show that inherent safety is the character of the molten salt reactor from the aspect of reactivity feedback and the fuel circulation has great influence on the kinetic characteristics of molten salt reactor. (author)

  15. Crushed Salt Constitutive Model

    International Nuclear Information System (INIS)

    Callahan, G.D.

    1999-01-01

    The constitutive model used to describe the deformation of crushed salt is presented in this report. Two mechanisms -- dislocation creep and grain boundary diffusional pressure solution -- are combined to form the basis for the constitutive model governing the deformation of crushed salt. The constitutive model is generalized to represent three-dimensional states of stress. Upon complete consolidation, the crushed-salt model reproduces the Multimechanism Deformation (M-D) model typically used for the Waste Isolation Pilot Plant (WIPP) host geological formation salt. New shear consolidation tests are combined with an existing database that includes hydrostatic consolidation and shear consolidation tests conducted on WIPP and southeastern New Mexico salt. Nonlinear least-squares model fitting to the database produced two sets of material parameter values for the model -- one for the shear consolidation tests and one for a combination of the shear and hydrostatic consolidation tests. Using the parameter values determined from the fitted database, the constitutive model is validated against constant strain-rate tests. Shaft seal problems are analyzed to demonstrate model-predicted consolidation of the shaft seal crushed-salt component. Based on the fitting statistics, the ability of the model to predict the test data, and the ability of the model to predict load paths and test data outside of the fitted database, the model appears to capture the creep consolidation behavior of crushed salt reasonably well

  16. Metallic materials corrosion problems in molten salt reactors

    International Nuclear Information System (INIS)

    Chauvin, G.; Dixmier, J.; Jarny, P.

    1977-01-01

    The USA forecastings concerning the molten salt reactors are reviewed (mixtures of fluorides containing the fuel, operating between 560 and 700 0 C). Corrosion problems are important in these reactors. The effects of certain characteristic factors on corrosion are analyzed: humidity and metallic impurities in the salts, temperature gradients, speed of circulation of salts, tellurium from fission products, coupling. In the molten fluorides and experimental conditions, the materials with high Ni content are particularly corrosion resistant alloys (hastelloy N). The corrosion of this material is about 2.6 mg.cm -2 at 700 0 C [fr

  17. Sea salt and pollution inputs over the continental United States

    International Nuclear Information System (INIS)

    Li, Y.-H.

    1992-01-01

    The average deposition rate of sea salt chloride over the world continents is about 10 meq m -2 yr -1 . Only about 14±1% of chloride in the pollution-corrected world average river is contributed by sea salt aerosols and the rest from the dissolution of evaporites. The significant increase of the ion concentrations in the Mississippi river from the year 1905 to 1987 was caused by anthropogenic inputs such as fossil fuel burning, common salt consumption, and dissolution of carbonate and silicate rocks by acids derived from acid precipitation. 29 refs., 4 figs., 3 tabs

  18. Advanced Nuclear Fuels for More Capable and Sustainable Exploration

    Data.gov (United States)

    National Aeronautics and Space Administration — Molten salt reactors are a subtype of reactor that uses nuclear fuel dissolved in a molten salt liquid medium (such as LiF-BeF2-UF4) as both fuel and coolant. The...

  19. Application of lithium in molten-salt reduction processes

    International Nuclear Information System (INIS)

    Gourishankar, K. V.

    1998-01-01

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

  20. Gases in molten salts

    CERN Document Server

    Tomkins, RPT

    1991-01-01

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

  1. A model for recovery of scrap monolithic uranium molybdenum fuel by electrorefining

    Science.gov (United States)

    Van Kleeck, Melissa A.

    The goal of the Reduced Enrichment for Research and Test Reactors program (RERTR) is toreduce enrichment at research and test reactors, thereby decreasing proliferation risk at these facilities. A new fuel to accomplish this goal is being manufactured experimentally at the Y12 National Security Complex. This new fuel will require its own waste management procedure,namely for the recovery of scrap from its manufacture. The new fuel is a monolithic uraniummolybdenum alloy clad in zirconium. Feasibility tests were conducted in the Planar Electrode Electrorefiner using scrap U-8Mo fuel alloy. These tests proved that a uranium product could be recovered free of molybdenum from this scrap fuel by electrorefining. Tests were also conducted using U-10Mo Zr clad fuel, which confirmed that product could be recovered from a clad version of this scrap fuel at an engineering scale, though analytical results are pending for the behavior of Zr in the electrorefiner. A model was constructed for the simulation of electrorefining the scrap material produced in the manufacture of this fuel. The model was implemented on two platforms, Microsoft Excel and MatLab. Correlations, used in the model, were developed experimentally, describing area specific resistance behavior at each electrode. Experiments validating the model were conducted using scrap of U-10Mo Zr clad fuel in the Planar Electrode Electrorefiner. The results of model simulations on both platforms were compared to experimental results for the same fuel, salt and electrorefiner compositions and dimensions for two trials. In general, the model demonstrated behavior similar to experimental data but additional refinements are needed to improve its accuracy. These refinements consist of a function for surface area at anode and cathode based on charge passed. Several approximations were made in the model concerning areas of electrodes which should be replaced by a more accurate function describing these areas.

  2. Fuel Handbook[Wood and other renewable fuels

    Energy Technology Data Exchange (ETDEWEB)

    Stroemberg, Birgitta [TPS Termiska Processer AB, Nykoeping (SE)] (ed.)

    2006-03-15

    This handbook on renewable fuels is intended for power and heat producers in Sweden. This fuel handbook provides, from a plant owner's perspective, a method to evaluate different fuels on the market. The fuel handbook concerns renewable fuels (but does not include household waste) that are available on the Swedish market today or fuels that have potential to be available within the next ten years. The handbook covers 26 different fuels. Analysis data, special properties, operating experiences and literature references are outlined for each fuel. [Special properties, operating experiences and literature references are not included in this English version] The handbook also contains: A proposed methodology for introduction of new fuels. A recommendation of analyses and tests to perform in order to reduce the risk of problems is presented. [The recommendation of analyses and tests is not included in the English version] A summary of relevant laws and taxes for energy production, with references to relevant documentation. [Only laws and taxes regarding EU are included] Theory and background to evaluate a fuel with respect to combustion, ash and corrosion properties and methods that can be used for such evaluations. Summary of standards, databases and handbooks on biomass fuels and other solid fuels, and links to web sites where further information about the fuels can be found. The appendices includes: A methodology for trial firing of fuels. Calculations procedures for, amongst others, heating value, flue gas composition, key number and free fall velocity [Free fall velocity is not included in the English version]. In addition, conversion routines between different units for a number of different applications are provided. Fuel analyses are presented in the appendix. (The report is a translation of parts of the report VARMEFORSK--911 published in 2005)

  3. Electrochemical reprocessing of nuclear fuels

    International Nuclear Information System (INIS)

    Brambilla, G.; Sartorelli, A.

    1980-01-01

    A method is described for the reprocessing of irradiated nuclear fuel which is particularly suitable for use with fuel from fast reactors and has the advantage of being a dry process in which there is no danger of radiation damage to a solvent medium as in a wet process. It comprises the steps of dissolving the fuel in a salt melt under such conditions that uranium and plutonium therein are converted to sulphate form. The plutonium sulphate may then be thermally decomposed to PuO 2 and removed. The salt melt is then subjected to electrolysis conditions to achieve cathodic deposition of UO 2 (and possibly PuO 2 ). The salt melt can then be recycled or conditioned for final disposal. (author)

  4. Embrittlement data base, version 1

    International Nuclear Information System (INIS)

    Wang, J.A.

    1997-08-01

    The aging and degradation of light-water-reactor (LWR) pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel (RPV) materials depends on many different factors such as flux, fluence, fluence spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Based on embrittlement predictions, decisions must be made concerning operating parameters and issues such as low-leakage-fuel management, possible life extension, and the need for annealing the pressure vessel. Large amounts of data from surveillance capsules and test reactor experiments, comprising many different materials and different irradiation conditions, are needed to develop generally applicable damage prediction models that can be used for industry standards and regulatory guides. Version 1 of the Embrittlement Data Base (EDB) is such a comprehensive collection of data resulting from merging version 2 of the Power Reactor Embrittlement Data Base (PR-EDB). Fracture toughness data were also integrated into Version 1 of the EDB. For power reactor data, the current EDB lists the 1,029 Charpy transition-temperature shift data points, which include 321 from plates, 125 from forgoings, 115 from correlation monitor materials, 246 from welds, and 222 from heat-affected-zone (HAZ) materials that were irradiated in 271 capsules from 101 commercial power reactors. For test reactor data, information is available for 1,308 different irradiated sets (352 from plates, 186 from forgoings, 303 from correlation monitor materials, 396 from welds and 71 from HAZs) and 268 different irradiated plus annealed data sets

  5. The fuel performance code future

    International Nuclear Information System (INIS)

    Ronchi, C.; Van de Laar, J.

    1988-01-01

    The paper describes the LWR version of the fuel performance code FUTURE, which was recently developed to calculate the fuel response (swelling, cladding deformation, release) to reactor transient conditions, starting from a broad-based description of the processes of major concern. The main physical models assumed are presented together with the scheme of the computer program

  6. What Are Bath Salts?

    Science.gov (United States)

    ... bath salts can produce: feelings of joy increased social interaction increased sex drive paranoia nervousness hallucinations (see or ... Institutes of Health; U.S. Department of Health and Human Services. Cite this article APA Style MLA Style ...

  7. Synthetic Cathinones ("Bath Salts")

    Science.gov (United States)

    ... Alcohol Club Drugs Cocaine Fentanyl Hallucinogens Inhalants Heroin Marijuana MDMA (Ecstasy/Molly) Methamphetamine Opioids Over-the-Counter Medicines Prescription Medicines Steroids (Anabolic) Synthetic Cannabinoids (K2/Spice) Synthetic Cathinones (Bath Salts) Tobacco/ ...

  8. Mass transport in bedded salt and salt interbeds

    International Nuclear Information System (INIS)

    Hwang, Y.; Pigford, T.H.; Chambre, P.L.; Lee, W.W.L.

    1989-08-01

    Salt is the proposed host rock for geologic repositories of nuclear waste in several nations because it is nearly dry and probably impermeable. Although experiments and experience at potential salt sites indicate that salt may contain brine, the low porosity, creep, and permeability of salt make it still a good choice for geologic isolation. In this paper we summarize several mass-transfer and transport analyses of salt repositories. The mathematical details are given in our technical reports

  9. Thermal diffusivity measurement of molten fluoride salt containing ThF4 (improvement of the simple ceramic cell)

    International Nuclear Information System (INIS)

    Kato, Y.; Araki, N.; Kobayashi, K.; Makino, A.

    1985-01-01

    Design conditions of a cylindrical ceramic cell are estimated which can be used to measure the absolute value of thermal diffusivity of molten salts by applying the stepwise heating method. Molten salt is expected to be used in nuclear systems such as the Molten-Salt Reactor, the Accelerator Molten-Salt Breeder, the Fusion Reactor Blanket Coolant, the Fuel Reprocessing System, and so on

  10. Fuel development studies

    International Nuclear Information System (INIS)

    Michel, F.

    1986-12-01

    This paper describes the main lines of the studies carried out to develop the Fast Neutron Fuel Element, from the ''SPX1-first load'' version, to progress to high performance which will be required for the project 1500 and for the fast neutron series [fr

  11. Simulation of Molten Salt Reactor dynamics

    International Nuclear Information System (INIS)

    Krepel, J.; Rohde, U.; Grundmann, U.

    2005-01-01

    Dynamics of the Molten Salt Reactor - one of the 'Generation IV' concepts - was studied in this paper. The graphite-moderated channel type MSR was selected for the numerical simulation of the reactor with liquid fuel. The MSR dynamics is very specific because of two physical peculiarities of the liquid fueled reactor: the delayed neutrons precursors are drifted by the fuel flow and the fission energy is immediately released directly into the coolant. Presently, there are not many accessible numerical codes appropriate for the MSR simulation, therefore the DYN3D-MSR code was developed based on the FZR in-house code DYN3D. It allows calculating of full 3D transient neutronics in combination with parallel channel type thermal-hydraulics. By means of DYN3D-MSR, several transients typical for the liquid fuel system were analyzed. Those transients were initiated by reactivity insertion, by overcooling of fuel at the core inlet, by the fuel pump start-up or coast-down, or by the blockage of selected fuel channels. In these considered transients, the response of the MSR is characterized by the immediate change of the fuel temperature with changing power and fast negative temperature feedback to the power. The response through the graphite temperature is slower. Furthermore, for big MSR cores fueled with U233 the graphite feedback coefficient can be positive. In this case the addition of erbium to the graphite can ensure the inherent safety features. The DYN3D-MSR code has been shown to be an effective tool for MSR dynamics studies. (author)

  12. Fuel Exhaling Fuel Cell.

    Science.gov (United States)

    Manzoor Bhat, Zahid; Thimmappa, Ravikumar; Devendrachari, Mruthyunjayachari Chattanahalli; Kottaichamy, Alagar Raja; Shafi, Shahid Pottachola; Varhade, Swapnil; Gautam, Manu; Thotiyl, Musthafa Ottakam

    2018-01-18

    State-of-the-art proton exchange membrane fuel cells (PEMFCs) anodically inhale H 2 fuel and cathodically expel water molecules. We show an unprecedented fuel cell concept exhibiting cathodic fuel exhalation capability of anodically inhaled fuel, driven by the neutralization energy on decoupling the direct acid-base chemistry. The fuel exhaling fuel cell delivered a peak power density of 70 mW/cm 2 at a peak current density of 160 mA/cm 2 with a cathodic H 2 output of ∼80 mL in 1 h. We illustrate that the energy benefits from the same fuel stream can at least be doubled by directing it through proposed neutralization electrochemical cell prior to PEMFC in a tandem configuration.

  13. Cooperativity of complex salt bridges

    OpenAIRE

    Gvritishvili, Anzor G.; Gribenko, Alexey V.; Makhatadze, George I.

    2008-01-01

    The energetic contribution of complex salt bridges, in which one charged residue (anchor residue) forms salt bridges with two or more residues simultaneously, has been suggested to have importance for protein stability. Detailed analysis of the net energetics of complex salt bridge formation using double- and triple-mutant cycle analysis revealed conflicting results. In two cases, it was shown that complex salt bridge formation is cooperative, i.e., the net strength of the complex salt bridge...

  14. Alternative Fuels Data Center: Natural Gas Vehicles

    Science.gov (United States)

    Natural Gas Printable Version Share this resource Send a link to Alternative Fuels Data Center : Natural Gas Vehicles to someone by E-mail Share Alternative Fuels Data Center: Natural Gas Vehicles on Facebook Tweet about Alternative Fuels Data Center: Natural Gas Vehicles on Twitter Bookmark Alternative

  15. Fast Thorium Molten Salt Reactors Started with Plutonium

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  16. Version pressure feedback mechanisms for speculative versioning caches

    Science.gov (United States)

    Eichenberger, Alexandre E.; Gara, Alan; O& #x27; Brien, Kathryn M.; Ohmacht, Martin; Zhuang, Xiaotong

    2013-03-12

    Mechanisms are provided for controlling version pressure on a speculative versioning cache. Raw version pressure data is collected based on one or more threads accessing cache lines of the speculative versioning cache. One or more statistical measures of version pressure are generated based on the collected raw version pressure data. A determination is made as to whether one or more modifications to an operation of a data processing system are to be performed based on the one or more statistical measures of version pressure, the one or more modifications affecting version pressure exerted on the speculative versioning cache. An operation of the data processing system is modified based on the one or more determined modifications, in response to a determination that one or more modifications to the operation of the data processing system are to be performed, to affect the version pressure exerted on the speculative versioning cache.

  17. The Molten Salt Reactor option for beneficial use of fissile material from dismantled weapons

    International Nuclear Information System (INIS)

    Gat, U.; Engel, J.R.; Dodds, H.L.

    1991-01-01

    The Molten Salt Reactor (MSR) option for burning fissile fuel from dismantled weapons is examined. It is concluded that MSRs are very suitable for beneficial utilization of the dismantled fuel. The MSRs can utilize any fissile fuel in continuous operation with no special modifications, as demonstrated in the Molten Salt Reactor Experiment. Thus MSRs are flexible while maintaining their economy. MSRs further require a minimum of special fuel preparation and can tolerate denaturing and dilution of the fuel. Fuel shipments can be arbitrarily small, all of which supports nonproliferation and averts diversion. MSRs have inherent safety features which make them acceptable and attractive. They can burn a fuel type completely and convert it to other fuels. MSRs also have the potential for burning the actinides and delivering the waste in an optimal form, thus contributing to the solution of one of the major remaining problems for deployment of nuclear power. 19 refs

  18. Process technology for the molten-salt reactor 233U--Th cycle

    International Nuclear Information System (INIS)

    Hightower, J.R. Jr.

    1975-01-01

    After a brief description of the design features of the molten-salt breeder reactor, fuel processing for removal of 233 Pa and fission products is examined. Some recent developments in processing technology are discussed

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

    NARCIS (Netherlands)

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

    2018-01-01

    The present study describes the full thermodynamic assessment of the Li,Cs,Th//F,I system. The existing database for the relevant fluoride salts considered as fuel for the Molten Salt Reactor (MSR) has been extended with two key fission products, cesium and iodine. A complete evaluation of all

  20. Fuel Cycle Concept with Advanced METMET and Composite Fuel in LWRs

    International Nuclear Information System (INIS)

    Savchenko, A.; Skupov, M.; Vatulin, A.; Glushenkov, A.; Kulakov, G.; Lipkina, K.

    2014-01-01

    The basic factor that limits the serviceability of fuel elements developing in the framework of RERTR Program (transition from HEU to LEU fuel of research reactors) is interaction between U10Mo fuel and aluminium matrix . Interaction results in extra swelling of fuels, disappearance of a heat conducting matrix, a temperature rise in the fuel centre, penetration porosity, etc. Several methods exist to prevent fuel-matrix interaction. In terms of simplifying fuel element fabrication technology and reducing interaction, doping of fuel is the most optimal version

  1. Waste form dissolution in bedded salt

    International Nuclear Information System (INIS)

    Kaufman, A.M.

    1980-01-01

    A model was devised for waste dissolution in bedded salt, a hydrologically tight medium. For a typical Spent UnReprocessed Fuel (SURF) emplacement, the dissolution rate wll be diffusion limited and will rise to a steady state value after t/sub eq/ approx. = 250 (1+(1-epsilon 0 ) K/sub D//epsilon 0 ) (years) epsilon 0 is the overpack porosity and K/sub d/ is the overpack sorption coefficient. The steady state dissolution rate itself is dominated by the solubility of UO 2 . Steady state rates between 5 x 10 -5 and .5 (g/year) are achievable by SURF emplacements in bedded salt without overpack, and rates between 5 x 10 -7 and 5 x 10 -3 (g/year) with an overpack having porosity of 10 -2

  2. Molten salt breeder reactor

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Tsukada, Kineo; Nakahara, Yasuaki; Oomichi, Toshihiko; Oono, Hideo.

    1982-01-01

    Purpose: To simplify the structure, as well as improve the technical reliability and safety by the elimination of a proton beam entering window. Constitution: The nuclear reactor container main body is made of Hastelloy N and provided at the inner surface with two layers of graphite shields except for openings. An aperture was formed in the upper surface of the container, through which protons accelerated by a linear accelerator are directly entered to the liquid surface of molten salts such as 7LiF-BeF 2 -ThF 4 , 7LiF-NaF-ThF 4 , 7LiF-Rb-UF 4 , NaF-KF-UF 4 and the like. The heated molten salts are introduced by way of a pipeway into a heat exchanger where the heat is transferred to coolant salts and electric generation is conducted by way of heated steams. (Furukawa, Y.)

  3. Used nuclear fuel separations process simulation and testing

    International Nuclear Information System (INIS)

    Pereira, C.; Krebs, J.F.; Copple, J.M.; Frey, K.E.; Maggos, L.E.; Figueroa, J.; Willit, J.L.; Papadias, D.D.

    2013-01-01

    Recent efforts in separations process simulation at Argonne have expanded from the traditional focus on solvent extraction flowsheet design in order to capture process dynamics and to simulate other components, processing and systems of a used nuclear fuel reprocessing plant. For example, the Argonne Model for Universal Solvent Extraction (AMUSE) code has been enhanced to make it both more portable and more readily extensible. Moving away from a spreadsheet environment makes the addition of new species and processes simpler for the expert user, which should enable more rapid implementation of chemical models that simulate evolving processes. The dyAMUSE (dynamic AMUSE) version allows the simulation of transient behavior across an extractor. Electrochemical separations have now been modeled using spreadsheet codes that simulate the electrochemical recycle of fast reactor fuel. The user can follow the evolution of the salt, products, and waste compositions in the electro-refiner, cathode processors, and drawdown as a function of fuel batches treated. To further expand capabilities in integrating multiple unit operations, a platform for linking mathematical models representing the different operations that comprise a reprocessing facility was adapted to enable systems-level analysis and optimization of facility functions. (authors)

  4. Used nuclear fuel separations process simulation and testing

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, C.; Krebs, J.F.; Copple, J.M.; Frey, K.E.; Maggos, L.E.; Figueroa, J.; Willit, J.L.; Papadias, D.D. [Argonne National Laboratory: 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2013-07-01

    Recent efforts in separations process simulation at Argonne have expanded from the traditional focus on solvent extraction flowsheet design in order to capture process dynamics and to simulate other components, processing and systems of a used nuclear fuel reprocessing plant. For example, the Argonne Model for Universal Solvent Extraction (AMUSE) code has been enhanced to make it both more portable and more readily extensible. Moving away from a spreadsheet environment makes the addition of new species and processes simpler for the expert user, which should enable more rapid implementation of chemical models that simulate evolving processes. The dyAMUSE (dynamic AMUSE) version allows the simulation of transient behavior across an extractor. Electrochemical separations have now been modeled using spreadsheet codes that simulate the electrochemical recycle of fast reactor fuel. The user can follow the evolution of the salt, products, and waste compositions in the electro-refiner, cathode processors, and drawdown as a function of fuel batches treated. To further expand capabilities in integrating multiple unit operations, a platform for linking mathematical models representing the different operations that comprise a reprocessing facility was adapted to enable systems-level analysis and optimization of facility functions. (authors)

  5. Mineral resource of the month: salt

    Science.gov (United States)

    Kostick, Dennis S.

    2010-01-01

    The article presents information on various types of salt. Rock salt is either found from underground halite deposits or near the surface. Other types of salt include solar salt, salt brine, and vacuum pan salt. The different uses of salt are also given including its use as a flavor enhancer, as a road deicing agent, and to manufacture sodium hydroxide.

  6. Definition of breeding gain for molten salt reactors - 147

    International Nuclear Information System (INIS)

    Nagy, K.; Kloosterman, J.L.; Lathouwers, D.; Van der Hagen, T.H.J.J.

    2010-01-01

    The graphite-moderated Molten Salt Reactor (MSR) is a potential breeder reactor using the thorium fuel cycle. The MSR has unique properties due to the possibility of making changes to the salt composition during operation. Most important is the extraction of protactinium, which separates the fissile uranium production into two volumes: the reactor core and the external stockpile. The paper focuses on the definition of breeding gain in such a system. The prospects of using breeding gain expressions defined for solid fuel reactors are investigated and new definitions are given which incorporate the processes occurring in the reactor core and the external stockpile. The difference of the growth rate of the mass of fissile material and breeding gain is pointed out. The new definitions are applied to an optimization study of the graphite-salt lattice of a breeder MSR. (authors)

  7. Determining Optimal Decision Version

    Directory of Open Access Journals (Sweden)

    Olga Ioana Amariei

    2014-06-01

    Full Text Available In this paper we start from the calculation of the product cost, applying the method of calculating the cost of hour- machine (THM, on each of the three cutting machines, namely: the cutting machine with plasma, the combined cutting machine (plasma and water jet and the cutting machine with a water jet. Following the calculation of cost and taking into account the precision of manufacturing of each machine, as well as the quality of the processed surface, the optimal decisional version needs to be determined regarding the product manufacturing. To determine the optimal decisional version, we resort firstly to calculating the optimal version on each criterion, and then overall using multiattribute decision methods.

  8. Version 2 of RSXMULTI

    International Nuclear Information System (INIS)

    Heinicke, P.; Berg, D.; Constanta-Fanourakis, P.; Quigg, E.K.

    1985-01-01

    MULTI is a general purpose, high speed, high energy physics interface to data acquisition and data investigation system that runs on PDP-11 and VAX architecture. This paper describes the latest version of MULTI, which runs under RSX-11M version 4.1 and supports a modular approach to the separate tasks that interface to it, allowing the same system to be used in single CPU test beam experiments as well as multiple interconnected CPU, large scale experiments. MULTI uses CAMAC (IEE-583) for control and monitoring of an experiment, and is written in FORTRAN-77 and assembler. The design of this version, which simplified the interface between tasks, and eliminated the need for a hard to maintain homegrown I/O system is also discussed

  9. Versioning of printed products

    Science.gov (United States)

    Tuijn, Chris

    2005-01-01

    During the definition of a printed product in an MIS system, a lot of attention is paid to the production process. The MIS systems typically gather all process-related parameters at such a level of detail that they can determine what the exact cost will be to make a specific product. This information can then be used to make a quote for the customer. Considerably less attention is paid to the content of the products since this does not have an immediate impact on the production costs (assuming that the number of inks or plates is known in advance). The content management is typically carried out either by the prepress systems themselves or by dedicated workflow servers uniting all people that contribute to the manufacturing of a printed product. Special care must be taken when considering versioned products. With versioned products we here mean distinct products that have a number of pages or page layers in common. Typical examples are comic books that have to be printed in different languages. In this case, the color plates can be shared over the different versions and the black plate will be different. Other examples are nation-wide magazines or newspapers that have an area with regional pages or advertising leaflets in different languages or currencies. When considering versioned products, the content will become an important cost factor. First of all, the content management (and associated proofing and approval cycles) becomes much more complex and, therefore, the risk that mistakes will be made increases considerably. Secondly, the real production costs are very much content-dependent because the content will determine whether plates can be shared across different versions or not and how many press runs will be needed. In this paper, we will present a way to manage different versions of a printed product. First, we will introduce a data model for version management. Next, we will show how the content of the different versions can be supplied by the customer

  10. COSY INFINITY Version 9

    International Nuclear Information System (INIS)

    Makino, Kyoko; Berz, Martin

    2006-01-01

    In this paper, we review the features in the newly released version of COSY INFINITY, which currently has a base of more than 1000 registered users, focusing on the topics which are new and some topics which became available after the first release of the previous versions 8 and 8.1. The recent main enhancements of the code are devoted to reliability and efficiency of the computation, to verified integration, and to rigorous global optimization. There are various data types available in COSY INFINITY to support these goals, and the paper also reviews the feature and usage of those data types

  11. Graphite and carbonaceous materials in a molten salt nuclear reactor

    International Nuclear Information System (INIS)

    Rousseau, Ginette; Lecocq, Alfred; Hery, Michel.

    1982-09-01

    A project for a molten salt 1000 MWe reactor is studied by EDF-CEA teams. The design provides for a chromesco 3 vessel housing graphite structures in which the salt circulates. The salt (Th, U, Be and Li fluorides) is cooled by direct contact with lead. The graphites and carbonated materials, inert with respect to lead and the fuel salt, are being considered not only as moderators, but as reflectors and in the construction of the sections where the heat exchange takes place. On the basis of the problems raised in the operation of the reactor, a study programme on French experimental materials (Le Carbone Lorraine, SERS, SEP) has been defined. Hence, depending on the function or functions that the material is to ensure in the structure, the criteria of choice which follow will have to be examined: behaviour under irradiation, insertion of a fluid in the material, thermal properties required, mechanical properties required, utilization [fr

  12. Some metallic materials and fluoride salts for high temperature applications

    International Nuclear Information System (INIS)

    Hosnedl, P.; Hron, M.; Matal, O.

    2009-01-01

    There has been a special Ni base alloy MONICR for high temperature applications in fluoride salt environments developed in the framework of the complex R and D program for the Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept development in the Czech Republic. Selected results of MONICR alloy tests and results of semi products fabrication from this alloy are discussed in the paper. The results of the structural materials tests are applied on semi-products and for the design of the testing devices as the autoclave in loop arrangement for high temperature fluoride salts applications. Material properties other Ni base alloys are compared to those of MONICR. Corrosion test results of the alloy A686 in the LiF - NaF - ZrF 4 molten salt are provided and compared to the measured values of the polarizing resistance. (author)

  13. Salt Removal from the Uranium Deposits of Electrorefiner

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, S. W.; Park, K. M.; Lee, S. J.; Park, S. B.; Cho, C. H.; Choi, S. Y.; Lee, H. S.; Kim, J. G. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Electrorefining is a key step in pyroprocessing. The electrorefining process is generally composed of two recovery steps. The deposit of uranium onto a solid cathode and the recovery of the remaining uranium and TRU elements simultaneously by a liquid cadmium cathode. The solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. In the liquid cathode, cadmium metal should be removed to recover actinide product. A physical separation process, such as distillation separation, is more attractive than a chemical or dissolution process because physical processes generate much less secondary process. Distillation process was employed for the cathode processing due to the advantages of minimal generation of secondary waste, compact unit process, simple and low cost equipment. The basis for vacuum distillation separation is the difference in vapor pressures between salt and uranium. A solid cathode deposit is heated in a heating region and salt vaporizes, while non volatile uranium remains behind. It is very important to increase the throughput of the salt separation system due to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. The evaporation rate of the LiCl-KCl eutectic salt in vacuum distiller is not so high to come up with the generation capacity of uranium dendrites in electro-refiner. Therefore, wide evaporation area or high distillation temperature is necessary for the successful salt separation. In this study, the solid-liquid separation was proposed prior to distillation of salt and a feasibility of the separation of the liquid salt by a metallic wire mesh (sieve) was tested for the reduction of the burden of the following vacuum distillation process

  14. Salt Removal from the Uranium Deposits of Electrorefiner

    International Nuclear Information System (INIS)

    Kwon, S. W.; Park, K. M.; Lee, S. J.; Park, S. B.; Cho, C. H.; Choi, S. Y.; Lee, H. S.; Kim, J. G.

    2010-01-01

    Electrorefining is a key step in pyroprocessing. The electrorefining process is generally composed of two recovery steps. The deposit of uranium onto a solid cathode and the recovery of the remaining uranium and TRU elements simultaneously by a liquid cadmium cathode. The solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. In the liquid cathode, cadmium metal should be removed to recover actinide product. A physical separation process, such as distillation separation, is more attractive than a chemical or dissolution process because physical processes generate much less secondary process. Distillation process was employed for the cathode processing due to the advantages of minimal generation of secondary waste, compact unit process, simple and low cost equipment. The basis for vacuum distillation separation is the difference in vapor pressures between salt and uranium. A solid cathode deposit is heated in a heating region and salt vaporizes, while non volatile uranium remains behind. It is very important to increase the throughput of the salt separation system due to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. The evaporation rate of the LiCl-KCl eutectic salt in vacuum distiller is not so high to come up with the generation capacity of uranium dendrites in electro-refiner. Therefore, wide evaporation area or high distillation temperature is necessary for the successful salt separation. In this study, the solid-liquid separation was proposed prior to distillation of salt and a feasibility of the separation of the liquid salt by a metallic wire mesh (sieve) was tested for the reduction of the burden of the following vacuum distillation process

  15. New rational nuclear energy system composed of accelerator molten-salt breeder (AMSB) and molten-salt power stations (MSCR)

    International Nuclear Information System (INIS)

    Furukawa, K.

    1985-01-01

    For the next century, it was predicted that some rational fission energy system breeding in significantly short doubling time less than 10 years should be developed replacing the fossil fuels. In practice, this rationality, that is, simplicity and high economy could be realized by the natural combination of: molten salt fuel concept; accelerator (spallation) breeding concept; and Thorium fuel cycle concept, in the symbiont system of Accelerator Molten-Salt breeders and Molten-Salt Power Stations. The economy of this system might significantly become better than the other breeder systems, although the prediction in Chapter 6 was too much conservative. Its more important aspect is the low cost of future R and D, which depend on the rational character of Molten-Fluoride Technology and really is verified by the basic R and D cost (only $0.13 B) in Oak Ridge N.L. It is interesting that molten-salt technology will be able to apply to chemical processing of U-Pu oxide fuels by the developing effort by USSR in near future. This fact and the demand of small power stations such as 150MWe MSCR presented here will be able to bridge between the present and the next century

  16. 21 CFR 100.155 - Salt and iodized salt.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 2 2010-04-01 2010-04-01 false Salt and iodized salt. 100.155 Section 100.155 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) FOOD FOR HUMAN CONSUMPTION GENERAL Specific Administrative Rulings and Decisions § 100.155 Salt and iodized salt. (a) For the purposes of this section, the...

  17. MARS: Story on Molten Salt Actinide Recycler and Transmuter Development by Rosatom in Co-operation with Euratom

    International Nuclear Information System (INIS)

    Ignatiev, V.; Feynberg, O.; Gnidoi, I.; Konakov, S.; Kormilitsyn, M.; Merzliakov, A.; Surenkov, A.; Uglov, V.; Zagnitko, A.

    2015-01-01

    New design options of MOSART and MSFR systems without and with U-Th support fuelled with different compositions of transuranic elements trifluorides from spent LWR fuel both based on homogeneous cores and used fuel salts with high enough solubility for transuranic elements trifluorides are being examined within MARS (Rosatom) and EVOL (Euratom) parallel coordinated projects. The paper has the main objective of presenting the fuel cycle flexibility of the mentioned above systems while accounting technical constrains and experimental data received in this study. A brief description is given of the calculation core neutronics properties and fuel cycle scenarios as well as experimental results on key fuel salt properties, salt chemistry control and combined materials compatibility to satisfy MOSART and MSFR systems requirements. Measurements described mainly concern phase behaviour and transport properties data for selected fuel salts. As for fuel salt clean-up operations in MOSART and MSFR fuel cycles, the most uncertain its part concerning rare earth removal is discussed. Last section is focused on the compatibility of special Ni-based alloys with fuel salt selected at temperatures required for MOSART and MSFR operation. The major achievements are: (1) ability to produce and maintain a high level of purity in fuel salt, (2) effective control of the Redox potential of the salt medium in order to minimize corrosion, (3) understanding of basic corrosion mechanisms in MOSART and MSFR systems. HN80MTY alloy can be recommended for further consideration as the main container material for the fuel circuit with operating temperature up to 1 023 K required for MOSART and MSFR designs. (authors)

  18. Premixing of corium into water during a Fuel-Coolant Interaction. The models used in the 3 field version of the MC3D code and two examples of validation on Billeau and FARO experiments

    Energy Technology Data Exchange (ETDEWEB)

    Berthoud, G.; Crecy, F. de; Duplat, F.; Meignen, R.; Valette, M. [CEA/Grenoble, DRN/DTP, 17 Avenue des Martyrs, 38054 Grenoble Cedex 9 (France)

    1998-01-01

    This paper presents the <> application of the multiphasic 3D computer code MC3D. This application is devoted to the premixing phase of a Fuel Coolant Interaction (FCI) when large amounts of molten corium flow into water and interact with it. A description of the new features of the model is given (a more complete description of the full model is given in annex). Calculations of Billeau experiments (cold or hot spheres dropped into water) and of a FARO test (<> corium dropped into 5 MPa saturated water) are presented. (author)

  19. Treatment of waste salts by oxygen sparging and vacuum distillation

    International Nuclear Information System (INIS)

    Cho, Y.J.; Yang, H.C.; Kim, E.H.; Kin, I.T.; Eun, H.C.

    2007-01-01

    Full text of publication follows. During the electrorefining process of the oxide spent fuel from LWR, amounts of waste salts containing some metal chloride species such as rare earths and actinide chlorides are generated, where the reuse of the waste salts is very important from the standpoint of an economical as well as an environmental aspect. In order to reuse the waste salts, a salt vacuum distillation method can be used. For the best separation by a vacuum distillation, the metal chloride species involved in the waste salts must be converted into their oxide(or oxychloride) forms due to the their low volatility compared to that of LiCl-KCl. In this study, an oxygen sparging process was adopted for the oxidation (or precipitation) of rare earth chlorides. The effects of oxygen flow rate and molten salt temperature on the conversion of rare earth chlorides to the precipitate phase (i.e. oxide or oxychloride) were investigated. In addition, distillation characteristics of LiCl-KCl molten salt with system pressure and temperature were studied. (authors)

  20. Inert matrix fuel in dispersion type fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Savchenko, A.M. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)]. E-mail: sav@bochvar.ru; Vatulin, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Morozov, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Sirotin, V.L. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Dobrikova, I.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kulakov, G.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Ershov, S.A. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kostomarov, V.P. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Stelyuk, Y.I. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)

    2006-06-30

    The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg{sup -1} (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

  1. Inert matrix fuel in dispersion type fuel elements

    Science.gov (United States)

    Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

    2006-06-01

    The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

  2. Molten salt oxidation as a technique for decommissioning: selection of low melting point salt mixtures

    International Nuclear Information System (INIS)

    Lainetti, Paulo E.O.; Garcia, Vitor F.; Benvegnu, Guilherme

    2013-01-01

    During the 70 and 80 years, IPEN built several facilities in pilot scale, destined to the technological domain of the Nuclear Fuel Cycle. In the nineties, radical changes in the Brazilian nuclear policy determined the interruption of the activities and the shut-down of pilot plants. Nowadays, IPEN has been facing the problem of the dismantling and decommissioning of its Nuclear Fuel Cycle old facilities. The facility CELESTE-I of the IPEN is a laboratory where reprocessing studies were accomplished during the decade of 80 and in the beginning of the 90s. The last operations occurred in 92-93. The research activities generated radioactive wastes in the form of organic and aqueous solutions of different compositions and concentrations. For the treatment of these liquid wastes it was proposed a study of waste thermal decomposition based on the molten salt oxidation process.Decomposition tests of different organic wastes have been performed in laboratory equipment developed at IPEN, in the range of temperatures of 900 to 1020 deg C, demonstrating the complete oxidation of the compounds. The reduction of the process temperatures would be of crucial importance. Besides this, the selection of lower melting point salt mixtures would have an important impact in the reduction of equipment costs. Several experiments were performed to determine the most suitable salt mixtures, optimizing costs and melting temperatures as low as possible. This paper describes the main characteristics of the molten salt oxidation process, besides the selection of salt mixtures of binary and ternary compositions, respectively Na 2 CO 3 - NaOH and Na 2 CO 3 - K 2 CO 3 -Li 2 CO 3 . (author)

  3. Safe actinide disposition in molten salt reactors

    International Nuclear Information System (INIS)

    Gat, U.

    1997-01-01

    Safe molten salt reactors (MSR) can readily accommodate the burning of all fissile actinides. Only minor compromises associated with plutonium are required. The MSRs can dispose safely of actinides and long lived isotopes to result in safer and simpler waste. Disposing of actinides in MSRs does increase the source term of a safety optimized MSR. It is concluded that the burning and transmutation of actinides in MSRs can be done in a safe manner. Development is needed for the processing to handle and separate the actinides. Calculations are needed to establish the neutron economy and the fuel management. 9 refs

  4. to salt stress

    African Journals Online (AJOL)

    Tony

    2012-02-14

    Feb 14, 2012 ... 3Inner Mongolia Industrial Engineering Research, Center of University for Castor, Tongliao 028042, ... strengthen and improve salt stress tolerance in plants. .... 2 µl cDNA, 1 µl each of 4 µM forward and reverse primer, 0.2 µl.

  5. Borehole closure in salt

    International Nuclear Information System (INIS)

    Fuenkajorn, K.; Daemen, J.J.K.

    1988-12-01

    Constitutive law parameters are determined from salt behavior characterization experiments. The results are applied to predict creep (time-dependent) closure of boreholes in salt specimens subjected to various loading configurations. Rheological models (linear and nonlinear viscoelastic and viscoplastic models), empirical models, and physical theory models have been formulated from the results of uniaxial creep tests, strain and stress rate controlled uniaxial tests, constant strain rate triaxial tests, cyclic loading tests, and seismic velocity measurements. Analytical solutions for a thick-walled cylinder subjected to internal and external pressures and for a circular hole in an infinite plate subjected to a biaxial or uniaxial stressfield have been derived from each of the linear viscoelastic models and from one of the empirical laws. The experimental results indicate that the salt samples behave as an elastic-viscoplastic material. The elastic behavior tends to be linear and time-independent. The plastic deformation is time-dependent. The stress increment to strain rate increment ratio gradually decreases as the stress level increases. The transient potential creep law seems to give the simplest satisfactory governing equation describing the viscoplastic behavior of salt during the transient phase. 204 refs., 27 figs., 29 tabs

  6. Salt repository design approach

    International Nuclear Information System (INIS)

    Matthews, S.C.

    1983-01-01

    This paper presents a summary discussion of the approaches that have been and will be taken in design of repository facilities for use with disposal of radioactive wastes in salt formations. Since specific sites have yet to be identified, the discussion is at a general level, supplemented with illustrative examples where appropriate. 5 references, 1 figure

  7. Learning SaltStack

    CERN Document Server

    Myers, Colton

    2015-01-01

    If you are a system administrator who manages multiple servers, then you know how difficult it is to keep your infrastructure in line. If you've been searching for an easier way, this book is for you. No prior experience with SaltStack is required.

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

    International Nuclear Information System (INIS)

    Gat, U.; Engel, J.R.

    1991-01-01

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

  9. PYRO, a system for modeling fuel reprocessing

    International Nuclear Information System (INIS)

    Ackerman, J.P.

    1989-01-01

    Compact, on-site fuel reprocessing and waste management for the Integral Fast Reactor are based on the pyrochemical reprocessing of metal fuel. In that process, uranium and plutonium in spent fuel are separated from fission products in an electrorefiner using liquid cadmium and molten salt solvents. Quantitative estimates of the distribution of the chemical elements among the metal and salt phases are essential for development of both individual pyrochemical process steps and the complete process. This paper describes the PYRO system of programs used to generate reliable mass flows and compositions

  10. In-vessel source term analysis code TRACER version 2.3. User's manual

    International Nuclear Information System (INIS)

    Toyohara, Daisuke; Ohno, Shuji; Hamada, Hirotsugu; Miyahara, Shinya

    2005-01-01

    A computer code TRACER (Transport Phenomena of Radionuclides for Accident Consequence Evaluation of Reactor) version 2.3 has been developed to evaluate species and quantities of fission products (FPs) released into cover gas during a fuel pin failure accident in an LMFBR. The TRACER version 2.3 includes new or modified models shown below. a) Both model: a new model for FPs release from fuel. b) Modified model for FPs transfer from fuel to bubbles or sodium coolant. c) Modified model for bubbles dynamics in coolant. Computational models, input data and output data of the TRACER version 2.3 are described in this user's manual. (author)

  11. Fuel conditioning facility electrorefiner cadmium vapor trap operation

    International Nuclear Information System (INIS)

    Vaden, D. E.

    1998-01-01

    Processing sodium-bonded spent nuclear fuel at the Fuel Conditioning Facility at Argonne National Laboratory-West involves an electrometallurgical process employing a molten LiCl-KCl salt covering a pool of molten cadmium. Previous research has shown that the cadmium dissolves in the salt as a gas, diffuses through the salt layer and vaporizes at the salt surface. This cadmium vapor condenses on cool surfaces, causing equipment operation and handling problems. Using a cadmium vapor trap to condense the cadmium vapors and reflux them back to the electrorefiner has mitigated equipment problems and improved electrorefiner operations

  12. Salt ingestion caves.

    Directory of Open Access Journals (Sweden)

    Lundquist Charles A.

    2006-01-01

    Full Text Available Large vertebrate herbivores, when they find a salt-bearing layer of rock, say in a cliff face, can produce sizable voids where, overgenerations, they have removed and consumed salty rock. The cavities formed by this natural animal process constitute a uniqueclass of caves that can be called salt ingestion caves. Several examples of such caves are described in various publications. Anexample in Mississippi U.S.A., Rock House Cave, was visited by the authors in 2000. It seems to have been formed by deer orbison. Perhaps the most spectacular example is Kitum Cave in Kenya. This cave has been excavated to a length over 100 metersby elephants. An ancient example is La Cueva del Milodon in Chile, which is reported to have been excavated by the now extinctmilodon, a giant ground sloth. Still other possible examples can be cited. This class of caves deserves a careful definition. First, thecavity in rock should meet the size and other conventions of the locally accepted definition of a cave. Of course this requirement differsin detail from country to country, particularly in the matter of size. The intent is to respect the local conventions. The characteristicthat human entry is possible is judged to be a crucial property of any recognized cave definition. Second, the cavity should besignificantly the result of vertebrate animal consumption of salt-bearing rock. The defining process is that rock removed to form thecave is carried away in the digestive track of an animal. While sodium salts are expected to be the norm, other salts for which thereis animal hunger are acceptable. Also some other speleogenesis process, such as solution, should not be excluded as long as it issecondary in formation of a cave in question.

  13. Thorium converter (ThorCon) - a doable molten salt reactor

    International Nuclear Information System (INIS)

    Myneni, Ganapati

    2015-01-01

    ThorCon mass-producible nuclear power plants are being built to generate electricity cheaper than coal, at a scale to make a real improvement in world poverty and environment, now. ThorCon irradiated materials and fuel salt are designed to be replaced in four-year cycles with no impact on electricity generation. This flex-fuel plant and its replaceable reactor cans can operate with mixtures of thorium and uranium at multiple enrichments. Fuel salt can be NaF/BeF 2 or LiF/BeF 2 if available. ThorCon's design exceeds current nuclear power safety practice. The team calls for regulatory participation in rigorous testing of a full-scale prototype to develop licensing guidance

  14. Reprocessing method for spent fuel

    International Nuclear Information System (INIS)

    Fujie, Makoto; Shoji, Yuichi; Kobayashi, Tsuguyuki.

    1997-01-01

    After reducing oxides of uranium (U), plutonium (Pu) and miner actinides in spent fuels by magnesium (Mg) in a molten salt, rear earth element oxides and salts of alkali metals and alkaline earth metals contained in the molten salt phase are separated and removed. Further, the Mg phase containing the reduced metals is evaporated to separate and remove Mg, thereby recovering U, Pu and minor actinides. In a lithium (Li) process, Li 2 O also generated in the reduction step is regenerated to Li simultaneously, and the reduction is conducted while suppressing the Li 2 O concentration in the molten salt low. This can improve the reduction rate of oxides of U, Pu and minor actinides compared with conventional cases. Since Li 2 O is regenerated into Li in the reduction step of the Li process, deposited Li 2 O is not carried to an electrolysis purification step, and recovering rate of U, Pu and minor actinides is not lowered. (T.M.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, Dane; Eapen, Jacob

    2013-10-01

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

  16. Characteristics analysis of salt vacuum distillation equipment

    International Nuclear Information System (INIS)

    Im, Hun Suk; Oh, Seung Chul; Hong, Sun Seok; Hur, Jin Mok; Lee, Hyo Jik

    2016-01-01

    A new technique for pyroprocessing was designed by adding an oxide reduction process to the previous one. It is regarded as a promising process to treat and recycle oxide spent fuels owing to its enhanced nuclear proliferation resistance and the simplified process equipment and the low process costing. Spent oxide fuel is reduced into a metal by an electrochemical method while using a high-temperature molten salt as the reaction medium. After being subjected to electrorefining and electrowinning processes, the reduced metal fuel can be used in sodium-cooled fast reactors. The salt vacuum distillation process termed cathode processing follows the oxide reduction stage and has been developed to remove the residual salt, allowing for clear fuel metal to be supplied to the next step, which is electrorefining. KAERI has manufactured this apparatus in several sizes and has been able to achieve a fuel recovery rate of 95%. However it is very difficult to scale up the equipment. Because all transport phenomena, including heat transfer and fluid flow, depend on the size and structure of the apparatus used. The ideal method for overcoming this issue is nondimensionalization, which allows one to determine the characteristic properties of a system. A comparison of the dimensionless variables corresponding to the M-type and P-type apparatuses performed on the basis of phase-transition phenomena as well as the results of the above-mentioned analysis elucidated the differences between the two apparatuses. It also means that the structure of the nozzle throat can be the one of the several causes for the recovery performance. First, the standard model (i.e., the M-type apparatus) was analyzed using dimensionless parameters. The characteristics of this apparatus were the following: 1) the diameter of the outlet of the nozzle throat was twice that of the inlet, 2) the ratio of the length to the diameter (L/D) was 8, and 3) the modified heat-transfer factor was 220-270. It indicates

  17. Characteristics analysis of salt vacuum distillation equipment

    Energy Technology Data Exchange (ETDEWEB)

    Im, Hun Suk; Oh, Seung Chul; Hong, Sun Seok; Hur, Jin Mok; Lee, Hyo Jik [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    A new technique for pyroprocessing was designed by adding an oxide reduction process to the previous one. It is regarded as a promising process to treat and recycle oxide spent fuels owing to its enhanced nuclear proliferation resistance and the simplified process equipment and the low process costing. Spent oxide fuel is reduced into a metal by an electrochemical method while using a high-temperature molten salt as the reaction medium. After being subjected to electrorefining and electrowinning processes, the reduced metal fuel can be used in sodium-cooled fast reactors. The salt vacuum distillation process termed cathode processing follows the oxide reduction stage and has been developed to remove the residual salt, allowing for clear fuel metal to be supplied to the next step, which is electrorefining. KAERI has manufactured this apparatus in several sizes and has been able to achieve a fuel recovery rate of 95%. However it is very difficult to scale up the equipment. Because all transport phenomena, including heat transfer and fluid flow, depend on the size and structure of the apparatus used. The ideal method for overcoming this issue is nondimensionalization, which allows one to determine the characteristic properties of a system. A comparison of the dimensionless variables corresponding to the M-type and P-type apparatuses performed on the basis of phase-transition phenomena as well as the results of the above-mentioned analysis elucidated the differences between the two apparatuses. It also means that the structure of the nozzle throat can be the one of the several causes for the recovery performance. First, the standard model (i.e., the M-type apparatus) was analyzed using dimensionless parameters. The characteristics of this apparatus were the following: 1) the diameter of the outlet of the nozzle throat was twice that of the inlet, 2) the ratio of the length to the diameter (L/D) was 8, and 3) the modified heat-transfer factor was 220-270. It indicates

  18. Salt supply to and significance of asymmetric salt diapirs

    DEFF Research Database (Denmark)

    Koyi, H.; Burliga, S.; Chemia, Zurab

    2012-01-01

    Salt diapirs can be asymmetric both internally and externally reflecting their evolution history. As such, this asymmetry bear a significant amount of information about the differential loading (± lateral forces) and in turn the salt supply that have shaped the diapir. In two dimensions......, In this study we compare results of analogue and numerical models of diapirs with two natural salt diapris (Klodawa and Gorleben diapirs) to explain their salt supply and asymmetric evolution. In a NW-SE section, the Gorleben salt diapir possesses an asymmetric external geometry represented by a large...... southeastern overhang due to salt extrusion during Middle Cretaceous followed by its burial in Tertiary. This external asymmetry is also reflected in the internal configuration of the diapir which shows different rates of salt flow on the two halves of the structure. The asymmetric external and internal...

  19. Fuel assemblies

    International Nuclear Information System (INIS)

    Mukai, Hideyuki

    1987-01-01

    Purpose: To prevent bending of fuel rods caused by the difference of irradiation growth between coupling fuel rods and standards fuel rods thereby maintain the fuel rod integrity. Constitution: The f value for a fuel can (the ratio of pole of zirconium crystals in the entire crystals along the axial direction of the fuel can) of a coupling fuel rod secured by upper and lower tie plates is made smaller than the f value for the fuel can of a standard fuel rod not secured by the upper and the lower tie plates. This can make the irradiation growth of the fuel can of the coupling fuel rod greater than the irradiation growth of the fuel can of the standard fuel rod and, accordingly, since the elongation of the standard fuel rod can always by made greater, bending of the standard fuel rod can be prevented. (Yoshihara, M.)

  20. Version control with Git

    CERN Document Server

    Loeliger, Jon

    2012-01-01

    Get up to speed on Git for tracking, branching, merging, and managing code revisions. Through a series of step-by-step tutorials, this practical guide takes you quickly from Git fundamentals to advanced techniques, and provides friendly yet rigorous advice for navigating the many functions of this open source version control system. This thoroughly revised edition also includes tips for manipulating trees, extended coverage of the reflog and stash, and a complete introduction to the GitHub repository. Git lets you manage code development in a virtually endless variety of ways, once you understand how to harness the system's flexibility. This book shows you how. Learn how to use Git for several real-world development scenarios ; Gain insight into Git's common-use cases, initial tasks, and basic functions ; Use the system for both centralized and distributed version control ; Learn how to manage merges, conflicts, patches, and diffs ; Apply advanced techniques such as rebasing, hooks, and ways to handle submodu...

  1. Global Historical Climatology Network (GHCN), Version 1 (Version Superseded)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Please note, this dataset has been superseded by a newer version (see below). Users should not use this version except in rare cases (e.g., when reproducing previous...

  2. COSY INFINITY version 8

    International Nuclear Information System (INIS)

    Makino, Kyoko; Berz, Martin

    1999-01-01

    The latest version of the particle optics code COSY INFINITY is presented. Using Differential Algebraic (DA) methods, the code allows the computation of aberrations of arbitrary field arrangements to in principle unlimited order. Besides providing a general overview of the code, several recent techniques developed for specific applications are highlighted. These include new features for the direct utilization of detailed measured fields as well as rigorous treatment of remainder bounds

  3. EASI graphics - Version II

    International Nuclear Information System (INIS)

    Allensworth, J.A.

    1984-04-01

    EASI (Estimate of Adversary Sequence Interruption) is an analytical technique for measuring the effectiveness of physical protection systems. EASI Graphics is a computer graphics extension of EASI which provides a capability for performing sensitivity and trade-off analyses of the parameters of a physical protection system. This document reports on the implementation of the Version II of EASI Graphics and illustrates its application with some examples. 5 references, 15 figures, 6 tables

  4. Test report : alternative fuels propulsion durability evaluation

    Science.gov (United States)

    2012-08-28

    This document, prepared by Honeywell Aerospace, Phoenix, AZ (Honeywell), contains the final : test report (public version) for the U.S. Department of Transportation/Federal Aviation : Administration (USDOT/FAA) Alternative Fuels Propulsion Engine Dur...

  5. Test procedures for salt rock

    International Nuclear Information System (INIS)

    Dusseault, M.B.

    1985-01-01

    Potash mining, salt mining, design of solution caverns in salt rocks, disposal of waste in salt repositories, and the use of granular halite backfill in underground salt rock mines are all mining activities which are practised or contemplated for the near future. Whatever the purpose, the need for high quality design parameters is evident. The authors have been testing salt rocks in the laboratory in a number of configurations for some time. Great care has been given to the quality of sample preparation and test methodology. This paper describes the methods, presents the elements of equipment design, and shows some typical results

  6. User's manual and guide to SALT3 and SALT4: two-dimensional computer codes for analysis of test-scale underground excavations for the disposal of radioactive waste in bedded salt deposits

    International Nuclear Information System (INIS)

    Lindner, E.N.; St John, C.M.; Hart, R.D.

    1984-02-01

    SALT3 and SALT4 are two-dimensional analytical/displacement-discontinuity codes designed to evaluate temperatures, deformation, and stresses associated with underground disposal of radioactive waste in bedded salt. These codes were developed by the University of Minnesota for the Office of Nuclear Waste Isolation in 1979. The present documentation describes the mathematical equations of the physical system being modeled, the numerical techniques utilized, and the organization of these computer codes. The SALT3 and SALT4 codes can simulate: (a) viscoelastic behavior in pillars adjacent to excavations; (b) transversely isotropic elastic moduli such as those exhibited by bedded or stratified rock; and (c) excavation sequence. Major advantages of these codes are: (a) computational efficiency; (b) the small amount of input data required; and (c) a creep law based on laboratory experimental data for salt. The main disadvantage is that some of the assumptions in the formulation of the codes, i.e., the homogeneous elastic half-space and temperature-independent material properties, render it unsuitable for canister-scale analysis or analysis of lateral deformation of the pillars. The SALT3 and SALT4 codes can be used for parameter sensitivity analyses of two-dimensional, repository-scale, thermomechanical response in bedded salt during the excavation, operational, and post-closure phases. It is especially useful in evaluating alternative patterns and sequences of excavation or waste canister placement. SALT3 is a refinement of an earlier code, SALT, and includes a fully anelastic creep model and thermal stress routine. SALT4 is a later version, and incorporates a revised creep model which is strain-hardening

  7. Engineering development studies for molten-salt breeder reactor processing No. 22

    International Nuclear Information System (INIS)

    Hightower, J.R. Jr.

    1976-06-01

    Processing methods are being developed for use in a close-coupled facility for removing fission products, corrosion products, and fissile materials from the MSBR fuel. This report discusses the autoresistance heating for the continuous fluorinator, the metal transfer experiment, experiments for the salt-metal contactor, and fuel reconstitution. 10 fig

  8. Thermochemical Properties of Nicotine Salts

    Directory of Open Access Journals (Sweden)

    Riggs DM

    2014-12-01

    Full Text Available The thermal gravimetric analysis (TGA and differential scanning calorimetry (DSC results presented in this report clearly show that the thermal stability and the endothermic peak nicotine release temperatures are different for different nicotine salts and these temperatures appear to be linked to the general microstructural details of the salt itself. In addition, the peak nicotine release temperatures are highly dependent upon the sample size used. The heat of vaporization for neat (non-protonated nicotine is also sample-size dependent. The TGA data showed that the least stable of the salts tested at elevated temperatures was the liquid salt nicotine triacetate followed by the crystalline materials (e.g., nicotine gallate and finally, the amorphous salts (e.g., nicotine alginate. The DSC results revealed that the liquid and crystalline salts exhibit nicotine release endotherms that are strongly related to the sample weight being tested. The amorphous salts show nicotine endotherm peak temperatures that are nearly independent of the sample weight. The range of peak nicotine release temperatures varied depending upon the specific salts and the sample size from 83 oC to well over 200 oC. Based on these results, the evolution of nicotine from the nicotine salt should be expected to vary based on the composition of the salt, the details of its microstructure, and the amount of nicotine salt tested.

  9. A guide to introducing burnup credit, preliminary version (English translation)

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Suyama, Kenya; Ryufuku, Susumu

    2017-06-01

    There is an ongoing discussion on the application of burnup credit to the criticality safety controls of facilities that treat spent fuels. With regard to such application of burnup credit in Japan, this document summarizes the current technical status of the prediction of the isotopic composition and criticality of spent fuels, as well as safety evaluation concerns and the current status of legal affairs. This report is an English translation of A Guide to Introducing Burnup Credit, Preliminary Version, originally published in Japanese as JAERI-Tech 2001-055 by the Nuclear Fuel Cycle Facility Safety Research Committee. (author)

  10. The FARC fuel archive of VVER

    International Nuclear Information System (INIS)

    Zizin, M.N.; Parfenova, N.A.; Proselkov, V.N.; Shishkov, L.K.

    1998-01-01

    The principles of organisation are explained and the structure of the FARC fuel archive for VVER reactors is described. The objective of the archive is accumulation of fuel data, data storage and obtaining the fuel using characteristics. The working version of fuel archive on 01.07.98 is realised, in which the data tables for fuel assemblies for 169 VVER-440 cycles and 35 VVER-1000 cycles are stored. There are two different versions of fuel archive - for VVER-440 (FARC) and for VVER-1000 (FARC1000). A structure of some tables and the texts of programs for them differ. The algorithms and codes for checking integrity, reasonableness and reliability of fuel archive data are developed. (author)

  11. Roadmap for disposal of Electrorefiner Salt as Transuranic Waste.

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, Robert P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Trone, Janis R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kalinina, Elena Arkadievna [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wang, Yifeng [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Lawrence C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-12-01

    The experimental breeder reactor (EBR-II) used fuel with a layer of sodium surrounding the uranium-zirconium fuel to improve heat transfer. Disposing of EBR-II fuel in a geologic repository without treatment is not prudent because of the potentially energetic reaction of the sodium with water. In 2000, the US Department of Energy (DOE) decided to treat the sodium-bonded fuel with an electrorefiner (ER), which produces metallic uranium product, a metallic waste, mostly from the cladding, and the salt waste in the ER, which contains most of the actinides and fission products. Two waste forms were proposed for disposal in a mined repository; the metallic waste, which was to be cast into ingots, and the ER salt waste, which was to be further treated to produce a ceramic waste form. However, alternative disposal pathways for metallic and salt waste streams may reduce the complexity. For example, performance assessments show that geologic repositories can easily accommodate the ER salt waste without treating it to form a ceramic waste form. Because EBR-II was used for atomic energy defense activities, the treated waste likely meets the definition of transuranic waste. Hence, disposal at the Waste Isolation Pilot Plant (WIPP) in southern New Mexico, may be feasible. This report reviews the direct disposal pathway for ER salt waste and describes eleven tasks necessary for implementing disposal at WIPP, provided space is available, DOE decides to use this alternative disposal pathway in an updated environmental impact statement, and the State of New Mexico grants permission.

  12. Liquid fuel concept benefits

    International Nuclear Information System (INIS)

    Hron, M.

    1996-01-01

    There are principle drawbacks of any kind of solid nuclear fuel listed and analyzed in the first part of the paper. One of the primary results of the analyses performed shows that the solid fuel concept, which was to certain degree advantageous in the first periods of a nuclear reactor development and operation, has guided this branch of a utilization of atomic nucleus energy to a death end. On the background of this, the liquid fuel concept and its benefits are introduced and briefly described in the first part of the paper, too. As one of the first realistic attempts to utilize the advantages of liquid fuels, the reactor/blanket system with molten fluoride salts in the role of fuel and coolant simultaneously, as incorporated in the accelerator-driven transmutation technology (ADTT) being proposed and currently having been under development in the Los Alamos National Laboratory, will be studied both theoretically and experimentally. There is a preliminary design concept of an experimental assembly LA-O briefly introduced in the paper which is under preparation in the Czech Republic for such a project. Finally, there will be another very promising concept of a small low power ADTT system introduced which is characterized by a high level of safety and economical efficiency. In the conclusion, the overall survey of principal benefits which may be expected by introducing liquid nuclear fuel in nuclear power and research reactor systems is given and critically analyzed. 7 refs, 4 figs

  13. The material flow of salt

    International Nuclear Information System (INIS)

    Kostick, D.S.

    1993-01-01

    Salt (NaCl) is a universal mineral commodity used by virtually every person in the world. Although a very common mineral today, at one time it was considered as precious as gold in certain cultures. This study traces the material flow of salt from its origin through the postconsumer phase of usage. The final disposition of salt in the estimated 14,000 different uses, grouped into several macrocategories, is traced from the dispersive loss of salt into the environment to the ultimate disposal of salt-base products into the waste stream after consumption. The base year for this study is 1990, in which an estimated 196 million short tons of municipal solid waste was discarded by the US population. Approximately three-fourths of domestic salt consumed is released to the environment and unrecovered while about one-fourth is discharged to landfills and incinerators as products derived from salt. Cumulative historical domestic production, trade, and consumption data have been compiled to illustrate the long-term trends within the US salt industry and the cumulative contribution that highway deicing salt has had on the environment. Salt is an important component of drilling fluids in well drilling. It is used to flocculate and to increase the density of the drilling fluid in order to overcome high down-well gas pressures. Whenever drilling activities encounter salt formations, salt is added to the drilling fluid to saturate the solution and minimize the dissolution within the salt strata. Salt is also used to increase the set rate of concrete in cemented casings. This subsector includes companies engaged in oil, gas, and crude petroleum exploration and in refining and compounding lubricating oil. It includes SIC major groups 13 and 29. 13 refs., 14 figs., 6 tabs

  14. Tritium control and capture in salt-cooled fission and fusion reactors: Status, challenges, and path forward

    International Nuclear Information System (INIS)

    Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.; Whyte, Dennis G.; Scarlat, Raluca

    2017-01-01

    Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The Fluoride-salt-cooled High-temperature Reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the base-line salts contain lithium where isotopically separated "7Li is proposed to minimize tritium production from neutron interactions with the salt. The Chinese Academy of Science plans to start operation of a 2-MWt molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in "6Li is proposed to maximize tritium generation the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700 °C liquid salt systems. We describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data is the primary constraint for designing efficient cost-effective methods of tritium control.

  15. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael Mejias

    2016-01-01

    This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

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

    International Nuclear Information System (INIS)

    Shafi, Suheel; Prabhakara Reddy, B.; Perumal, S.V.; Nagarajan, K.

    2014-01-01

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

  17. The Unified Extensional Versioning Model

    DEFF Research Database (Denmark)

    Asklund, U.; Bendix, Lars Gotfred; Christensen, H. B.

    1999-01-01

    Versioning of components in a system is a well-researched field where various adequate techniques have already been established. In this paper, we look at how versioning can be extended to cover also the structural aspects of a system. There exist two basic techniques for versioning - intentional...

  18. Rehabilitation of saline ecosystems through cultivation of salt tolerant plants

    International Nuclear Information System (INIS)

    Abdul, R.; Mahmood, K.

    2012-01-01

    In Pakistan, salt-affected regions have been drastically disturbed by unchecked activities of local populations. Removal of deep-rooted perennials and overgrazing destroy the native vegetation leading to rapid desertification. Shallow-rooted agricultural crops are grown on marginal soils on limited area that is not enough with respect to the spread of salinity problem. Sustainable restoration of these ecosystems requires a large scale integration of perennial plants (trees, shrubs and herbs) back in to farming systems. However, selenization processes continue because the available options for cultivation of perennial plants prove less profitable than agricultural crops. This study relates to resort the salt-affected lands for plant production and develop a technology for sustainable saline ecosystem. Plants, having salt tolerance potential, have been identified and introduced on salt-affected wastelands to develop a sustainable ecosystem with increased productivity. The biomass so produced can be used directly as forage, fuel, and even as food or feed. In addition, fish aquaculture, and some value-added products make this ecosystem more sustainable. This technology is practically demonstrated at Biosaline Research Station of Nuclear Institute for Agriculture and Biology (NIAB), Pakka Anna, Faisalabad, Pakistan. The marginally saline soils and wastelands ameliorated as a result of growing salt tolerant perennials can also be used for growing salt tolerant cultivars of conventional crops like wheat, barley and mustard. So, through proper management the saline ecosystem can become economical and profitable. (author)

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

    Science.gov (United States)

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

    2018-04-01

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

  20. PVWatts Version 5 Manual

    Energy Technology Data Exchange (ETDEWEB)

    Dobos, A. P.

    2014-09-01

    The NREL PVWatts calculator is a web application developed by the National Renewable Energy Laboratory (NREL) that estimates the electricity production of a grid-connected photovoltaic system based on a few simple inputs. PVWatts combines a number of sub-models to predict overall system performance, and makes includes several built-in parameters that are hidden from the user. This technical reference describes the sub-models, documents assumptions and hidden parameters, and explains the sequence of calculations that yield the final system performance estimate. This reference is applicable to the significantly revised version of PVWatts released by NREL in 2014.

  1. Where Does Road Salt Go - a Static Salt Model

    Science.gov (United States)

    Yu, C. W.; Liu, F.; Moriarty, V. W.

    2017-12-01

    Each winter, more than 15 million tons of road salt is applied in the United States for the de-icing purpose. Considerable amount of chloride in road salt flows into streams/drainage systems with the snow melt runoff and spring storms, and eventually goes into ecologically sensitive low-lying areas in the watershed, such as ponds and lakes. In many watersheds in the northern part of US, the chloride level in the water body has increased significantly in the past decades, and continues an upward trend. The environmental and ecological impact of the elevated chloride level can no longer be ignored. However although there are many studies on the biological impact of elevated chloride levels, there are few investigations on how the spatially distributed road salt application affects various parts of the watershed. In this presentation, we propose a static road salt model as a first-order metric to address spacial distribution of salt loading. Derived from the Topological Wetness Index (TWI) in many hydrological models, this static salt model provides a spatial impact as- sessment of road salt applications. To demonstrate the effectiveness of the static model, National Elevation Dataset (NED) of ten-meter resolution of Lake George watershed in New York State is used to generate the TWI, which is used to compute a spatially dis- tributed "salt-loading coefficient" of the whole watershed. Spatially varying salt applica- tion rate is then aggregated, using the salt-loading coefficients as weights, to provide salt loading assessments of streams in the watershed. Time-aggregated data from five CTD (conductivity-temperature-depth) sensors in selected streams are used for calibration. The model outputs and the sensor data demonstrate a strong linear correlation, with the R value of 0.97. The investigation shows that the static modeling approach may provide an effective method for the understanding the input and transport of road salt to within watersheds.

  2. Pyroelectrochemical process for reprocessing irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Brambilla, G.; Sartorelli, A.

    1982-01-01

    A pyroelectrochemical process for reprocessing irradiated fast reactor mixed oxide or carbide fuels is described. The fuel is dissolved in a bath of molten alkali metal sulfates. The Pu(SO 4 ) 2 formed in the bath is thermally decomposed, leaving crystalline PuO 2 on the bottom of the reaction vessel. Electrodes are then introduced into the bath, and UO 2 is deposited on the cathode. Alternatively, both UO 2 and PuO 2 may be electrodeposited. The molten salts, after decontamination by precipitating the fission products dissolved in the bath by introducing basic agents such as oxides, carbonates, or hydroxides, may be recycled. Since it is not possible to remove cesium from the molten salt bath, periodic disposal and partial renewal with fresh salts is necessary. The melted salts that contain the fission products are conditioned for disposal by embedding them in a metallic matrix

  3. Multilayer Porous Crucibles for the High Throughput Salt Separation from Uranium Deposits

    International Nuclear Information System (INIS)

    Kwon, S. W.; Park, K. M.; Kim, J. G.; Kim, I. T.; Seo, B. K.; Moon, J. G.

    2013-01-01

    Solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. A physical separation process, such as a distillation separation, is more attractive than a chemical or dissolution process because physical processes generate much less secondary process. Distillation process was employed for the cathode processsing due to the advantages of minimal generation of secondary waste, compact unit process, simple and low cost equipment. The basis for vacuum distillation separation is the difference in vapor pressures between salt and uranium. A solid cathode deposit is heated in a heating region and salt vaporizes, while nonvolatile uranium remains behind. It is very important to increase the throughput of the salt separation system owing to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. The evaporation rate of the LiCl-KCl eutectic salt in vacuum distiller is not so high to come up with the generation capacity of uranium dendrites in an electro-refiner. Therefore, a wide evaporation area or high distillation temperature is necessary for the successful salt separation. In this study, it was attempted to enlarge a throughput of the salt distiller with a multilayer porous crucibles for the separation of adhered salt in the uranium deposits generated from the electrorefiner. The feasibility of the porous crucibles was tested by the salt distillation experiments. In this study, the salt distiller with multilayer porous crucibles was proposed and the feasibility of liquid salt separation was examined to increase a throughput. It was found that the effective separation of salt from uranium deposits was possible by the multilayer porous crucibles

  4. Multilayer Porous Crucibles for the High Throughput Salt Separation from Uranium Deposits

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, S. W.; Park, K. M.; Kim, J. G.; Kim, I. T.; Seo, B. K.; Moon, J. G. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    Solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. A physical separation process, such as a distillation separation, is more attractive than a chemical or dissolution process because physical processes generate much less secondary process. Distillation process was employed for the cathode processsing due to the advantages of minimal generation of secondary waste, compact unit process, simple and low cost equipment. The basis for vacuum distillation separation is the difference in vapor pressures between salt and uranium. A solid cathode deposit is heated in a heating region and salt vaporizes, while nonvolatile uranium remains behind. It is very important to increase the throughput of the salt separation system owing to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. The evaporation rate of the LiCl-KCl eutectic salt in vacuum distiller is not so high to come up with the generation capacity of uranium dendrites in an electro-refiner. Therefore, a wide evaporation area or high distillation temperature is necessary for the successful salt separation. In this study, it was attempted to enlarge a throughput of the salt distiller with a multilayer porous crucibles for the separation of adhered salt in the uranium deposits generated from the electrorefiner. The feasibility of the porous crucibles was tested by the salt distillation experiments. In this study, the salt distiller with multilayer porous crucibles was proposed and the feasibility of liquid salt separation was examined to increase a throughput. It was found that the effective separation of salt from uranium deposits was possible by the multilayer porous crucibles.

  5. Nuclear data covariances and sensitivity analysis, validation of a methodology based on the perturbation theory; application to an innovative concept: the molten thorium salt fueled reactor; Analyses de sensibilite et d'incertitude de donnees nucleaires. Contribution a la validation d'une methodologie utilisant la theorie des perturbations; application a un concept innovant: reacteur a sels fondus thorium a spectre epithermique

    Energy Technology Data Exchange (ETDEWEB)

    Bidaud, A

    2005-10-15

    Neutron transport simulation of nuclear reactors is based on the knowledge of the neutron-nucleus interaction (cross-sections, fission neutron yields and spectra...) for the dozens of nuclei present in the core over a very large energy range (fractions of eV to several MeV). To obtain the goal of the sustainable development of nuclear power, future reactors must have new and more strict constraints to their design: optimization of ore materials will necessitate breeding (generation of fissile material from fertile material), and waste management will require transmutation. Innovative reactors that could achieve such objectives - generation IV or ADS (accelerator driven system) - are loaded with new fuels (thorium, heavy actinides) and function with neutron spectra for which nuclear data do not benefit from 50 years of industrial experience, and thus present particular challenges. After validation on an experimental reactor using an international benchmark, we take classical reactor physics tools along with available nuclear data uncertainties to calculate the sensitivities and uncertainties of the criticality and temperature coefficient of a thorium molten salt reactor. In addition, a study based on the important reaction rates for the calculation of cycle's equilibrium allows us to estimate the efficiency of different reprocessing strategies and the contribution of these reaction rates on the uncertainty of the breeding and then on the uncertainty of the size of the reprocessing plant. Finally, we use this work to propose an improvement of the high priority experimental request list. (author)

  6. URGENCES NOUVELLE VERSION

    CERN Multimedia

    Medical Service

    2002-01-01

    The table of emergency numbers that appeared in Bulletin 10/2002 is out of date. The updated version provided by the Medical Service appears on the following page. Please disregard the previous version. URGENT NEED OF A DOCTOR GENEVAPATIENT NOT FIT TO BE MOVED: Call your family doctor Or SOS MEDECINS (24H/24H) 748 49 50 Or ASSOC. OF GENEVA DOCTORS (7H-23H) 322 20 20 PATIENT CAN BE MOVED: HOPITAL CANTONAL 24 Micheli du Crest 372 33 11 382 33 11 CHILDREN'S HOSPITAL 6 rue Willy Donzé 382 68 18 382 45 55 MATERNITY 24 Micheli du Crest 382 68 16 382 33 11 OPHTALMOLOGY 22 Alcide Jentzer 382 84 00 HOPITAL DE LA TOUR Meyrin 719 61 11 CENTRE MEDICAL DE MEYRIN Champs Fréchets 719 74 00 URGENCES : FIRE BRIGADE 118 FIRE BRIGADE CERN 767 44 44 BESOIN URGENT D'AMBULANCE (GENEVE ET VAUD) : 144 POLICE 117 ANTI-POISON CENTRE 24H/24H 01 251 51 510 EUROPEAN EMERGENCY CALL: 112 FRANCE PATIENT NOT FIT TO BE MOVED: call your family doctor PATIENT CAN BE MOVED: ST. JULIE...

  7. Recent development of BOA version 3

    International Nuclear Information System (INIS)

    Deshon, J.; Hussey, D.; Westacott, J.; Young, M.; Secker, J.; Epperson, K.; McGurk, J.; Henshaw, J.

    2010-01-01

    BOA is the EPRI code developed to help understand material transport issues in PWR systems. In particular, BOA was developed to investigate the problem of crud induced power shifts (CIPS) or axial offset anomalies (AOA) that are related to crud accumulation on fuel clad and the subsequent uptake of boron within this crud. Over the last few years significant developments of BOA have taken place; these include new models for release and deposition of soluble and particulate material as well as new models for boron capture in the crud. The new version of BOA (version 3), as well as predicting the extent of AOA a plant might expect during a cycle, also predicts soluble and particulate Ni/Fe levels in the coolant, end of cycle Ni shut-down releases and the relative amounts of Ni/Fe within the fuel crud. The model is being validated against plant data and this paper briefly discusses the recent developments, the fundamental scientific basis for these changes and some of the validation studies. (author)

  8. Experimental studies on natural circulation in molten salt loops

    International Nuclear Information System (INIS)

    Srivastava, A.K.; Borgohain, A.; Maheshwari, N.K.; Vijayan, P.K.

    2015-01-01

    Molten salts are increasingly getting attention as a coolant and storage medium in solar thermal power plants and as a liquid fuel, blanket and coolant in Molten Salt Reactors (MSR’s). Two different test facilities named Molten Salt Natural Circulation Loop (MSNCL) and Molten Active Fluoride salt Loop (MAFL) have been setup for thermal hydraulics, instrument development and material related studies relevant to MSR and solar power plants. The working medium for MSNCL is a molten nitrate salt which is a mixture of NaNO 3 and KNO 3 in 60:40 ratio and proposed as one of the coolant option for molten salt based reactor and coolant as well as storage medium for solar thermal power application. On the other hand, the working medium for MAFL is a eutectic mixture of LiF and ThF 4 and proposed as a blanket salt for Indian Molten Salt Breeder Reactor (MSBR). Steady state natural circulation experiments at different power level have been performed in the MSNCL. Transient studies for startup of natural circulation, loss of heat sink, heater trip and step change in heater power have also been carried out in the same. A 1D code LeBENC, developed in-house to simulate the natural circulation characteristics in closed loops, has been validated with the experimental data obtained from MSNCL. Further, LeBENC has been used for Pretest analysis of MAFL. This paper deals with the description of both the loops and experimental studies carried out in MSNCL. Validation of LeBENC along with the pretest analysis of MAFL using the same are also reported in this paper. (author)

  9. Generic aspects of salt repositories

    International Nuclear Information System (INIS)

    Laughon, R.B.

    1979-01-01

    The history of geological disposal of radioactive wastes in salt is presented from 1957 when a panel of the National Academy of Sciences-National Research Council recommended burial in bedded salt deposits. Early work began in the Kansas, portion of the Permian Basin where simulated wastes were placed in an abandoned salt mine at Lyons, Kansas, in the late 1960's. This project was terminated when the potential effect of nearby solution mining activities could not be resolved. Evaluation of bedded salts resumed a few years later in the Permian Basin in southeastern New Mexico, and search for suitable sites in the 1970's resulted in the formation of the National Waste Terminal Storage Program in 1976. Evaluation of salt deposits in many regions of the United States has been virtually completed and has shown that deposits having the greatest potential for radioactive waste disposal are those of the largest depositional basins and salt domes of the Gulf Coast region

  10. Immobilization of LiCl-Li 2 O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Peterson, Jacob A.; Kroll, Jared O.; Frank, Steven M.

    2018-04-01

    In this study, salt occlusion and hydrothermal processes were used to make chlorosodalite through reaction with a high-LiCl salt simulating a waste stream following pyrochemical treatment of oxide-based used nuclear fuel. Some products were reacted with glass binders to increase chlorosodalite yield through alkali ion exchange and aide in densification. Hydrothermal processes included reaction of the salt simulant in an acid digestion vessel with either zeolite 4A or sodium aluminate and colloidal silica. Chlorosodalite yields in the crystalline products were nearly complete in the glass-bonded materials at values of 100 mass% for the salt-occlusion method, up to 99.0 mass% for the hydrothermal synthesis with zeolite 4A, and up to 96 mass% for the hydrothermal synthesis with sodium aluminate and colloidal silica. These results show promise for using chemically stable chlorosodalite to immobilize oxide reduction salt wastes.

  11. Immobilization of LiCl-Li2O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods

    Science.gov (United States)

    Riley, Brian J.; Peterson, Jacob A.; Kroll, Jared O.; Frank, Steven M.

    2018-04-01

    In this study, hydrothermal and salt-occlusion processes were used to make chlorosodalite through reactions with a high-LiCl salt simulating a waste stream generated from pyrochemical treatment of oxide-based used nuclear fuel. Some products were reacted with glass binders to increase chlorosodalite yield through alkali ion exchange and to aid in densification. Hydrothermal processes included reaction of the salt simulant in an autoclave with either zeolite 4A or sodium aluminate and colloidal silica. Chlorosodalite yields in the crystalline products were nearly complete in the glass-bonded materials at values of 100 mass% for the salt-occlusion method, up to 99.0 mass% for the hydrothermal synthesis with zeolite 4A, and up to 96 mass% for the hydrothermal synthesis with sodium aluminate and colloidal silica. These results show promise for using chemically stable chlorosodalite to immobilize oxide reduction salt wastes.

  12. Alternative Fuels

    Science.gov (United States)

    Alternative fuels include gaseous fuels such as hydrogen, natural gas, and propane; alcohols such as ethanol, methanol, and butanol; vegetable and waste-derived oils; and electricity. Overview of alternative fuels is here.

  13. Development of High Throughput Salt Separation System with Integrated Liquid Salt Separation - Salt Distillation Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Sangwoon; Park, K. M.; Kim, J. G.; Jeong, J. H.; Lee, S. J.; Park, S. B.; Kim, S. S.

    2013-01-15

    The capacity of a salt distiller should be sufficiently large to reach the throughput of uranium electro-refining process. In this study, an assembly composing a liquid separation sieve and a distillation crucible was developed for the sequential operation of a liquid salt separation and a vacuum distillation in the same tower. The feasibility of the sequential salt separation was examined by the rotation test of the sieve-crucible assembly and sequential operation of a liquid salt separation and a vacuum distillation. The adhered salt in the uranium deposits was removed successfully. The salt content in the deposits was below 0.1 wt% after the sequential operation of the liquid salt separation - salt distillation. From the results of this study, it could be concluded that efficient salt separation can be realized by the sequential operation of liquid salt separation and vacuum distillation in one distillation tower since the operation procedures are simplified and no extra operation of cooling and reheating is necessary.

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

    International Nuclear Information System (INIS)

    Williamson, M.A.; Venneri, F.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    1976-11-01

    Molten-salt corrosion loops MSR-FCL-3 and MSR-FCL-4 are high-temperature test facilities designed to evaluate corrosion and mass transfer of modified Hastelloy N alloys for future use in Molten-Salt Breeder Reactors. Salt is circulated by a centrifugal sump pump to evaluate material compatibility with LiF-BeF 2 -ThF 4 -UF 4 fuel salt at velocities up to 6 m/s (20 fps) and at salt temperatures from 566 to 705 0 C (1050 to 1300 0 F). The report presents the design description of the various components and systems that make up each corrosion facility, such as the salt pump, corrosion specimens, salt piping, main heaters, salt coolers, salt sampling equipment, and helium cover-gas system, etc. The electrical systems and instrumentation and controls are described, and operational procedures, system limitations, and maintenance philosophy are discussed

  16. LHCb - SALT, a dedicated readout chip for strip detectors in the LHCb Upgrade experiment

    CERN Multimedia

    Swientek, Krzysztof Piotr

    2015-01-01

    Silicon strip detectors in the upgraded Tracker of LHCb experiment will require a new readout 128-channel ASIC called SALT. It will extract and digitise analogue signals from the sensor, perform digital processing and transmit serial output data. SALT is designed in CMOS 130 nm process and uses a novel architecture comprising of analogue front-end and ultra-low power ($<$0.5 mW) fast (40 MSps) sampling 6-bit ADC in each channel. A prototype of first 8-channel version of SALT chip, comprising all important functionalities, was submitted. Its design and possibly first tests results will be presented.

  17. Zechstein salt Denmark. Vol. 1

    International Nuclear Information System (INIS)

    Lyngsie Jacobsen, F.; Soenderholm, M.; Springer, N.; Gutzon Larsen, J.; Lagoni, P.; Fabricius, J.

    1984-01-01

    The Salt Research Project EFP-81 has mainly been aiming upon an elucidation of the stratigraphy of the Danish Zechstein evaporites. Also an attempt to clarify the connection between the fabric and the strength of the strongly deformed domal rock salt is performed. The unravelling of the stratigraphy is carried out by means of renewed interpretations of new and old data from all the wells drilling in the Danish Permian basin in connection with a revaluation of the core descriptions. By means of trace elements analysis it is possible to some extent to distinguish between Zestein 1 and 2 ''grey salt''. A description of the transition zone between Zechstein 1 and 2 is carried out. New methods of fabric analyses are introduced and the strength measurements of the rock salt are treated statistically in connection with new defined rock salt parameters. An investigation of fluid inclusions in halite and quartz crystals from dome salt has resulted in the determination of salinity and chemical composition of the brines present in the salt. Temperatures and corresponding pressures during the evolution of the salt pillow and salt dome have been established. The dehydration conditions of natural carnallite in situ are clarified. (author)

  18. Salt formations offer disposal alternative

    International Nuclear Information System (INIS)

    Funderburk, R.

    1990-01-01

    This paper discusses how three U.S. firms are spending millions to permit and build underground disposal sites in salt formations. These companies claim salt is the ideal geological medium for holding hazardous wastes. Two Texas locations and one in Michigan have been targeted as future sites for hazardous waste disposal. The Michigan site, outside Detroit, is a former salt mine 2,000 feet beneath the Ford Motor Co. (Detroit) assembly works in Dearborn. Both Texas sites are atop salt domes---one east and one west of Houston

  19. Fuel conditioning facility electrorefiner start-up results

    International Nuclear Information System (INIS)

    Goff, K.M.; Mariani, R.D.; Vaden, D.; Bonomo, N.L.; Cunningham, S.S.

    1996-01-01

    At ANL-West, there are several thousand kilograms of metallic spent nuclear fuel containing bond sodium. This fuel will be treated in the Fuel Conditioning Facility (FCF) at ANL-West to produce stable waste forms for storage and disposal. The treatment operations will make use of an electrometallurgical process employing molten salts and liquid metals. The treatment equipment is presently undergoing testing with depleted uranium. Operations with irradiated fuel will commence when the environmental evaluation for FCF is complete

  20. Fuel assembly

    International Nuclear Information System (INIS)

    Chaki, Masao; Nishida, Koji; Karasawa, Hidetoshi; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Kashiwai, Shin-ichi; Masuhara, Yasuhiro

    1998-01-01

    The present invention concerns a fuel assembly, for a BWR type nuclear reactor, comprising fuel rods in 9 x 9 matrix. The inner width of the channel box is about 132mm and the length of the fuel rods which are not short fuel rods is about 4m. Two water rods having a circular cross section are arranged on a diagonal line in a portion of 3 x 3 matrix at the center of the fuel assembly, and two fuel rods are disposed at vacant spaces, and the number of fuel rods is 74. Eight fuel rods are determined as short fuel rods among 74 fuel rods. Assuming the fuel inventory in the short fuel rod as X(kg), and the fuel inventory in the fuel rods other than the short fuel rods as Y(kg), X and Y satisfy the relation: X + Y ≥ 173m, Y ≤ - 9.7X + 292, Y ≤ - 0.3X + 203 and X > 0. Then, even when the short fuel rods are used, the fuel inventory is increased and fuel economy can be improved. (I.N.)

  1. Fuel assembly

    International Nuclear Information System (INIS)

    Yamazaki, Hajime.

    1995-01-01

    In a fuel assembly having fuel rods of different length, fuel pellets of mixed oxides of uranium and plutonium are loaded to a short fuel rod. The volume ratio of a pellet-loaded portion to a plenum portion of the short fuel rod is made greater than the volume ratio of a fuel rod to which uranium fuel pellets are loaded. In addition, the volume of the plenum portion of the short fuel rod is set greater depending on the plutonium content in the loaded fuel pellets. MOX fuel pellets are loaded on the short fuel rods having a greater degree of freedom relevant to the setting for the volume of the plenum portion compared with that of a long rod fuel, and the volume of the plenum portion is ensured greater depending on the plutonium content. Even if a large amount of FP gas and He gas are discharged from the MOX fuels compared with that from the uranium fuels, the internal pressure of the MOX fuel rod during operation is maintained substantially identical with that of the uranium fuel rod, so that a risk of generating excess stresses applied to the fuel cladding tubes and rupture of fuels are greatly reduced. (N.H.)

  2. Process for removal of sulfur compounds from fuel gases

    Science.gov (United States)

    Moore, Raymond H.; Stegen, Gary E.

    1978-01-01

    Fuel gases such as those produced in the gasification of coal are stripped of sulfur compounds and particulate matter by contact with molten metal salt. The fuel gas and salt are intimately mixed by passage through a venturi or other constriction in which the fuel gas entrains the molten salt as dispersed droplets to a gas-liquid separator. The separated molten salt is divided into a major and a minor flow portion with the minor flow portion passing on to a regenerator in which it is contacted with steam and carbon dioxide as strip gas to remove sulfur compounds. The strip gas is further processed to recover sulfur. The depleted, minor flow portion of salt is passed again into contact with the fuel gas for further sulfur removal from the gas. The sulfur depleted, fuel gas then flows through a solid absorbent for removal of salt droplets. The minor flow portion of the molten salt is then recombined with the major flow portion for feed to the venturi.

  3. Neutronic study of a nuclear reactor of fused salts

    International Nuclear Information System (INIS)

    Garcia B, F. B.; Francois L, J. L.

    2012-10-01

    The reactors of fused salts called Molten Salt Reactor have presented a resurgence of interest in the last decade, due to they have a versatility in particular to operate, either with a thermal or fast neutrons spectrum. The most active development was by the middle of 1950 and principles of 1970 in the Oak Ridge National Laboratory. In this work some developed models are presented particularly and studied with the help of the MCNPX code, for the development of the neutronic study of this reactor, starting of proposed models and from a simple and homogeneous geometry until other more complex models and approximate to more real cases. In particular the geometry conditions and criticality of each model were analyzed, the isotopic balance, as well as the concentrations of the salts and different assigned fuel types. (Author)

  4. Status Report on Laboratory Testing and International Collaborations in Salt.

    Energy Technology Data Exchange (ETDEWEB)

    Kuhlman, Kristopher L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Matteo, Edward N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Reedlunn, Benjamin [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sobolik, Steven R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mills, Melissa Marie [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kirkes, Leslie Dawn [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Xiong, Yongliang [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Icenhower, Jonathan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-01

    This report is a summary of the international collaboration and laboratory work funded by the US Department of Energy Office of Nuclear Energy Spent Fuel and Waste Science & Technology (SFWST) as part of the Sandia National Laboratories Salt R&D work package. This report satisfies milestone levelfour milestone M4SF-17SN010303014. Several stand-alone sections make up this summary report, each completed by the participants. The first two sections discuss international collaborations on geomechanical benchmarking exercises (WEIMOS) and bedded salt investigations (KOSINA), while the last three sections discuss laboratory work conducted on brucite solubility in brine, dissolution of borosilicate glass into brine, and partitioning of fission products into salt phases.

  5. Molten salt treatment to minimize and optimize waste

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  6. Fused salt electrolysis

    International Nuclear Information System (INIS)

    Ares, Osvaldo; Botbol, Jose.

    1989-01-01

    Working conditions for zirconium preparation by fused salt electrolysis were studied. For such purpose, a cell was built for operation under argon atmosphere. A graphite crucible served as anode, with steel cathodes. Proper design allowed cathode rechange under the inert atmosphere. Cathodic deposits of zirconium powder occluded salts from the bath. After washing with both water and hydrochloric acid, the metallic powder was consolidated by fusion. Optimum operating conditions were found to arise from an electrolyte of 12% potassium hexafluorzirconate -88% sodium chloride, at 820 deg C and 5 A/cm 2 cathodic current density. Deposits contained 35% of metal and current efficiency reached 66%. The powder contained up to 600 ppm of chlorine and 1.700 ppm of fluorine; after fusion, those amounts decreased to 2 ppm and 3 ppm respectively, with low proportion of metallic impurities. Though oxygen proportion was 4.500 ppm, it should be lowered by improving working conditions, as well as working on an ampler scale. (Author)

  7. An optimized symbiotic fusion and molten-salt fission reactor system

    International Nuclear Information System (INIS)

    Blinkin, V.L.; Novikov, V.M.

    A symbiotic fusion-fission reactor system which breeds nuclear fuel is discussed. In the blanket of the controlled thermonuclear reactor (CTR) uranium-233 is generated from thorium, which circulates in the form of ThF 4 mixed with molten sodium and beryllium fluorides. The molten-salt fission reactor (MSR) burns up the uranium-233 and generates tritium for the fusion reactor from lithium, which circulates in the form of LiF mixed with BeF 2 and 233 UF 4 through the MSR core. With a CTR-MSR thermal power ratio of 1:11 the system can produce electrical energy and breed fuel with a doubling time of 4-5 years. The system has the following special features: (1) Fuel reprocessing is much simpler and cheaper than for contemporary fission reactors; reprocessing consists simply in continuous removal of 233 U from the salt circulating in the CTR blanket by the fluorination method and removal of xenon from the MSR fuel salt by gas scavenging; the MSR fuel salt is periodically exchanged for fresh salt and the 233 U is then removed from it; (2) Tritium is produced in the fission reactor, which is a much simpler system than the fusion reactor; (3) The CTR blanket is almost ''clean''; no tritium is produced in it and fission fragment activity does not exceed the activity induced in the structural materials; (4) Almost all the thorium introduced into the CTR blanket can be used for producing 233 U

  8. Nuclear fuels

    International Nuclear Information System (INIS)

    Gangwani, Saloni; Chakrabortty, Sumita

    2011-01-01

    Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

  9. Development of advanced spent fuel management process

    International Nuclear Information System (INIS)

    Shin, Young Joon; Cho, S. H.; You, G. S.

    2001-04-01

    Currently, the economic advantage of any known approach to the back end fuel cycle of a nuclear power reactor has not been well established. Thus the long term storage of the spent fuel in a safe manner is one of the important issues to be resolved in countries where the nuclear power has a relatively heavy weight in power production of that country. At KAERI, as a solution to this particular issue midterm storage of the spent fuel, an alternative approach has been developed. This approach includes the decladding and pulverization process of the spent PWR fuel rod, the reducing process from the uranium oxide to a metallic uranium powder using Li metal in a LiCl salt, the continuous casting process of the reduced metal, and the recovery process of Li from mixed salts by the electrolysis. We conducted the laboratory scale tests of each processes for the technical feasibility and determination for the operational conditions for this approach. Also, we performed the theoretical safety analysis and conducted integral tests for the equipment integration through the Mock-up facility with non-radioactive samples. There were no major issues in the approach, however, material incompatibility of the alkaline metal and oxide in a salt at a high temperature and the reactor that contains the salt became a show stopper of the process. Also the difficulty of the clear separation of the salt with metals reduced from the oxide became a major issue

  10. Fuel and nuclear fuel cycle

    International Nuclear Information System (INIS)

    Prunier, C.

    1998-01-01

    The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

  11. Experiments in connection with Salt Domes

    NARCIS (Netherlands)

    Escher, B.G.; Kuenen, Ph.H.

    1928-01-01

    The different theories concerning the origin of Salt Domes in Roumania, Germany, Texas, Louisiana, Colorado and Utah are discussed. In Roumania the salt occurs in cores of “Diapir” anticlines. The existance of hills of salt indicates, that the salt is still pushing upwards. In Germany the salt

  12. Salt disposal: Paradox Basin, Utah

    International Nuclear Information System (INIS)

    1983-04-01

    This report presents the findings of a study conducted for the National Waste Terminal Storage (NWTS) Program. Permanent disposal options are examined for salt resulting from the excavation of a waste repository in the bedded salt deposits of the Paradox Basin of southeastern Utah. The study is based on a repository salt backfill compaction of 60% of the original density which leaves a total of 8 million tons of 95% pure salt to be disposed of over a 30-year period. The feasibility, impacts, and mitigation methods are examined for five options: commercial disposal, permanent onsite surface disposal, permanent offsite disposal, deepwell injection, and ocean and Great Salt Lake disposal. The study concludes the following: Commercial marketing of all repository salt would require a subsidy for transportation to major salt markets. Permanent onsite surface storage is both economically and technically feasible. Permanent offsite disposal is technically feasible but would incur additional transportation costs. Selection of an offsite location would provide a means of mitigating impacts associated with surface storage at the repository site. Deepwell injection is an attractive disposal method; however, the large water requirement, high cost of development, and poor performance of similar operating brine disposal wells eliminates this option from consideration as the primary means of disposal for the Paradox Basin. Ocean disposal is expensive because of high transportation cost. Also, regulatory approval is unlikely. Ocean disposal should be eliminated from further consideration in the Paradox Basin. Great Salt Lake disposal appears to be technically feasible. Great Salt Lake disposal would require state approval and would incur substantial costs for salt transportation. Permanent onsite disposal is the least expensive method for disposal of all repository salt

  13. Surveillance instrumentation for spent-fuel safeguards

    International Nuclear Information System (INIS)

    McKenzie, J.M.; Holmes, J.P.; Gillman, L.K.; Schmitz, J.A.; McDaniel, P.J.

    1978-01-01

    The movement, in a facility, of spent reactor fuel may be tracked using simple instrumentation together with a real time unfolding algorithm. Experimental measurements, from multiple radiation monitors and crane weight and position monitors, were obtained during spent fuel movements at the G.E. Morris Spent-Fuel Storage Facility. These data and a preliminary version of an unfolding algorithm were used to estimate the position of the centroid and the magnitude of the spent fuel radiation source. Spatial location was estimated to +-1.5 m and source magnitude to +-10% of their true values. Application of this surveillance instrumentation to spent-fuel safeguards is discussed

  14. Fission product behavior in the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

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

    1975-10-01

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

  15. Areal thermal loading recommendations for nuclear waste repositories in salt

    International Nuclear Information System (INIS)

    Russell, J.E.

    1979-06-01

    This document gives a wider understanding of the history of the recommended thermal loadings in salt for both high-level waste (HLW) from fresh UO 2 -fueled, light-water reactors (LWR) with no recycle and spent unreprocessed fuel (SURF) from LWRs. Aspects of the current recommendations that need further study are identified. Finally, an interim set of design thermal-loading recommendations are given that have a common rationale of satisfying performance limits within our current state of knowledge. These recommendations are made on a generic rather than a site-specific basis. 11 figures, 5 tables

  16. Melting of Uranium Metal Powders with Residual Salts

    International Nuclear Information System (INIS)

    Jin-Mok Hur; Dae-Seung Kang; Chung-Seok Seo

    2007-01-01

    The Advanced Spent Fuel Conditioning Process (ACP) of the Korea Atomic Energy Research Institute focuses on the conditioning of Pressurized Water Reactor spent oxide nuclear fuel. After the oxide reduction step of the ACP, the resultant metal powders containing ∼ 30 wt% residual LiCl-Li 2 O should be melted for a consolidation of the fine metal powders. In this study, we investigated the melting behaviors of uranium metal powders considering the effects of a LiCl-Li 2 O residual salt. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

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

    International Nuclear Information System (INIS)

    Jeong, Yongjin; Choi, Sooyoung; Lee, Deokjung

    2014-01-01

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

  19. Corrosion Behavior of Superalloys in Hot Lithium Molten Salt

    International Nuclear Information System (INIS)

    Cho, Soo-Haeng; Hur, Jin-Mok; Seo, Chung-Seok; Park, Seoung-Won

    2006-01-01

    The Li-reduction process involves the chemical reduction of spent fuel oxides by liquid lithium metal in a molten LiCl salt bath at 650 .deg. C followed by a separate electrochemical reduction of lithium oxide (Li 2 O), which builds up in the salt bath. This process requires a high purity inert gas atmosphere inside remote hot cell nuclear facility to prevent unwanted Li oxidation and fires during the handling of chemically active Li metal. In light of the limitations of the Li-reduction process, a direct electrolytic reduction technology is being developed by KAERI to enhance process safety and economic viability. The electrolytic reduction of spent oxide fuel involves the liberation of oxygen in a molten LiCl electrolyte, which results in a chemically aggressive environment that is too corrosive for typical structural materials. Even so, the electrochemical process vessel must be resilient at ∼ 650 .deg. C in the presence of oxygen to enable high processing rates and an extended service life. But, the mechanism and the rate of the corrosion of metals in LiCl-Li 2 O molten salt under oxidation condition are not clear. In the present work, the corrosion behavior and corrosion mechanism of superalloys have been studied in the molten salt of LiCl-Li 2 O under oxidation condition

  20. Electrometallurgical treatment of oxide spent fuels

    International Nuclear Information System (INIS)

    Karell, E. J.

    1999-01-01

    The Department of Energy (DOE) inventory of spent nuclear fuel contains a wide variety of oxide fuel types that may be unsuitable for direct repository disposal in their current form. The molten-salt electrometallurgical treatment technique developed by Argonne National Laboratory (ANL) has the potential to simplify preparing and qualifying these fuels for disposal by converting them into three uniform product streams: uranium metal, a metal waste form, and a ceramic waste form. This paper describes the major steps in the electrometallurgical treatment process for oxide fuels and provides the results of recent experiments performed to develop and scale up the process

  1. The METEOR/TRANSURANUS fuel performance code

    International Nuclear Information System (INIS)

    Struzik, C.; Guerin, Y.

    1996-01-01

    The first calculations for the FUMEX exercise were performed using version 1.1 of the METEOR/TRANSURANUS code. Since then, important improvements have been implemented on several models. In its present state, the code describes fuel rod behaviour in standard PWR conditions. Its validity extends to UO 2 and MOX fuels clad in Zircaloy-4. Power transient calculations for UO 2 and Gd doped fuel calculations are possible, but further developments are in progress, and the applications will be fully qualified in version 2.0. A considerable effort is made to replace semi-empirical models with models that have a sounder physical basis. (authors). 14 refs

  2. Recommendations for a restart of Molten Salt Reactor development

    International Nuclear Information System (INIS)

    Moir, R. W.

    2007-01-01

    The concept of the molten salt reactor (MSR) refuses to go away. The Generation-IV process lists the MSR as one of the six concepts to be considered for extending fuel resources. Good fuel utilization and good economics are required to meet the often cited goal of 10 TWe globally and 1 TWe for the US by non-carbon energy sources in this century by nuclear fission. A strong incentive for the molten salt reactor design is its good fuel utilization, good economics, amazing flexibility and promised large benefits. It can: - use thorium or uranium; o be designed with lots of graphite to have a fairly thermal neutron spectrum or without graphite moderator to have a fast neutron spectrum reactor; - fission uranium isotopes and plutonium isotopes; - operate with non-weapon grade fissile fuel, or in suitable sites it can operate with enrichment between reactor-grade and weapon-grade fissile fuel; - be a breeder or near breeder; - operate at temperature >1100 degree C if carbon composites are successfully employed. Enhancing 2 32U content in the uranium to over 500 pm makes the fuel undesirable for weapons, but it should not detract from its economic use in liquid fuel reactors: a big advantage in nonproliferation. Economics of the MSR is enhanced by operating at low pressure and high temperature and may even lead to the preferred route to hydrogen production. The cost of the electricity produced from low enriched fuel averaged over the life of the entire process, has been predicted to be about 10% lower than that from LWRs, and 20% lower for high enriched fuel, with uncertainties of about 10%. The development cost has been estimated at about 1 B$ (e.g., a 100 M$/y base program for ten years) not including construction of a series of reactors leading up to the deployment of multiple commercial units at an assumed cost of 9 B$ (450 M$/y over 20 years). A benefit of liquid fuel is that smaller power reactors can faithfully test features of larger reactors, thereby reducing the

  3. Improvement to molten salt reactors

    International Nuclear Information System (INIS)

    Bienvenu, Claude.

    1975-01-01

    The invention proposes a molten salt nuclear reactor whose core includes a mass of at least one fissile element salt to which can be added other salts to lower the melting temperature of the mass. This mass also contains a substance with a low neutron capture section that does not give rise to a chemical reaction or to an azeotropic mixture with these salts and having an atmospheric boiling point under that of the mass in operation. Means are provided for collecting this substance in the vapour state and returning it as a liquid to the mass. The kind of substance chosen will depend on that of the molten salts (fissile element salts and, where required, salts to lower the melting temperature). In actual practice, the substance chosen will have an atmospheric pressure boiling point of between 600 and 1300 0 C and a melting point sufficiently below 600 0 C to prevent solidification and clogging in the return line of the substance from the exchanger. Among the materials which can be considered for use, mention is made of magnesium, rubidium, cesium and potassium but metal cesium is not employed in the case of many fissile salts, such as fluorides, which it would reduced to the planned working temperatures [fr

  4. Caenorhabditis elegans response to salt

    NARCIS (Netherlands)

    O.O. Umuerri (Oluwatoroti Omowayewa)

    2012-01-01

    textabstractThis thesis describes my work, where I used genetic methods to identify new genes involved in salt taste in C. elegans. In addition, I used calcium imaging to characterize the cellular response of C. elegans to salt. The thesis is divided into five sections and each section is summarized

  5. Detection of Fluorescence for Lanthanides in LiCl-KCl Molten Salt Medium

    International Nuclear Information System (INIS)

    Im, Hee Jung; Kim, Tack Jin; Song, Kyu Seok; Jee, Kwang Yong

    2007-01-01

    In the electrorefining step of the pyrochemical process, actinide ions dissolved in the LiCl-KCl eutectic salt are recovered as pure actinide metals at a cathode for a re-use as a nuclear fuel from the aspect of its nonproliferation of the nuclear fuel cycles. The lanthanide species dissolved in the LiCl-KCl eutectic salt play an important role in an effective metal purification during the electrorefining step, so it is necessary to understand the chemical and physical behaviors of lanthanides in molten salt. The in situ spectroscopic measurement system and studies according to temperature changes are essential for better understandable information. To our knowledge, the absorption studies of lanthanides at high temperatures have been reported before, but the fluorescence studies of those at high temperature are not reported yet. We will discuss here the fluorescence behaviors of lanthanides in LiCl-KCl molten salt medium according to a changing temperature

  6. Fuel assembly

    International Nuclear Information System (INIS)

    Sakuyama, Tadashi; Mukai, Hideyuki.

    1988-01-01

    Purpose: To prevent the bending of a fuel rod caused by the difference in the elongation between a joined fuel rod and a standard fuel rod thereby maintain the fuel rod integrity. Constitution: A joined fuel rod is in a thread engagement at its lower end plug thereof with a lower plate, while passed through at its upper end plug into an upper tie plate and secured with a nut. Further, a standard fuel rod is engaged at its upper end plug and lower end plug with the upper tie plate and the lower tie plate respectively. Expansion springs are mounted to the upper end plugs of these bonded fuel rods and the standard fuel rods for preventing this lifting. Each of the fuel rods comprises a plurality of sintered pellets of nuclear fuel materials laminated in a zircaloy fuel can. The content of the alloy ingredient in the fuel can of the bonded fuel rod is made greater than that of the alloy ingredient of the standard fuel rod. this can increase the elongation for the bonded fuel rod, and the spring of the standard fuel rod is tightly bonded to prevent the bending. (Yoshino, Y.)

  7. R and D activities on the management of waste chloride salts in KAERI

    International Nuclear Information System (INIS)

    In-Tae, Kim; Hwan-Seo, Park; Jeong-Gook, Kim; Hee-Chul, Yang; Yong-Joon, Cho; Eung-Ho Kim

    2007-01-01

    Full text of publication follows. Electrochemical treatment of spent oxide fuels has been intensively studied in KAERI to reduce the volume, heat load and radiotoxicity of high-level wastes. It consists of an electrolytic reduction process to convert the oxide fuel into a metallic form and an electro-refining process to separate TRU elements from the electro-reduced metal ingot. Two types of waste salts are expected to generate from the electrochemical pyro-processes, that is, LiCl salt from the reduction process and LiCl+KCl eutectic salt form the refining process. The R and D strategy of the waste salt management in KAERI can be categorized into two parts: 1) enhancement of safety by the stabilisation/solidification of waste salt that is to be finally disposed of and 2) reduction of the waste generation by the regeneration/recycle of the spent salt after removal of radionuclides in it. A sol-gel technique and a zeolite occlusion technique are under development to stabilize the waste salt. The LiCl salt is stabilised by a low-temperature sol-gel process and then the gel product is solidified into a ceramic-like waste form with an addition of glass frit. Another method uses Zeolite-4A to occlude the LiCl salt into its cage and adsorption site to immobilize the radionuclides. The product, salt-occluded zeolite, is fabricated into another type of a ceramic waste form. For the regeneration and recycle of the spent salt, the radionuclides in the salt are removed by a zeolite process for the LiCl salt and by an oxidation/distillation process for the eutectic salt. The target nuclides to be removed in each process are Cs/Sr and rare earth (RE) elements, respectively. In the oxidation/ distillation process, the rare earth chloride nuclides are oxidised by an oxygen sparging method, and the products are precipitated in the form of oxide or oxychloride REs. After separation of the RE elements from the precipitates by distillation, the refined spent salt with a low content

  8. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. This part describes the MSBR core (data presented are from ORNL 4541). The principal characteristics of the core are presented in tables together with plane and elevation drawings, stress being put upon the reflector, and loading and unloading. Neutronic, and thermal and hydraulic characteristics (core and reflectors) are more detailed. The reasons why a graphite with a tight graphite layer has been chosen are briefly exposed. The physical properties of the standard graphite (irradiation behavior) have been determined for an isotropic graphite with fine granulometry; its dimensional variations largely ressemble that of Gilsonite. The mechanical stresses computed (Wigner effect) do not implicate in any way the graphite stack [fr

  9. The Salt II Treaty

    International Nuclear Information System (INIS)

    Caldwell, D.

    1991-01-01

    The first strategic arms limitation talks resulted in two agreements: the Anti-Ballistic Missile Treaty and the Interim Agreement to Limit Strategic Offensive Arms. Senator Henry M. (Scoop) Jackson (D-Wa.) was concerned about the numerical advantage granted to the USSR by the Latter agreement and proposed an amendment that would prohibit future negotiators from granting the Soviet Union similar terms. This paper discusses the second round of SALT negotiations which opened in November 1972 and continued under presidents Richard M. Nixon, Gerald Ford, and Jimmy Carter. As the negotiators met, U.S. and Soviet scientists and engineers continued their work to develop new nuclear weapons and launchers. Particularly problematic were modern, large ballistic missiles, cruise missiles, and the Soviet Backfire bomber

  10. Salt resistant crop plants

    KAUST Repository

    Roy, Stuart J.

    2014-04-01

    Soil salinity is a major constraint to agriculture. To improve salinity tolerance of crops, various traits can be incorporated, including ion exclusion, osmotic tolerance and tissue tolerance. We review the roles of a range of genes involved in salt tolerance traits. Different tissues and cells are adapted for specific and often diverse function, so it is important to express the genes in specific cell-types and to pyramid a range of traits. Modern biotechnology (marker- assisted selection or genetic engineering) needs to be increasingly used to introduce the correct combination of genes into elite crop cultivars. Importantly, the effects of introduced genes need to be evaluated in the field to determine their effect on salinity tolerance and yield improvement.

  11. Behaviour of conductivity improvers in jet fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dacre, B.; Hetherington, J.I. [Cranfield Univ., Wiltshire (United Kingdom)

    1995-05-01

    Dangerous accumulation of electrostatic charge can occur due to high speed pumping and microfiltration of fuel. This can be avoided by increasing the electrical conductivity of the fuel using conductivity improver additives. However, marked variations occur in the conductivity response of different fuels when doped to the same level with conductivity improver. This has been attributed to interactions of the conductivity improver with other fuel additives or fuel contaminants. The present work concentrates on the effects of fuel contaminants, in particular polar compounds, on the performance of the conductivity improver. Conductivity is the fuel property of prime interest. The conductivity response of model systems of the conductivity improver STADIS 450 in dodecane has been measured and the effect on this conductivity of additions of model polar contaminants sodium naphthenate, sodium dodecyl benzene sulphonate, and sodium phenate have been measured. The sodium salts have been found to have a complex effect on the performance of STADIS 450, reducing the conductivity at low concentrations to a minimum value and then increasing the conductivity at high concentrations of sodium salts. This work has focused on characterising this minimum in the conductivity values and on understanding the reason for its occurrence. The effects on the minimum conductivity value of the following parameters are investigated: (a) time, (b) STADIS 450 concentration, (c) sodium salt concentration, (d) mixed sodium salts, (e) experimental method, (f) a phenol, (g) individual components of STADIS 450. The complex conductivity response of the STADIS 450 to sodium salt impurities is discussed in terms of possible inter-molecular interactions.

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

    International Nuclear Information System (INIS)

    Zhang Dalin; Qiu Suizheng; Su Guanghui

    2009-01-01

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

  13. Modelling strategic responses to car and fuel taxation

    NARCIS (Netherlands)

    Heijnen, P.; Kooreman, P.

    We develop a model to analyse the interactions between actors involved in car and fuel taxation: consumers, car producers, fuel producers and the government. Heterogeneous consumers choose between two versions of a car that differ in engine type (diesel or gasoline). Car manufacturers and fuel

  14. Alternative Fuels Data Center: Diesel Vehicles Using Biodiesel

    Science.gov (United States)

    Biodiesel Printable Version Share this resource Send a link to Alternative Fuels Data Center : Diesel Vehicles Using Biodiesel to someone by E-mail Share Alternative Fuels Data Center: Diesel Vehicles Using Biodiesel on Facebook Tweet about Alternative Fuels Data Center: Diesel Vehicles Using Biodiesel

  15. Alternative Fuels Data Center: Indiana Transportation Data for Alternative

    Science.gov (United States)

    (nameplate, MW) 1,430 Source: BioFuels Atlas from the National Renewable Energy Laboratory Case Studies Video Alternative Fuels Save Money in Indy April 1, 2012 More Case Studies Videos Text Version More Indiana Videos on YouTube Video thumbnail for Indiana Beverage Company Invests in Alternative Fuels Indiana Beverage

  16. Alternative Fuels Data Center: Alabama Transportation Data for Alternative

    Science.gov (United States)

    Public Private Biodiesel (B20 and above) 2 8 Compressed Natural Gas (CNG) 10 23 Electric 84 67 Ethanol Boasts 200-Plus Flex Fuel Vehicles May 24, 2013 Video thumbnail for Biodiesel Fuels Education in Alabama Biodiesel Fuels Education in Alabama May 1, 2012 More Case Studies Videos Text Version More Alabama Videos

  17. Fuel processing

    International Nuclear Information System (INIS)

    Allardice, R.H.

    1990-01-01

    The technical and economic viability of the fast breeder reactor as an electricity generating system depends not only upon the reactor performance but also on a capability to recycle plutonium efficiently, reliably and economically through the reactor and fuel cycle facilities. Thus the fuel cycle is an integral and essential part of the system. Fuel cycle research and development has focused on demonstrating that the challenging technical requirements of processing plutonium fuel could be met and that the sometimes conflicting requirements of the fuel developer, fuel fabricator and fuel reprocessor could be reconciled. Pilot plant operation and development and design studies have established both the technical and economic feasibility of the fuel cycle but scope for further improvement exists through process intensification and flowsheet optimization. These objectives and the increasing processing demands made by the continuing improvement to fuel design and irradiation performance provide an incentive for continuing fuel cycle development work. (author)

  18. Salt brickwork as long-term sealing in salt formations

    International Nuclear Information System (INIS)

    Walter, F.; Yaramanci, U.

    1993-01-01

    Radioactive wastes can be disposed of in deep salt formations. Rock salt is a suitable geologic medium because of its unique characteristics. Open boreholes, shafts and drifts are created to provide physical access to the repository. Long-term seals must be emplaced in these potential pathways to prevent radioactive release into the biosphere. The sealing materials must be mechanically and, most important, geochemically stable within the host rock. Salt bricks made from compressed salt-powder are understood to be the first choice long-term sealing material. Seals built of salt bricks will be ductile. Large sealing systems are built by combining the individual bricks with mortar. Raw materials for mortar are fine-grained halite powder and ground saliferous clay. This provides for the good adhesive strength of the mortar to the bricks and the high shear-strength of the mortar itself. To test the interaction of rock salt with an emplaced long-term seal, experiments will be carried out in situ, in the Asse salt mine in Germany. Simple borehole sealing experiments will be performed in horizontal holes and a complicated drift sealing experiment is planned, to demonstrate the technology of sealing a standard size drift or shaft inside a disturbed rock mass. Especially, the mechanical stability of the sealing system has to be demonstrated

  19. ERRATUM - French version only

    CERN Multimedia

    Le texte suivant remplace la version française de l'encadré paru en page 2 du Bulletin 28/2003 : Le 1er juillet 1953, les représentants des douze Etats Membres fondateurs du CERN signèrent la convention de l'Organisation. Aujourd'hui, le CERN compte vingt Etats Membres Européens : l'Allemagne, l'Autriche, la Belgique, la Bulgarie, le Danemark, l'Espagne, la Finlande, la France, la Grèce, la Hongrie, l'Italie, la Norvège, les Pays-Bas, la Pologne, le Portugal, la République Slovaque, la République Tchèque, le Royaume-Uni, la Suède, et la Suisse. Les Etats-Unis, l'Inde, l'Israël, le Japon, la Fédération Russe, la Turquie, la Commission Européenne et l'UNESCO ont un statut d'Etat observateur.

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

    Science.gov (United States)

    Willit, James L [Batavia, IL

    2010-09-21

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

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

    Science.gov (United States)

    Willit, James L.

    2007-09-11

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

  2. Waste isolation facility description: bedded salt

    Energy Technology Data Exchange (ETDEWEB)

    1976-09-01

    The waste isolation facility is designed to receive and store three basic types of solidified wastes: high-level wastes, intermediate level high-gamma transuranic waste, and low-gamma transuranic wastes. The facility under consideration in this report is designed for bedded salt at a depth of approximately 1800 ft. The present design for the facility includes an area which would be used initially as a pilot facility to test the viability of the concept, and a larger facility which would constitute the final storage area. The total storage area in the pilot facility is planned to be 77 acres and in the fuel facility 1601 acres. Other areas for shaft operations and access would raise the overall size of the total facility to slightly less than 2,000 acres. The following subjects are discussed in detail: surface facilities, shaft design and characteristics, design and construction of the underground waste isolation facility, ventilation systems, and design requirements and criteria. (LK)

  3. Waste isolation facility description: bedded salt

    International Nuclear Information System (INIS)

    1976-09-01

    The waste isolation facility is designed to receive and store three basic types of solidified wastes: high-level wastes, intermediate level high-gamma transuranic waste, and low-gamma transuranic wastes. The facility under consideration in this report is designed for bedded salt at a depth of approximately 1800 ft. The present design for the facility includes an area which would be used initially as a pilot facility to test the viability of the concept, and a larger facility which would constitute the final storage area. The total storage area in the pilot facility is planned to be 77 acres and in the fuel facility 1601 acres. Other areas for shaft operations and access would raise the overall size of the total facility to slightly less than 2,000 acres. The following subjects are discussed in detail: surface facilities, shaft design and characteristics, design and construction of the underground waste isolation facility, ventilation systems, and design requirements and criteria

  4. Pyroprocessing of IFR Metal Fuel

    International Nuclear Information System (INIS)

    Laidler, J.J.

    1993-01-01

    The Integral Fast Reactor (IFR) fuel cycle features the use of an innovative reprocessing method, known as open-quotes pyroprocessingclose quotes featuring fused-salt electrofining of the spent fuel. Electrofining of IFR spent fuel involves uranium recovery by electro-transport to a solid steel cathode. The thermodynamics of the system preclude plutonium recovery in the same way, so a liquid cadmium cathode located in the electrolyte salt phase is utilized. The deposition of Pu, Am, Np, and Cm takes place at the liquid cadmium cathode in the form of cadmium intermetallic compounds (e.g, PuCd 6 ), and uranium deposits as the pure metal when cadmium saturation is reached. A small amount of rare earth fission products deposit together with the heavy metals at both the solid and liquid cadmium cathodes, providing a significant degree of self-protection. A full scope demonstration of the IFR fuel cycle will begin in 1993, using fuel irradiated in EBR-II

  5. Systematic study on Thorium fuel

    International Nuclear Information System (INIS)

    Shibata, Toshikazu; Kimura, Itsuro; Iwata, Shiro; Furuya, Hirotaka; Suzuki, Susumu.

    1988-01-01

    Introduced is the activities of the Joint Research Project Team on Thorium Fuel organized by mainly university researchers in Japan and supported by the Ministry of Education, Science and Culture for seven years since 1980. Four major groups were organized; (1) nuclear data, reactor physics and design, (2) nuclear fuel, (3) down stream and (4) biological effects of thorium. The first group covered measurements and analysis on nuclear data of thorium related nuclides, experiment and analysis on nuclear characteristics of thorium containing cores, basic engineering on a thorium molten salt reactor, and designs of several types of reactors. Fabrication and irradiation tests of thorium oxide fuel, and basic studies on new type thorium fuels (e.g. carbide and nitride) were studied by the second group. The third group covered the use of solutions in reprocessing of spent fuel, behavior of fission products, immobilization of high level radioactive waste, and continuous reprocessing for a molten salt reactor. The fourth group performed the trace study for patients who had been intravascularly injected with thorotrast for diagnosis of war injuries during the Second World War. (author)

  6. Properties and thermal decomposition of the double salts of uranyl nitrate-ammonium nitrate

    International Nuclear Information System (INIS)

    Notz, K.J.; Haas, R.A.

    1989-01-01

    The formation of ammonium nitrate-uranyl nitrate double salts has important effects on the thermal denitration process for the preparation of UO 3 and on the physical properties of the resulting product. Analyses were performed, and properties and decomposition behavior were determined for three double salts: NH 4 UO 2 (NO 3 ) 3 , (NH 4 ) 2 UO 2 (NO 3 ) 4 , and (NH 4 ) 2 UO 2 (NO 3 ) 4 ·2H 2 O. The tinitrate salt decomposes without melting at 270-300 C to give a γ-UO 3 powder of ∼3-μm average size, with good ceramic properties for fabrication into UO 2 nuclear fuel pellets. The tetranitrate dihydrate melts at 48 C; it also dehydrates to the anhydrous salt. The anhydrous tetranitrate decomposes exothermically, without melting, at 170-270 C by losing one mole of ammonium nitrate to form the trinitrate salt

  7. Evaluation of a molten salt electrolyte for direct reduction of actinides

    International Nuclear Information System (INIS)

    Alangi, Nagaraj; Anupama, P.; Mukherjee, Jaya; Gantayet, L.M.

    2011-01-01

    Use of molten fluoride salt towards direct reduction of actinides and lanthanides by molten salt electrolysis is of interest for problems related to metallic nuclear fuels. The performance of the molten salt bath is dependent on the pre-conditioning of the molten salt. A procedure for conditioning of LiF-BaF 2 salt mixtures has been developed based on systematic electrochemical experimental investigations using voltammetry with graphite and platinum as electrode materials. We utilize the linear sweep voltammetry (LSV) as a diagnostic tool for assessment of the electrolyte condition. This technique is fast and offers the advantage of in-situ/online measurement eliminating the need for sampling. The conditioning procedure that was developed was tried on LiF-CaF 2

  8. Development of spent salt treatment technology by zeolite column system. Performance evaluation of zeolite column

    International Nuclear Information System (INIS)

    Miura, Hidenori; Uozumi, Koichi

    2009-01-01

    At electrorefining process, fission products(FPs) accumulate in molten salt. To avoid influence on heating control by decay heat and enlargement of FP amount in the recovered fuel, FP elements must be removed from the spent salt of the electrorefining process. For the removal of the FPs from the spent salt, we are investigating the availability of zeolite column system. For obtaining the basic data of the column system, such as flow property and ion-exchange performance while high temperature molten salt is passing through the column, and experimental apparatus equipped with fraction collector was developed. By using this apparatus, following results were obtained. 1) We cleared up the flow parameter of column system with zeolite powder, such as flow rate control by argon pressure. 2) Zeolite 4A in the column can absorb cesium that is one of the FP elements in molten salt. From these results, we got perspective on availability of the zeolite column system. (author)

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

    International Nuclear Information System (INIS)

    Cheng Maosong; Dai Zhimin

    2014-01-01

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

  10. High Efficiency Advanced Lightweight Fuel Cell (HEAL-FC), Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Infinity's High Efficiency Advanced Lightweight Fuel Cell (HEAL FC) is an improved version of its current fuel cell technology developed for space applications. The...

  11. Concept of the demonstration molten salt unit for the transuranium elements transmutations

    International Nuclear Information System (INIS)

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

    1999-01-01

    Fluorine reprocessing is discussed of spent fuel and of fluoride molten salt reactor in critical and subcritical modes for plutonium and minor actinides burning. International collaboration for creation of such system is proposed. Additional neutron source in the core will have positive influence on the transmutation processes in the reactor. Demonstration critical molten salt reactor of small power capacity will permit to decide the most part of problems inherent to large critical reactors and subcritical drivers. It could be expected that fluoride molten salt transmuter can work without accelerator as a critical reactor. (author)

  12. Proposed Guidance for Preparing and Reviewing Molten Salt Nonpower Reactor Licence Applications (NUREG-1537)

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [ORNL; Flanagan, George F. [ORNL; Voth, Marcus [Boston Government Services, LLC

    2018-05-01

    Development of non-power molten salt reactor (MSR) test facilities is under consideration to support the analyses needed for development of a full-scale MSR. These non-power MSR test facilities will require review by the US Nuclear Regulatory Commission (NRC) staff. This report proposes chapter adaptations for NUREG-1537 in the form of interim staff guidance to address preparation and review of molten salt non-power reactor license applications. The proposed adaptations are based on a previous regulatory gap analysis of select chapters from NUREG-1537 for their applicability to non-power MSRs operating with a homogeneous fuel salt mixture.

  13. Coupled study of the Molten Salt Fast Reactor core physics and its associated reprocessing unit

    International Nuclear Information System (INIS)

    Doligez, X.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Ghetta, V.

    2014-01-01

    Highlights: • The limit on the reprocessing is due to the redox potential control. • Alkali and Earth-alkaline elements do not have to be extracted. • Criticality risks have to be studied in the reprocessing unit. • The neutronics properties are not sensitive to chemical data. • The reprocessing chemistry, from a pure numerical point of view, is an issue. - Abstract: Molten Salt Reactors (MSRs) are liquid-fuel reactors, in which the fuel is also the coolant and flows through the core. A particular configuration presented in this paper called the Molten Salt Fast Reactor consists in a Molten Salt Reactor with no moderator inside the core and a salt composition that leads to a fast neutron spectrum. Previous studies showed that this concept (previously called Thorium Molten Salt Reactor – Nonmoderated) has very promising characteristics. The liquid fuel implies a special reprocessing. Each day a small amount of the fuel salt is extracted from the core for on-site reprocessing. To study such a reactor, the materials evolution within the core has to be coupled to the reprocessing unit, since the latter cleans the salt quasi continuously and feeds the reactor. This paper details the issues associated to the numerical coupling of the core and the reprocessing. It presents how the chemistry is introduced inside the classical Bateman equation (evolution of nuclei within a neutron flux) in order to carry a numerical coupled study. To achieve this goal, the chemistry has to be modeled numerically and integrated to the equations of evolution. This paper presents how is it possible to describe the whole concept (reactor + reprocessing unit) by a system of equations that can be numerically solved. Our program is a connection between MCNP and a homemade evolution code called REM. Thanks to this tool; constraints on the fuel reprocessing were identified. Limits are specified to preserve the good neutronics properties of the MSFR. In this paper, we show that the limit

  14. Fusion reactor fuel processing

    International Nuclear Information System (INIS)

    Johnson, E.F.

    1972-06-01

    For thermonuclear power reactors based on the continuous fusion of deuterium and tritium the principal fuel processing problems occur in maintaining desired compositions in the primary fuel cycled through the reactor, in the recovery of tritium bred in the blanket surrounding the reactor, and in the prevention of tritium loss to the environment. Since all fuel recycled through the reactor must be cooled to cryogenic conditions for reinjection into the reactor, cryogenic fractional distillation is a likely process for controlling the primary fuel stream composition. Another practical possibility is the permeation of the hydrogen isotopes through thin metal membranes. The removal of tritium from the ash discharged from the power system would be accomplished by chemical procedures to assure physiologically safe concentration levels. The recovery process for tritium from the breeder blanket depends on the nature of the blanket fluids. For molten lithium the only practicable possibility appears to be permeation from the liquid phase. For molten salts the process would involve stripping with inert gas followed by chemical recovery. In either case extremely low concentrations of tritium in the melts would be desirable to maintain low tritium inventories, and to minimize escape of tritium through unwanted permeation, and to avoid embrittlement of metal walls. 21 refs

  15. Nuclear fuels

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

    2009-01-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO 2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO 2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation

  16. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

  17. Transient response of small molten salt reactor at duct blockage accident

    International Nuclear Information System (INIS)

    Yamamoto, Takahisa; Mitachi, Koshi; Ikeuchi, Koji; Suzuki, Takashi

    2005-01-01

    This paper performed transient core analysis of a small Molten Salt Reactor (MSR) at the time of a duct blockage accident. The numerical model employed in this study consists of continuity and momentum conservation equations for fuel salt flow, two group diffusion equations for fast and thermal neutron fluxes, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The analysis shows that (1) the effective multiplication factor and reactor power after the blockage accident hardly change because of the self-control performance of the MSR, (2) fuel salt and graphite moderator temperatures rise at the blockage point and its vicinity, drastically but locally, (3) the highest temperature after the blockage accident is 1 363 K, very lower than the boiling point of fuel salt and melt point of reactor vessel, (4) fast and thermal neutron fluxes distributions after the blockage accident hardly change, and (5) delayed neutron precursors accumulate at the blockage point, especially 1st delayed neutron precursor due to is large decay constant. These results lead that the safety of MSR is assured in the blockage accident. (author)

  18. Assessment of the Capability of Molten Salt Reactors as a Next Generation High Temperature Reactors

    International Nuclear Information System (INIS)

    Elsheikh, B.M.

    2017-01-01

    Molten Salt Reactor according to Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment (MSRE) programs, was designed to be the first full-scale, commercial nuclear power plant utilizing molten salt liquid fuels that can be used for producing electricity, and producing fissile fuels (breeding)burning actinides. The high temperature in the primary cycle enables the realization of efficient thermal conversion cycles with net thermal efficiencies reach in some of the designs of nuclear reactors greater than 45%. Molten salts and liquid salt because of their low vapor pressure are excellent candidates for meeting most of the requirements of these high temperature reactors. There is renewed interest in MSRs because of changing goals and new technologies in the use of high-temperature reactors. Molten Salt Reactors for high temperature create substantial technical challenges to have high effectiveness intermediate heat transfer loop components. This paper will discuss and investigate the capability and compatibility of molten salt reactors, toward next generation high temperature energy system and its technical challenges

  19. Impacts of using salt and salt brine for roadway deicing.

    Science.gov (United States)

    2014-06-01

    Idaho Transportation Department (ITD) uses a variety of methods to help ensure safe travel on the state highway system : following winter storm events. These methods include plowing, use of sand to improve traction, and use of salt and chemical : com...

  20. Molten salts in nuclear reactors

    International Nuclear Information System (INIS)

    Dirian, J.; Saint-James

    1959-01-01

    Collection of references dealing with the physicochemical studies of fused salts, in particular the alkali and alkali earth halides. Numerous binary, ternary and quaternary systems of these halides with those of uranium and thorium are examined, and the physical properties, density, viscosity, vapour pressure etc... going from the halides to the mixtures are also considered. References relating to the corrosion of materials by these salts are included and the treatment of the salts with a view to recuperation after irradiation in a nuclear reactor is discussed. (author) [fr