WorldWideScience

Sample records for bio-solids fuel materials

  1. Characteristics of sustainable bio-solid fuel produced from sewage sludge as a conventional fuel substitute

    International Nuclear Information System (INIS)

    Jung, Bongjin; Nam, Wonjun; Lee, Na-Yeon; Kim, Kyung-Hoon

    2010-01-01

    Safely final disposal of sewage sludge which is being increased every year has already become serious problems. As one of the promising technologies to solve this problem, thermal drying method has been attracting wide attention due to energy recovery from sewage sludge. This paper describes several characteristics of sustainable bio-solid fuel, as a conventional fuel substitute, produced from sewage sludge drying and granulation plant having the treatment capacity of 10 ton/ day. This plant has been successfully operated many times and is now designing for scale-up. Average moisture content of twelve kinds of bio-solid fuels produced from the plant normally less than 10 wt% and average shape of them is mainly composed of granular type having a diameter of 2-8 mm for easy handling and transportation to the final market destinations. Average higher heating value, which is one of the important properties to estimate the possibility of available energy, of bio-solid fuels is about 3800 kcal/ kg as dry basis. So they can be utilized to supply energy in the coal power plant and cement kiln etc. as a conventional fuel substitute for a beneficial reuse. Characteristics including proximate analysis, ultimate analysis, contents of heavy metals, wettability etc. of bio-solid fuels have been also analyzed for the environmentally safe re utilization. (author)

  2. FUEL CELL ELECTRODE MATERIALS

    Science.gov (United States)

    FUEL CELL ELECTRODE MATERIALS. RAW MATERIAL SELECTION INFLUENCES POLARIZATION BUT IS NOT A SINGLE CONTROLLING FACTOR. AVAILABLE...DATA INDICATES THAT AN INTERRELATIONSHIP OF POROSITY, AVERAGE PORE VOLUME, AND PERMEABILITY CONTRIBUTES TO ELECTRODE FUEL CELL BEHAVIOR.

  3. Material input of nuclear fuel

    International Nuclear Information System (INIS)

    Rissanen, S.; Tarjanne, R.

    2001-01-01

    The Material Input (MI) of nuclear fuel, expressed in terms of the total amount of natural material needed for manufacturing a product, is examined. The suitability of the MI method for assessing the environmental impacts of fuels is also discussed. Material input is expressed as a Material Input Coefficient (MIC), equalling to the total mass of natural material divided by the mass of the completed product. The material input coefficient is, however, only an intermediate result, which should not be used as such for the comparison of different fuels, because the energy contents of nuclear fuel is about 100 000-fold compared to the energy contents of fossil fuels. As a final result, the material input is expressed in proportion to the amount of generated electricity, which is called MIPS (Material Input Per Service unit). Material input is a simplified and commensurable indicator for the use of natural material, but because it does not take into account the harmfulness of materials or the way how the residual material is processed, it does not alone express the amount of environmental impacts. The examination of the mere amount does not differentiate between for example coal, natural gas or waste rock containing usually just sand. Natural gas is, however, substantially more harmful for the ecosystem than sand. Therefore, other methods should also be used to consider the environmental load of a product. The material input coefficient of nuclear fuel is calculated using data from different types of mines. The calculations are made among other things by using the data of an open pit mine (Key Lake, Canada), an underground mine (McArthur River, Canada) and a by-product mine (Olympic Dam, Australia). Furthermore, the coefficient is calculated for nuclear fuel corresponding to the nuclear fuel supply of Teollisuuden Voima (TVO) company in 2001. Because there is some uncertainty in the initial data, the inaccuracy of the final results can be even 20-50 per cent. The value

  4. Materials for low-temperature fuel cells

    CERN Document Server

    Ladewig, Bradley; Yan, Yushan; Lu, Max

    2014-01-01

    There are a large number of books available on fuel cells; however, the majority are on specific types of fuel cells such as solid oxide fuel cells, proton exchange membrane fuel cells, or on specific technical aspects of fuel cells, e.g., the system or stack engineering. Thus, there is a need for a book focused on materials requirements in fuel cells. Key Materials in Low-Temperature Fuel Cells is a concise source of the most important and key materials and catalysts in low-temperature fuel cells. A related book will cover key materials in high-temperature fuel cells. The two books form part

  5. Materials for high-temperature fuel cells

    CERN Document Server

    Jiang, San Ping; Lu, Max

    2013-01-01

    There are a large number of books available on fuel cells; however, the majority are on specific types of fuel cells such as solid oxide fuel cells, proton exchange membrane fuel cells, or on specific technical aspects of fuel cells, e.g., the system or stack engineering. Thus, there is a need for a book focused on materials requirements in fuel cells. Key Materials in High-Temperature Fuel Cells is a concise source of the most important and key materials and catalysts in high-temperature fuel cells with emphasis on the most important solid oxide fuel cells. A related book will cover key mater

  6. ALD coating of nuclear fuel actinides materials

    Science.gov (United States)

    Yacout, A. M.; Pellin, Michael J.; Yun, Di; Billone, Mike

    2017-09-05

    The invention provides a method of forming a nuclear fuel pellet of a uranium containing fuel alternative to UO.sub.2, with the steps of obtaining a fuel form in a powdered state; coating the fuel form in a powdered state with at least one layer of a material; and sintering the powdered fuel form into a fuel pellet. Also provided is a sintered nuclear fuel pellet of a uranium containing fuel alternative to UO.sub.2, wherein the pellet is made from particles of fuel, wherein the particles of fuel are particles of a uranium containing moiety, and wherein the fuel particles are coated with at least one layer between about 1 nm to about 4 nm thick of a material using atomic layer deposition, and wherein the at least one layer of the material substantially surrounds each interfacial grain barrier after the powdered fuel form has been sintered.

  7. Fuel conditioning facility material accountancy

    International Nuclear Information System (INIS)

    Yacout, A.M.; Bucher, R.G.; Orechwa, Y.

    1995-01-01

    The operation of the Fuel conditioning Facility (FCF) is based on the electrometallurgical processing of spent metallic reactor fuel. It differs significantly, therefore, from traditional PUREX process facilities in both processing technology and safeguards implications. For example, the fissile material is processed in FCF only in batches and is transferred within the facility only as solid, well-characterized items; there are no liquid steams containing fissile material within the facility, nor entering or leaving the facility. The analysis of a single batch lends itself also to an analytical relationship between the safeguards criteria, such as alarm limit, detection probability, and maximum significant amount of fissile material, and the accounting system's performance, as it is reflected in the variance associated with the estimate of the inventory difference. This relation, together with the sensitivity of the inventory difference to the uncertainties in the measurements, allows a thorough evaluation of the power of the accounting system. The system for the accountancy of the fissile material in the FCF has two main components: a system to gather and store information during the operation of the facility, and a system to interpret this information with regard to meeting safeguards criteria. These are described and the precision of the inventory closure over one batch evaluated

  8. Study of bioavailability of Mn in soil manured with bio solids using neutron activation analysis (NAA)

    International Nuclear Information System (INIS)

    Mateus, Natalina de Fatima; Madi Filho, Tufic

    2010-01-01

    Full text: This work evaluated, using neutron activation analysis (NAA), the behavior of Mn absorption by Eucalyptus manured with bio solids. Manganese is an important micro nutrient because it is an activator of enzymes, controller of oxy reduction reactions, essential to the photosynthesis and synthesis of chlorophyll and protein. Its lack causes a decrease in photosynthesis, which reduces growth and productivity. In alkaline soils there is manganese deficiency. The critical level of Mn is 675 μg:g -1 , above this value manganese is toxic to the plant, but the bio solids applied had a concentration of 300 μg:g -1 , below the critical level. 16 samples were analyzed, and various elements were detected. But the interest in this work was to evaluate the behavior of Mn with increasing the amount of bio solids used as manure. Three different concentrations of bio solids: 10 kg:ha -1 ; 20 kg:ha -1 and 40 kg:ha -1 were applied. Alkaline bio solids provides an increase of the soil pH, which cause the reduction of the manganese availability. The bio solid is a mud resulting from the biological treatment of wasted liquids. It is considered as a profitable alternative and important to minimize the environmental impact generated by the sewage thrown into sanitary lands. The level of pH above 5.5 causes the Mn reduction. The values obtained showed this reduction of Mn phytoavailability with the bio solids increase. (author)

  9. Bio solids Application on Banana Production: Soil Chemical Properties and Plant Nutrition

    International Nuclear Information System (INIS)

    Teixeira, L.A.J; Berton, R.S.B; Coscione, A.R; Saes, L.A

    2011-01-01

    Bio solids are relatively rich in N, P, and S and could be used to substitute mineral fertilization for banana crop. A field experiment was carried out in a Yellow Oxisol to investigate the effects of bio solids application on soil chemical properties and on banana leaf's nutrient concentration during the first cropping cycle. Soil analysis (ph, organic matter, resin P, exchangeable Ca and K, available B, DTPA-extracted micro nutrients, and heavy metals) and index-leaf analysis (B, Cu, Fe, Mn, Zn, Cd, Cr, Ni, and Pb) were evaluated. Bio solids can completely substitute mineral N and P fertilizer to banana growth. Soil exchangeable K and leaf-K concentration must be monitored in order to avoid K deficiency in banana plants. No risk of heavy metal (Cr, Ni, Pb, and Cd) concentration increase in the index leaf was observed when bio solids were applied at the recommended N rate.

  10. Development of metallic fuel materials

    International Nuclear Information System (INIS)

    Kang, Young Ho; Lee, Chong Tak; Yang, Yeoung Seok; Kim, Ki Hwan; Hwang, Sung Chan; Joo, Keun Sik; Ann, Hyun Suk; Chang, Sae Jung.

    1997-09-01

    Through the control of melting and casting parameters, the sound and homogenous U-10wt.%Zr alloy could be fabricated. The yield and segregation of Zr elements were 85% and ±0.1wt.%, and the density of the alloy was about 16.6 g/cm 3 . The major phase were α-U and δ-UZr 2 . The microstructure showed the laminar structure with fiber morphology which was arranged alternatively with uranium and Zr-rich phase. This alloy will be used for KALIMER fuel material through developing the fabrication technology and the characteristics analysis. And electrorefining study was performed to separate uranium from uranium-neodymium and uranium-zirconium alloy by their different free energy for chloride formation. The liquid cadmium phase becomes the anode of the electrorefining cell. Uranium is electrolytically transported through a molten salt electrolyte to a low carbon steel cathode. The electrolyte is composed of KCl-LiCl eutectic and some UCl 3 , which are installed in the salt to facilitate the electrotransport of uranium. In pyrochemical process the reaction condition of chlorination and the maintenance its purity in preparing UCl 4 by chlorination of UO 2 is strongly dependent on the reaction temperature and time. (author).52 refs., 40 tabs., 129 figs

  11. Prevention of nuclear fuel cladding materials corrosion

    International Nuclear Information System (INIS)

    Yang, K.R.; Yang, J.C.; Lee, I.C.; Kang, H.D.; Cho, S.W.; Whang, C.K.

    1983-01-01

    The only way which could be performed by the operator of nuclear power plant to minimizing the degradation of nuclear fuel cladding material is to control the water quality of primary coolant as specified standard conditions which dose not attack the cladding material. If the water quality of reactor coolant does not meet far from the specification, the failure will occure not only cladding material itself but construction material of primary system which contact with the coolant. The corrosion product of system material are circulate through the whole primary system with the coolant and activated by the neutron near the reactor core. The activated corrosion products and fission products which released from fuel rod to the coolant, so called crud, will repeate deposition and redeposition continuously on the fuel rod and construction material surface. As a result we should consider heat transfer problem. In this study following activities were performed; 1. The crud sample was taken from the spent fuel rod surface of Kori unit one and analized for radioactive element and non radioactive chemical species. 2. The failure mode of nuclear fuel cladding material was estimated by the investigation of releasing type of fission products from the fuel rod to the reactor coolant using the iodine isotopes concentration of reactor coolants. 3. A study was carried out on the sipping test results of spent fuel and a discussion was made on the water quality control records through the past three cycle operation period of Kori unit one plant. (Author)

  12. Nuclear fuel elements made from nanophase materials

    Science.gov (United States)

    Heubeck, Norman B.

    1998-01-01

    A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000.degree. F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics.

  13. Materials and processes for solar fuel production

    CERN Document Server

    Viswanathan, Balasubramanian; Lee, Jae Sung

    2014-01-01

    This book features different approaches to non-biochemical pathways for solar fuel production. This one-of-a-kind book addresses photovoltaics, photocatalytic water splitting for clean hydrogen production and CO2 conversion to hydrocarbon fuel through in-depth comprehensive contributions from a select blend of established and experienced authors from across the world. The commercial application of solar based systems, with particular emphasis on non-PV based devices have been discussed. This book intends to serve as a primary resource for a multidisciplinary audience including chemists, engineers and scientists providing a one-stop location for all aspects related to solar fuel production. The material is divided into three sections: Solar assisted water splitting to produce hydrogen; Solar assisted CO2 utilization to produce green fuels and Solar assisted electricity generation. The content strikes a balance between theory, material synthesis and application with the central theme being solar fuels.

  14. Corrosion of fuel assembly materials

    International Nuclear Information System (INIS)

    Noe, M.; Frejaville, G.; Beslu, P.

    1985-08-01

    Corrosion of zircaloy-4 is reviewed in relation with previsions of improvement in PWRs performance: higher fuel burnup; increase coolant temperature, implying nucleate boiling on the hot clad surfaces; increase duration of the cycle due to load-follow operation. Actual knowledge on corrosion rates, based partly on laboratory tests, is insufficient to insure that external clad corrosion will not constitute a limitation to these improvements. Therefore, additional testing within representative conditions is felt necessary [fr

  15. Novel materials for fuel cells operating on liquid fuels

    Directory of Open Access Journals (Sweden)

    César A. C. Sequeira

    2017-05-01

    Full Text Available Towards commercialization of fuel cell products in the coming years, the fuel cell systems are being redefined by means of lowering costs of basic elements, such as electrolytes and membranes, electrode and catalyst materials, as well as of increasing power density and long-term stability. Among different kinds of fuel cells, low-temperature polymer electrolyte membrane fuel cells (PEMFCs are of major importance, but their problems related to hydrogen storage and distribution are forcing the development of liquid fuels such as methanol, ethanol, sodium borohydride and ammonia. In respect to hydrogen, methanol is cheaper, easier to handle, transport and store, and has a high theoretical energy density. The second most studied liquid fuel is ethanol, but it is necessary to note that the highest theoretically energy conversion efficiency should be reached in a cell operating on sodium borohydride alkaline solution. It is clear that proper solutions need to be developed, by using novel catalysts, namely nanostructured single phase and composite materials, oxidant enrichment technologies and catalytic activity increasing. In this paper these main directions will be considered.

  16. Chemical characterization of nuclear fuel materials

    International Nuclear Information System (INIS)

    Ramakumar, K.L.

    2011-01-01

    India is fabricating nuclear fuels for various types of reactors, for example, (U-Pu) MOX fuel of varying Pu content for boiling water reactors (BWRs), pressurized heavy water reactors (PHWRs), prototype fast breeder reactors (PFBRs), (U-Pu) carbide fuel fast breeder test reactor (FBTR), and U-based fuels for research reactors. Nuclear fuel being the heart of the reactor, its chemical and physical characterisation is an important component of this design. Both the fuel materials and finished fuel products are to be characterised for this purpose. Quality control (both chemical and physical) provides a means to ensure that the quality of the fabricated fuel conforms to the specifications for the fuel laid down by the fuel designer. Chemical specifications are worked out for the major and minor constituents which affect the fuel properties and hence its performance under conditions prevailing in an operating reactor. Each fuel batch has to be subjected to comprehensive chemical quality control for trace constituents, stoichiometry and isotopic composition. A number of advanced process and quality control steps are required to ensure the quality of the fuels. Further more, in the case of Pu-based fuels, it is necessary to extract maximum quality data by employing different evaluation techniques which would result in minimum scrap/waste generation of valuable plutonium. The task of quality control during fabrication of nuclear fuels of various types is both challenging and difficult. The underlying philosophy is total quality control of the fuel by proper mix of process and quality control steps at various stages of fuel manufacture starting from the feed materials. It is also desirable to adapt more than one analytical technique to increase the confidence and reliability of the quality data generated. This is all the most required when certified reference materials are not available. In addition, the adaptation of non-destructive techniques in the chemical quality

  17. Waste Plastic Converting into Hydrocarbon Fuel Materials

    Energy Technology Data Exchange (ETDEWEB)

    Sarker, Moinuddin; Mamunor Rashid, Mohammad; Molla, Mohammad

    2010-09-15

    The increased demand and high prices for energy sources are driving efforts to convert organic compounds into useful hydrocarbon fuels. Although much of this work has focused on biomass, there are strong benefits to deriving fuels from waste plastic material. Natural State Research Inc. (NSR) has invented a simple and economically viable process to decompose the hydrocarbon polymers of waste plastic into the shorter chain hydrocarbon of liquid fuel (patent pending). The method and principle of the production / process will be discussed. Initial tests with several widely used polymers indicate a high potential for commercialization.

  18. Advanced materials for solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, T.; Stevenson, J.

    1995-12-31

    The purpose of this research is to improve the properties of the current state-of-the-art materials used for solid oxide fuel cells (SOFCs). The objectives are to: (1) develop materials based on modifications of the state-of-the-art materials; (2) minimize or eliminate stability problems in the cathode, anode, and interconnect; (3) Electrochemically evaluate (in reproducible and controlled laboratory tests) the current state-of-the-art air electrode materials and cathode/electrolyte interfacial properties; (4) Develop accelerated electrochemical test methods to evaluate the performance of SOFCs under controlled and reproducible conditions; and (5) Develop and test materials for use in low-temperature SOFCs.

  19. Nuclear Fuels & Materials Spotlight Volume 5

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-10-01

    As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • Evaluation and modeling of light water reactor accident tolerant fuel concepts • Status and results of recent TRISO-coated particle fuel irradiations, post-irradiation examinations, high-temperature safety testing to demonstrate the accident performance of this fuel system, and advanced microscopy to improve the understanding of fission product transport in this fuel system. • Improvements in and applications of meso and engineering scale modeling of light water reactor fuel behavior under a range of operating conditions and postulated accidents (e.g., power ramping, loss of coolant accident, and reactivity initiated accidents) using the MARMOT and BISON codes. • Novel measurements of the properties of nuclear (actinide) materials under extreme conditions, (e.g. high pressure, low/high temperatures, high magnetic field) to improve the scientific understanding of these materials. • Modeling reactor pressure vessel behavior using the GRIZZLY code. • New methods using sound to sense temperature inside a reactor core. • Improved experimental capabilities to study the response of fusion reactor materials to a tritium plasma. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at Idaho National Laboratory, and hope that you find this issue informative.

  20. Stability of solid oxide fuel cell materials

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, T.R.; Bates, J.L.; Chick, L.A. [Pacific Northwest Lab., Richland, WA (United States)

    1996-04-01

    Interconnection materials in a solid oxide fuel cell are exposed to both highly oxidizing conditions at the cathode and to highly reducing conditions at the anode. The thermal expansion characteristics of substituted lanthanum and yttrium chromite interconnect materials were evaluated by dilatometry as a function of oxygen partial pressures from 1 atm to 10{sup -18} atm, controlled using a carbon dioxide/hydrogen buffer.

  1. Final Report: Conceptual Design of an Electron Accelerator for Bio-Solid Waste Treatment

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, Charles [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States)

    2017-09-20

    Several studies have identified electron beam (EB) irradiation of municipal wastewater and bio-solids as an effective and promising approach to the environmental remediation of the enormous quantities of human waste created by a growing world-wide population and increased urbanization. However, despite the technical success of experimental and pilot programs over the last several decades, the technique is still not in commercial use anywhere in the world. In addition, the report also identifies the need for “Financial and infrastructure participation from a utility for demonstration project” and “Education and awareness of safety of utilizing electron beam technology” as two additional roadblocks preventing technology adoption of EB treatment for bio-solids. In this concept design, we begin to address these barriers by working with Metropolitan Water Reclamation District of Greater Chicago (MWRD) and by the applying the latest accelerator technologies developed at Fermilab and within the DOE Office of Science laboratory complex.

  2. Nuclear Fuels & Materials Spotlight Volume 4

    Energy Technology Data Exchange (ETDEWEB)

    I. J. van Rooyen,; T. M. Lillo; Y. Q. WU; P.A. Demkowicz; L. Scott; D.M. Scates; E. L. Reber; J. H. Jackson; J. A. Smith; D.L. Cottle; B.H. Rabin; M.R. Tonks; S.B. Biner; Y. Zhang; R.L. Williamson; S.R. Novascone; B.W. Spencer; J.D. Hales; D.R. Gaston; C.J. Permann; D. Anders; S.L. Hayes; P.C. Millett; D. Andersson; C. Stanek; R. Ali; S.L. Garrett; J.E. Daw; J.L. Rempe; J. Palmer; B. Tittmann; B. Reinhardt; G. Kohse; P. Ramuhali; H.T. Chien; T. Unruh; B.M. Chase; D.W. Nigg; G. Imel; J. T. Harris

    2014-04-01

    As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • The first identification of silver and palladium migrating through the SiC layer in TRISO fuel • A description of irradiation assisted stress corrosion testing capabilities that support commercial light water reactor life extension • Results of high-temperature safety testing on coated particle fuels irradiated in the ATR • New methods for testing the integrity of irradiated plate-type reactor fuel • Description of a 'Smart Fuel' concept that wirelessly provides real time information about changes in nuclear fuel properties and operating conditions • Development and testing of ultrasonic transducers and real-time flux sensors for use inside reactor cores, and • An example of a capsule irradiation test. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps to spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at INL, and hope that you find this issue informative.

  3. Materials Approach to Fuel Efficient Tires

    Energy Technology Data Exchange (ETDEWEB)

    Votruba-Drzal, Peter [PPG Industries, Monroeville, PA (United States); Kornish, Brian [PPG Industries, Monroeville, PA (United States)

    2015-06-30

    The objective of this project was to design, develop, and demonstrate fuel efficient and safety regulation compliant tire filler and barrier coating technologies that will improve overall fuel efficiency by at least 2%. The program developed and validated two complementary approaches to improving fuel efficiency through tire improvements. The first technology was a modified silica-based product that is 15% lower in cost and/or enables a 10% improvement in tread wear while maintaining the already demonstrated minimum of 2% improvement in average fuel efficiency. The second technology was a barrier coating with reduced oxygen transmission rate compared to the state-of-the-art halobutyl rubber inner liners that will provide extended placarded tire pressure retention at significantly reduced material usage. A lower-permeance, thinner inner liner coating which retains tire pressure was expected to deliver the additional 2% reduction in fleet fuel consumption. From the 2006 Transportation Research Board Report1, a 10 percent reduction in rolling resistance can reduce consumer fuel expenditures by 1 to 2 percent for typical vehicles. This savings is equivalent to 6 to 12 gallons per year. A 1 psi drop in inflation pressure increases the tire's rolling resistance by about 1.4 percent.

  4. Fugitive binder for nuclear fuel materials

    International Nuclear Information System (INIS)

    Gallivan, T.J.

    1980-01-01

    A compound consisting of ammonium cations and carbonate, bicarbonate, or carbamate anions, or a mixture of such compounds, is useful as a binder for uranium dioxide fuel pellets for which it is desired to maintain a certain degree of porosity, uniformity of pore size, a lack of interconnections between the pores, and the shape or configuration of the base material particles in the final article after sintering. Upon heating, these binders decompose into gases and leave substantially no impurities. A process for sintering green nuclear fuel pellets using these binders is provided. (LL)

  5. 49 CFR 392.51 - Reserve fuel; materials of trade.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 5 2010-10-01 2010-10-01 false Reserve fuel; materials of trade. 392.51 Section... COMMERCIAL MOTOR VEHICLES Fueling Precautions § 392.51 Reserve fuel; materials of trade. Small amounts of...) may be designated as materials of trade (see 49 CFR 171.8). (a) The aggregate gross weight of all...

  6. Recovery of weapon plutonium as feed material for reactor fuel

    International Nuclear Information System (INIS)

    Armantrout, G.A.; Bronson, M.A.; Choi, Jor-Shan

    1994-01-01

    This report presents preliminary considerations for recovering and converting weapon plutonium from various US weapon forms into feed material for fabrication of reactor fuel elements. An ongoing DOE study addresses the disposition of excess weapon plutonium through its use as fuel for nuclear power reactors and subsequent disposal as spent fuel. The spent fuel would have characteristics similar to those of commercial power spent fuel and could be similarly disposed of in a geologic repository

  7. MMSNF 2005. Materials models and simulations for nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Freyss, M.; Durinck, J.; Carlot, G.; Sabathier, C.; Martin, P.; Garcia, P.; Ripert, M.; Blanpain, P.; Lippens, M.; Schut, H.; Federov, A.V.; Bakker, K.; Osaka, M.; Miwa, S.; Sato, I.; Tanaka, K.; Kurosaki, K.; Uno, M.; Yamanaka, S.; Govers, K.; Verwerft, M.; Hou, M.; Lemehov, S.E.; Terentyev, D.; Govers, K.; Kotomin, E.A.; Ashley, N.J.; Grimes, R.W.; Van Uffelen, P.; Mastrikov, Y.; Zhukovskii, Y.; Rondinella, V.V.; Kurosaki, K.; Uno, M.; Yamanaka, S.; Minato, K.; Phillpot, S.; Watanabe, T.; Shukla, P.; Sinnott, S.; Nino, J.; Grimes, R.; Staicu, D.; Hiernaut, J.P.; Wiss, T.; Rondinella, V.V.; Ronchi, C.; Yakub, E.; Kaye, M.H.; Morrison, C.; Higgs, J.D.; Akbari, F.; Lewis, B.J.; Thompson, W.T.; Gueneau, C.; Gosse, S.; Chatain, S.; Dumas, J.C.; Sundman, B.; Dupin, N.; Konings, R.; Noel, H.; Veshchunov, M.; Dubourg, R.; Ozrin, C.V.; Veshchunov, M.S.; Welland, M.T.; Blanc, V.; Michel, B.; Ricaud, J.M.; Calabrese, R.; Vettraino, F.; Tverberg, T.; Kissane, M.; Tulenko, J.; Stan, M.; Ramirez, J.C.; Cristea, P.; Rachid, J.; Kotomin, E.; Ciriello, A.; Rondinella, V.V.; Staicu, D.; Wiss, T.; Konings, R.; Somers, J.; Killeen, J

    2006-07-01

    The MMSNF Workshop series aims at stimulating research and discussions on models and simulations of nuclear fuels and coupling the results into fuel performance codes.This edition was focused on materials science and engineering for fuel performance codes. The presentations were grouped in three technical sessions: fundamental modelling of fuel properties; integral fuel performance codes and their validation; collaborations and integration of activities. (A.L.B.)

  8. Sealing materials for solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, P.H.

    1999-02-01

    A major obstacle in the achievement of high electrical efficiency for planar solid oxide fuel cell stacks (SOFC) is the need for long term stable seals at the operational temperature between 850 and 1000 deg. C. In the present work the formation and properties of sealing materials for SOFC stacks that fulfil the necessary requirements were investigated. The work comprises analysis of sealing material properties independently, in simple systems as well as tests in real SOFC stacks. The analysed sealing materials were based on pure glasses or glass-ceramic composites having B{sub 2}O{sub 3}, P{sub 2}O{sub 5} or siO{sub 2} as glass formers, and the following four glass systems were investigated: MgO/caO/Cr{sub 2}O{sub 3}-Al{sub 2}O{sub 3}B{sub 2}O{sub 3}-P{sub 2}O{sub 5}, MgO-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}, MgO-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}-SiO{sub 2} and BaO/Na{sub 2}O-Al{sub 2}O{sub 3}-SiO{sub 2}. (au) 32 tabs., 106 ills., 107 refs.

  9. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    International Nuclear Information System (INIS)

    Kim, Kyu-Tae

    2013-01-01

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10 −6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

  10. Aerospace Fuels From Nonpetroleum Raw Materials

    Science.gov (United States)

    Palaszewski, Bryan A.; Hepp, Aloysius F.; Kulis, Michael J.; Jaworske, Donald A.

    2013-01-01

    Recycling human metabolic and plastic wastes minimizes cost and increases efficiency by reducing the need to transport consumables and return trash, respectively, from orbit to support a space station crew. If the much larger costs of transporting consumables to the Moon and beyond are taken into account, developing waste recycling technologies becomes imperative and possibly mission enabling. Reduction of terrestrial waste streams while producing energy and/or valuable raw materials is an opportunity being realized by a new generation of visionary entrepreneurs; several relevant technologies are briefly compared, contrasted and assessed for space applications. A two-step approach to nonpetroleum raw materials utilization is presented; the first step involves production of supply or producer gas. This is akin to synthesis gas containing carbon oxides, hydrogen, and simple hydrocarbons. The second step involves production of fuel via the Sabatier process, a methanation reaction, or another gas-to-liquid technology, typically Fischer-Tropsch processing. Optimization to enhance the fraction of product stream relevant to transportation fuels via catalytic (process) development at NASA Glenn Research Center is described. Energy utilization is a concern for production of fuels whether for operation on the lunar or Martian surface, or beyond. The term green relates to not only mitigating excess carbon release but also to the efficiency of energy usage. For space, energy usage can be an essential concern. Another issue of great concern is minimizing impurities in the product stream(s), especially those that are potential health risks and/or could degrade operations through catalyst poisoning or equipment damage; technologies being developed to remove heteroatom impurities are discussed. Alternative technologies to utilize waste fluids, such as a propulsion option called the resistojet, are discussed. The resistojet is an electric propulsion technology with a powered

  11. Center for Intelligent Fuel Cell Materials Design

    Energy Technology Data Exchange (ETDEWEB)

    Santurri, P.R., (Chemsultants International); Hartmann-Thompson, C.; Keinath, S.E. (Michigan Molecular Inst.)

    2008-08-26

    The goal of this work was to develop a composite proton exchange membrane utilizing 1) readily available, low cost materials 2) readily modified and 3) easily processed to meet the chemical, mechanical and electrical requirements of high temperature PEM fuel cells. One of the primary goals was to produce a conducting polymer that met the criteria for strength, binding capability for additives, chemical stability, dimensional stability and good conductivity. In addition compatible, specialty nanoparticles were synthesized to provide water management and enhanced conductivity. The combination of these components in a multilayered, composite PEM has demonstrated improved conductivity at high temperatures and low humidity over commercially available polymers. The research reported in this final document has greatly increased the knowledge base related to post sulfonation of chemically and mechanically stable engineered polymers (Radel). Both electrical and strength factors for the degree of post sulfonation far exceed previous data, indicating the potential use of these materials in suitable proton exchange membrane architectures for the development of fuel cells. In addition compatible, hydrophilic, conductive nano-structures have been synthesized and incorporated into unique proton exchange membrane architectures. The use of post sulfonation for the engineered polymer and nano-particle provide cost effective techniques to produce the required components of a proton exchange membrane. The development of a multilayer proton exchange membrane as described in our work has produced a highly stable membrane at 170°C with conductivities exceeding commercially available proton exchange membranes at high temperatures and low humidity. The components and architecture of the proton exchange membrane discussed will provide low cost components for the portable market and potentially the transportation market. The development of unique components and membrane architecture

  12. Composting of bio solids by composting tunnels; Compostaje de biosolidos mediante tunes de compostado

    Energy Technology Data Exchange (ETDEWEB)

    Varo, P.; Rodriguez, M.; Prats, D.; Soto, R.; Pastor, B.; Monges, M.

    2003-07-01

    The objective of this work is to study the bio-solid composting process carried out in the composting plant of Aspe (Alicante) by means of open composting tunnels, and to determine the quality of the resulting compost. The parameters under control are temperature. humidity, density, pH, conductivity, organic matter, C/N ratio, ammonium nitride and organic nitrogen. The concentrations of cadmium, chromium, nickel, lead and copper were monitored during the composting process. Observing the parameters analyzed we can conclude that the composting process of the sewage sludge from Aspe procedures a product suitable for agricultural use. The values obtained allow the product resulting from the process to be designated as compost. (Author)

  13. Development of new membrane materials for direct methanol fuel cells

    NARCIS (Netherlands)

    Yildirim, M.H.

    2009-01-01

    Development of new membrane materials for direct methanol fuel cells Direct methanol fuel cells (DMFCs) can convert the chemical energy of a fuel directly into electrical energy with high efficiency and low emission of pollutants. DMFCs can be used as the power sources to portable electronic devices

  14. Fuel cell electrode interconnect contact material encapsulation and method

    Energy Technology Data Exchange (ETDEWEB)

    Derose, Anthony J.; Haltiner, Jr., Karl J.; Gudyka, Russell A.; Bonadies, Joseph V.; Silvis, Thomas W.

    2016-05-31

    A fuel cell stack includes a plurality of fuel cell cassettes each including a fuel cell with an anode and a cathode. Each fuel cell cassette also includes an electrode interconnect adjacent to the anode or the cathode for providing electrical communication between an adjacent fuel cell cassette and the anode or the cathode. The interconnect includes a plurality of electrode interconnect protrusions defining a flow passage along the anode or the cathode for communicating oxidant or fuel to the anode or the cathode. An electrically conductive material is disposed between at least one of the electrode interconnect protrusions and the anode or the cathode in order to provide a stable electrical contact between the electrode interconnect and the anode or cathode. An encapsulating arrangement segregates the electrically conductive material from the flow passage thereby, preventing volatilization of the electrically conductive material in use of the fuel cell stack.

  15. Status and development of RBMK fuel rods and reactor materials

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Reshetnikov, F.G.; Ioltukhovsky, A.G.

    1998-01-01

    The paper presents current status and development of RBMK fuel rods and reactor materials. With regard to fuel rod cladding the following issues have been discussed: corrosion, tensile properties, welding technology and testing of an alternative cladding alloy with a composition of Zr-Nb-Sn-Fe. Erbium doped fuel has been suggested for safety improvement. Also analysis of fuel reliability is presented in the paper. (author)

  16. The regulations concerning the uses of nuclear fuel materials

    International Nuclear Information System (INIS)

    1978-01-01

    The Regulations are established on the basis of ''The law for the regulations of nuclear source materials, nuclear fuel materials and reactors'' and the ''Provisions concerning the usage of nuclear fuel materials'' in the Enforcement Ordinance of the Law, to enforce such provisions. Terms are explained, such as exposure radiation dose, cumulative dose, control area, surrounding inspection area, persons engaging in works, area for incoming and outgoing of materials, batch, real stocks, effective value and main measuring points. In the applications for the permission to use nuclear fuel materials, the expected period and quantity of usage of each kind of such materials and the other party and the method of selling, lending and returning spent fuel or the process of disposal of such fuel must be written. Explanations concerning the technical ability required for the usage of nuclear fuel materials shall be attached to the applications. Applications shall be filed for the inspection of facilities for use, in which the name and the address of the applicant, the name and the address of the factory or the establishment, the range of the facilities for use, the maximum quantity of nuclear fuel materials to be used or stocked, and the date, the place and the kind of the expected inspection are written. Prescriptions cover the records to be held, safety regulations, the technical standards for usage, the disposal, transport and storage of nuclear fuel materials and the reports to be filed. (Okada, K.)

  17. Review of the IAEA nuclear fuel cycle and material section activities connected with nuclear fuel including WWER fuel

    International Nuclear Information System (INIS)

    Sokolov, F.

    2001-01-01

    Program activities on Nuclear Fuel Cycle and Materials cover the areas of: 1) raw materials (B.1.01); 2) fuel performance and technology (B.1.02); 3) pent fuel (B.1.03); 4) fuel cycle issues and information system (B.1.04); 5) support to technical cooperation activities (B.1.05). The IAEA activities in fuel performance and technology in 2001 include organization of the fuel experts meetings and completion of the Co-ordinate Research Projects (CRP). The special attention is given to the advanced post-irradiation examination techniques for water reactor fuel and fuel behavior under transients and LOCA conditions. An international research program on modeling of activity transfer in primary circuit of NPP is finalized in 2001. A new CRP on fuel modeling at extended burnup (FUMEX II) has planed to be carried out during the period 2002-2006. In the area of spent fuel management the implementation of burnup credit (BUC) in spent fuel management systems has motivated to be used in criticality safety applications, based on economic consideration. An overview of spent fuel storage policy accounting new fuel features as higher enrichment and final burnup, usage of MOX fuel and prolongation of the term of spent fuel storage is also given

  18. A Path Forward to Advanced Nuclear Fuels: Spectroscopic Calorimetry of Nuclear Fuel Materials

    International Nuclear Information System (INIS)

    Tobin, J.G.

    2009-01-01

    The goal is to relieve the shortage of thermodynamic and kinetic information concerning the stability of nuclear fuel alloys. Past studies of the ternary nuclear fuel UPuZr have demonstrated constituent redistribution when irradiated or with thermal treatment. Thermodynamic data is key to predicting the possibilities of effects such as constituent redistribution within the fuel rods and interaction with cladding materials

  19. Nuclear fuels for material test reactors

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Durazzo, M.; Freitas, C.T. de

    1982-01-01

    Experimental results related do the development of nuclear fuels for reactors cooled and moderated by water have been presented cylindrical and plate type fuels have been described in which the core consists of U compouns dispersed in an Al matrix and is clad with aluminium. Fabrication details involving rollmilling, swaging or hot pressing have been described. Corrosion and irradiation test results are also discussed. The performance of the different types of fuels indicates that it is possible to locally fabricate fuel plates with U 3 O 8 +Al cores (20% enriched U) for use in operating Brazilian research reactors. (Author) [pt

  20. Multiscale NTP Fuel Element Materials Simulation

    Data.gov (United States)

    National Aeronautics and Space Administration — Project will leverage a multiscale modeling approach pioneered for light water reactor (LWR) fuels to simulate performance in a prototypical environment. The...

  1. Sulfur Release from Cement Raw Materials during Solid Fuel Combustion

    DEFF Research Database (Denmark)

    Nielsen, Anders Rooma; Larsen, Morten B.; Glarborg, Peter

    2011-01-01

    deposit formation in the kiln system. SO2 release from cement raw materials during combustion of solid fuels has been studied experimentally in a high temperature rotary drum. The fuels were tire rubber, pine wood, petcoke, sewage sludge, and polypropylene. The SO2 release from the raw materials...... was observed to increase when (a) the inlet oxygen concentration decreased, (b) the temperature increased, and (c) when the total surface area of the fuel particles increased. The type of fuel also had a significant effect on the SO2 release. The sequence of SO2 release for fuel particles in the size interval...... 1–2 mm was found to be tire rubber granulate > sewage sludge > pine wood sawdust > petcoke > polypropylene flakes. The SO2 release was generally observed to increase when formation of local reducing conditions near the raw material bed increased, as indicated by elevated levels of CO during the fuel...

  2. Regulation on the transport of nuclear fuel materials by vehicles

    International Nuclear Information System (INIS)

    1984-01-01

    The regulations applying to the transport of nuclear fuel materials by vehicles, mentioned in the law for the regulations of nuclear source materials, nuclear fuel materials and reactors. The transport is for outside of the factories and the site of enterprises by such modes of transport as rail, trucks, etc. Covered are the following: definitions of terms, places of fuel materials handling, loading methods, limitations on mix loading with other cargo, radiation dose rates concerning the containers and the vehicles, transport indexes, signs and indications, limitations on train linkage during transport by rail, security guards, transport of empty containers, etc. together with ordinary rail cargo and so on. (Mori, K.)

  3. PEM fuel cell bipolar plate material requirements for transportation applications

    Energy Technology Data Exchange (ETDEWEB)

    Borup, R.L.; Stroh, K.R.; Vanderborgh, N.E. [Los Alamos National Lab., NM (United States)] [and others

    1996-04-01

    Cost effective bipolar plates are currently under development to help make proton exchange membrane (PEM) fuel cells commercially viable. Bipolar plates separate individual cells of the fuel cell stack, and thus must supply strength, be electrically conductive, provide for thermal control of the fuel stack, be a non-porous materials separating hydrogen and oxygen feed streams, be corrosion resistant, provide gas distribution for the feed streams and meet fuel stack cost targets. Candidate materials include conductive polymers and metal plates with corrosion resistant coatings. Possible metals include aluminium, titanium, iron/stainless steel and nickel.

  4. Regulations concerning the fabricating business of nuclear fuel materials

    International Nuclear Information System (INIS)

    1978-01-01

    The Regulation is revised on the basis of ''The law for the regulations of nuclear source materials, nuclear fuel materials and reactors'' and the ''Provisions concerning the enterprises processing nuclear fuel materials'' in the Enforcement Ordinance for the Law, to enforce such provisions. This is the complete revision of the regulation of the same name in 1957. Terms are explained, such as exposure radiation dose, cumulative dose, control area, surrounding inspection area, persons engaged in works, radioactive wastes, area for incoming and outgoing of materials, fluctuation of stocks, batch, real stocks, effective value and main measuring points. For the applications for the permission of the enterprises processing nuclear fuel materials, the location of an enterprise, the construction of buildings and the construction of and the equipments for facilities of chemical processing, forming, coating, assembling, storage of nuclear fuel materials, disposal of radioactive wastes and radiation control must be written. Records shall be made and maintained for the periods specified on the inspection of processing facilities, nuclear fuel materials, radiation control, operation, maintainance, accidents of processing facilities and weather. Limit to entrance into the control area, measures for exposure radiation dose, patrol and inspection, operation of processing facilities, transport of materials, disposal of radioactive wastes, safety regulations are provided for. Reports to be filed by the persons engaging in the enterprises processing nuclear fuel materials are prescribed. (Okada, K.)

  5. Device and materials modeling in PEM fuel cells

    CERN Document Server

    Promislow, Keith

    2009-01-01

    Device and Materials Modeling in PEM Fuel Cells is a specialized text that compiles the mathematical details and results of both device and materials modeling in a single volume. Proton exchange membrane (PEM) fuel cells will likely have an impact on our way of life similar to the integrated circuit. The potential applications range from the micron scale to large scale industrial production. Successful integration of PEM fuel cells into the mass market will require new materials and a deeper understanding of the balance required to maintain various operational states. This book contains articles from scientists who contribute to fuel cell models from both the materials and device perspectives. Topics such as catalyst layer performance and operation, reactor dynamics, macroscopic transport, and analytical models are covered under device modeling. Materials modeling include subjects relating to the membrane and the catalyst such as proton conduction, atomistic structural modeling, quantum molecular dynamics, an...

  6. Effect of heavy metals from soils amended with bio solids and sowed with forages on the abundance and biodiversity of edaphic arthropods

    International Nuclear Information System (INIS)

    Flores-Pardave, L.; Flores-Tena, F. J.; Hernandez-Sanchez, A. J.

    2009-01-01

    There are many studies about positive effects of bio solids application to ameliorate grain and forage production. However it is necessary to know more about the effects of this by-product on edaphic biota. Therefore the goal of this study was to know the effects of heavy metals from bio solids if the wastewater treatment plant of Aguascalientes city (Mexico) on adaphic arthropods in soils sowed with lucerne and corn amended with bio solids at 200 (low), 400 middle) and 800 (high) ton/ha wet weight. (Author)

  7. Fusion fuel cycle: material requirements and potential effluents

    International Nuclear Information System (INIS)

    Teofilo, V.L.; Bickford, W.E.; Long, L.W.; Price, B.A.; Mellinger, P.J.; Willingham, C.E.; Young, J.K.

    1980-10-01

    Environmental effluents that may be associated with the fusion fuel cycle are identified. Existing standards for controlling their release are summarized and anticipated regulatory changes are identified. The ability of existing and planned environmental control technology to limit effluent releases to acceptable levels is evaluated. Reference tokamak fusion system concepts are described and the principal materials required of the associated fuel cycle are analyzed. These materials include the fusion fuels deuterium and tritium; helium, which is used as a coolant for both the blanket and superconducting magnets; lithium and beryllium used in the blanket; and niobium used in the magnets. The chemical and physical processes used to prepare these materials are also described

  8. Fusion fuel cycle: material requirements and potential effluents

    Energy Technology Data Exchange (ETDEWEB)

    Teofilo, V.L.; Bickford, W.E.; Long, L.W.; Price, B.A.; Mellinger, P.J.; Willingham, C.E.; Young, J.K.

    1980-10-01

    Environmental effluents that may be associated with the fusion fuel cycle are identified. Existing standards for controlling their release are summarized and anticipated regulatory changes are identified. The ability of existing and planned environmental control technology to limit effluent releases to acceptable levels is evaluated. Reference tokamak fusion system concepts are described and the principal materials required of the associated fuel cycle are analyzed. These materials include the fusion fuels deuterium and tritium; helium, which is used as a coolant for both the blanket and superconducting magnets; lithium and beryllium used in the blanket; and niobium used in the magnets. The chemical and physical processes used to prepare these materials are also described.

  9. Technical limitations of nuclear fuel materials and structures

    International Nuclear Information System (INIS)

    Hansson, L.; Planman, T.; Vitikainen, E.

    1993-05-01

    This report gives a summary of the tasks carried out within the project 'Technical limitations of nuclear fuel materials and structures' which belongs to the Finnish national research programme called 'Systems behaviour and operational aspects of safety'. The duration of the project was three years from 1990 to 1992. Most western LWR utilities, including the two Finnish ones have an incentive to implement extended burnup fuel cycles in their nuclear power plants. The aim of this project has been authorities to support them in the assessment and licensing of new fuel designs and materials. The research work of the project was focused on collecting and qualifying fuel performance data and on performing laboratory tests on fresh and irradiated cladding and structural materials. Moreover, knowledge of the high burnup phenomena was obtained through participation in international research projects such as OECD Halden Project and several Studsvik projects. Experimental work within the framework of the VVER fuel cooperative effort was also continued. (orig.)

  10. Development of alternative materials for BWR fuel springs

    International Nuclear Information System (INIS)

    Uruma, Y.; Osato, T.; Yamazaki, K.

    2002-01-01

    Major sources of radioactivity introduced into reactor water of BWR were estimated fuel crud and in-core materials (especially, fuel springs). Fuel springs are used for fixation of fuel cladding tubes with spacer grid. Those are small parts (total length is only within 25 mm) and so many numbers are loaded simultaneously and then total surfaces area are calculated up to about 200 m 2 . Fuel springs are located under high radiation field and high oxidative environment. Conventional fuel spring is made of alloy-X750 which is one of nickel-based alloy and is reported to show relatively higher corrosion release rate. 58 Co and 60 Co will be released directly into reactor water from intensely radio-activated fuel springs surface and increase radioactivity concentrations in primary coolant. Corrosion release control from fuel springs is an important technical item and a development of alternative material instead of alloy-X750 for fuel spring is a key subject to achieve ultra low man-rem exposure BWR plant. In present work, alloy-X718 which started usage for PWR fuel springs and stainless steel type 316L which has many mechanical property data are picked up for alternative materials and compared their corrosion behaviors with conventional material. Corrosion experiment was conducted under vapor-water two phases flow which is simulated fuel cladding surface boiling condition. After exposure, corrosion film formed under corrosion test was analyzed in detail and corrosion film amount and corrosion release amount are estimated among three materials. (authors)

  11. Characterization of spent fuel approved testing material---ATM-105

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01

    The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report.

  12. Development of fuels and structural materials for fast breeder reactors

    Indian Academy of Sciences (India)

    Fast breeder reactors (FBRs) are destined to play a crucial role inthe Indian nuclear power programme in the foreseeable future. FBR technology involves a multi-disciplinary approach to solve the various challenges in the areas of fuel and materials development. Fuels for FBRs have significantly higher concentration of ...

  13. Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2009-01-01

    The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U 3 Si 2 -Al followed by 0.03% for U 3 Si-Al, and 0.01% for U 3 O 8 -Al fuel. The U 3 O 8 -Al fueled reactor gave the maximum ρ excess at BOL which was 21.67% more than the original fuel followed by U 3 Si-Al which was 2.55% more, while that of U 3 Si 2 -Al was 2.50% more than the original UAl x -Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U 3 O 8 -Al followed by U 3 Si-Al and then U 3 Si 2 -Al fuel.

  14. Advances in nuclear fuel cycle materials and concepts. Vol. 1

    International Nuclear Information System (INIS)

    El-Sayed, A.A.

    1996-01-01

    This presentation gives an overview of the new trends in the materials used in various steps of the nuclear fuel cycle. This will cover fuels for various types of reactors (PWRs, HTRs, ... etc.) cladding materials, control rod materials, reactor structural materials, as well as materials used in the back end of the fuel cycle. Problems associated with corrosion of fuel cladding materials as well as those in control rod materials (B 4 C swelling...etc.), and approaches for combating these influences are reviewed. For the case of reactor pressure vessel materials issues related to the influences of alloy composition, design approaches including the use of more forged parts and minimizing, as for as possible, longitudinal welds especially in the central region, are discussed. Furthermore the application of techniques for recovery of pre-irradiation mechanical properties of PVS components is also covered. New candidate materials for the construction of high level waste containers including modified types of stainless steel (high Ni and high MO), nickel-base alloys and titanium alloys are also detailed. Finally, nuclear fuel cycle concepts involving plutonium and actinides recycling shall be reviewed. 28 figs., 6 tabs

  15. Packaging, carriage and dispatching fuel and radioactive materials, IAEA regulations

    International Nuclear Information System (INIS)

    White, M.

    1981-01-01

    The need to bring fuel and other radioactive substances into a nuclear power plant and to send out irradiated or contaminated materials: spent fuel, activated equipment, used filters, resin and clothing, etc. gives rise to the question: How can these materials be transported safely and economically. The purpose of this paper is to answer that question by providing information on the regulatory requirements that have been developed for packaging, labelling and handling and on the containers which are being employed. (orig./RW)

  16. Research of fuel temperature control in fuel pipeline of diesel engine using positive temperature coefficient material

    Directory of Open Access Journals (Sweden)

    Xiaolu Li

    2016-01-01

    Full Text Available As fuel temperature increases, both its viscosity and surface tension decrease, and this is helpful to improve fuel atomization and then better combustion and emission performances of engine. Based on the self-regulated temperature property of positive temperature coefficient material, this article used a positive temperature coefficient material as electric heating element to heat diesel fuel in fuel pipeline of diesel engine. A kind of BaTiO3-based positive temperature coefficient material, with the Curie temperature of 230°C and rated voltage of 24 V, was developed, and its micrograph and element compositions were also analyzed. By the fuel pipeline wrapped in six positive temperature coefficient ceramics, its resistivity–temperature and heating characteristics were tested on a fuel pump bench. The experiments showed that in this installation, the surface temperature of six positive temperature coefficient ceramics rose to the equilibrium temperature only for 100 s at rated voltage. In rated power supply for six positive temperature coefficient ceramics, the temperature of injection fuel improved for 21°C–27°C within 100 s, and then could keep constant. Using positive temperature coefficient material to heat diesel in fuel pipeline of diesel engine, the injection mass per cycle had little change, approximately 0.3%/°C. This study provides a beneficial reference for improving atomization of high-viscosity liquids by employing positive temperature coefficient material without any control methods.

  17. Characterization of spent fuel approved testing material: ATM-106

    International Nuclear Information System (INIS)

    Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thornhill, C.K.

    1988-10-01

    The characterization data obtained to date are described for Approved Testing Material (ATM)-106 spent fuel from Assembly BT03 of pressurized-water reactor Calvert Cliffs No. 1. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well- characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCWRM) program. ATM-106 consists of 20 full-length irradiated fuel rods with rod-average burnups of about 3700 GJ/kgM (43 MWd/kgM) and expected fission gas release of /approximately/10%. Characterization data include (1) as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) calculated nuclide inventories and radioactivities in the fuel and cladding; and (6) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel rod are being conducted and will be included in planned revisions of this report. 12 refs., 110 figs., 81 tabs

  18. Transient assembly of active materials fueled by a chemical reaction

    Science.gov (United States)

    Boekhoven, Job; Hendriksen, Wouter E.; Koper, Ger J. M.; Eelkema, Rienk; van Esch, Jan H.

    2015-09-01

    Fuel-driven self-assembly of actin filaments and microtubules is a key component of cellular organization. Continuous energy supply maintains these transient biomolecular assemblies far from thermodynamic equilibrium, unlike typical synthetic systems that spontaneously assemble at thermodynamic equilibrium. Here, we report the transient self-assembly of synthetic molecules into active materials, driven by the consumption of a chemical fuel. In these materials, reaction rates and fuel levels, instead of equilibrium composition, determine properties such as lifetime, stiffness, and self-regeneration capability. Fibers exhibit strongly nonlinear behavior including stochastic collapse and simultaneous growth and shrinkage, reminiscent of microtubule dynamics.

  19. Design and construction of the Fuels and Materials Examination Facility

    International Nuclear Information System (INIS)

    Burgess, C.A.

    1979-01-01

    Final design is more than 85 percent complete on the Fuels and Materials Examination Facility, the facility for post-irradiation examination of the fuels and materials tests irradiated in the FFTF and for fuel process development, experimental test pin fabrication and supporting storage, assay, and analytical chemistry functions. The overall facility is generally described with specific information given on some of the design features. Construction has been initiated and more than 10% of the construction contracts have been awarded on a fixed price basis

  20. Nuclear and materials aspects of the thorium fuel cycle

    Science.gov (United States)

    Rodriguez, P.; Sundaram, C. V.

    1981-09-01

    This paper is an attempt to assess and review the materials aspects of the thorium fuel cycle. It starts with an examination of the nuclear aspects of the thorium fuel cycle, meant as an introduction for materials scientists and engineers who may not normally be familiar with the concepts and terms involved. After defining and describing the thorium and uranium fuel cycles, the reasons for the resurgence of interest in the thorium fuel cycle and the technical and economic considerations that support its early adoption are examined. The reactor physics and fissile economics aspects of the thorium and uranium cycles are then compared. The specific reactor types suitable for the adoption of the thorium cycle are briefly examined and described. Subsequent sections of the paper are devoted to a detailed discussion of the materials aspects of the thorium fuel cycle. Available information on fabrication, refabrication and irradiation performance of thorium-based fuels for light water reactors, heavy water reactors, high temperature gas-cooled reactors, molten salt breeder reactors and fast breeder reactors is critically reviewed and analysed. Materials problems related to cladding and structural materials are also discussed whenever these are unique to the thorium cycle.

  1. Transfer of pollution from municipal wastewater to bio solids: their chemical characterization; Transferencia de contaminacion desde el agua residual urbana a los lodos de depuracion: caracterizacion de biosolidos

    Energy Technology Data Exchange (ETDEWEB)

    Marin Galvin, R.

    2007-07-01

    Production of bio solids depends on the amount of wastewater treated according the expression y=ax''2,5575, being y the log of TM/year of bio solids produced, and X the log then m''3/year of wastewater treated. Quality of bio solids generated by three WWTP does not seem function either of the amount of treated water and neither of the treatment process applied (active sludges or bio discs). The bio solids exhibited values of 20-25% in dehidradation, and those of organic matter, nitrogen and phosp hour being equal to 60-75%, 7% and 4%, respectively. Moreover, the total of metals there present were 11-19 g/kg over dried matter, supposing Fe, Zn, Cu and Mn the 97% of all metals, and being Hg the minority metal. Bio solids can be used in agricultural practices (they agree with the Spanish normative here applied) and they concentrated the organic matter and metals found in wastewater up to 417 and 869 times, respectively. At the same time, we have estimated that each 4841 of wastewater produced 1 kg of bio solid. (Author)

  2. High Temperature Gas Cooled Reactor Fuels and Materials

    International Nuclear Information System (INIS)

    2010-03-01

    At the third annual meeting of the technical working group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), held in Vienna, in 2004, it was suggested 'to develop manuals/handbooks and best practice documents for use in training and education in coated particle fuel technology' in the IAEA's Programme for the year 2006-2007. In the context of supporting interested Member States, the activity to develop a handbook for use in the 'education and training' of a new generation of scientists and engineers on coated particle fuel technology was undertaken. To make aware of the role of nuclear science education and training in all Member States to enhance their capacity to develop innovative technologies for sustainable nuclear energy is of paramount importance to the IAEA Significant efforts are underway in several Member States to develop high temperature gas cooled reactors (HTGR) based on either pebble bed or prismatic designs. All these reactors are primarily fuelled by TRISO (tri iso-structural) coated particles. The aim however is to build future nuclear fuel cycles in concert with the aim of the Generation IV International Forum and includes nuclear reactor applications for process heat, hydrogen production and electricity generation. Moreover, developmental work is ongoing and focuses on the burning of weapon-grade plutonium including civil plutonium and other transuranic elements using the 'deep-burn concept' or 'inert matrix fuels', especially in HTGR systems in the form of coated particle fuels. The document will serve as the primary resource materials for 'education and training' in the area of advanced fuels forming the building blocks for future development in the interested Member States. This document broadly covers several aspects of coated particle fuel technology, namely: manufacture of coated particles, compacts and elements; design-basis; quality assurance/quality control and characterization techniques; fuel irradiations; fuel

  3. IAEA programme on nuclear fuel cycle and materials technologies - 2009

    International Nuclear Information System (INIS)

    Killeen, J.

    2009-01-01

    In this paper a brief description and the main objectives of IAEA Programme B on Nuclear fuel cycle are given. The following Coordinated Research Projects: 1) Delayed Hydride Cracking (DHC); 2) Structural Materials Radiation Effects (SMoRE); 3) Water Chemistry (FUWAC) and 4) Fuel Modelling (FUMEX-III) are shortly described. The data collected by the IAEA Expert Group of Fuel Failures in Water Cooled Reactors including information about fuel assembly damage that did not result in breach of the fuel rod cladding, such as assembly bow or crud deposition an the experience with these unexpected fuel issues shows that they can seriously affect plant operations, and it is clear that concerns about reliability in this area are of similar importance today as fuel rod failures, at least for LWR fuel are discussed. Detection, examination and analysis of fuel failures and description of failures and mitigation measures as well as preparation of a Monograph on Zirconium including an overview of Zirconium for nuclear applications, including extraction, forming, properties and irradiation experience are presented

  4. Advanced materials for solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, T.R.; Stevenson, J.; Paulik, S.

    1996-12-31

    Purpose of the research is to improve the properties of current state- of-the-art materials used for SOFCs. The project includes interconnect development, high-performance cathode, electrochemical testing, and accelerated testing. This document reports results of mechanical tests (bend strength, elastic modulus, fracture strength) of acceptor-substituted lanthanum chromite (interconnect material).

  5. Solid oxide fuel cell (sofc) materials

    CERN Document Server

    Saravanan, R

    2018-01-01

    Developing materials for SOFC applications is one of the key topics in energy research. The book focuses on manganite structured materials, such as doped lanthanum chromites and lanthanum manganites, which have interesting properties: thermal and chemical stability, mixed ionic and electrical conductivity, electrocatalytic activity, magnetocaloric property and colossal magnetoresistance (CMR).

  6. Solid oxide fuel cell (SOFC) materials

    CERN Document Server

    Saravanan, R

    2018-01-01

    Developing materials for SOFC applications is one of the key topics in energy research. The book focuses on manganite structured materials, such as doped lanthanum chromites and lanthanum manganites, which have interesting properties: thermal and chemical stability, mixed ionic and electrical conductivity, electrocatalytic activity, magnetocaloric property and colossal magnetoresistance (CMR).

  7. Experience of air transport of nuclear fuel material in Japan

    International Nuclear Information System (INIS)

    Yamashita, T.; Toguri, D.; Kawasaki, M.

    2004-01-01

    Certified Reference Materials (hereafter called as to CRMs), which are indispensable for Quality Assurance and Material Accountability in nuclear fuel plants, are being provided by overseas suppliers to Japanese nuclear entities as Type A package (non-fissile) through air transport. However, after the criticality accident at JCO in Japan, special law defining nuclear disaster countermeasures (hereafter called as to the LAW) has been newly enforced in June 2000. Thereafter, nuclear fuel materials must meet not only to the existing transport regulations but also to the LAW for its transport

  8. FISSILE MATERIAL AND FUEL ELEMENTS FOR NEUTRONIC REACTORS

    Science.gov (United States)

    Shaner, B.E.

    1961-08-15

    The fissile material consists of about 64 to 70% (weight) zirconium dioxide, 15 to 19% uranium dioxide, and 8 to 17% calcium oxide. The fissile material is formed into sintered composites which are disposed in a compartmented fuel element, comprising essentially a flat filler plate having a plurality of compartments therein, enclosed in cladding plates of the same material as the filler plate. The resultant fuel has good resistance to corrosion in high temperature pressurized water, good dimensional stability to elevated temperatures, and good resistance to thermal shock. (AEC)

  9. The radioactivity of construction materials, fuel peat and peat ash

    International Nuclear Information System (INIS)

    1994-01-01

    The guide presents safety requirements related to radiation exposure from building materials and materials used in road and related construction. It also covers the safety requirements for the production of fuel peat and the handling and disposal of peat ash. Guidelines are also given for maintaining the required safety level. (5 refs.)

  10. Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2009-01-01

    The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 deg. C to 50 deg. C and 100 deg. C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U 3 O 8 -Al was about 2% more than the original UAl x -Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.

  11. 21st Century Renewable Fuels, Energy, and Materials

    Energy Technology Data Exchange (ETDEWEB)

    Berry, K. Joel [Kettering Univ., Flint, MI (United States); Das, Susanta K. [Kettering Univ., Flint, MI (United States)

    2012-11-29

    The objectives of this project were multi-fold: (i) conduct fundamental studies to develop a new class of high temperature PEM fuel cell material capable of conducting protons at elevated temperature (180°C), (ii) develop and fabricate a 5k We novel catalytic flat plate steam reforming process for extracting hydrogen from multi-fuels and integrate with high-temperature PEM fuel cell systems, (iii) research and develop improved oxygen permeable membranes for high power density lithium air battery with simple control systems and reduced cost, (iv) research on high energy yield agriculture bio-crop (Miscanthus) suitable for reformate fuel/alternative fuel with minimum impact on human food chain and develop a cost analysis and production model, and (v) develop math and science alternative energy educator program to include bio-energy and power.

  12. Advances in Materials and System Technology for Portable Fuel Cells

    Science.gov (United States)

    Narayanan, Sekharipuram R.

    2007-01-01

    This viewgraph presentation describes the materials and systems engineering used for portable fuel cells. The contents include: 1) Portable Power; 2) Technology Solution; 3) Portable Hydrogen Systems; 4) Direct Methanol Fuel Cell; 5) Direct Methanol Fuel Cell System Concept; 6) Overview of DMFC R&D at JPL; 7) 300-Watt Portable Fuel Cell for Army Applications; 8) DMFC units from Smart Fuel Cell Inc, Germany; 9) DMFC Status and Prospects; 10) Challenges; 11) Rapid Screening of Well-Controlled Catalyst Compositions; 12) Screening of Ni-Zr-Pt-Ru alloys; 13) Issues with New Membranes; 14) Membranes With Reduced Methanol Crossover; 15) Stacks; 16) Hybrid DMFC System; 17) Small Compact Systems; 18) Durability; and 19) Stack and System Parameters for Various Applications.

  13. Segmented fuel irradiation program: investigation on advanced materials

    International Nuclear Information System (INIS)

    Uchida, H.; Goto, K.; Sabate, R.; Abeta, S.; Baba, T.; Matias, E. de; Alonso, J.

    1999-01-01

    The Segmented Fuel Irradiation Program, started in 1991, is a collaboration between the Japanese organisations Nuclear Power Engineering Corporation (NUPEC), the Kansai Electric Power Co., Inc. (KEPCO) representing other Japanese utilities, and Mitsubishi Heavy Industries, Ltd. (MHI); and the Spanish Organisations Empresa Nacional de Electricidad, S.A. (ENDESA) representing A.N. Vandellos 2, and Empresa Nacional Uranio, S.A. (ENUSA); with the collaboration of Westinghouse. The objective of the Program is to make substantial contribution to the development of advanced cladding and fuel materials for better performance at high burn-up and under operational power transients. For this Program, segmented fuel rods were selected as the most appropriate vehicle to accomplish the aforementioned objective. Thus, a large number of fuel and cladding combinations are provided while minimising the total amount of new material, at the same time, facilitating an eventual irradiation extension in a test reactor. The Program consists of three major phases: phase I: design, licensing, fabrication and characterisation of the assemblies carrying the segmented rods (1991 - 1994); phase II: base irradiation of the assemblies at Vandellos 2 NPP, and on-site examination at the end of four cycles (1994-1999). Phase III: ramp testing at the Studsvik facilities and hot cell PIE (1996-2001). The main fuel design features whose effects on fuel behaviour are being analysed are: alloy composition (MDA and ZIRLO vs. Zircaloy-4); tubing texture; pellet grain size. The Program is progressing satisfactorily as planned. The base irradiation is completed in the first quarter of 1999, and so far, tests and inspections already carried out are providing useful information on the behaviour of the new materials. Also, the Program is delivering a well characterized fuel material, irradiated in a commercial reactor, which can be further used in other fuel behaviour experiments. The paper presents the main

  14. The quest for safe and reliable fuel cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Pino, Eddy S.; Abe, Alfredo Y., E-mail: eddypino132@hotmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

    2015-07-01

    The tragic Fukushima Daiichi Nuclear Plant accident of March, 2011, has brought great unrest and challenge to the nuclear industry, which, in collaboration with universities and nuclear research institutes, is making great efforts to improve the safety in nuclear reactors developing accident tolerant fuels (ATF). This involves the study of different materials to be applied as cladding and, also, the improvement in the fuel properties in order to enhance the fuel performance and safety, specifically under accident conditions. Related to the cladding, iron based alloys and silicon carbide (SiC) materials have been studied as a good alternative. In the case of austenitic stainless steel, there is the advantage that the austenitic stainless steel 304 was used as cladding material in the first PWR (Pressurized Water Reactor) registering a good performance. Then, alternated cladding materials such as iron based alloys (304, 310, 316, 347) should be used to replace the zirconium-based alloys in order to improve safety. In this paper, these cladding materials are evaluated in terms of their physical and chemical properties; among them, strength and creep resistance, thermal conductivity, thermal stability and corrosion resistance. Additionally, these properties are compared with those of conventional zirconium-based alloys, the most used material in actual PWR, to assess the advantages and disadvantages of each material concerning to fuel performance and safety contribution. (author)

  15. The quest for safe and reliable fuel cladding materials

    International Nuclear Information System (INIS)

    Pino, Eddy S.; Abe, Alfredo Y.; Giovedi, Claudia

    2015-01-01

    The tragic Fukushima Daiichi Nuclear Plant accident of March, 2011, has brought great unrest and challenge to the nuclear industry, which, in collaboration with universities and nuclear research institutes, is making great efforts to improve the safety in nuclear reactors developing accident tolerant fuels (ATF). This involves the study of different materials to be applied as cladding and, also, the improvement in the fuel properties in order to enhance the fuel performance and safety, specifically under accident conditions. Related to the cladding, iron based alloys and silicon carbide (SiC) materials have been studied as a good alternative. In the case of austenitic stainless steel, there is the advantage that the austenitic stainless steel 304 was used as cladding material in the first PWR (Pressurized Water Reactor) registering a good performance. Then, alternated cladding materials such as iron based alloys (304, 310, 316, 347) should be used to replace the zirconium-based alloys in order to improve safety. In this paper, these cladding materials are evaluated in terms of their physical and chemical properties; among them, strength and creep resistance, thermal conductivity, thermal stability and corrosion resistance. Additionally, these properties are compared with those of conventional zirconium-based alloys, the most used material in actual PWR, to assess the advantages and disadvantages of each material concerning to fuel performance and safety contribution. (author)

  16. Precise measurement of fuel content of irradiated and nonirradiated materials

    International Nuclear Information System (INIS)

    Harker, Y.D.; Napper, P.R.; Proctor, A.E.

    1984-01-01

    This paper discusses the application of precise reactivity measurements in the Advanced Reactivity Measurement Facility at Idaho National Engineering Laboratory (INEL) to determine th fuel content in irradiated and nonirradiated materials. Different methods of reactivity measurements and examples of how they have been are presented, which provides an insight in capabilities available to analyze samples with different geometrical sizes from small volumes approx. 100 cc to 12 ft long fuel pins and also samples with different fuel content ranges from approx. 2 mg to approx. 600 g. The overall accuracy of these measurements is approx. 0.5% (1sigma)

  17. Electrode materials for microbial fuel cells: nanomaterial approach

    KAUST Repository

    Mustakeem, Mustakeem

    2015-11-05

    Microbial fuel cell (MFC) technology has the potential to become a major renewable energy resource by degrading organic pollutants in wastewater. The performance of MFC directly depends on the kinetics of the electrode reactions within the fuel cell, with the performance of the electrodes heavily influenced by the materials they are made from. A wide range of materials have been tested to improve the performance of MFCs. In the past decade, carbon-based nanomaterials have emerged as promising materials for both anode and cathode construction. Composite materials have also shown to have the potential to become materials of choice for electrode manufacture. Various transition metal oxides have been investigated as alternatives to conventional expensive metals like platinum for oxygen reduction reaction. In this review, different carbon-based nanomaterials and composite materials are discussed for their potential use as MFC electrodes.

  18. Regulations concerning the fabricating business of nuclear fuel materials

    International Nuclear Information System (INIS)

    1985-01-01

    In the Law for the Regulations of Nuclear Source Material, Nuclear Fuel Material and Reactors, the regulations have all been revised on the fabrication business of nuclear fuel materials. The revised regulations are given : application for permission of the fabrication business, application for permission of the alteration, application for approval of the design and the construction methods, application for approval of the alteration, application for the facilities inspection, facilities inspection, recordings, entry limitations etc. for controlled areas, measures concerning exposure radiation doses etc., operation of the fabrication facilities, transport within the site of the business, storage, disposal within the site of the business, security regulations, designation etc. of the licensed engineer of nuclear fuels, collection of reports, etc. (Mori, K.)

  19. New JMTR irradiation test plan on fuels and materials

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Nishiyama, Yutaka; Chimi, Yasuhiro; Sasajima, Hideo; Ogiyanagi, Jin; Nakamura, Jinichi; Suzuki, Masahide; Kawamura, Hiroshi

    2009-01-01

    In order to maintain and enhance safety of light water reactors (LWRs) in long-term and up-graded operations, proper understanding of irradiation behavior of fuels and materials is essentially important. Japanese government and the Japan Atomic Energy Agency (JAEA) have decided to refurbish the Japan Materials Testing Reactor (JMTR) and to install new tests rigs, in order to play an active role for solving irradiation related issues on plant aging and high-duty uses of the current LWRs and on development of next-generation reactors. New tests on fuel integrity under simulated abnormal transients and high-duty irradiation conditions are planned in the JMTR. Power ramp tests of newdesign fuel rods will also be performed in the first stage of the program, which is expected to start in year 2011 after refurbishment of the JMTR. Combination of the JMTR tests with simulated reactivity initiated accident tests in the Nuclear Safety Research Reactor (NSRR) and loss of coolant accident tests in hot laboratories would serve as the integrated fuel safety research on the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients. For the materials irradiation, fracture toughness of reactor vessel steels and stress corrosion cracking behavior of stainless steels are being studied in addition to basic irradiation behavior of nuclear materials such as hafnium. The irradiation studies would contribute not only to solve the current problems but also to identify possible seeds of troubles and to make proactive responses. (author)

  20. Advanced materials for alternative fuel capable directly fired heat engines

    Energy Technology Data Exchange (ETDEWEB)

    Fairbanks, J.W.; Stringer, J. (eds.)

    1979-12-01

    The first conference on advanced materials for alternative fuel capable directly fired heat engines was held at the Maine Maritime Academy, Castine, Maine. It was sponsored by the US Department of Energy, (Assistant Secretary for Fossil Energy) and the Electric Power Research Institute, (Division of Fossil Fuel and Advanced Systems). Forty-four papers from the proceedings have been entered into EDB and ERA and one also into EAPA; three had been entered previously from other sources. The papers are concerned with US DOE research programs in this area, coal gasification, coal liquefaction, gas turbines, fluidized-bed combustion and the materials used in these processes or equipments. The materials papers involve alloys, ceramics, coatings, cladding, etc., and the fabrication and materials listing of such materials and studies involving corrosion, erosion, deposition, etc. (LTN)

  1. Fission product behaviors in spent fuel materials during DUPIC fuel fabrication process

    International Nuclear Information System (INIS)

    Kim, J. H.; Na, S. H.; Lee, J. H.; Yang, M. S.

    2002-01-01

    In order to obtain the fundamental data for the analysis of fission product behaviors during DUPIC fuel fabrication process, which is to convert spent PWR fuel into CANDU reactor fuel, the measurement system of radioactivity in spent fuel materials by gamma spectrometry technique was installed in IMEF M6 hot cell,and the preliminary analysis on the release behaviors of fission gas during the DUPIC fuel fabrication process were conducted. Based on the radioactivity measurement for the spent oxidized powder, green pellet and the sintered pellet produced from DUPIC fabrication process, it was found that little Cs-137 was released during OREOX process, but almost 99% of Cs-137 was released during sintering process. The release rate of both Zr-95 and Ru-103 was not so high during sintering process

  2. Proton Conducting Fuel Cells where Electrochemistry Meets Material Science

    DEFF Research Database (Denmark)

    Li, Qingfeng

    Fuel cells are electrochemical devices which directly convert the chemical energy of fuels into electrical energy. They are featured of high energy conversion efficiency and minimized pollutant emission. Proton conducting electrolytes are primarily used as separator materials for low and intermed...... science point of view including novel proton conducting materials and non-precious metal catalysts. The discussion will be made with highlights of DTU´s recent research and of course addressing a diverse technical audience.......Fuel cells are electrochemical devices which directly convert the chemical energy of fuels into electrical energy. They are featured of high energy conversion efficiency and minimized pollutant emission. Proton conducting electrolytes are primarily used as separator materials for low...... followed by a review of the state-of-the-art in terms of performance, lifetime and cost. Technically faced challenges are then outlined on a system level and traced back to fundamental issues of the proton conducting mechanisms and materials. Perspectives and future research are sketched from a materials...

  3. Ordinance concerning the filing of transport of nuclear fuel materials

    International Nuclear Information System (INIS)

    1987-01-01

    This Order provides provisions concerning nuclear fuel substances requiring notification (nuclear fuel substance, material contaminated with nuclear fuel substances, fissionable substances, etc.), procedure for notification (to prefectural public safety commission), certificate of transpot (issued via public safety commission), instructions (speed of vehicle for transporting nuclear fuel substances, parking of vehicle, place for loading and unloading of nuclear fuel substances, method for loading and unloading, report to police, measures for disaster prevention during transport, etc.), communication among members of public safety commission (for smooth transport), notification of alteration of data in transport certificate (application to be submitted to public safety commission), application of reissue of transport certificate, return of transport certificate, inspection concerning transport (to be performed by police), submission of report (to be submitted by refining facilities manager, processing facilities manager, nuclear reactor manager, master of foreign nuclear powered ship, reprocessing facilities manager, waste disposal facilities manager; concerning stolen or missing nuclear fuel substances, traffic accident, unusual leakage of nuclear fuel substances, etc.). (Nogami, K.)

  4. Leaching of spent fuel in the presence of environmental material

    International Nuclear Information System (INIS)

    Le Lous, Karine

    1997-01-01

    The aim of this work is the study of the alteration kinetics of spent fuels and the making of a status of the radioactivity released by spent fuels in conditions of direct disposal in deep underground. A system has been fitted inside a shielded cell to study the leaching by synthetic groundwater of fuel powder irradiated at 60 GWJ.tU -1 in the presence of environmental material (clay or granite) at 40 bars and 90 deg. C. This system allows to reach and keep reductive conditions characteristic of the redox conditions of a deep geological repository. The preparation of calibrated spent fuel powders and the recovery of the activity fixed by the environmental materials has required the implementation of specific procedures. Similar experiments have been performed in parallel with Simfuel in a controlled area. A first series of experiments has been carried out in 4 environments for each fuel. Important sorption phenomena take place in the environmental materials and the actinide concentrations stabilize rapidly at low values: 10 -8 mol/l for U, 10 -12 mol/l for Pu and 10 -13 -10 -14 mol/l for Cm. The activity released by 90 Sr at the end of each experiment is about two times higher in the presence of clay than in the presence of granite. The average alteration rates are of about 0.2 mg.m -2 /day in the presence of granite and 0.4 to 0.6 mg.m -2 /day in the presence of clay. They are comparable to those reported in the literature for reducing conditions. Such tests are necessary to determine the leaching rate of spent fuels in reducing conditions and in the absence of environmental materials in order to show the possible effects of these materials. (J.S.)

  5. Material Control and Accountability Experience at the Fuel Conditioning Facility

    International Nuclear Information System (INIS)

    Vaden, D.; Fredrickson, G.L.

    2007-01-01

    The Fuel Conditioning Facility (FCF) at the Idaho National Laboratory (INL) treats spent nuclear fuel using an electrometallurgical process that separates the uranium from the fission products, sodium thermal bond, and cladding materials. Material accountancy is necessary at FCF for two reasons: 1) it provides a mechanism for detecting a potential loss of nuclear material for safeguards and security, and 2) it provides a periodic check of inventories to ensure that processes and materials are within control limits. Material Control and Accountability is also a Department of Energy (DOE) requirement (DOE Order 474.1). The FCF employs a computer based Mass Tracking (MTG) System to collect, store, retrieve, and process data on all operations that directly affect the flow of materials through the FCF. The MTG System is important for the operations of the FCF because it supports activities such as material control and accountability, criticality safety, and process modeling. To conduct material control and accountability checks and to monitor process performance, mass balances are routinely performed around the process equipment. The equipment used in FCF for pyro-processing consists of two mechanical choppers and two electro-refiners (the Mark-IV with the accompanying element chopper and Mark-V with the accompanying blanket chopper for processing driver fuel and blanket, respectively), and a cathode processor (used for processing both driver fuel and blanket) and casting furnace (mostly used for processing driver fuel). Performing mass balances requires the measurement of the masses and compositions of several process streams and equipment inventories. The masses of process streams are obtained via in-cell balances (i.e., load cells) that weigh containers entering and leaving the process equipment. Samples taken at key locations are analyzed to determine the composition of process streams and equipment inventories. In cases where equipment or containers cannot be

  6. Review of the IAEA Nuclear Fuel Cycle Materials Section activities related to WWER fuel

    International Nuclear Information System (INIS)

    Killeen, J.

    2003-01-01

    The IAEA Nuclear Fuel Cycle Programme, designated as Programme B, has the main objective of supporting Member States in policy making, strategic planning, developing technology and addressing issues with respect to safe, reliable, economically efficient, proliferation resistant and environmentally sound nuclear fuel cycle. This paper is concentrated on describing the work within Sub-programme B.2 'Fuel Performance and Technology'. Two Technical Working Groups assist in the preparation of the IAEA programme in the nuclear fuel cycle area - Technical Working Group on Water Reactor Fuel Performance and Technology and Technical Working Group on Nuclear Fuel Cycle Options. The activities of the Unit within the Nuclear Fuel Cycle and Materials Section working on Fuel Performance and Technology are given, based on the sub-programme structure of the Agency programme and budget for 2002-2003. Within the framework of Co-ordinated Research Projects a study of the delayed hydride cracking (DHC) of the zirconium alloys used in pressurised heavy water reactors (PHWR) involving 10 countries has been completed. It achieved very effective transfer of know-how at the laboratory level in three technologically important areas: 1) Controlled hydriding of samples to predetermined levels; 2) Accurate measurement of hydrogen concentrations at the relatively low levels found in pressure tubes and RBMK channel tubes; and 3) In the determination of DHC rates under various conditions of temperature and stress. A new project has been started on the 'Improvement of Models used for Fuel Behaviour Simulation' (FUMEX II) to assist Member States in improving the predictive capabilities of computer codes used in modelling fuel behaviour for extended burnup. The IAEA also collaborates with organisations in the Member States to support activities and meetings on nuclear fuel cycle related topics

  7. Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

    1976-04-01

    Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

  8. Materials management in an internationally safeguarded fuels reprocessing plant

    International Nuclear Information System (INIS)

    Hakkila, E.A.; Cobb, D.D.; Dayem, H.A.; Dietz, R.J.; Kern, E.A.; Markin, J.T.; Shipley, J.P.; Barnes, J.W.; Scheinman, L.

    1980-04-01

    The first volume of this report summarizes the results and conclusions for this study of conventional and advanced nuclear materials accounting systems applicable for both large (1500 MTHM/y) and small (210 MTHM/y) spent-fuel reprocessing facilities subject to international verification

  9. Comparing the Energy Content of Batteries, Fuels, and Materials

    Science.gov (United States)

    Balsara, Nitash P.; Newman, John

    2013-01-01

    A methodology for calculating the theoretical and practical specific energies of rechargeable batteries, fuels, and materials is presented. The methodology enables comparison of the energy content of diverse systems such as the lithium-ion battery, hydrocarbons, and ammonia. The methodology is relevant for evaluating the possibility of using…

  10. Improved Membrane Materials for PEM Fuel Cell Application

    Energy Technology Data Exchange (ETDEWEB)

    Kenneth A. Mauritz; Robert B. Moore

    2008-06-30

    The overall goal of this project is to collect and integrate critical structure/property information in order to develop methods that lead to significant improvements in the durability and performance of polymer electrolyte membrane fuel cell (PEMFC) materials. This project is focused on the fundamental improvement of PEMFC membrane materials with respect to chemical, mechanical and morphological durability as well as the development of new inorganically-modified membranes.

  11. Behavior of organic materials with regard to oxygenated fuels

    Energy Technology Data Exchange (ETDEWEB)

    Puisais, M.; Mollard, M.; Dawans, F.

    The effects of using oxygenated products in gasolines have been determined concerning the resistance of the leading polymer materials used in vehicles and distribution equipment. Fuels containing varying amounts of ethanol or methanol, possibly in conjunction with cosolvents such as tertiobutanol or a butanol-acetone mixture, have been examined. The test methods applied have determined the long-term behavior of materials with regard to their permeability, swelling, aging, stress cracking, extraction and ozone resistance.

  12. Material Selection for Accident Tolerant Fuel Cladding

    International Nuclear Information System (INIS)

    Pint, Bruce A.; Terrani, Kurt A.; Yamamoto, Yukinori; Snead, Lance Lewis

    2015-01-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as > 100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H 2 environments at ≥1473 K (1200°C) for short times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti 2 AlC form a protective alumina scale in steam. However, commercial Ti 2 AlC that was not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO 2 , and therefore Ti 2 AlC may be challenging to form as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation-assisted α' formation. The composition effects and critical limits to retaining protective scale formation at > 1400°C are still being evaluated.

  13. Material Selection for Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as > 100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H2 environments at ≥ 1200°C for short times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. However, commercial Ti2AlC that was not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO2, and therefore Ti2AlC may be challenging to form as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation-assisted α´ formation. The composition effects and critical limits to retaining protective scale formation at > 1400°C are still being evaluated.

  14. Material accountancy in an electrometallurgical Fuel Conditioning Facility

    International Nuclear Information System (INIS)

    Vaden, D.; Benedict, R.W.; Goff, K.M.; Keyes, R.W.; Mariani, R.D.; Bucher, R.G.; Yacout, A.M.

    1996-01-01

    The Fuel Conditioning Facility (FCF) treats spent nuclear fuel using an electrometallurgical process that separates the uranium from the fission products, sodium thermal bond and cladding materials. Material accountancy is necessary at FCF for two reasons: first, it provides a mechanism for detecting a potential loss of nuclear material for safeguards and security; second, it provides a periodic check of inventories to ensure that processes and material are under control. By weighing material entering and leaving a process, and using sampling results to determine composition, an inventory difference (ID) results when the measured inventory is compared to the predicted inventory. The ID and its uncertainty, based on error propagation, determines the degree of assurance that an operation proceeded according to expectations. FCF uses the ID calculation in two ways: closeout, which is the ID and uncertainty for a particular operational step, and material accountancy, which determines an ID and its associated uncertainty for a material balance area through several operational steps. Material accountancy over the whole facility for a specified time period assists in detecting diversion of nuclear material. Data from depleted uranium operations are presented to illustrate the method used in FCF

  15. Fuels and materials testing capabilities in Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Baker, R.B.; Chastain, S.A.; Culley, G.E.; Ethridge, J.L.; Lovell, A.J.; Newland, D.J.; Pember, L.A.; Puigh, R.J.; Waltar, A.E.

    1989-01-01

    The Fast Flux Test Facility (FFTF) reactor, which started operating in 1982, is a 400 MWt sodium-cooled fast neutron reactor located in Hanford, Washington State, and operated by Westinghouse Hanford Co. under contract with U.S. Department of Energy. The reactor has a wide variety of functions for irradiation tests and special tests, and its major purpose is the irradiation of fuel and material for liquid metal reactor, nuclear reactor and space reactor projects. The review first describes major technical specifications and current conditions of the FFTF reactor. Then the plan for irradiation testing is outlined focusing on general features, fuel pin/assembly irradiation tests, and absorber irradiation tests. Assemblies for special tests include the material open test assembly (MOTA), fuel open test assembly (FOTA), closed loop in-reactor assembly (CLIRA), and other special fuel assemblies. An interim examination and maintenance cell (FFTF/IEM cell) and other hot cells are used for nondestructive/destructive tests and physical/mechanical properties test of material after irradiation. (N.K.)

  16. Regeneration of ammonia borane from spent fuel materials.

    Science.gov (United States)

    Summerscales, Owen T; Gordon, John C

    2013-07-28

    A shift to the hydrogen economy requires the development of an effective hydrogen fuel carrier with high volumetric and gravimetric storage capacity. Ammonia borane (AB) has emerged as a leading candidate due to its light weight and multiple protic (N-H) and hydridic (B-H) hydrogens. As a consequence, much work has been directed towards fine tuning the release of H2 from AB, in addition to its regeneration from the dehydrogenated "spent fuel" materials. This review summarizes the development of these regeneration methodologies.

  17. Technology development of nuclear material safeguards for DUPIC fuel cycle

    International Nuclear Information System (INIS)

    Hong, Jong Sook; Kim, Ho Dong; Kang, Hee Young; Lee, Young Gil; Byeon, Kee Ho; Park, Young Soo; Cha, Hong Ryul; Park, Ho Joon; Lee, Byung Doo; Chung, Sang Tae; Choi, Hyung Rae; Park, Hyun Soo.

    1997-07-01

    During the second phase of research and development program conducted from 1993 to 1996, nuclear material safeguards studies system were performed on the technology development of DUPIC safeguards system such as nuclear material measurement in bulk form and product form, DUPIC fuel reactivity measurement, near-real-time accountancy, and containment and surveillance system for effective and efficient implementation of domestic and international safeguards obligation. By securing in advance a optimized safeguards system with domestically developed hardware and software, it will contribute not only to the effective implementation of DUPIC safeguards, but also to enhance the international confidence build-up in peaceful use of spent fuel material. (author). 27 refs., 13 tabs., 89 figs

  18. Technology development of nuclear material safeguards for DUPIC fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jong Sook; Kim, Ho Dong; Kang, Hee Young; Lee, Young Gil; Byeon, Kee Ho; Park, Young Soo; Cha, Hong Ryul; Park, Ho Joon; Lee, Byung Doo; Chung, Sang Tae; Choi, Hyung Rae; Park, Hyun Soo

    1997-07-01

    During the second phase of research and development program conducted from 1993 to 1996, nuclear material safeguards studies system were performed on the technology development of DUPIC safeguards system such as nuclear material measurement in bulk form and product form, DUPIC fuel reactivity measurement, near-real-time accountancy, and containment and surveillance system for effective and efficient implementation of domestic and international safeguards obligation. By securing in advance a optimized safeguards system with domestically developed hardware and software, it will contribute not only to the effective implementation of DUPIC safeguards, but also to enhance the international confidence build-up in peaceful use of spent fuel material. (author). 27 refs., 13 tabs., 89 figs.

  19. Density determination of sintered ceramic nuclear fuel materials

    International Nuclear Information System (INIS)

    Landspersky, H.; Medek, J.

    1980-01-01

    The feasibility was tested of using solids for pycnometric determination of the density of uranium dioxide-based sintered ceramic fuel materials manufactured by the sol-gel method in the shape of spherical particles of 0.7 to 1.0 mm in size and of particles smaller than 200 μm. For fine particles, this is the only usable method of determining their density which is a very important parameter of the fine fraction when it is employed for the manufacture of fuel elements by vibration compacting. The method consists in compacting a mixture of pycnometric material and dispersed particles of uranium dioxide, determining the size and weight of the compact, and in calculating the density of the material measured from the weight of the oxide sample in the mixture. (author)

  20. Cornell Fuel Cell Institute: Materials Discovery to Enable Fuel Cell Technologies

    Energy Technology Data Exchange (ETDEWEB)

    Abruna, H.D.; DiSalvo, Francis J.

    2012-06-29

    The discovery and understanding of new, improved materials to advance fuel cell technology are the objectives of the Cornell Fuel Cell Institute (CFCI) research program. CFCI was initially formed in 2003. This report highlights the accomplishments from 2006-2009. Many of the grand challenges in energy science and technology are based on the need for materials with greatly improved or even revolutionary properties and performance. This is certainly true for fuel cells, which have the promise of being highly efficient in the conversion of chemical energy to electrical energy. Fuel cells offer the possibility of efficiencies perhaps up to 90 % based on the free energy of reaction. Here, the challenges are clearly in the materials used to construct the heart of the fuel cell: the membrane electrode assembly (MEA). The MEA consists of two electrodes separated by an ionically conducting membrane. Each electrode is a nanocomposite of electronically conducting catalyst support, ionic conductor and open porosity, that together form three percolation networks that must connect to each catalyst nanoparticle; otherwise the catalyst is inactive. This report highlights the findings of the three years completing the CFCI funding, and incudes developments in materials for electrocatalyts, catalyst supports, materials with structured and functional porosity for electrodes, and novel electrolyte membranes. The report also discusses developments at understanding electrocatalytic mechanisms, especially on novel catalyst surfaces, plus in situ characterization techniques and contributions from theory. Much of the research of the CFCI continues within the Energy Materials Center at Cornell (emc2), a DOE funded, Office of Science Energy Frontier Research Center (EFRC).

  1. Encapsulation of spent nuclear fuel in ceramic materials

    International Nuclear Information System (INIS)

    Forberg, S.; Westermark, T.

    1983-03-01

    The international situation with regard to deposition of spent nuclear fuel is surveyed, with emphasis on encapsulation in ceramic materials. The feasibility and advantages of ceramic containers, thermodynamic stable in groundwater, are discussed as well as the possibility to ensure that stability for longevity by engineered measures. The design prerequisite are summarized and suggestions are made for a conceptual design, comprising rutile containers with stacks of coiled fuel pins. A novel technique is suggested for the homogeneous sealing of rutile containers at low temperatures. acceptable also for the fuel pin package. Key points are given for research, demonstration and verifications of the design foundations and for future improvements. Of which a few ideas are exemplified. (author)

  2. Advanced proton-exchange materials for energy efficient fuel cells.

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Cy H.; Grest, Gary Stephen; Hickner, Michael A.; Cornelius, Christopher James; Staiger, Chad Lynn; Hibbs, Michael R.

    2005-12-01

    The ''Advanced Proton-Exchange Materials for Energy Efficient Fuel Cells'' Laboratory Directed Research and Development (LDRD) project began in October 2002 and ended in September 2005. This LDRD was funded by the Energy Efficiency and Renewable Energy strategic business unit. The purpose of this LDRD was to initiate the fundamental research necessary for the development of a novel proton-exchange membranes (PEM) to overcome the material and performance limitations of the ''state of the art'' Nafion that is used in both hydrogen and methanol fuel cells. An atomistic modeling effort was added to this LDRD in order to establish a frame work between predicted morphology and observed PEM morphology in order to relate it to fuel cell performance. Significant progress was made in the area of PEM material design, development, and demonstration during this LDRD. A fundamental understanding involving the role of the structure of the PEM material as a function of sulfonic acid content, polymer topology, chemical composition, molecular weight, and electrode electrolyte ink development was demonstrated during this LDRD. PEM materials based upon random and block polyimides, polybenzimidazoles, and polyphenylenes were created and evaluated for improvements in proton conductivity, reduced swelling, reduced O{sub 2} and H{sub 2} permeability, and increased thermal stability. Results from this work reveal that the family of polyphenylenes potentially solves several technical challenges associated with obtaining a high temperature PEM membrane. Fuel cell relevant properties such as high proton conductivity (>120 mS/cm), good thermal stability, and mechanical robustness were demonstrated during this LDRD. This report summarizes the technical accomplishments and results of this LDRD.

  3. The regulations concerning refining business of nuclear source material and nuclear fuel materials

    International Nuclear Information System (INIS)

    1979-01-01

    The regulations are provided for under the law for the regulations of nuclear source materials, nuclear fuel materials and reactors and provisions concerning refining business in the enforcement order for the law. The basic concepts and terms are defined, such as: exposure dose, accumulative dose; controlled area; inspected surrounding area and employee. Refining facilities listed in the application for designation shall be classified into clushing and leaching, thickning, refining facilities, storage facilities of nuclear source materials and nuclear fuel materials, disposal facilities of contaminated substances and building for refining, etc. Business program attached to the application shall include expected time of beginning of refining, estimated production amount of nuclear source materials or nuclear fuel materials for the first three years and funds necessary for construction, etc. Records shall be made and kept for particular periods on delivery and storage of nuclear source materials and nuclear fuel materials, control of radiation, maintenance and accidents of refining facilities. Safety securing, application of internationally regulated substances and measures in dangerous situations are stipulated respectively. Exposure dose of employees and other specified matters shall be reported by the refiner yearly to the Director General of Science and Technology Agency and the Minister of International Trade and Industry. (Okada, K.)

  4. The regulations concerning refining business of nuclear source material and nuclear fuel materials

    International Nuclear Information System (INIS)

    1981-01-01

    This rule is established under the provisions concerning refining business in the law concerning the regulation of nuclear raw materials, nuclear fuel materials and nuclear reactors and the ordinance for the execution of this law, and to enforce them. Basic terms are defined, such as: exposure radiation dose, cumulative dose, control area, surrounding monitoring area and worker. The application for the designation for refining business under the law shall be classified into the facilities for crushing and leaching-filtration, thikening, and refining, the storage facilities for nuclear raw materials and nuclear fuel materials, and the disposal facilities for radioactive wastes, etc. To the application, shall be attached business plans, the explanations concerning the technical abilities of applicants and the prevention of hazards by nuclear raw materials and nuclear fuel materials regarding refining facilities, etc. Records shall be made on the accept, delivery and stock of each kind of nuclear raw materials and nuclear fuel materials, radiation control, the maintenance of and accidents in refining facilities, and kept for specified periods, respectively. Security regulations shall be enacted for each works or enterprise on the functions and organizations of persons engaged in the control of refining facilities, the operation of the apparatuses which must be controlled for the prevention of accidents, and the establishment of control area and surrounding monitoring area, etc. The report on the usage of internationally regulated goods and the measures taken at the time of danger are defined particularly. (Okada, K.)

  5. Principles and Materials Aspects of Direct Alkaline Alcohol Fuel Cells

    Directory of Open Access Journals (Sweden)

    Eileen Hao Yu

    2010-08-01

    Full Text Available Direct alkaline alcohol fuel cells (DAAFCs have attracted increasing interest over the past decade because of their favourable reaction kinetics in alkaline media, higher energy densities achievable and the easy handling of the liquid fuels. In this review, principles and mechanisms of DAAFCs in alcohol oxidation and oxygen reduction are discussed. Despite the high energy densities available during the oxidation of polycarbon alcohols they are difficult to oxidise. Apart from methanol, the complete oxidation of other polycarbon alcohols to CO2 has not been achieved with current catalysts. Different types of catalysts, from conventional precious metal catalyst of Pt and Pt alloys to other lower cost Pd, Au and Ag metal catalysts are compared. Non precious metal catalysts, and lanthanum, strontium oxides and perovskite-type oxides are also discussed. Membranes like the ones used as polymer electrolytes and developed for DAAFCs are reviewed. Unlike conventional proton exchange membrane fuel cells, anion exchange membranes are used in present DAAFCs. Fuel cell performance with DAAFCs using different alcohols, catalysts and membranes, as well as operating parameters are summarised. In order to improve the power output of the DAAFCs, further developments in catalysts, membrane materials and fuel cell systems are essential.

  6. Improved moulding material for addition to nuclear fuel particles to produce nuclear fuel elements

    International Nuclear Information System (INIS)

    Miertschin, G.N.; Leary, D.F.

    1976-01-01

    A suggestion is made to improve the moulding materials used to produce carbon-contained nuclear fuel particles by a coke-reducing added substance. The nuclear fuel particles are meant for the formation of fuel elements for gas-cooled high-temperature nuclear reactors. The moulding materials are above all for the formation of coated particles which are burnt in situ in nuclear fuel element chambers out of 'green' nuclear fuel bodies. The added substance improves the shape stability of the particles forming and prevents a stiding or bridge formation between the particles or with the surrounding walls. The following are named as added substances: 1) Polystyrene and styrene-butadiene-Co polymers (mol. wt. between 5oo and 1,000,000), 2) aromatic compounds (mol. wt. 75 to 300), 3) saturated hydrocarbon polymers (mol. wt. 5,000 to 1,000,000). Additional release agents further improve the properties in the same direction (e.g. alcohols, fatty acids, amines). (orig.) [de

  7. LIFE Materials: Fuel Cycle and Repository Volume 11

    International Nuclear Information System (INIS)

    Shaw, H.; Blink, J.A.

    2008-01-01

    The fusion-fission LIFE engine concept provides a path to a sustainable energy future based on safe, carbon-free nuclear power with minimal nuclear waste. The LIFE design ultimately offers many advantages over current and proposed nuclear energy technologies, and could well lead to a true worldwide nuclear energy renaissance. When compared with existing and other proposed future nuclear reactor designs, the LIFE engine exceeds alternatives in the most important measures of proliferation resistance and waste minimization. The engine needs no refueling during its lifetime. It requires no removal of fuel or fissile material generated in the LIFE engine. It leaves no weapons-attractive material at the end of life. Although there is certainly a need for additional work, all indications are that the 'back end' of the fuel cycle does not to raise any 'showstopper' issues for LIFE. Indeed, the LIFE concept has numerous benefits: (1) Per unit of electricity generated, LIFE engines would generate 20-30 times less waste (in terms of mass of heavy metal) requiring disposal in a HLW repository than does the current once-through fuel cycle. (2) Although there may be advanced fuel cycles that can compete with LIFE's low mass flow of heavy metal, all such systems require reprocessing, with attendant proliferation concerns; LIFE engines can do this without enrichment or reprocessing. Moreover, none of the advanced fuel cycles can match the low transuranic content of LIFE waste. (3) The specific thermal power of LIFE waste is initially higher than that of spent LWR fuel. Nevertheless, this higher thermal load can be managed using appropriate engineering features during an interim storage period, and could be accommodated in a Yucca-Mountain-like repository by appropriate 'staging' of the emplacement of waste packages during the operational period of the repository. The planned ventilation rates for Yucca Mountain would be sufficient for LIFE waste to meet the thermal constraints of

  8. An analysis of the drying process in forest fuel material

    Science.gov (United States)

    G.M. Byram; R.M. Nelson

    2015-01-01

    It is assumed that the flow of moisture in forest fuels and other woody materials is determined by the gradient of a quantity g which is a function of some property, or properties, of the moisture content. There appears to be no preferred choice for this function, hence moisture transfer equations can be based on a number of equally valid definitions of g. The physical...

  9. Concerning enactment of regulations on burying of waste of nuclear fuel material or waste contaminated with nuclear fuel material

    International Nuclear Information System (INIS)

    1988-01-01

    The Atomic Safety Commission of Japan, after examining a report submitted by the Science and Technology Agency concerning the enactment of regulations on burying of waste of nuclear fuel material or waste contaminated with nuclear fuel material, has approved the plan given in the report. Thus, laws and regulations concerning procedures for application for waste burying business, technical standards for implementation of waste burying operation, and measures to be taken for security should be established to ensure the following. Matters to be described in the application for the approval of such business and materials to be attached to the application should be stipulated. Technical standards concerning inspection of waste burying operation should be stipulated. Measures to be taken for the security of waste burying facilities and security concerning the transportation and disposal of nuclear fuel material should be stipulated. Matters to be specified in the security rules should be stipulated. Matters to be recorded by waste burying business operators, measures to be taken to overcome dangers and matters to be reported to the Science and Technology Agency should be stipulated. (Nogami, K.)

  10. Container materials in environments of corroded spent nuclear fuel

    Science.gov (United States)

    Huang, F. H.

    1996-07-01

    Efforts to remove corroded uranium metal fuel from the K Basins wet storage to long-term dry storage are underway. The multi-canister overpack (MCO) is used to load spent nuclear fuel for vacuum drying, staging, and hot conditioning; it will be used for interim dry storage until final disposition options are developed. Drying and conditioning of the corroded fuel will minimize the possibility of gas pressurization and runaway oxidation. During all phases of operations the MCO is subjected to radiation, temperature and pressure excursions, hydrogen, potential pyrophoric hazard, and corrosive environments. Material selection for the MCO applications is clearly vital for safe and efficient long-term interim storage. Austenitic stainless steels (SS) such as 304L SS or 316L SS appear to be suitable for the MCO. Of the two, Type 304L SS is recommended because it possesses good resistance to chemical corrosion, hydrogen embrittlement, and radiation-induced corrosive species. In addition, the material has adequate strength and ductility to withstand pressure and impact loading so that the containment boundary of the container is maintained under accident conditions without releasing radioactive materials.

  11. Graphitic matrix materials for spherical HTR fuel elements

    International Nuclear Information System (INIS)

    Schulze, R.E.; Schulze, H.A.

    1981-02-01

    The report comprises the graphical documentation of irradiation results on graphitic matrix materials for spherical HTR fuel elements. The plotted results are based on data analyses of the series of exposures in the High Flux Reactor Petten (HFR). The documentation includes information about the changes of - the dimensions - the dynamic modulus of elasticity - the coefficient of thermal expansion of the materials after irradiation with fast neutrons. The irradiation experiments and the data analyses are part of the matrix development and irradiation programme, whose objective, realization and results obtained are summarized. (orig./IHOE) [de

  12. Atomic scale modelling of materials of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Bertolus, M.

    2011-10-01

    This document written to obtain the French accreditation to supervise research presents the research I conducted at CEA Cadarache since 1999 on the atomic scale modelling of non-metallic materials involved in the nuclear fuel cycle: host materials for radionuclides from nuclear waste (apatites), fuel (in particular uranium dioxide) and ceramic cladding materials (silicon carbide). These are complex materials at the frontier of modelling capabilities since they contain heavy elements (rare earths or actinides), exhibit complex structures or chemical compositions and/or are subjected to irradiation effects: creation of point defects and fission products, amorphization. The objective of my studies is to bring further insight into the physics and chemistry of the elementary processes involved using atomic scale modelling and its coupling with higher scale models and experimental studies. This work is organised in two parts: on the one hand the development, adaptation and implementation of atomic scale modelling methods and validation of the approximations used; on the other hand the application of these methods to the investigation of nuclear materials under irradiation. This document contains a synthesis of the studies performed, orientations for future research, a detailed resume and a list of publications and communications. (author)

  13. The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1980-01-01

    The law intends under the principles of the atomic energy act to regulate the refining, processing and reprocessing businesses of nuclear raw and fuel metarials and the installation and operation of reactors for the peaceful and systematic utilization of such materials and reactors and for securing public safety by preventing disasters, as well as to control internationally regulated things for effecting the international agreements on the research, development and utilization of atomic energy. Basic terms are defined, such as atomic energy; nuclear fuel material; nuclear raw material; nuclear reactor; refining; processing; reprocessing; internationally regulated thing. Any person who is going to engage in refining businesses other than the Power Reactor and Nuclear Fuel Development Corporation shall get the special designation by the Prime Minister and the Minister of International Trade Industry. Any person who is going to engage in processing businesses shall get the particular admission of the Prime Minister. Any person who is going to establish reactors shall get the particular admission of the Prime Minister, The Minister of International Trade and Industry or the Minister of Transportation according to the kinds of specified reactors, respectively. Any person who is going to engage in reprocessing businesses other than the Power Reactor and Nuclear Fuel Development Corporation and the Japan Atomic Energy Research Institute shall get the special designation by the Prime Minister. The employment of nuclear fuel materials and internationally regulated things is defined in detail. (Okada, K.)

  14. The regulations concerning refining business of nuclear source material and nuclear fuel materials

    International Nuclear Information System (INIS)

    1987-01-01

    Regulations specified here cover application for designation of undertakings of refining (spallation and eaching filtration facilities, thickening facilities, refining facilities, nuclear material substances or nuclear fuel substances storage facilities, waste disposal facilities, etc.), application for permission for alteration (business management plan, procurement plan, fund raising plan, etc.), application for approval of merger (procedure, conditions, reason and date of merger, etc.), submission of report on alteration (location, structure, arrangements processes and construction plan for refining facilities, etc.), revocation of designation, rules for records, rules for safety (personnel, organization, safety training for employees, handling of important apparatus and tools, monitoring and removal of comtaminants, management of radioactivity measuring devices, inspection and testing, acceptance, transport and storage of nuclear material and fuel, etc.), measures for emergency, submission of report on abolition of an undertaking, submission of report on disorganization, measures required in the wake of revocation of designation, submission of information report (exposure to radioactive rays, stolen or missing nuclear material or nuclear fuel, unusual leak of nuclear fuel or material contaminated with nuclear fuel), etc. (Nogami, K.)

  15. Review on MIEC Cathode Materials for Solid Oxide Fuel Cells

    Science.gov (United States)

    Burnwal, Suman Kumar; Bharadwaj, S.; Kistaiah, P.

    2016-11-01

    The cathode is one of the most important components of solid oxide fuel cells (SOFCs). The reduction of oxygen at the cathode (traditional cathodes like LSM, LSGM, etc.) is the slow step in the cell reaction at intermediate temperature (600-800∘C) which is one of the key obstacles to the development of SOFCs. The mixed ionic and electronic conducting cathode (MIEC) like LSCF, BSCF, etc., has recently been proposed as a promising cathode material for SOFC due to the improvement of the kinetic of the cathode reaction. The MIEC materials provide not only the electrons for the reduction of oxygen, but also the ionic conduction required to ensure the transport of the formed oxygen ions and thereby improves the overall electrochemical performance of SOFC system. The characteristics of MIEC cathode materials and its comparison with other traditional cathode materials is studied and presented in the paper.

  16. 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.

  17. Characterization of Catalyst Materials for Production of Aerospace Fuels

    Science.gov (United States)

    Best, Lauren M.; De La Ree, Ana B.; Hepp, Aloysius F.

    2012-01-01

    Due to environmental, economic, and security issues, there is a greater need for cleaner alternative fuels. There will undoubtedly be a shift from crude oil to non-petroleum sources as a feedstock for aviation (and other transportation) fuels. Additionally, efforts are concentrated on reducing costs coupled with fuel production from non-conventional sources. One solution to this issue is Fischer-Tropsch gas-to-liquid technology. Fischer-Tropsch processing of synthesis gas (CO/H2) produces a complex product stream of paraffins, olefins, and oxygenated compounds such as alcohols and aldehydes. The Fisher-Tropsch process can produce a cleaner diesel oil fraction with a high cetane number (typically above 70) without any sulfur or aromatic compounds. This process is most commonly catalyzed by heterogeneous (in this case, silver and platinum) catalysts composed of cobalt supported on alumina or unsupported alloyed iron powders. Physisorption, chemisorptions, scanning electron microscopy (SEM), and energy dispersive spectroscopy (EDS) are described to better understand the potential performance of Fischer-Tropsch cobalt on alumina catalysts promoted with silver and platinum. The overall goal is to preferentially produce C8 to C18 paraffin compounds for use as aerospace fuels. Progress towards this goal will eventually be updated and achieved by a more thorough understanding of the characterization of catalyst materials. This work was supported by NASA s Subsonic Fixed Wing and In-situ Resource Utilization projects.

  18. LIFE Materials: Overview of Fuels and Structural Materials Issues Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J

    2008-09-08

    The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission

  19. Advanced Fuel Pellet Materials and Fuel Rod Design for Water Cooled Reactors. Proceedings of a Technical Committee Meeting

    International Nuclear Information System (INIS)

    2010-10-01

    The economics of current nuclear power plants have improved through increased fuel burnup and longer fuel cycles, i.e. increasing the effective time that fuel remains in the reactor core and the amount of energy it generates. Efficient consumption of fissile material in the fuel element before it is discharged from the reactor means that less fuel is required over the reactor's life cycle, which results in lower amounts of fresh fuel, lower spent fuel storage costs, and less waste for ultimate disposal. Better utilization of fissile nuclear materials, as well as more flexible power manoeuvring, place challenging operational demands on materials used in reactor components, and first of all, on fuel and cladding materials. It entails increased attention to measures ensuring desired in-pile fuel performance parameters that require adequate improvements in fuel material properties and fuel rod designs. These are the main reasons that motivated the IAEA Technical Working Group on Fuel Performance and Technology (TWG-FPT) to recommend the organization of a Technical Committee Meeting on Advanced Fuel Pellet Materials and Fuel Rod Designs for Power Reactors. The proposal was supported by the IAEA TWGs on Advanced Technologies for Light and Heavy Water-Cooled Reactors (TWG-LWR and TWG-HWR), and the meeting was held at the invitation of the Government of Switzerland at the Paul Scherrer Institute in Villigen, from 23 to 26 November 2009. This was the third IAEA meeting on these subjects (the first was held in 1996 in Tokyo, Japan, and the second in 2003 in Brussels, Belgium), which reflects the continuous interest in the above issues among Member States. The purpose of the meeting was to review the current status in the development of fuel pellet materials and to explore recent improvements in fuel rod designs for light and heavy water cooled power reactors. The meeting was attended by 45 specialists representing fuel vendors, nuclear utilities, research and development

  20. Graphitic matrix materials for spherical HTR fuel elements

    International Nuclear Information System (INIS)

    Rind, W.; Schulze, R.E.; Schulze, H.A.

    1981-12-01

    The present report comprises the essential results of material development and irradiation testing of graphitic matrix materials for spherical HTR fuel elements and completes the documentation of the irradiation data for 20 matrix materials (Juel-1702). The main emphasis is given to the matrices A3-3 (standard matrix) and A3-27 (matrix synthesized resin), both of which are being used as structural materials for the fuel elements of the AVR and the THTR respectively. In addition, comparisons are made between 18 A3-variants and the standard matrix A3-3. It is shown that three of the variants come into question as a potential for use. The results described were obtained in the framework of the HTR project 'Hochtemperaturreaktor-Brennstoffkreislauf' (HBK), in which are involved the Gesellschaft fuer Hochtemperaturreaktor-Technik mbH, Hochtemperaturreaktor-Brennelemente GmbH, Hochtemperatur-Reaktorbau GmbH, Kernforschungsanlage Juelich GmbH, NUKEM GmbH, and Sigri Elektrographit GmbH/Ringsdorff-Werke GmbH. The project is sponsored by the 'Bundesministerium fuer Forschung und Technologie' and by the state of 'Nordrhein-Westfalen'. (orig.) [de

  1. QUARTERLY PROGRESS REPORT JANUARY, FEBRUARY, MARCH, 1968 REACTOR FUELS AND MATERIALS DEVELOPMENT PROGRAMS FOR FUELS AND MATERIALS BRANCH OF USAEC DIVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Cadwell, J. J.; de Halas, D. R.; Nightingale, R. E.; Worlton, D. C.

    1968-06-01

    Progress is reported in these areas: nuclear graphite; fuel development for gas-cooled reactors; HTGR graphite studies; nuclear ceramics; fast-reactor nitrides research; non-destructive testing; metallic fuels; basic swelling studies; ATR gas and water loop operation and maintenance; reactor fuels and materials; fast reactor dosimetry and damage analysis; and irradiation damage to reactor metals.

  2. Thermal durability of modified Synroc material as reactor fuel matrix

    International Nuclear Information System (INIS)

    Kikuchi, Akira; Kanazawa, Hiroyuki; Togashi, Yoshihiro; Matumoto, Seiichiro; Nishino, Yasuharu; Ohwada, Isao; Nakata, Masahito; Amano, Hidetoshi; Mitamura, Hisayoshi

    1994-08-01

    A Synroc, a polyphase titanate ceramics composed of three mineral phases (perovskite, hollandite and zirconolite), has an excellent performance of immobilization of high level nuclear waste. A working group in the Department of Hot Laboratories paid special attention to this merit and started a development study on a LWR fuel named 'Waste Disposal Possible (WDP) Fuel', which has the two functions of a reactor fuel and a waste form. The present paper mainly describes thermal durability of a modified Synroc material, which is essentially important for applying the material to a fuel matrix. The two kinds of Synroc specimens, designated 'SM' as modified and 'SB' as a reference, were prepared by hot-pressing and annealed at 1200degC to 1500degC for 30 min in air. Unexpected and peculiar spherical voids were observed in the specimen SM at 1400degC and 1500degC, which caused the specimen swelling. The formation of the voids depends significantly on the existence of spherical precipitates seen in the as-fabricated specimen including latent micropores with high pressure. On the other hand, the heat treatment at 1500degC formed additional new phases, designated 'Phase A' for the specimen SB and 'Phase X' for SM. Phase A is a decomposition product of hollandite and Phase X a reaction product of Phase A and perovskite in the spherical voids. Furthermore, additional information and thermal properties examined are presented in Appendix 1 and Appendix 2, respectively. It was recognized that the modified Synroc specimen SM had excellent thermal properties. (author)

  3. Microbial fuel cells: recent developments in design and materials

    Science.gov (United States)

    Bhargavi, G.; Venu, V.; Renganathan, S.

    2018-03-01

    Microbial Fuel Cells (MFCs) are the promising devices which can produce electricity by anaerobic fermentation of organic / inorganic matter from easily metabolized biomass to complex wastewater using microbes as biocatalysts. MFC technology has been found as a potential technology for electricity generation and concomitant wastewater treatment. However, the high cost of the components and low efficiency are barricading the commercialization of MFC when compared with other energy generating systems. The performance of an MFC is largely relying on the reactor design and electrode materials. On the way to improve the efficiency of an MFC, tremendous exercises have been carried out to explore new electrode materials and reactor designs in recent decades. The current review is excogitated to amass the progress in design and electrode materials, which could bolster further investigations on MFCs to improve their performance, mitigate the cost and successful implementation of technology in field applications as well.

  4. Determination of internationally controlled materials according to provisions of the law for the regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1977-01-01

    According to the provisions of The Law, those stipulated as internationally controlled materials are nuclear source materials, nuclear fuel materials, moderating materials, reactors and facilities, transferred from such as the U.S.A., the U.K. and Canada on the agreements of peaceful uses of atomic energy, and nuclear fuel materials accruing therefrom. (Mori, K.)

  5. Hibiscus fiber carbon for fuel cell device material

    International Nuclear Information System (INIS)

    Nanik Indayaningsih; Anne Zulfia; Dedi Priadi; Suprapedi

    2010-01-01

    The objective of this research is carbon of hibiscus fibers for the application as basic material of fuel cell device. The carbon is made using a pyrolysis process in inert gas (nitrogen) for 1 hour at temperature of 500 °C, 700 °C and 900 °C. The X-Ray Diffractometer (XRD), Scanning Electron Microscope (SEM) and Impedance-Capacitance-Resistance-meter are used to find out the microstructure, morphology and electrical properties respectively. The results of the experiment showed that the carbon had a structure of amorphous, and as the semiconductor material the electrical conductivity was 5 x 10 -5 S.cm -1 to 4.9 x 10 -5 S.cm -1 increasing in accordance with the pyrolysis temperature. The morphology resembled to plaited mats constructed by porous fibers having width of 50 µm to 300 µm, thickness of 25 µm to 35 µm, and the porous size of 0.5 µm to 5 µm. This morphology enables carbon to be applied as a candidate for a basic material of the Proton Exchange Membrane Fuel Cell. (author)

  6. NMSS handbook for decommissioning fuel cycle and materials licensees

    International Nuclear Information System (INIS)

    Orlando, D.A.; Hogg, R.C.; Ramsey, K.M.

    1997-03-01

    The US Nuclear Regulatory Commission amended its regulations to set forth the technical and financial criteria for decommissioning licensed nuclear facilities. These regulations were further amended to establish additional recordkeeping requirements for decommissioning; to establish timeframes and schedules for the decommissioning; and to clarify that financial assurance requirements must be in place during operations and updated when licensed operations cease. Reviews of the Site Decommissioning Management Plan (SDMP) program found that, while the NRC staff was overseeing the decommissioning program at nuclear facilities in a manner that was protective of public health and safety, progress in decommissioning many sites was slow. As a result NRC determined that formal written procedures should be developed to facilitate the timely decommissioning of licensed nuclear facilities. This handbook was developed to aid NRC staff in achieving this goal. It is intended to be used as a reference document to, and in conjunction with, NRC Inspection Manual Chapter (IMC) 2605, ''Decommissioning Inspection Program for Fuel Cycle and Materials Licensees.'' The policies and procedures discussed in this handbook should be used by NRC staff overseeing the decommissioning program at licensed fuel cycle and materials sites; formerly licensed sites for which the licenses were terminated; sites involving source, special nuclear, or byproduct material subject to NRC regulation for which a license was never issued; and sites in the NRC's SDMP program. NRC staff overseeing the decommissioning program at nuclear reactor facilities subject to regulation under 10 CFR Part 50 are not required to use the procedures discussed in this handbook

  7. Regulations concerning the fabricating business of nuclear fuel materials

    International Nuclear Information System (INIS)

    1979-01-01

    The regulations are entirely revised under the law for the regulations of nuclear materials, nuclear fuel materials and reactors and provisions concerning the fabricating business in the order for execution of the law. Basic concepts and terms are defined, such as: exposure dose; accumulative dose; controlled area; inspected surrounding area; employee and radioactive waste. The application for permission of the fabricating business shall include: location of processing facilities; structure of building structure and equipment of chemical processing facilities; molding facilities; structure and equipment of covering and assembling facilities, storage facilities of nuclear fuel materials and disposal facilities of radioactive waste, etc. Records shall be made and kept for particular periods in each works and place of enterprise on inspection of processing facilities, control of dose, operation, maintenance, accident of processing facilities and weather. Specified measures shall be taken in controlled area and inspected surrounding area to restrict entrance. Measures shall be made not to exceed permissible exposure dose for employees defined by the Director General of Science and Technology Agency. Inspection and check up of processing facilities shall be carried on by employees more than once a day. Operation of processing facilities, transportation in the works and enterprise, storage, disposal, safety securing, report and measures in dangerous situations, etc. are in detail prescribed. (Okada, K.)

  8. NMSS handbook for decommissioning fuel cycle and materials licensees

    Energy Technology Data Exchange (ETDEWEB)

    Orlando, D.A.; Hogg, R.C.; Ramsey, K.M. [and others

    1997-03-01

    The US Nuclear Regulatory Commission amended its regulations to set forth the technical and financial criteria for decommissioning licensed nuclear facilities. These regulations were further amended to establish additional recordkeeping requirements for decommissioning; to establish timeframes and schedules for the decommissioning; and to clarify that financial assurance requirements must be in place during operations and updated when licensed operations cease. Reviews of the Site Decommissioning Management Plan (SDMP) program found that, while the NRC staff was overseeing the decommissioning program at nuclear facilities in a manner that was protective of public health and safety, progress in decommissioning many sites was slow. As a result NRC determined that formal written procedures should be developed to facilitate the timely decommissioning of licensed nuclear facilities. This handbook was developed to aid NRC staff in achieving this goal. It is intended to be used as a reference document to, and in conjunction with, NRC Inspection Manual Chapter (IMC) 2605, ``Decommissioning Inspection Program for Fuel Cycle and Materials Licensees.`` The policies and procedures discussed in this handbook should be used by NRC staff overseeing the decommissioning program at licensed fuel cycle and materials sites; formerly licensed sites for which the licenses were terminated; sites involving source, special nuclear, or byproduct material subject to NRC regulation for which a license was never issued; and sites in the NRC`s SDMP program. NRC staff overseeing the decommissioning program at nuclear reactor facilities subject to regulation under 10 CFR Part 50 are not required to use the procedures discussed in this handbook.

  9. Numerical simulation of tensile behaviour of nuclear fuel cladding materials

    International Nuclear Information System (INIS)

    Hosbons, R.R.; Coleman, C.E.; Holt, R.A.

    1975-10-01

    We have developed an expression to describe the true-stress true-strain behaviour of irradiated Zircaloy, the usual fuel sheathing material. The expression includes strain rate sensitivity, work hardening, irradiation hardening and strain softening due to the annihilation of irradiation damage. A cylindrical specimen, containing a slight taper, was used to simulate the development of plastic instability. By combining the phenomenological expression with the model specimen, numerical calculation of tensile tests on irradiated Zircaloy accurately described load-elongation curves, true-stress true-strain behaviour during necking, the effect of irradiation on ductility, and stress relaxation. (author)

  10. Resilient Sealing Materials for Solid Oxide Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Signo T. Reis; Richard K. Brow

    2006-09-30

    This report describes the development of ''invert'' glass compositions designed for hermetic seals in solid oxide fuel cells (SOFC). Upon sealing at temperatures compatible with other SOFC materials (generally {le}900 C), these glasses transform to glass-ceramics with desirable thermo-mechanical properties, including coefficients of thermal expansion (CTE) over 11 x 10{sup -6}/C. The long-term (>four months) stability of CTE under SOFC operational conditions (e.g., 800 C in wet forming gas or in air) has been evaluated, as have weight losses under similar conditions. The dependence of sealant properties on glass composition are described in this report, as are experiments to develop glass-matrix composites by adding second phases, including Ni and YSZ. This information provides design-guidance to produce desirable sealing materials.

  11. 46 CFR 58.50-15 - Alternate material for construction of independent fuel tanks.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Alternate material for construction of independent fuel... Alternate material for construction of independent fuel tanks. (a) Materials other than those specifically... tank construction only if the tank as constructed meets material and testing requirements approved by...

  12. Drop detachment and motion on fuel cell electrode materials.

    Science.gov (United States)

    Gauthier, Eric; Hellstern, Thomas; Kevrekidis, Ioannis G; Benziger, Jay

    2012-02-01

    Liquid water is pushed through flow channels of fuel cells, where one surface is a porous carbon electrode made up of carbon fibers. Water drops grow on the fibrous carbon surface in the gas flow channel. The drops adhere to the superficial fiber surfaces but exhibit little penetration into the voids between the fibers. The fibrous surfaces are hydrophobic, but there is a substantial threshold force necessary to initiate water drop motion. Once the water drops begin to move, however, the adhesive force decreases and drops move with minimal friction, similar to motion on superhydrophobic materials. We report here studies of water wetting and water drop motion on typical porous carbon materials (carbon paper and carbon cloth) employed in fuel cells. The static coefficient of friction on these textured surfaces is comparable to that for smooth Teflon. But the dynamic coefficient of friction is several orders of magnitude smaller on the textured surfaces than on smooth Teflon. Carbon cloth displays a much smaller static contact angle hysteresis than carbon paper due to its two-scale roughness. The dynamic contact angle hysteresis for carbon paper is greatly reduced compared to the static contact angle hysteresis. Enhanced dynamic hydrophobicity is suggested to result from the extent to which a dynamic contact line can track topological heterogeneities of the liquid/solid interface.

  13. The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The law aims to perform regulations on enterprises of refining, processing and reprocessing of nuclear source and fuel materials and on establishment and operation of reactors to realize the peaceful and deliberate utilization of atomic energy according to the principle of the atomic energy basic law. Regulations of use of internationally regulated substances are also envisaged to observe international agreements. Basic concepts and terms are defined, such as: atomic energy; nuclear fuel material; nuclear source material; reactor; refining; processing; reprocessing and internationally regulated substance. Any person besides the Power Reactor and Nuclear Fuel Material Developing Corporation who undertakes refining shall be designated by the Prime Minister and the Minister of International Trade and Industry. An application shall be filed to the ministers concerned, listing name and address of the person, name and location of the refining works, equipment and method of refining, etc. The permission of the Prime Minister is necessary for any person who engages in processing. An application shall be filed to the Prime Minister, listing name and address of the person, name and location of the processing works and equipment and method of processing, etc. Permission of the Prime Minister, the Minister of International Trade and Industry or the Minister of Transport is necessary for any person who sets up reactors. An application shall be filed to the minister concerned, listing name and address of the person, purpose of operation, style, thermal output of reactor and number of units, etc. (Okada, K.)

  14. Compacting biomass waste materials for use as fuel

    Science.gov (United States)

    Zhang, Ou

    Every year, biomass waste materials are produced in large quantity. The combustibles in biomass waste materials make up over 70% of the total waste. How to utilize these waste materials is important to the nation and the world. The purpose of this study is to test optimum processes and conditions of compacting a number of biomass waste materials to form a densified solid fuel for use at coal-fired power plants or ordinary commercial furnaces. Successful use of such fuel as a substitute for or in cofiring with coal not only solves a solid waste disposal problem but also reduces the release of some gases from burning coal which cause health problem, acid rain and global warming. The unique punch-and-die process developed at the Capsule Pipeline Research Center, University of Missouri-Columbia was used for compacting the solid wastes, including waste paper, plastics (both film and hard products), textiles, leaves, and wood. The compaction was performed to produce strong compacts (biomass logs) under room temperature without binder and without preheating. The compaction conditions important to the commercial production of densified biomass fuel logs, including compaction pressure, pressure holding time, back pressure, moisture content, particle size, binder effects, and mold conditions were studied and optimized. The properties of the biomass logs were evaluated in terms of physical, mechanical, and combustion characteristics. It was found that the compaction pressure and the initial moisture content of the biomass material play critical roles in producing high-quality biomass logs. Under optimized compaction conditions, biomass waste materials can be compacted into high-quality logs with a density of 0.8 to 1.2 g/cm3. The logs made from the combustible wastes have a heating value in the range 6,000 to 8,000 Btu/lb which is only slightly (10 to 30%) less than that of subbituminous coal. To evaluate the feasibility of cofiring biomass logs with coal, burn tests were

  15. Nuclear Materials Characterization in the Materials and Fuels Complex Analytical Hot Cells

    International Nuclear Information System (INIS)

    Rodriquez, Michael

    2009-01-01

    As energy prices skyrocket and interest in alternative, clean energy sources builds, interest in nuclear energy has increased. This increased interest in nuclear energy has been termed the 'Nuclear Renaissance'. The performance of nuclear fuels, fuels and reactor materials and waste products are becoming a more important issue as the potential for designing new nuclear reactors is more immediate. The Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) Analytical Laboratory Hot Cells (ALHC) are rising to the challenge of characterizing new reactor materials, byproducts and performance. The ALHC is a facility located near Idaho Falls, Idaho at the INL Site. It was built in 1958 as part of the former Argonne National Laboratory West Complex to support the operation of the second Experimental Breeder Reactor (EBR-II). It is part of a larger analytical laboratory structure that includes wet chemistry, instrumentation and radiochemistry laboratories. The purpose of the ALHC is to perform analytical chemistry work on highly radioactive materials. The primary work in the ALHC has traditionally been dissolution of nuclear materials so that less radioactive subsamples (aliquots) could be transferred to other sections of the laboratory for analysis. Over the last 50 years though, the capabilities within the ALHC have also become independent of other laboratory sections in a number of ways. While dissolution, digestion and subdividing samples are still a vitally important role, the ALHC has stand alone capabilities in the area of immersion density, gamma scanning and combustion gas analysis. Recent use of the ALHC for immersion density shows that extremely fine and delicate operations can be performed with the master-slave manipulators by qualified operators. Twenty milligram samples were tested for immersion density to determine the expansion of uranium dioxide after irradiation in a nuclear reactor. The data collected confirmed modeling analysis with very tight

  16. Some aspects of the chemistry of fast reactor fuel, structural material and decontamination

    International Nuclear Information System (INIS)

    Ganesan, V.

    2012-01-01

    The chemistry of materials pertaining to fast reactors is both fascinating and challenging considering the nature of materials involved such as the fuel, coolant, control and shielding materials in addition to the interactions between the structural materials and the fuel/coolant depending on the nature and conditions involved. The different chemical forms of fuel materials, the need to operate up to high burnups with consequent interactions of the fuel with clad materials, the need to close the fuel cycle by recovery of the fuel materials from spent fuels for refabrication and the necessity to manage the waste, throw a host of challenges which make their study scientifically interesting and technologically important. The use of liquid sodium as coolant in fast reactor heat transport systems combined with its inherent chemical reactivity opens up an interesting branch of chemistry involving liquid sodium especially in contact with structural materials during normal operation of the reactor and with fuels in the event of fuel pin failure. The phenomenon of sodium wetting and the associated corrosion of structural materials in contact with it combined with the need to carryout decontamination of such materials make it interesting to examine and evaluate their suitability for reuse without compromising on their structural integrity. Boron being the material of choice for control and shielding applications in fast reactors with varying isotopic enrichment and the technological challenge to produce large quantities of boron carbide makes it unique. Some of these aspects are addressed in this paper. (author)

  17. Catalysts and methods for converting carbonaceous materials to fuels

    Science.gov (United States)

    Hensley, Jesse; Ruddy, Daniel A.; Schaidle, Joshua A.; Behl, Mayank

    2017-07-25

    Catalysts and processes designed to convert DME and/or methanol and hydrogen (H.sub.2) to desirable liquid fuels are described. These catalysts produce the fuels efficiently and with a high selectivity and yield, and reduce the formation of aromatic hydrocarbons by incorporating H.sub.2 into the products. Also described are process methods to further upgrade these fuels to higher molecular weight liquid fuel mixtures, which have physical properties comparable with current commercially used liquid fuels.

  18. Catalysts and methods for converting carbonaceous materials to fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hensley, Jesse; Ruddy, Daniel A.; Schaidle, Joshua A.; Behl, Mayank

    2017-10-31

    Catalysts and processes designed to convert DME and/or methanol and hydrogen (H.sub.2) to desirable liquid fuels are described. These catalysts produce the fuels efficiently and with a high selectivity and yield, and reduce the formation of aromatic hydrocarbons by incorporating H.sub.2 into the products. Also described are process methods to further upgrade these fuels to higher molecular weight liquid fuel mixtures, which have physical properties comparable with current commercially used liquid fuels.

  19. Catalysts and methods for converting carbonaceous materials to fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hensley, Jesse; Ruddy, Daniel A.; Schaidle, Joshua A.; Behl, Mayank

    2017-10-24

    This disclosure relates to catalysts and processes designed to convert DME and/or methanol and hydrogen (H.sub.2) to desirable liquid fuels. These catalysts produce the fuels efficiently and with a high selectivity and yield, and reduce the formation of aromatic hydrocarbons by incorporating H.sub.2 into the products. This disclosure also describes process methods to further upgrade these fuels to higher molecular weight liquid fuel mixtures, which have physical properties comparable with current commercially used liquid fuels.

  20. Development of eutectic free cladding materials for metallic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tokiwai, Moriyasu; Yuda, Ryoichi [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan); Ohuchi, Atsushi [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan); Amaya, Masaki [Global Nuclear Fuel-Japan Co., Ltd, Oarai, Ibaraki (Japan)

    2002-11-01

    Historically, it is well known that U base metallic fuel has a lower eutectic temperature with stainless steel cladding. In the phase diagram for the U-Fe binary system, the eutectic temperature is 998K. The eutectic reaction is a limiting factor for raising reactor operation temperature. For the purpose of development of eutectic-free cladding materials, three kinds of diffusion-couple tests with 10 mass%Zr alloy were conducted at a temperature of 1027K for 2250 hrs. We selected the following materials: (a) nitrogen charged zirconium foils, (b) vanadium foils of commercial grade, and (c) nitrogen charged ferritic stainless steel (HT-9). The results showed that typical Zr with layer was observed in all of these materials. Zr with layer appeared to act as a barrier against inter-diffusion of U, Fe. The barrier provided immunity to the eutectic reaction. Discussion was made on C-14 problems in relation to another desirable thermodynamic characteristics of Zr such as carbon-14 immobilization. EPMA analysis indicated relatively high nitrogen concentration at the barrier. The barrier is probably composed of ZrN. (author)

  1. An assessment of materials for nuclear fuel immobilization containers

    International Nuclear Information System (INIS)

    Nuttall, K.; Urbanic, V.F.

    1981-09-01

    A wide range of engineering metals and alloys was assessed for their suitability as container materials for irradiated nuclear fuel intended for permanent disposal in a deep, underground hard-rock vault. The container must last at least 500 years without being breached. Materials were assessed for their physical and mechanical metallurgy, weldability, potential embrittlement mechanisms, and economics. A study of the possible mechanisms of metallic corrosion for the various engineering alloys and the expected range of environmental conditons in the vault showed that localized corrosion and delayed fracture processes are the most likely to limit container lifetime. Thus such processes either must be absent or proceed at an insignificant rate. Three groups of alloys are recommended for further study: AISI 300 series austenitic stainless steels, high nickel-base alloys and very dilute titanium-base alloys. Specific alloys from each group are indicated as having the optimum combination of required properties, including cost. For container designs where the outer container shell does not independently support the service loads, copper should also be considered. The final material selection will depend primarily on the environmental conditions in the vault

  2. Process for converting cellulosic materials into fuels and chemicals

    Science.gov (United States)

    Scott, C.D.; Faison, B.D.; Davison, B.H.; Woodward, J.

    1994-09-20

    A process is described for converting cellulosic materials, such as waste paper, into fuels and chemicals utilizing enzymatic hydrolysis of the major constituent of paper, cellulose. A waste paper slurry is contacted by cellulase in an agitated hydrolyzer. The cellulase is produced from a continuous, columnar, fluidized-bed bioreactor utilizing immobilized microorganisms. An attrition mill and a cellobiase reactor are coupled to the agitated hydrolyzer to improve reaction efficiency. The cellulase is recycled by an adsorption process. The resulting crude sugars are converted to dilute product in a fluidized-bed bioreactor utilizing microorganisms. The dilute product is concentrated and purified by utilizing distillation and/or a biparticle fluidized-bed bioreactor system. 1 fig.

  3. The characteristics of bioethanol fuel made of vegetable raw materials

    Science.gov (United States)

    Muhaji; Sutjahjo, D. H.

    2018-01-01

    The aim of this research is to identify the most potential vegetable raw as the material to make a bioethanol fuel as the alternative energy for gasoline. This study used experimental method. The high-level bioethanol was obtained through the process of saccharification, fermentation and stratified distillation. ASTM standards were used as the method of testing the chemical element (D 5501, D 1744, D 1688, D 512, D 2622, D 381), and physical test (D 1613, D 240, D 1298-99, D 445, and D 93). The result of the analysis showed that from the seven bioethanols being studied there is one bioethanol from Saccharum of icinarum linn that has physical and chemical properties close to the standard of bioethanol. Meanwhile, the others only meet some of the physical and chemical properties of the standard bioethanol.

  4. Research reactors for power reactor fuel and materials testing - Studsvik's experience

    International Nuclear Information System (INIS)

    Grounes, M.

    1998-01-01

    Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

  5. Principal organic materials in a repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    Hallbeck, Lotta

    2010-01-01

    the redox potential within the repository. The products of cellulose degradation may help enhance the complexing capacity of the groundwater around the repository, so the amount of cellulose left in the repository should be minimised. 4. Fuels and engine emissions. No important effects are expected from these organic materials in the repository. Although the presence of aromatic compounds and PAHs in groundwater is not desirable in itself, these compounds are of no consequence for long-term repository performance. 5. Detergents and lubricants. The same reasoning as for fuels and engine emissions can be applied to these materials. The amount of detergents should be minimised, although in the amounts in which they are expected to occur, no important impact is foreseen. 6. Materials from human activities. Of these materials, fibres from clothes could have a more important effect, due to the presence of cellulose. Accordingly, human-related wastes should me minimised, although no large amounts of these materials are expected to be present after repository closure. Three processes are considered to have the largest potential impact on repository performance: i) Increasing the reducing capacity and reducing the redox potential in the short term, and increasing the depletion rate of oxygen trapped during the repository operation stage. ii) Increasing the complexing capacity of the groundwater due to the presence of organic complexants, which is expected to be a more relevant process in the long term. Many organic molecules with complexing capacity, for example, short-chain organic acids such as acetate, however, will be oxidised due to microbial metabolism. The projected acetate concentration in groundwater is below the detection limit of available analytical methods. The amount of organic material in groundwater is usually only being a few mg L-1, and 25-75% of this material is non-humic material, i.e. short-chain acids. iii) Production of HS- from the oxidation of short

  6. Principal organic materials in a repository for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hallbeck, Lotta (Microbial Analytics Sweden AB, Moelnlycke (Sweden))

    2010-01-15

    the redox potential within the repository. The products of cellulose degradation may help enhance the complexing capacity of the groundwater around the repository, so the amount of cellulose left in the repository should be minimised. 4. Fuels and engine emissions. No important effects are expected from these organic materials in the repository. Although the presence of aromatic compounds and PAHs in groundwater is not desirable in itself, these compounds are of no consequence for long-term repository performance. 5. Detergents and lubricants. The same reasoning as for fuels and engine emissions can be applied to these materials. The amount of detergents should be minimised, although in the amounts in which they are expected to occur, no important impact is foreseen. 6. Materials from human activities. Of these materials, fibres from clothes could have a more important effect, due to the presence of cellulose. Accordingly, human-related wastes should me minimised, although no large amounts of these materials are expected to be present after repository closure. Three processes are considered to have the largest potential impact on repository performance: i) Increasing the reducing capacity and reducing the redox potential in the short term, and increasing the depletion rate of oxygen trapped during the repository operation stage. ii) Increasing the complexing capacity of the groundwater due to the presence of organic complexants, which is expected to be a more relevant process in the long term. Many organic molecules with complexing capacity, for example, short-chain organic acids such as acetate, however, will be oxidised due to microbial metabolism. The projected acetate concentration in groundwater is below the detection limit of available analytical methods. The amount of organic material in groundwater is usually only being a few mg L-1, and 25-75% of this material is non-humic material, i.e. short-chain acids. iii) Production of HS- from the oxidation of short

  7. Technical meeting on advanced fuel pellet materials and fuel rod designs for water cooled reactors. Book of abstracts

    International Nuclear Information System (INIS)

    2009-01-01

    The economics of current nuclear power plants have improved through increasing fuel burnups and fuel cycles, i.e. the effective time that fuel remains in the reactor core and the amount of energy it generates. Increasing the consumption of fissile material in the fuel element before it is discharged from the reactor means less fuel is required over the reactor's life cycle, which results in lower amounts of fresh fuel, lower spent fuel storage costs, and less waste for ultimate disposal. There has been a continuous historical increase in fuel burnup from 20-25 GWd/tU in Generation I reactors to 50-60 GWd/tU in today's light water reactors and this tendency continues in as much as technological and operational improvements make it possible. In parallel, higher enrichments are discussed, leading to a higher energy yield. For heavy water reactors slight enrichment of fuels and correspondingly growing burnups in CANDU/PHWR are driven by the same economical incentives. Higher burnups and better utilization of fissile nuclear materials (including use of MOX fuel and burnable neutron absorbers), as well as more flexible power manoeuvring, place challenging operational demands on materials used in reactor components, and first of all on fuel and cladding materials. It defines a need for increased attention to measures ensuring compliance to safety criteria related to fission gas release (to limit the internal rod pressure), pellet-cladding interaction (to avoid clad cracking combined with stress and aggressive chemical environment) and pellet-cladding mechanical interaction (to avoid clad mechanical fracture). These measures that secure desired in-pile fuel performance parameters include adequate improvements in fuel material properties and fuel rod designs. That is why the subject Technical Meeting was recommended to the IAEA in 2007 by the Technical Working Group on Fuel Performance and Technology (TWGFPT), and the recommendation was supported by the TWG on Light and

  8. Materials System for Intermediate Temperature Solid Oxide Fuel Cell

    Energy Technology Data Exchange (ETDEWEB)

    Uday B. Pal; Srikanth Gopalan

    2006-01-12

    The objective of this work was to obtain a stable materials system for intermediate temperature solid oxide fuel cell (SOFC) capable of operating between 600-800 C with a power density greater than 0.2 W/cm{sup 2}. The solid electrolyte chosen for this system was La{sub 0.9}Sr{sub 0.1}Ga{sub 0.8}Mg{sub 0.2}O{sub 3}, (LSGM). To select the right electrode materials from a group of possible candidate materials, AC complex impedance spectroscopy studies were conducted between 600-800 C on symmetrical cells that employed the LSGM electrolyte. Based on the results of the investigation, LSGM electrolyte supported SOFCs were fabricated with La{sub 0.6}Sr{sub 0.4}Co{sub 0.8}Fe{sub 0.2}O{sub 3}-La{sub 0.9}Sr{sub 0.1}Ga{sub 0.8}Mg{sub 0.2}O{sub 3} (LSCF-LSGM) composite cathode and Nickel-Ce{sub 0.6}La{sub 0.4}O{sub 3} (Ni-LDC) composite anode having a barrier layer of Ce{sub 0.6}La{sub 0.4}O{sub 3} (LDC) between the LSGM electrolyte and the Ni-LDC anode. Electrical performance and stability of these cells were determined and the electrode polarization behavior as a function of cell current was modeled between 600-800 C. The electrical performance of the anode-supported SOFC was simulated assuming an electrode polarization behavior identical to the LSGM-electrolyte-supported SOFC. The simulated electrical performance indicated that the selected material system would provide a stable cell capable of operating between 600-800 C with a power density between 0.2 to 1 W/cm{sup 2}.

  9. IAEA programme on nuclear fuel cycle and materials technologies

    International Nuclear Information System (INIS)

    Killeen, J.

    2006-01-01

    In this paper a brief description and the main objectives of IAEA Programme B on Nuclear fuel cycle are given. The coordinated research project on Improvement of Models Used For Fuel Behaviour Simulation (FUMEX II) is also presented

  10. The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1977-01-01

    Concerning refining, fabrication and reprocessing operations of such materials as well as the installation and operation of reactors, necessary regulations are carried out. Namely, in case of establishing the business of refining, fabricating and reprocessing nuclear materials as well as installing nuclear reactors, applications for the permission of the Prime Minister and the Minister of International Trade and Industry should be filed. Change of such operations should be permitted after filing applications. These permissions are retractable. As regards the reactors installed aboard foreign ships, it must be reported to enter Japanese waters and the permission by the Prime Minister must be obtained. In case of nuclear fuel fabricators, a chief technician of nuclear fuel materials (qualified) must be appointed per each fabricator. In case of installing nuclear reactors, the design and methods of construction should be permitted by the Prime Minister. The standard for such permission is specified, and a chief engineer for operating reactors (qualified) must be appointed. Successors inherit the positions of ones who have operated nuclear material refining, fabrication and reprocessing businesses or operated nuclear reactors. (Rikitake, Y.)

  11. Advanced Materials for PEM-Based Fuel Cell Systems

    Energy Technology Data Exchange (ETDEWEB)

    James E. McGrath; Donald G. Baird; Michael von Spakovsky

    2005-10-26

    Proton exchange membrane fuel cells (PEMFCs) are quickly becoming attractive alternative energy sources for transportation, stationary power, and small electronics due to the increasing cost and environmental hazards of traditional fossil fuels. Two main classes of PEMFCs include hydrogen/air or hydrogen/oxygen fuel cells and direct methanol fuel cells (DMFCs). The current benchmark membrane for both types of PEMFCs is Nafion, a perfluorinated sulfonated copolymer made by DuPont. Nafion copolymers exhibit good thermal and chemical stability, as well as very high proton conductivity under hydrated conditions at temperatures below 80 degrees C. However, application of these membranes is limited due to their high methanol permeability and loss of conductivity at high temperatures and low relative humidities. These deficiencies have led to the search for improved materials for proton exchange membranes. Potential PEMs should have good thermal, hydrolytic, and oxidative stability, high proton conductivity, selective permeability, and mechanical durability over long periods of time. Poly(arylene ether)s, polyimides, polybenzimidazoles, and polyphenylenes are among the most widely investigated candidates for PEMs. Poly(arylene ether)s are a promising class of proton exchange membranes due to their excellent thermal and chemical stability and high glass transition temperatures. High proton conductivity can be achieved through post-sulfonation of poly(arylene ether) materials, but this most often results in very high water sorption or even water solubility. Our research has shown that directly polymerized poly(arylene ether) copolymers show important advantages over traditional post-sulfonated systems and also address the concerns with Nafion membranes. These properties were evaluated and correlated with morphology, structure-property relationships, and states of water in the membranes. Further improvements in properties were achieved through incorporation of inorganic

  12. Advanced Materials for PEM-Based Fuel Cell Systems

    Energy Technology Data Exchange (ETDEWEB)

    James E. McGrath

    2005-10-26

    Proton exchange membrane fuel cells (PEMFCs) are quickly becoming attractive alternative energy sources for transportation, stationary power, and small electronics due to the increasing cost and environmental hazards of traditional fossil fuels. Two main classes of PEMFCs include hydrogen/air or hydrogen/oxygen fuel cells and direct methanol fuel cells (DMFCs). The current benchmark membrane for both types of PEMFCs is Nafion, a perfluorinated sulfonated copolymer made by DuPont. Nafion copolymers exhibit good thermal and chemical stability, as well as very high proton conductivity under hydrated conditions at temperatures below 80 °C. However, application of these membranes is limited due to their high methanol permeability and loss of conductivity at high temperatures and low relative humidities. These deficiencies have led to the search for improved materials for proton exchange membranes. Potential PEMs should have good thermal, hydrolytic, and oxidative stability, high proton conductivity, selective permeability, and mechanical durability over long periods of time. Poly(arylene ether)s, polyimides, polybenzimidazoles, and polyphenylenes are among the most widely investigated candidates for PEMs. Poly(arylene ether)s are a promising class of proton exchange membranes due to their excellent thermal and chemical stability and high glass transition temperatures. High proton conductivity can be achieved through post-sulfonation of poly(arylene ether) materials, but this most often results in very high water sorption or even water solubility. Our research has shown that directly polymerized poly(arylene ether) copolymers show important advantages over traditional post-sulfonated systems and also address the concerns with Nafion membranes. These properties were evaluated and correlated with morphology, structure-property relationships, and

  13. The use of graphene based materials for fuel cell, photovoltaics, and supercapacitor electrode materials

    Science.gov (United States)

    Tsang, Alpha C. H.; Kwok, Holly Y. H.; Leung, Dennis Y. C.

    2017-05-01

    This manuscript presents the methodology of the production of 2D and 3D graphene based material, and their applications in fuel cell, supercapacitor, and photovoltic in recent years. Due to the uniqueness and attractive properties of graphene nanosheets, a large number of techniques have been developed for raw graphene preparation, from a chemical method to a physical deposition of carbon vapor under extreme conditions. A variety of graphene based materials were also prepared from raw graphene or graphene oxide, including the metal loaded, metal oxides loaded, to the foreign elements doped graphene. Both two-dimensional (2D) to three-dimensional (3D) structured graphene were covered. These materials included the bulk or template hybrid composite, containing graphene hydrogel, graphene aerogel, or graphene foam and its derived products. They were widely used in green energy device research, which exhibited strong activity, and developed some special usage in recent research.

  14. Determination of internationally controlled materials according to provisions of the law for the regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1984-01-01

    The internationally controlled materials determined according to the law for nuclear source materials, etc. are the following: nuclear source materials, nuclear fuel materials, moderating materials, facilities including reactors, etc. sold, transferred, etc. to Japan according to the agreements for peaceful uses of atomic energy between Japan, and the United States, the United Kingdom, Canada, Australia and France by the respective governments and those organs under them; nuclear fuel materials resulting from usage of the above sold and transferred materials, facilities; nuclear fuel materials sold to Japan according to agreements set by the International Atomic Energy Agency; nuclear fuel materials involved with the safeguards in nuclear weapons non-proliferation treaty with IAEA. (Mori, K.)

  15. Findings and Recommendations from the NIST Workshop on Alternative Fuels and Materials: Biocorrosion.

    Science.gov (United States)

    Mansfield, Elisabeth; Sowards, Jeffrey W; Crookes-Goodson, Wendy J

    2015-01-01

    In 2013, the Applied Chemicals and Materials Division of the National Institute of Standards and Technology (NIST) hosted a workshop to identify and prioritize research needs in the area of biocorrosion. Materials used to store and distribute alternative fuels have experienced an increase in corrosion due to the unique conditions caused by the presence of microbes and the chemistry of biofuels and biofuel precursors. Participants in this workshop, including experts from the microbiological, fuel, and materials communities, delved into the unique materials and chemical challenges that occur with production, transport, and storage of alternative fuels. Discussions focused on specific problems including: a) the changing composition of "drop-in" fuels and the impact of that composition on materials; b) the influence of microbial populations on corrosion and fuel quality; and c) state-of-the-art measurement technologies for monitoring material degradation and biofilm formation.

  16. Potential electrode materials for symmetrical Solid Oxide Fuel Cells

    Directory of Open Access Journals (Sweden)

    Ruiz Morales, J. C.

    2008-08-01

    Full Text Available Chromites, titanates and Pt-YSZ-CeO2 cermets have been investigated as potential electrode materials for an alternative concept of Solid Oxide Fuel Cell (SOFC, the symmetrical SOFCs (SFC. In this configuration, the same electrode material is used simultaneously as anode and cathode. Interconnector materials, such as chromites, could be considered as potential SFC electrodes, at least under pure hydrogen-fed at relatively high temperatures, as they do not exhibit significant catalytic activity towards hydrocarbon oxidation. This may be overcome by partially substituting Cr in the perovskite B-sites by other transition metal cations such as Mn. La0.75Sr0.25Cr0.5Mn0.5O3-δ (LSCM is a good candidate for such SFCs, rendering fuel cell performances in excess of 500 and 300mW/cm2 using pure H2 and CH4 as fuel, at 950 oC. Similarly, typical n-type electronic conductors traditionally regarded as anode materials, such as strontium titanates, may also operate under oxidising conditions as cathodes by substituting some Ti content for Fe to introduce p-type conductivity. Preliminary electrochemical experiments on La4Sr8Ti12-xFexO38-δ-based SFCs show that they perform reasonably well under humidified H2, at high temperatures. A third group of materials is the support material of any typical cermet anode, i.e. YSZ, CeO2 plus a current collector. It has been found that this combination could be optimised to operate as SFC electrodes, rendering performances of 400mW/cm2 under humidified pure H2 at 950oC.

    Cromitas, titanatos y cermets de Pt-YSZ-CeO2 han sido investigados como potenciales materiales de electrodo para un concepto alternativo de Pilas de Combustible de Óxidos Sólidos (SOFC, las pilas SOFC simétricas (SFC. En

  17. The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1987-01-01

    General provisions specify the purpose of the Law and definitions of terms used in it. Provisions relating to control of business management for refining cover designation of business operation, requirements for designation, permission and report of alteration, report of commencement of business operation, revocation of designation, recording, and measures for wastes. Provisions relating to control of business management for processing cover permission of operation, requirements for permission, approval of design and construction plan, inspection of facilities, report of commencement of business management, measures for maintenance, suspension of use of facilities, responsible personnel for handling nuclear fuel, and permit, obligations, etc. of responsible personnel for handing nuclear fuel. Provisions relating to control of construction and operation of nuclear reactor cover permission of construction, permission concerning nuclear reactor mounted on foreign nuclear powered ships, requirements for permission, etc. Other articles stipulate provisions relating to control of business management for reprocessing, use of nuclear fuel substances, use of materials and substances covered by international regulations, designation of inspection organizations, and other rules. (Nogami, K.)

  18. CONVERSION OF LIGNOCELLULOSIC MATERIAL TO CHEMICALS AND FUELS; TOPICAL

    International Nuclear Information System (INIS)

    Edwin S. Olson

    2001-01-01

    A direct conversion of cellulosic wastes, including resin-bonded furniture and building waste, to levulinate esters is being investigated with the view to producing fuels, solvents, and chemical intermediates as well as other useful by-products in an inexpensive process. The acid-catalyzed reaction of cellulosic materials with ethanol or methanol at 200 C gives good yields of levulinate and formate esters, as well as useful by-products, such as a solid residue (charcoal) and a resinous lignin residue. An initial plant design showed reasonable rates of return for production of purified ethyl levulinate and by-products. In this project, investigations have been performed to identify and develop reactions that utilize esters of levulinic acid produced during the acid-catalyzed ethanolysis reaction. We wish to develop uses for levulinate esters that allow their marketing at prices comparable to inexpensive polymer intermediates. These prices will allow a sufficient rate of return to justify building plants for utilizing the waste lignocellulosics. If need is demonstrated for purified levulinate, the initial plant design work may be adequate, at least until further pilot-scale work on the process is performed

  19. CONVERSION OF LIGNOCELLULOSIC MATERIAL TO CHEMICALS AND FUELS

    Energy Technology Data Exchange (ETDEWEB)

    Edwin S. Olson

    2001-06-30

    A direct conversion of cellulosic wastes, including resin-bonded furniture and building waste, to levulinate esters is being investigated with the view to producing fuels, solvents, and chemical intermediates as well as other useful by-products in an inexpensive process. The acid-catalyzed reaction of cellulosic materials with ethanol or methanol at 200 C gives good yields of levulinate and formate esters, as well as useful by-products, such as a solid residue (charcoal) and a resinous lignin residue. An initial plant design showed reasonable rates of return for production of purified ethyl levulinate and by-products. In this project, investigations have been performed to identify and develop reactions that utilize esters of levulinic acid produced during the acid-catalyzed ethanolysis reaction. We wish to develop uses for levulinate esters that allow their marketing at prices comparable to inexpensive polymer intermediates. These prices will allow a sufficient rate of return to justify building plants for utilizing the waste lignocellulosics. If need is demonstrated for purified levulinate, the initial plant design work may be adequate, at least until further pilot-scale work on the process is performed.

  20. Regulations concerning the fabricating business of nuclear fuel materials

    International Nuclear Information System (INIS)

    1977-01-01

    As regards an application for permission of an fabricating business of nuclear fuel materials, it should describe the site of the fabricating facilities and the structure and equipments of buildings (fire-resistant, aseismatic, waterproof, ventilating and air-tight structures), etc. The business plan to be attached to the foregoing application should contain 1) scheduled date when the fabricating business starts, 2) scheduled amounts of products classified by the kinds in each business year within 5 years since the business starts, 3) the amount and the procurement plan of funds necessary for the operation, etc. For the permission of change of a fabricating business, an application must be filed. One who wants to obtain the permission of design and construction of fabricating facilities must file an application. One who wants to undergo inspection of the construction of fabricating facilities must file an application in which various items must be written. After such inspection has been done and it is regarded as passable, a certificate of passing inspection will be given. (Rikitake, Y.)

  1. A Review of Materials for Gas Turbines Firing Syngas Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Gibbons, Thomas [ORNL; Wright, Ian G [ORNL

    2009-05-01

    Following the extensive development work carried out in the 1990's, gas turbine combined-cycle (GTCC) systems burning natural gas represent a reliable and efficient power generation technology widely used in many parts of the world. A critical factor was that, in order to operate at the high turbine entry temperatures required for high efficiency operation, aero-engine technology, i.e., single-crystal blades, thermal barrier coatings, and sophisticated cooling techniques had to be rapidly scaled up and introduced into these large gas turbines. The problems with reliability that resulted have been largely overcome, so that the high-efficiency GTCC power generation system is now a mature technology, capable of achieving high levels of availability. The high price of natural gas and concern about emission of greenhouse gases has focused attention on the desirability of replacing natural gas with gas derived from coal (syngas) in these gas turbine systems, since typical systems analyses indicate that IGCC plants have some potential to fulfil the requirement for a zero-emissions power generation system. In this review, the current status of materials for the critical hot gas path parts in large gas turbines is briefly considered in the context of the need to burn syngas. A critical factor is that the syngas is a low-Btu fuel, and the higher mass flow compared to natural gas will tend to increase the power output of the engine. However, modifications to the turbine and to the combustion system also will be necessary. It will be shown that many of the materials used in current engines will also be applicable to units burning syngas but, since the combustion environment will contain a greater level of impurities (especially sulfur, water vapor, and particulates), the durability of some components may be prejudiced. Consequently, some effort will be needed to develop improved coatings to resist attack by sulfur-containing compounds, and also erosion.

  2. IAEA programme on nuclear fuel cycle and materials technologies

    International Nuclear Information System (INIS)

    Killeen, J.

    2008-01-01

    In this paper a brief description and the main objectives of IAEA Programme B on Nuclear fuel cycle are given. The coordinated research project on Improvement of Models Used For Fuel Behaviour Simulation (FUMEX II) as well as the changes, trends and main outputs of Sub-programme B.2 for 2006/2007 are discussed. The aim, composition and activities within the International Fuel Performance Experiments (IFPE) Database project are also presented

  3. Evaluation of conceptual flowsheets for incorporating Light Water Reactor (LWR) fuel materials in an advanced nuclear fuel cycle

    International Nuclear Information System (INIS)

    Bell, J.T.; Burch, W.D.; Collins, E.D.; Forsberg, C.W.; Prince, B.E.; Bond, W.D.; Campbell, D.O.; Delene, J.G.; Mailen, J.C.

    1990-08-01

    A preliminary study by a group of experts at ORNL has generated and evaluated a number of aqueous and non-aqueous flowsheets for recovering transuranium actinides from LWR fuel for use as fuel in an LMR and, at the same time, for transmutation of the wastes to less hazardous materials. The need for proliferation resistance was a consideration in the flowsheets. The current state of development of the flowsheets was evaluated and recommendations for additional study were made. 3 refs., 6 figs

  4. Fabrication and Testing of CERMET Fuel Materials for Nuclear Thermal Propulsion

    Science.gov (United States)

    Hickman, Robert; Broadway, Jeramie; Mireles, Omar

    2012-01-01

    A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on Nuclear Thermal Propulsion (NTP) is currently being developed for Advanced Space Exploration Systems. The overall goal of the project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of NTP systems. The current technology roadmap for NTP identifies the development of a robust fuel form as a critical near term need. The lack of a qualified nuclear fuel is a significant technical risk that will require a considerable fraction of program resources to mitigate. Due to these risks and the cost for qualification, the development and selection of a primary fuel must begin prior to Authority to Proceed (ATP) for a specific mission. The fuel development is a progressive approach to incrementally reduce risk, converge the fuel materials, and mature the design and fabrication process of the fuel element. A key objective of the current project is to advance the maturity of CERMET fuels. The work includes fuel processing development and characterization, fuel specimen hot hydrogen screening, and prototypic fuel element testing. Early fuel materials development is critical to help validate requirements and fuel performance. The purpose of this paper is to provide an overview and status of the work at Marshall Space Flight Center (MSFC).

  5. Fuel Cells for Portable Power: 1. Introduction to DMFCs; 2. Advanced Materials and Concepts for Portable Power Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Zelenay, Piotr [Los Alamos National Laboratory

    2012-07-16

    Thanks to generally less stringent cost constraints, portable power fuel cells, the direct methanol fuel cell (DMFC) in particular, promise earlier market penetration than higher power polymer electrolyte fuel cells (PEFCs) for the automotive and stationary applications. However, a large-scale commercialization of DMFC-based power systems beyond niche applications already targeted by developers will depend on improvements to fuel cell performance and performance durability as well as on the reduction in cost, especially of the portable systems on the higher end of the power spectrum (100-250 W). In this part of the webinar, we will focus on the development of advanced materials (catalysts, membranes, electrode structures, and membrane electrode assemblies) and fuel cell operating concepts capable of fulfilling two key targets for portable power systems: the system cost of $5/W and overall fuel conversion efficiency of 2.0-2.5 kWh/L. Presented research will concentrate on the development of new methanol oxidation catalysts, hydrocarbon membranes with reduced methanol crossover, and improvements to component durability. Time permitted, we will also present a few highlights from the development of electrocatalysts for the oxidation of two alternative fuels for the direct-feed fuel cells: ethanol and dimethyl ether.

  6. Development of fuels and structural materials for fast breeder reactors

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    technology for the energy security of India in the 21st century. Keywords. Fast breeder reactor .... The design of the nuclear fuel is an important aspect which has to be optimised for efficient, economic and safe ..... Recycling of plutonium present in the irradiated fuel with minimum delay is important to ensure rapid growth of ...

  7. Subcritical Measurements Research Program for Fresh and Spent Materials Test Reactor Fuels

    International Nuclear Information System (INIS)

    Blanchard, A.

    1999-01-01

    'A series of subcritical noise measurements were performed on fresh and spent University of Missouri Research Reactor fuel assemblies. These experimental measurements were performed for the purposes of providing benchmark quality data for validating transport theory computer codes and nuclear cross-section data used to perform criticality safety analyses for highly enriched, uranium-aluminum Material Test Reactor fuel assemblies. A mechanical test rig was designed and built to hold up to four fuel assemblies and neutron detectors in a subcritical array. The rig provided researchers with the ability to evaluate the reactivity effects of variable fuel/detector spacing, fuel rotation, and insertion of metal reflector plates into the lattice.'

  8. Determination of internationally controlled materials according to provisions of the law for the regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1981-01-01

    This rule is established under the provisions of the law concerning the regulation of nuclear raw materials, nuclear fuel materials and reactors, and the former notification No. 26, 1961, is hereby abolished. Internationally regulated goods under the law are as follows: nuclear raw materials, nuclear fuel materials and moderator materials transferred by sale or other means from the governments of the U.S., U.K., Canada, Australia and France or the persons under their jurisdictions according to the agreements concluded between the governments of Japan and these countries, respectively, the nuclear fuel materials recovered from these materials or produced by their usage, nuclear reactors, the facilities and heavy water transferred by sale or other means from these governments or the persons under their jurisdictions, the nuclear fuel materials produced by the usage of such reactors, facilities and heavy water, the nuclear fuel materials sold by the International Atomic Energy Agency under the contract between the Japanese government and the IAEA, the nuclear fuel materials recovered from these materials or produced by their usage, the heavy water produced by the facilities themselves transferred from the Canadian government, Canadian governmental enterprises or the persons under the jurisdiction of the Canadian government or produced by the usage of these facilities, etc. (Okada, K.)

  9. Waste management issues and their potential impact on technical specifications of CANDU fuel materials

    International Nuclear Information System (INIS)

    Tait, J.C.; Johnson, L.H.

    1997-01-01

    The technical specifications for the composition of nuclear fuels and materials used in Canada's CANDU reactors have been developed by AECL and materials manufacturers, taking into account considerations specific to their manufacture and the effect of minor impurities on fuel behaviour in reactor. Nitrogen and chlorine are examples of UO 2 impurities, however, where there is no technical specification limit. These impurities are present in the source materials or introduced in the fabrication process and are neutron activated to 14 C and 36 C1, which after 129 I , are the two most significant contributors to dose in safety assessments for the disposal of used fuel. For certain impurities, environmental factors, particularly the safety of the disposal of used fuels, should be taken into consideration when deriving 'allowable' impurity limits for nuclear fuel materials. (author)

  10. The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1978-01-01

    This law has following two purposes. At first, it exercises necessary controls concerning nuclear source material, nuclear fuel material and reactors in order to: (a) limit their uses to those for the peaceful purpose; (b) ensure planned uses of them; and (c) ensure the public safety by preventing accidents from their uses. Necessary controls are to be made concerning the refining, fabricating and reprocessing businesses, as well as the construction and operation of reactors. The second purpose of the law is to exercise necessary controls concerning internationally controlled material in order to execute the treaties and other international agreements on the research, development and use of atomic energy (the first chapter). In the second and following chapters the law prescribes controls for the persons who wish to carry on the refining and fabricating businesses, to construct and operate reactors, and to conduct the reprocessing business, as well as for those who use the internationally controlled material, respectively in separate chapters by the category of those businesses. For example, the controls to the person who wishes to construct and operate reactors are: (a) the permission of the business after the examination; (b) the examination and approval of the design and methods of construction prior to the construction; (c) the inspection of the facilities prior to their use; (d) periodic inspections of the facilities; (e) the establishment of requirements for safety measures and punishments to their violations. (Matsushima, A.)

  11. Polymer electrolyte fuel cells physical principles of materials and operation

    CERN Document Server

    Eikerling, Michael

    2014-01-01

    The book provides a systematic and profound account of scientific challenges in fuel cell research. The introductory chapters bring readers up to date on the urgency and implications of the global energy challenge, the prospects of electrochemical energy conversion technologies, and the thermodynamic and electrochemical principles underlying the operation of polymer electrolyte fuel cells. The book then presents the scientific challenges in fuel cell research as a systematic account of distinct components, length scales, physicochemical processes, and scientific disciplines. The main part of t

  12. Fuel cells science and engineering. Materials, processes, systems and technology. Vol. 1

    Energy Technology Data Exchange (ETDEWEB)

    Stolten, Detlef; Emonts, Bernd (eds.) [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energieforschung (IEF), Brennstoffzellen (IEF-3)

    2012-07-01

    The first volume is divided in four parts and 22 chapters. It is structured as follows: PART I: Technology. Chapter 1: Technical Advancement of Fuel-Cell Research and Development (Dr. Bernd Emonts, Ludger Blum, Thomas Grube, Werner Lehnert, Juergen Mergel, Martin Mueller and Ralf Peters); 2: Single-Chamber Fuel Cells (Teko W. Napporn and Melanie Kuhn); 3: Technology and Applications of Molten Carbonate Fuel Cells (Barbara Bosio, Elisabetta Arato and Paolo Greppi); 4: Alkaline Fuel Cells (Erich Guelzow); 5: Micro Fuel Cells (Ulf Groos and Dietmar Gerteisen); 6: Principles and Technology of Microbial Fuel Cells (Jan B. A. Arends, Joachim Desloover, Sebastia Puig and Willy Verstraete); 7: Micro-Reactors for Fuel Processing (Gunther Kolb); 8: Regenerative Fuel Cells (Martin Mueller). PART II: Materials and Production Processes. Chapter 9: Advances in Solid Oxide Fuel Cell Development between 1995 and 2010 at Forschungszentrum Juelich GmbH, Germany (Vincent Haanappel); 10: Solid Oxide Fuel Cell Electrode Fabrication by Infiltration (Evren Gunen); 11: Sealing Technology for Solid Oxide Fuel Cells (K. Scott Weil); 12: Phosphoric Acid, an Electrolyte for Fuel Cells - Temperature and Composition Dependence of Vapor Pressure and Proton Conductivity (Carsten Korte); 13: Materials and Coatings for Metallic Bipolar Plates in Polymer Electrolyte Membrane Fuel Cells (Heli Wang and John A. Turner); 14: Nanostructured Materials for Fuel Cells (John F. Elter); 15: Catalysis in Low-Temperature Fuel Cells - An Overview (Sabine Schimpf and Michael Bron). PART III: Analytics and Diagnostics. Chapter 16: Impedance Spectroscopy for High-Temperature Fuel Cells (Ellen Ivers-Tiffee, Andre Leonide, Helge Schichlein, Volker Sonn and Andre Weber); 17: Post-Test Characterization of Solid Oxide Fuel-Cell Stacks (Norbert H. Menzler and Peter Batfalsky); 18: In Situ Imaging at Large-Scale Facilities (Christian Toetzke, Ingo Manke and Werner Lehnert); 19: Analytics of Physical Properties of Low

  13. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-01-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period

  14. Foreign materials in a deep repository for spent nuclear fuels

    International Nuclear Information System (INIS)

    Jones, C.; Christiansson, Aa.; Wiborgh, M.

    1999-12-01

    The effects of foreign substances introduced into a spent-fuel repository are reviewed. Possible impacts on processes and barrier-functions are examined, and the following areas are identified: Corrosion of the spent-fuel canister through the presence of sulfur and substances that favor microbial growth; impacts on the bentonite properties through the presence of cations as calcium, potassium and iron; radionuclide transport through the presence of complex-formers and surface-active substances

  15. An integrated approach to selecting materials for fuel cladding in advanced high-temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rangacharyulu, C., E-mail: chary.r@usask.ca [Univ. of Saskatchewan, Saskatoon, SK (Canada); Guzonas, D.A.; Pencer, J.; Nava-Dominguez, A.; Leung, L.K.H. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    An integrated approach has been developed for selection of fuel cladding materials for advanced high-temperature reactors. Reactor physics, thermalhydraulic and material analyses are being integrated in a systematic study comparing various candidate fuel-cladding alloys. The analyses established the axial and radial neutron fluxes, power distributions, axial and radial temperature distributions, rates of defect formation and helium production using AECL analytical toolsets and experimentally measured corrosion rates to optimize the material composition for fuel cladding. The project has just been initiated at University of Saskatchewan. Some preliminary results of the analyses are presented together with the path forward for the project. (author)

  16. Strategies for carbon and sulfur tolerant solid oxide fuel cell materials, incorporating lessons from heterogeneous catalysis

    OpenAIRE

    Boldrin, P; Ruiz Trejo, E; Mermelstein, J; Bermudez Menendez, J; Ramirez Reina, T; Brandon, N

    2016-01-01

    Solid oxide fuel cells (SOFCs) are a rapidly emerging energy technology for a low carbon world, providing high efficiency, potential to use carbonaceous fuels and compatibility with carbon capture and storage. However, current state-of-the-art materials have low tolerance to sulfur, a common contaminant of many fuels, and are vulnerable to deactivation due to carbon deposition when using carbon-containing compounds. In this review we first study the theoretical basis behind carbon and sulfur ...

  17. Development of materials for fuel cell application by radiation technology

    International Nuclear Information System (INIS)

    Rhee, Chang Kyu; Lee, Min Ku; Park, Junju; Lee, Gyoungja; Lee, Byung Cheol; Shin, Junhwa; Nho, Youngchang; Kang, Philhyun; Sohn, Joon Yong; Rang, Uhm Young

    2012-06-01

    The development of the single cell of SOFC with low operation temperature at and below 650 .deg. C(above 400 mW/cm 2 ) Ο The development of fabrication method for the single cell of solid oxide fuel cell (SOFC) by dip-coating of nanoparticles such as NiO, YSZ, Ag, and Ag/C, etc. Ο The optimization of the preparation and performance of SOFC by using nanoparticles. Ο The preparation of samples for SOFC with large dimension. The development of fluoropolymer-based fuel cell membranes with crosslinked structure by radiation grafting technique Ο The development of fuel cell membranes with low methanol permeability via the introduction of novel monomers (e. g. vinylbenzyl chloride and vinylether chloride) by radiation grafting technique Ο The development of hydrocarbon fuel cell membrane by radiation crosslinking technique Ο The structure analysis and the evaluations of the property, performance, and radiation effect of the prepared membranes Ο The optimization of the preparation and performance of DMFC fuel cell membrane via the structure-property analysis (power: above 130 mW/cm 2 /50 cm 2 at 5M methanol) Ο The preparation of samples for MEA stack assembly

  18. Modern methods of material accounting for mixed oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Eggers, R.F.; Pindak, J.L.; Brouns, R.J.; Williams, R.C.; Brite, D.W.; Kinnison, R.R.; Fager, J.E.

    1981-01-01

    The generic requirements loss detection, and response to alarms of a contemporary material control and accounting (MCandA) philosophy have been applied to a mixed oxide fuel fabrication plant to produce a detailed preliminary MCandA system design that is generally applicable to facilities of this type. This paper summarizes and discusses detailed results of the mixed oxide fuel fabrication plant study

  19. Fuel-cell applied research: Electrocatalysis and materials

    Science.gov (United States)

    Srinivasan, S.; Isaacs, H. S.; McBreen, J.; Ogrady, W. E.; Olender, H.; Olmer, L. J.; Schouler, E. J. L.; Kordesch, K. V.

    1980-09-01

    Phosphoric acid electrolyte fuel cells and high temperature solid electrolyte fuel cells are discussed. The corrosion behavior of five furnace black carbon supports for platinum electrocatalysts in phosphoric acid was evaluated using a cyclic voltammetric technique. The electrode kinetic studies of the hydrogen oxidation and the oxygen reduction reactions on smooth platinum in phosphoric acid revealed that the entire decrease in performance of a fuel cell in this impure electrolyte is entirely due to the impurity. A mathematical model which describes the slow oxygen adsorption step is compared with the experimental polarization and impedance results of uncontaminated platinum electrodes in contact with yttria stabilized zirconia electrolytes. Attempts to explain discrepancies are made by assuming that the adsorption of oxygen followed a Langmuir and then a Frumkin adsorption isotherm. It is concluded that a site variation occurred on the electrode surface when potentials were varied in the cathodic region.

  20. Testing of reactor fuel materials using nuclear techniques

    International Nuclear Information System (INIS)

    Khouri, M.T.F.C.

    1978-01-01

    The tests presented here apply to: the quantitative determination of uranium in the core of fuel element plates by the detection of the number of neutrons produced in photo induced reactions in uranium; the determination of 235 U proportion in uranium dioxide samples, in the form of uranyl nitrate, by the technique of the detection of tracks produced by fission fragments and in pellet samples by passive gamma spectrometry and the checking of uranium homogenization distribution in fuel plates and uranium dioxide pellets. (Author) [pt

  1. 78 FR 71532 - Amendments to Material Control and Accounting Regulations and Proposed Guidance for Fuel Cycle...

    Science.gov (United States)

    2013-11-29

    ... Accounting Regulations and Proposed Guidance for Fuel Cycle Facility Material Control and Accounting Plans... material control and accounting (MC&A) of special nuclear material (SNM) and the proposed guidance... INFORMATION CONTACT section of this document. NRC's Agencywide Documents Access and Management System (ADAMS...

  2. Corrosion resistance of metallic materials for use in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Pinard Legry, G.; Pelras, M.; Turluer, G.

    1988-01-01

    The main of this review is to reassess the corrosion resistance properties required from metallic materials to be used in the various developments of the Purex process applied for nuclear fuel reprocessing

  3. Microstructural evolution of nanograin nickel-zirconia cermet anode materials for solid oxide fuel cell applications

    International Nuclear Information System (INIS)

    Nayak, Bibhuti Bhusan

    2012-01-01

    The aim of the study is to study the structure, microstructure, porosity, thermal expansion, electrical conductivity and electrochemical behavior of the anode material thus synthesized in order to find its suitability for solid oxide fuel cell (SOFC) anode application

  4. Properties of materials for water reactor fuel elements and methods of measurement

    International Nuclear Information System (INIS)

    1987-05-01

    The present meeting was scheduled by the International Atomic Energy Agency upon the proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology and held in the IAEA Headquarters in Vienna from 13 to 16 October 1986. Nineteen participants from 11 countries attended the meeting. Fifteen papers were presented in three sessions covering the following subjects: cladding materials (6 papers), fuel materials (4 papers) and influence of material properties on fuel rod behaviour (5 papers). Three working group meetings covered the above-mentioned topics which were held in order to discuss, in depth, the presentation of papers and to develop the recommendations for future activities aimed at improving properties of cladding and fuel materials for water cooled power reactors. A separate abstract was prepared for each of these 15 papers

  5. Roll compaction and granulation system for nuclear fuel material

    International Nuclear Information System (INIS)

    Goldmann, L.H. Jr.; Holley, C.C.

    1981-01-01

    A roll compaction and roll granulation system has been designed and fabricated to replace conventional preslugging and crushing operations typically used in the fabrication of mixed oxide nuclear fuel pellets. This equipment will be of maintenance advantage with only the compaction and granulation rolls inside containment. The prototype is being tested and the results will be reported within a year

  6. Development of fuels and structural materials for fast breeder reactors

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    The latter occurs especially during transient over-power (TOP) incidents. The wrapper tubes of the ... generator a critical component in determining the efficient running of the plant and demands high integrity of steam ... diffusivity of the fuel, which was measured at BARC as well as at IGCAR, by a technique employing the ...

  7. Fuel Pellets from Biomass. Processing, Bonding, Raw Materials

    DEFF Research Database (Denmark)

    Stelte, Wolfgang

    The depletion of fossil fuels and the need to reduce green house gas emissions has resulted in a strong growth of biomass utilization for heat and power production. Attempts to overcome the poor handling properties of biomass, i.e. its low bulk density and inhomogeneous structure, have resulted...

  8. Current status of sea transport of nuclear fuel materials and LLW in Japan

    International Nuclear Information System (INIS)

    Kitagawa, Hiroshi; Akiyama, Hideo

    2000-01-01

    Along with the basic policy of the nuclear fuel cycle of Japan, many fuel cycle facilities have been already constructed in Rokkasho-Mura, Aomori prefecture, such as the uranium enrichment plant, the low level waste disposal center and the receiving pool of the spent nuclear fuels for reprocessing. These facilities belong to the Japan Nuclear Fuel Limited. (JNFL). Domestic sea transport of the spent nuclear fuels (SF) has been carried out since 1977 to the Tokai Reprocessing Plant, and the first sea transport of the SF to the fuel cycle facility in Rokkasho-Mura was done in Oct, 1998 using a new exclusive ship 'Rokuei-Maru'. Sea transport of the low level radioactive wastes (LLW) has been carried out since 1992 to the Rokkasho LLW Disposal Center, and about 130,000 LLW drams were transported from the nuclear power plant sites. These sea transport have demonstrated the safety of the transport of the nuclear fuel cycle materials. It is hoped that the safe sea transport of the nuclear fuel materials will contribute to the more progress of the nuclear fuel cycle activities of Japan. (author)

  9. Permeability of EVOH Barrier Material Used in Automotive Applications: Metrology Development for Model Fuel Mixtures

    Directory of Open Access Journals (Sweden)

    Zhao Jing

    2015-02-01

    Full Text Available EVOH (Ethylene-Vinyl Alcohol materials are widely used in automotive applications in multi-layer fuel lines and tanks owing to their excellent barrier properties to aromatic and aliphatic hydrocarbons. These barrier materials are essential to limit environmental fuel emissions and comply with the challenging requirements of fast changing international regulations. Nevertheless, the measurement of EVOH permeability to model fuel mixtures or to their individual components is particularly difficult due to the complexity of these systems and their very low permeability, which can vary by several orders of magnitude depending on the permeating species and their relative concentrations. This paper describes the development of a new automated permeameter capable of taking up the challenge of measuring minute quantities as low as 1 mg/(m2.day for partial fluxes for model fuel mixtures containing ethanol, i-octane and toluene at 50°C. The permeability results are discussed as a function of the model fuel composition and the importance of EVOH preconditioning is emphasized for accurate permeability measurements. The last part focuses on the influence of EVOH conditioning on its mechanical properties and its microstructure, and further illustrates the specific behavior of EVOH in presence of ethanol oxygenated fuels. The new metrology developed in this work offers a new insight in the permeability properties of a leading barrier material and will help prevent the consequences of (bioethanol addition in fuels on environmental emissions through fuel lines and tanks.

  10. Hybrid Direct Carbon Fuel Cell Performance with Anode Current Collector Material

    DEFF Research Database (Denmark)

    Deleebeeck, Lisa; Kammer Hansen, Kent

    2015-01-01

    The influence of the current collector on the performance of a hybrid direct carbon fuel cell (HDCFC), consisting of solid oxide fuel cell (SOFC) with a molten carbonate-carbon slurry in contact with the anode, has been investigated using current-voltage curves. Four different anode current...... collectors were studied: Au, Ni, Ag, and Pt. It was shown that the performance of the direct carbon fuel cell (DCFC) is dependent on the current collector materials, Ni and Pt giving the best performance, due to their catalytic activity. Gold is suggested to be the best material as an inert current collector...

  11. Device for separating ruthenium ion from spent fuel material

    International Nuclear Information System (INIS)

    Izumida, Tatsuo; Sasahira, Akira; Ozawa, Yoshihiro; Kawamura, Fumio.

    1988-01-01

    Purpose: To separate plutonium ions efficiently and selectively from organic solvent containing tributyl phosphate used in the main step of reprocessing process. Constitution: The device comprises, as the main constituent factor, a liquid-liquid contact device for bringing not water soluble organic solvent into contact with a nitric acid solution of spent fuel substances and a liquid-liquid contact-separation device for bringing an organic solvent solution containing spent fuel substances separated with nitric acid into contact again with nitric acid. Then, a device is disposed between two liquid-liquid contact devices for staying ruthenium ions and organic solvent for a sufficient time. In this way, ruthenium ions in the organic solvent containing butyl phosphate are gradually converted into complex compounds combined with tributyl phosphate thereby enabling to separate ruthenium ions efficiently and remarkably reduce the corrosion of equipments. (Horiuchi, T.)

  12. Environmental Degradation of Fuels, Fluids and Related Materials for Aircraft

    Science.gov (United States)

    1976-03-01

    residue or may be a hydrolysis product formed during; the water extraction step. 1. Conclusions The residue war shown to contain iron and, likely...three on the tar- nished s’irface. 1. Conclusions While an alkylbenzene sulfonamide has been clearly shown to be present in the fuel, its source is...acidity. Singly alkylsubstituted benzene sulfonamides , however, have an acid character which would potentially cause the observed corrosion. Factors which

  13. Fabrication and Characterizations of Materials and Components for Intermediate Temperature Fuel Cells and Water Electrolysers

    DEFF Research Database (Denmark)

    Jensen, Annemette Hindhede; Prag, Carsten Brorson; Li, Qingfeng

    The worldwide development of fuel cells and electrolysers has so far almost exclusively addressed either the low temperature window (20-200 °C) or the high temperature window (600-1000 °C). This work concerns the development of key materials and components of a new generation of fuel cells...... and electrolysers for operation in the intermediate temperature range from 200 to 400 °C. The intermediate temperature interval is of importance for the use of renewable fuels. Furthermore electrode kinetics is significantly enhanced compared to when operating at low temperature. Thus non-noble metal catalysts...... might be used. One of the key materials in the fuel cell and electrolyser systems is the electrolyte. Proton conducting materials such as cesium hydrogen phosphates, zirconium hydrogen phosphates and tin pyrophosphates have been investigated by others and have shown interesting potential....

  14. Raw material and market for biomass fuel; Raastoff og marked for biobrensel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    The report from a conference deals with raw material and market relations for biomass fuel in Norway. The proceedings cover themes like requirements concerning quality and purity, supply of raw materials, supply and production of chips, supply and market for industrial waste and wood waste, supply of raw materials and market relations for pellets, practical experience from a pelletizing plant, use of source selected paper as a biomass fuel, use of bio-carbon in the ferro-alloy industry, biomass fuel and waste in the cement industry - technical requirements and experience of utilization, processed biomass fuel from wastes - possible niches of marketing, and evaluation of a bio-energy project. 9 figs., 12 tabs.

  15. Development of materials for use in solid oxid fuel cells anodes using renewable fuels in direct operation

    International Nuclear Information System (INIS)

    Lima, D.B.P.L. de; Florio, D.Z. de; Bezerra, M.E.O.

    2016-01-01

    Fuel cells produce electrical current from the electrochemical combustion of a gas or liquid (H2, CH4, C2H5OH, CH3OH, etc.) inserted into the anode cell. An important class of fuel cells is the SOFC (Solid Oxide Cell Fuel). It has a ceramic electrolyte that transports protons (H +) or O-2 ions and operating at high temperatures (500-1000 °C) and mixed conductive electrodes (ionic and electronic) ceramics or cermets. This work aims to develop anodes for fuel cells of solid oxide (SOFC) in order to direct operations with renewable fuels and strategic for the country (such as bioethanol and biogas). In this context, it becomes important to study in relation to the ceramic materials, especially those that must be used in high temperatures. Some types of double perovskites such as Sr2MgMoO6 (or simply SMMO) have been used as anodes in SOFC. In this study were synthesized by the polymeric precursor method, analyzed and characterized different ceramic samples of families SMMO, doped with Nb, this is: Sr2 (MgMo)1-xNbxO6 with 0 ≤ x ≤ 0.2. The materials produced were characterized by various techniques such as, thermal analysis, X-ray diffraction and scanning electron microscopy, and electrical properties determined by dc and ac measurements in a wide range of temperature, frequency and partial pressure of oxygen. The results of this work will contribute to a better understanding of advanced ceramic properties with mixed driving (electronic and ionic) and contribute to the advancement of SOFC technology operating directly with renewable fuels. (author)

  16. Evaluation of MHD materials for use in high-temperature fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Guidotti, R.

    1978-06-15

    The MHD and high-temperature fuel cell literature was surveyed for data pertaining to materials properties in order to identify materials used in MHD power generation which also might be suitable for component use in high-temperature fuel cells. Classes of MHD-electrode materials evaluated include carbides, nitrides, silicides, borides, composites, and oxides. Y/sub 2/O/sub 3/-stabilized ZrO/sub 2/ used as a reference point to evaluate materials for use in the solid-oxide fuel cell. Physical and chemical properties such as electrical resistivity, coefficient of thermal expansion, and thermodynamic stability toward oxidation were used to screen candidate materials. A number of the non-oxide ceramic MHD-electrode materials appear promising for use in the solid-electrolyte and molten-carbonate fuel cell as anodes or anode constituents. The MHD-insulator materials appear suitable candidates for electrolyte-support tiles in the molten-carbonate fuel cells. The merits and possible problem areas for these applications are discussed and additional needed areas of research are delineated.

  17. Use of filler materials to aid spent nuclear fuel dry storage

    International Nuclear Information System (INIS)

    Anderson, K.J.

    1981-09-01

    The use of filler materials (also known as stabilizer or encapsulating materials) was investigated in conjunction with the dry storage of irradiated light water reactor (LWR) fuel. The results of this investigation appear to be equally valid for the wet storage of fuel. The need for encapsulation and suitable techniques for closing was also investigated. Various materials were reviewed (including solids, liquids, and gases) which were assumed to fill the void areas within a storage can containing either intact or disassembled spent fuel. Materials were reviewed and compared on the basis of cost, thermal characteristics, and overall suitability in the proposed environment. A thermal analysis was conducted to yield maximum centerline and surface temperatures of a design basis fuel encapsulated within various filler materials. In general, air was found to be the most likely choice as a filler material for the dry storage of spent fuel. The choice of any other filler material would probably be based on a desire, or need, to maximize specific selection criteria, such as surface temperatures, criticality safety, or confinement

  18. Determination of internationally controlled materials according to provisions of the law for the regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1978-01-01

    As the internationally controlled material is section 2, paragraph 8 of above mentioned law this notification designates at first: the nuclear source material, nuclear fuel material, moderator material and reactors supplied by the U. S. government or the person under the jurisdiction of it in accordance with the agreements for cooperation between Japanese and U. S. governments concerning peaceful uses of atomic energy; the nuclear fuel material recovered from or produced as the result of the use of above listed materials; and the nuclear fuel material produced as the result of the use of above mentioned reactors. After that, the notification designates in the same way those materials and reactors supplied by the relevant government or the person under the jurisdiction of it, or by the relevant agency in accordance with similar agreements between Japanese government and governments of Britain, Canada, France and Australia, and also International Atomic Energy Agency, respectively. The nuclear fuel material recovered from or produced as the result of the use of these materials or reactors are also included. (Matsushima, A.)

  19. Recovery of fissile materials from plutonium residues, miscellaneous spent nuclear fuel, and uranium fissile wastes

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1997-01-01

    A new process is proposed that converts complex feeds containing fissile materials into a chemical form that allows the use of existing technologies (such as PUREX and ion exchange) to recover the fissile materials and convert the resultant wastes to glass. Potential feed materials include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel, and (3) uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, and organics. 14 refs., 4 figs

  20. LIFE Materials: Phase Formation and Transformations in Transmutation Fuel Materials for the LIFE Engine Part I - Path Forward Volume 3

    International Nuclear Information System (INIS)

    Turchi, P.A.; Kaufman, L.; Fluss, M.

    2008-01-01

    The current specifications of the LLNL fusion-fission hybrid proposal, namely LIFE, impose severe constraints on materials, and in particular on the nuclear fissile or fertile nuclear fuel and its immediate environment. This constitutes the focus of the present report with special emphasis on phase formation and phase transformations of the transmutation fuel and their consequences on particle and pebble thermal, chemical, and mechanical integrities. We first review the work that has been done in recent years to improve materials properties under the Gen-IV project, and with in particular applications to HTGR and MSR, and also under GNEP and AFCI in the USA. Our goal is to assess the nuclear fuel options that currently exist together with their issues. Among the options, it is worth mentioning TRISO, IMF, and molten salts. The later option will not be discussed in details since an entire report (Volume 8 - Molten-salt Fuels) is dedicated to it. Then, in a second part, with the specific LIFE specifications in mind, the various fuel options with their most critical issues are revisited with a path forward for each of them in terms of research, both experimental and theoretical. Since LIFE is applicable to very high burn-up of various fuels, distinctions will be made depending on the mission, i.e., energy production or incineration. Finally a few conclusions are drawn in terms of the specific needs for integrated materials modeling and the in depth knowledge on time-evolution thermo-chemistry that controls and drastically affects the performance of the nuclear materials and their immediate environment. Although LIFE demands materials that very likely have not yet been fully optimized, the challenges are not insurmountable, and a well concerted experimental-modeling effort should lead to dramatic advances that should well serve other fission programs such as Gen-IV, GNEP, AFCI as well as the international fusion program, ITER

  1. LIFE Materials: Phase Formation and Transformations in Transmutation Fuel Materials for the LIFE Engine Part I - Path Forward Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    Turchi, P A; Kaufman, L; Fluss, M

    2008-12-19

    The current specifications of the LLNL fusion-fission hybrid proposal, namely LIFE, impose severe constraints on materials, and in particular on the nuclear fissile or fertile nuclear fuel and its immediate environment. This constitutes the focus of the present report with special emphasis on phase formation and phase transformations of the transmutation fuel and their consequences on particle and pebble thermal, chemical, and mechanical integrities. We first review the work that has been done in recent years to improve materials properties under the Gen-IV project, and with in particular applications to HTGR and MSR, and also under GNEP and AFCI in the USA. Our goal is to assess the nuclear fuel options that currently exist together with their issues. Among the options, it is worth mentioning TRISO, IMF, and molten salts. The later option will not be discussed in details since an entire report (Volume 8 - Molten-salt Fuels) is dedicated to it. Then, in a second part, with the specific LIFE specifications in mind, the various fuel options with their most critical issues are revisited with a path forward for each of them in terms of research, both experimental and theoretical. Since LIFE is applicable to very high burn-up of various fuels, distinctions will be made depending on the mission, i.e., energy production or incineration. Finally a few conclusions are drawn in terms of the specific needs for integrated materials modeling and the in depth knowledge on time-evolution thermo-chemistry that controls and drastically affects the performance of the nuclear materials and their immediate environment. Although LIFE demands materials that very likely have not yet been fully optimized, the challenges are not insurmountable, and a well concerted experimental-modeling effort should lead to dramatic advances that should well serve other fission programs such as Gen-IV, GNEP, AFCI as well as the international fusion program, ITER.

  2. Materials management in an internationally safeguarded fuels reprocessing plant

    International Nuclear Information System (INIS)

    Hakkila, E.A.; Baker, A.L.; Cobb, D.D.

    1980-04-01

    The following appendices are included: aqueous reprocessing and conversion technology, reference facilities, process design and operating features relevant to materials accounting, operator's safeguards system structure, design principles of dynamic materials accounting systems, modeling and simulation approach, optimization of measurement control, aspects of international verification problem, security and reliability of materials measurement and accounting system, estimation of in-process inventory in solvent-extraction contactors, conventional measurement techniques, near-real-time measurement techniques, isotopic correlation techniques, instrumentation available to IAEA inspectors, and integration of materials accounting and containment and surveillance

  3. Materials management in an internationally safeguarded fuels reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Hakkila, E.A.; Baker, A.L.; Cobb, D.D.

    1980-04-01

    The following appendices are included: aqueous reprocessing and conversion technology, reference facilities, process design and operating features relevant to materials accounting, operator's safeguards system structure, design principles of dynamic materials accounting systems, modeling and simulation approach, optimization of measurement control, aspects of international verification problem, security and reliability of materials measurement and accounting system, estimation of in-process inventory in solvent-extraction contactors, conventional measurement techniques, near-real-time measurement techniques, isotopic correlation techniques, instrumentation available to IAEA inspectors, and integration of materials accounting and containment and surveillance. (DLC)

  4. Device and materials modeling in PEM fuel cells

    National Research Council Canada - National Science Library

    Paddison, Stephen J; Promislow, Keith

    2009-01-01

    .... Materials modeling include subjects relating to the membrane and the catalyst such as proton conduction, atomistic structural modeling, quantum molecular dynamics, and molecular-level modeling...

  5. A Preliminary Evaluation of Using Fill Materials to Stabilize Used Nuclear Fuel During Storage and Transportation

    Energy Technology Data Exchange (ETDEWEB)

    Maheras, Steven J.; Best, Ralph; Ross, Steven B.; Lahti, Erik A.; Richmond, David J.

    2012-08-01

    This report contains a preliminary evaluation of potential fill materials that could be used to fill void spaces in and around used nuclear fuel contained in dry storage canisters in order to stabilize the geometry and mechanical structure of the used nuclear fuel during extended storage and transportation after extended storage. Previous work is summarized, conceptual descriptions of how canisters might be filled were developed, and requirements for potential fill materials were developed. Elements of the requirements included criticality avoidance, heat transfer or thermodynamic properties, homogeneity and rheological properties, retrievability, material availability and cost, weight and radiation shielding, and operational considerations. Potential fill materials were grouped into 5 categories and their properties, advantages, disadvantages, and requirements for future testing were discussed. The categories were molten materials, which included molten metals and paraffin; particulates and beads; resins; foams; and grout. Based on this analysis, further development of fill materials to stabilize used nuclear fuel during storage and transportation is not recommended unless options such as showing that the fuel remains intact or canning of used nuclear fuel do not prove to be feasible.

  6. Novel Materials for High Efficiency Direct Methanol Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Carson, Stephen [Arkema Inc.; Mountz, David [Arkema Inc.; He, Wensheng [Arkema Inc.; Zhang, Tao [Arkema Inc.

    2013-12-31

    Direct methanol fuel cell membranes were developed using blends of different polyelectrolytes with PVDF. The membranes showed complex relationships between polyelectrolyte chemistry, morphology, and processing. Although the PVDF grade was found to have little effect on the membrane permselectivity, it does impact membrane conductivity and methanol permeation values. Other factors, such as varying the polyelectrolyte polarity, using varying crosslinking agents, and adjusting the equivalent weight of the membranes impacted methanol permeation, permselectivity, and areal resistance. We now understand, within the scope of the project work completed, how these inter-related performance properties can be tailored to achieve a balance of performance.

  7. Advances in fuel materials for the transmutation of actinides

    Energy Technology Data Exchange (ETDEWEB)

    Prunier, C.

    1994-12-31

    The physical feasibility of actinides, spent fuels and fission products burning in fission reactors is well understood. In fast reactors, this operation is more favourable. The homogeneous recycling mode has had a preliminary validation in Phenix (the Super fact experiment). For the heterogenous recycling mode, past experience for {sup 238} Pu production in thermal spectrum was obtained with Np O{sub 2}-Mg O targets. An irradiation experiment in Phenix blanket is foreseen with the same type of target. The {sup 237} Np problem seems to be most conveniently treated, even in the short term, by homogeneous recycling with Pu in fast reactors. (author). 15 figs., 4 tabs.

  8. Dosimetry methods for fuels, cladding and structural materials

    International Nuclear Information System (INIS)

    Roettger, H.

    1980-01-01

    This volume of the proceedings of the symposium on reactor dosimetry covers the following topics: the metallurgy and dosimetry interface, radiation damage correlations of structural materials and damage analyses techniques, dosimetry for fusion materials, light water reactor pressure vessel surveillance in practice and irradiation experiments, fast reactor and reseach reactor characterization

  9. A Lifecycle Emissions Model (LEM): Lifecycle Emissions from Transportation Fuels, Motor Vehicles, Transportation Modes, Electricity Use, Heating and Cooking Fuels, and Materials, APPENDIX A: Energy Use and Emissions from the Lifecycle of Diesel-Like Fuels Derived From Biomass

    OpenAIRE

    Delucchi, Mark; Lipman, Timothy

    2003-01-01

    An Appendix to the Report, “A Lifecycle Emissions Model (LEM): Lifecycle Emissions From Transportation Fuels, Motor Vehicles, Transportation Modes, Electricity Use, Heating and Cooking Fuels, and Materialsâ€

  10. A viable electrode material for use in microbial fuel cells for tropical regions

    DEFF Research Database (Denmark)

    Offei, Felix; Thygesen, Anders; Mensah, Moses

    2016-01-01

    Electrode materials are critical for microbial fuel cells (MFC) since they influence the construction and operational costs. This study introduces a simple and efficient electrode material in the form of palm kernel shell activated carbon (AC) obtained in tropical regions. The novel introduction ...

  11. 78 FR 79328 - Amendments to Material Control and Accounting Regulations and Proposed Guidance for Fuel Cycle...

    Science.gov (United States)

    2013-12-30

    ..., 72, 74, and 150 [NRC-2009-0096 and NRC-2013-0195] RIN 3150-AI61 Amendments to Material Control and Accounting Regulations and Proposed Guidance for Fuel Cycle Facility Material Control and Accounting Plans... of this document. NRC's Agencywide Documents Access and Management System (ADAMS): You may access...

  12. Experience of work with radioactive materials and nuclear fuel at the reactor WWR-K

    International Nuclear Information System (INIS)

    Maltseva, R.M.; Petukhov, V.K.

    1998-01-01

    In the report there are considered questions concerning the handling with fresh and spent fuel, experimental devices, containing high enriched uranium, being fissile materials of the bulk form, radioisotopes, obtained in the reactor, and radioactive waste, formed during the operation of the reactor, and organization of storage, account and control of radioactive and fissile materials is described. (author)

  13. Regulations concerning the fabricating business of nuclear fuel materials

    International Nuclear Information System (INIS)

    1987-01-01

    Regulations specified here cover application for such matters as permission for an undertaking of processing, alteration (of location, structure, arrangements, processing method, etc.), approval of design and construction plan, approval of alteration (of design and construction plan of processing facilities), and inspection of the facilities. The regulations also cover execution of facilities inspection, certificate of facilities inspection, processing facilities subject to welding inspection, application for welding inspection, execution of welding inspection, facilities not subject to welding inspection, approval of welding method, welding inspection for imported equipment, certificate of welding inspection, application for approval of joint management, notice of alteration, etc., cancellation of permission, record keeping, restriction on access to areas under management measures concerning exposure to radioactive rays, patrol and checking in processing facilities, operation of processing equipment, transportation within plant or operation premises, storage, waste disposal within plant or operation premises, safety rules, public notification concerning examination and successful applicants, procedure for application for examination, reissue of certificate for nuclear fuel handling expert, return of certificate for nuclear fuel handling expert, submission of report, measures for emergency, notice of abolition of business, measures concerning cancellation of permission, identification card, etc. (Nogami, K.)

  14. FRAPCON-3: Modifications to fuel rod material properties and performance models for high-burnup application

    International Nuclear Information System (INIS)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L.

    1997-12-01

    This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs

  15. Preliminary Investigation of Candidate Materials for Use in Accident Resistant Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jason M. Harp; Paul A. Lessing; Blair H. Park; Jakeob Maupin

    2013-09-01

    As part of a Collaborative Research and Development Agreement (CRADA) with industry, Idaho National Laboratory (INL) is investigating several options for accident resistant uranium compounds including silicides, and nitrides for use in future light water reactor (LWR) fuels. This work is part of a larger effort to create accident tolerant fuel forms where changes to the fuel pellets, cladding, and cladding treatment are considered. The goal fuel form should have a resistance to water corrosion comparable to UO2, have an equal to or larger thermal conductivity than uranium dioxide, a melting temperature that allows the material to stay solid under power reactor conditions, and a uranium loading that maintains or improves current LWR power densities. During the course of this research, fuel fabricated at INL will be characterized, irradiated at the INL Advanced Test Reactor, and examined after irradiation at INL facilities to help inform industrial partners on candidate technologies.

  16. FRAPCON-3: Modifications to fuel rod material properties and performance models for high-burnup application

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L.

    1997-12-01

    This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs.

  17. Materials management in an internationally safeguarded fuels reprocessing plant

    International Nuclear Information System (INIS)

    Hakkila, E.A.; Cobb, D.D.; Dayem, H.A.; Dietz, R.J.; Kern, E.A.; Markin, J.T.; Shipley, J.P.; Barnes, J.W.; Scheinman, L.

    1980-04-01

    The second volume describes the requirements and functions of materials measurement and accounting systems (MMAS) and conceptual designs for an MMAS incorporating both conventional and near-real-time (dynamic) measurement and accounting techniques. Effectiveness evaluations, based on recently developed modeling, simulation, and analysis procedures, show that conventional accountability can meet IAEA goal quantities and detection times in these reference facilities only for low-enriched uranium. Dynamic materials accounting may meet IAEA goals for detecting the abrupt (1-3 weeks) diversion of 8 kg of plutonium. Current materials accounting techniques probably cannot meet the 1-y protracted-diversion goal of 8 kg for plutonium

  18. Notification determining technical details concerning measures for transportation of nuclear fuel materials

    International Nuclear Information System (INIS)

    1977-01-01

    These provisions are established on the basis of and to enforce ''The regulation for installation and operation of reactor'', ''The regulation concerning the fabricating business of nuclear fuel'' and ''The regulations concerning the reprocessing business of spent fuel''. The terms used hereinafter are according to those used in such regulations. The limit of radioactivity concentration of things contaminated by the nuclear fuel materials which are not required to be enclosed in vessels is defined in the lists attached. In the applications for the approval of the measures concerning the transport of things remarkably difficult to be enclosed in vessels, the name and the address of the applicant, the kind, quantity, form and constitution of the thing contaminated by the nuclear fuel materials to be transported, the date and route of the transport and the measures for the prevention of injuries during the transport must be written. The limit of quantity of nuclear fuel materials classifying the performance of vessels is defined respectively in the lists attached. The radiation dose rates provided for by the Director General of the Science and Technology Agency concerning transported things and transporting apparatuses are 200 millirem per hour on the surfaces of such things and containers. The nuclear fission materials specified, for which the measures for the prevention of criticality are especially required, include uranium 233, uranium 235, plutonium 238, plutonium 239, plutonium 241, and the chemical compounds of such substances, and the nuclear fuel materials containing one or two and more of such substances, excluding the nuclear fuel materials with less than 15 grams of such uranium and plutonium. (Okada, K.)

  19. Strategies for Carbon and Sulfur Tolerant Solid Oxide Fuel Cell Materials, Incorporating Lessons from Heterogeneous Catalysis.

    Science.gov (United States)

    Boldrin, Paul; Ruiz-Trejo, Enrique; Mermelstein, Joshua; Bermúdez Menéndez, José Miguel; Ramı Rez Reina, Tomás; Brandon, Nigel P

    2016-11-23

    Solid oxide fuel cells (SOFCs) are a rapidly emerging energy technology for a low carbon world, providing high efficiency, potential to use carbonaceous fuels, and compatibility with carbon capture and storage. However, current state-of-the-art materials have low tolerance to sulfur, a common contaminant of many fuels, and are vulnerable to deactivation due to carbon deposition when using carbon-containing compounds. In this review, we first study the theoretical basis behind carbon and sulfur poisoning, before examining the strategies toward carbon and sulfur tolerance used so far in the SOFC literature. We then study the more extensive relevant heterogeneous catalysis literature for strategies and materials which could be incorporated into carbon and sulfur tolerant fuel cells.

  20. Direct alcohol fuel cells materials, performance, durability and applications

    CERN Document Server

    Corti, Horacio R; Antolini, Ermete

    2014-01-01

    After an introductory overview of this emerging form of clean, portable energy, experts from industry and academia discuss the challenges in materials development, performance, and commercialization standing between DAFCs and widespread public use.

  1. Transportation impact analysis for shipment of irradiated N-reactor fuel and associated materials

    International Nuclear Information System (INIS)

    Daling, P.M.; Harris, M.S.

    1994-12-01

    An analysis of the radiological and nonradiological impacts of highway transportation of N-Reactor irradiated fuel (N-fuel) and associated materials is described in this report. N-fuel is proposed to be transported from its present locations in the 105-KE and 105-KW Basins, and possibly the PUREX Facility, to the 327 Building for characterization and testing. Each of these facilities is located on the Hanford Site, which is near Richland, Washington. The projected annual shipping quantity is 500 kgU/yr for 5 years for a total of 2500 kgU. It was assumed the irradiated fuel would be returned to the K- Basins following characterization, so the total amount of fuel shipped was assumed to be 5000 kgU. The shipping campaign may also include the transport and characterization of liquids, gases, and sludges from the storage basins, including fuel assembly and/or canister parts that may also be present in the basins. The impacts of transporting these other materials are bounded by the impacts of transporting 5000 kgU of N-fuel. This report was prepared to support an environmental assessment of the N-fuel characterization program. The RADTRAN 4 and GENII computer codes were used to evaluate the radiological impacts of the proposed shipping campaign. RADTRAN 4 was used to calculate the routine exposures and accident risks to workers and the general public from the N-fuel shipments. The GENII computer code was used to calculate the consequences of the maximum credible accident. The results indicate that the transportation of N-fuel in support of the characterization program should not cause excess radiological-induced latent cancer fatalities or traffic-related nonradiological accident fatalities. The consequences of the maximum credible accident are projected to be small and result in no excess latent cancer fatalities

  2. Recycle and reuse of materials and components from waste streams of nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    2000-01-01

    All nuclear fuel cycle processes utilize a wide range of equipment and materials to produce the final products they are designed for. However, as at any other industrial facility, during operation of the nuclear fuel cycle facilities, apart from the main products some byproducts, spent materials and waste are generated. A lot of these materials, byproducts or some components of waste have a potential value and may be recycled within the original process or reused outside either directly or after appropriate treatment. The issue of recycle and reuse of valuable material is important for all industries including the nuclear fuel cycle. The level of different materials involvement and opportunities for their recycle and reuse in nuclear industry are different at different stages of nuclear fuel cycle activity, generally increasing from the front end to the back end processes and decommissioning. Minimization of waste arisings and the practice of recycle and reuse can improve process economics and can minimize the potential environmental impact. Recognizing the importance of this subject, the International Atomic Energy Agency initiated the preparation of this report aiming to review and summarize the information on the existing recycling and reuse practice for both radioactive and non-radioactive components of waste streams at nuclear fuel cycle facilities. This report analyses the existing options, approaches and developments in recycle and reuse in nuclear industry

  3. Measurement and analysis of transient vaporization in oxide fuel materials

    International Nuclear Information System (INIS)

    Benson, D.A.; Bergeron, E.G.

    1979-01-01

    This paper describes a series of experiments in which samples are heated to produce high vapor pressure states in times of 10 -6 to 10 -3 seconds. Experimental measurements of vapor pressures over fresh UO 2 from the pulsed electron beam and pulsed reactor heating tests are presented and compared with other high temperature data. The interpretation of the vapor pressure measured in the tests is discussed in detail. Effects of original sample stoichiometry, chemical interactions with the container and non-equilibrium evaporation due to induced temperature gradients are discussed. Special attention is given to dynamic behavior in rapid heating and vaporization of the oxide due to chemical non-equilibrium. Finally, similar projected reactor experiments on irradiated fuel are described and vapor pressure predictions made using available equilibrium models. A discussion of information accessible from such future tests and its importance is presented. (orig.) [de

  4. Life Prediction of Spent Fuel Storage Canister Material

    Energy Technology Data Exchange (ETDEWEB)

    Ballinger, Ronald

    2018-04-16

    The original purpose of this project was to develop a probabilistic model for SCC-induced failure of spent fuel storage canisters, exposed to a salt-air environment in the temperature range 30-70°C for periods up to and exceeding 100 years. The nature of this degradation process, which involves multiple degradation mechanisms, combined with variable and uncertain environmental conditions dictates a probabilistic approach to life prediction. A final report for the original portion of the project was submitted earlier. However, residual stress measurements for as-welded and repair welds could not be performed within the original time of the project. As a result of this, a no-cost extension was granted in order to complete these tests. In this report, we report on the results of residual stress measurements.

  5. Materials and Components for Low Temperature Solid Oxide Fuel Cells – an Overview

    Directory of Open Access Journals (Sweden)

    D. Radhika

    2013-06-01

    Full Text Available This article summarizes the recent advancements made in the area of materials and components for low temperature solid oxide fuel cells (LT-SOFCs. LT-SOFC is a new trend in SOFCtechnology since high temperature SOFC puts very high demands on the materials and too expensive to match marketability. The current status of the electrolyte and electrode materials used in SOFCs, their specific features and the need for utilizing them for LT-SOFC are presented precisely in this review article. The section on electrolytes gives an overview of zirconia, lanthanum gallate and ceria based materials. Also, this review article explains the application of different anode, cathode and interconnect materials used for SOFC systems. SOFC can result in better performance with the application of liquid fuels such methanol and ethanol. As a whole, this review article discusses the novel materials suitable for operation of SOFC systems especially for low temperature operation.

  6. Influence of the starting materials on performance of high temperature oxide fuel cells devices

    Directory of Open Access Journals (Sweden)

    Emília Satoshi Miyamaru Seo

    2004-03-01

    Full Text Available High temperature solid oxide fuel cells (SOFCs offer an environmentally friendly technology to convert gaseous fuels such as hydrogen, natural gas or gasified coal into electricity at high efficiencies. Besides the efficiency, higher than those obtained from the traditional energy conversion systems, a fuel cell provides many other advantages like reliability, modularity, fuel flexibility and very low levels of NOx and SOx emissions. The high operating temperature (950-1000 °C used by the current generation of the solid oxide fuel cells imposes severe constraints on materials selection in order to improve the lifetime of the cell. Besides the good electrical, electrochemical, mechanical and thermal properties, the individual cell components must be stable under the fuel cell operating atmospheres. Each material has to perform not only in its own right but also in conjunction with other system components. For this reason, each cell component must fulfill several different criteria. This paper reviews the materials and the methods used to fabricate the different cell components, such as the cathode, the electrolyte, the anode and the interconnect. Some remarkable results, obtained at IPEN (Nuclear Energy Research Institute in São Paulo, have been presented.

  7. Comparison of Material Behavior of Matrix Graphite for HTGR Fuel Elements upon Irradiation: A literature Survey

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The fuel elements for the HTGRs (i.e., spherical fuel element in pebble-bed type core design and fuel compact in prismatic core design) consists of coated fuel particles dispersed and bonded in a closely packed array within a carbonaceous matrix. This matrix is generally made by mixing fully graphitized natural and needle- or pitchcoke originated powders admixed with a binder material (pitch or phenolic resin), The resulting resinated graphite powder mixture, when compacted, may influence a number of material properties as well as its behavior under neutron irradiation during reactor operation. In the fabrication routes of these two different fuel element forms, different consolidation methods are employed; a quasi-isostatic pressing method is generally adopted to make pebbles while fuel compacts are fabricated by uni-axial pressing mode. The result showed that the hardness values obtained from the two directions showed an anisotropic behavior: The values obtained from the perpendicular section showed much higher micro hardness (176.6±10.5MPa in average) than from the parallel section ((125.6±MPa in average). This anisotropic behavior was concluded to be related to the microstructure of the matrix graphite. This may imply that the uni-axial pressing method to make compacts influence the microstructure of the matrix and hence the material properties of the matrix graphite.

  8. Integrity assessment of research reactor fuel cladding and material testing using eddy current inspection

    International Nuclear Information System (INIS)

    Alencar, Donizete Anderson de

    2004-01-01

    A methodology to perform the integrity assessment of research reactors nuclear fuels cladding, such as those installed in IPR-Rl (TRIGA) and IEA-R1 (MTR), using nondestructive electromagnetic inspection (eddy current) is presented. This methodology is constituted by: the development of calibration reference standards, specific for each type of fuel; the development of special test probes; the recommendations for the inspection equipment calibration; the construction of voltage based evaluation curves and the inspection procedures developed for the characterization of detected flaws. The test probes development, specially those designed for the inspection of MTR fuels cladding, which present access difficulties due to the narrow gap between fuel plates (2,89 mm for IEAR-R1), constituted a challenge that demanded the introduction of unusual materials and constructive techniques. The operational performance of the developed resources, as well as the special operative characteristics of the test probes, such as their immunity to adjacent fuel plates interference and electrical resistivity changes of the fuels meat are experimentally demonstrated. The practical applicability of the developed methodology is verified in non radioactive environment, using a dummy MTR fuel element model, similar to an IEA-R1 reactor fuel element, produced and installed in IPEN, Sao Paulo. The efficacy of the proposed methodology was verified by the achieved results. (author)

  9. Corrosion-induced changes in pore-size distributions of fuel-matrix material

    International Nuclear Information System (INIS)

    Krautwasser, P.; Eatherly, W.P.

    1981-01-01

    In order to understand the mechanism of metallic fission-product adsorption and desorption as well as diffusion in graphitic materials, a detailed knowledge of the material microstructure is essential. Different types of grahitic matrix material used or to be used in fuel elements of the German HTR Program were measured at ORNL in cooperation with the Hahn-Meitner-Institut Berlin. Actual measurements of fission product diffusion and adsorption/desorption were performed at HMI Berlin

  10. Nanomaterials for Polymer Electrolyte Membrane Fuel Cells; Materials Challenges Facing Electrical Energy Storate

    Energy Technology Data Exchange (ETDEWEB)

    Gopal Rao, MRS Web-Editor; Yury Gogotsi, Drexel University; Karen Swider-Lyons, Naval Research Laboratory

    2010-08-05

    Symposium T: Nanomaterials for Polymer Electrolyte Membrane Fuel Cells Polymer electrolyte membrane (PEM) fuel cells are under intense investigation worldwide for applications ranging from transportation to portable power. The purpose of this seminar is to focus on the nanomaterials and nanostructures inherent to polymer fuel cells. Symposium topics will range from high-activity cathode and anode catalysts, to theory and new analytical methods. Symposium U: Materials Challenges Facing Electrical Energy Storage Electricity, which can be generated in a variety of ways, offers a great potential for meeting future energy demands as a clean and efficient energy source. However, the use of electricity generated from renewable sources, such as wind or sunlight, requires efficient electrical energy storage. This symposium will cover the latest material developments for batteries, advanced capacitors, and related technologies, with a focus on new or emerging materials science challenges.

  11. Nonuniform transformation field analysis of multiphase elasto viscoplastic materials: application to MOX fuels

    International Nuclear Information System (INIS)

    Roussette, S.

    2005-05-01

    The description of the overall behavior of nonlinear materials with nonlinear dissipative phases requires an infinity of internal variables. An approximate model involving only a finite number of internal variables, Nonuniform Transformation Field Analysis, is obtained by considering a decomposition of these variables on a finite set of nonuniform transformation fields, called plastic modes. The method is initially developed for incompressible elasto viscoplastic materials. Karhunen-Loeve expansion is proposed to optimize the plastic modes. Then the method is extended to porous elasto viscoplastic materials. Finally the transformation field analysis, developed by Dvorak, is applied to nuclear fuels MOX. This method enables to make sensitivity studies to determine the role of some microstructural parameters on the fuel behaviour. Moreover the adequacy of the nonuniform method for fuels MOX is shown, the final objective being to be able to apply the model to the MOX in 3D. (author)

  12. Survey test of canister, geology, and fuel cladding material interactions

    International Nuclear Information System (INIS)

    Krogness, J.C.; Almassy, M.Y.; Cantley, D.A.; Davis, R.B.

    1979-08-01

    A series of Material Interaction Test (MIT) is being conducted. The first test is being conducted in the Dry Surface Storage Demonstration at the EMAD facility on NTS. This paper discusses details of this first test and gives a status report on the MIT series. 17 figures

  13. ANODE CATALYST MATERIALS FOR USE IN FUEL CELLS

    DEFF Research Database (Denmark)

    2002-01-01

    Catalyst materials having a surface comprising a composition M¿x?/Pt¿3?/Sub; wherein M is selected from the group of elements Fe, Co, Rh and Ir; or wherein M represent two different elements selected from the group comprising Fe, CO, Rh, Ir, Ni, Pd, CU, Ag, Au and Sn; and wherein Sub represents...

  14. New Nuclear Materials Including Non Metallic Fuel Elements. Vol. II. Proceedings of the Conference on New Nuclear Materials Technology, Including Non Metallic Fuel Elements

    International Nuclear Information System (INIS)

    1963-01-01

    One of the major aims of the International Atomic Energy Agency in furthering the peaceful uses of atomic energy is to encourage the development of economical nuclear power. Certainly, one of the more obvious methods of producing economical nuclear power is the development of economical fuels that can be used at high temperatures for long periods of time, and which have sufficient strength and integrity to operate under these conditions without permitting the release of fission products. In addition it is desirable that after irradiation these new fuels be economically reprocessed to reduce further the cost of the fuel cycle. As nuclear power becomes more and more competitive with conventional power the interest in new and more efficient higher-temperature fuels naturally increases rapidly. For these reasons, the Agency organized a Conference on New Nuclear Materials Technology, Including Non-Metallic Fuel Elements, which was held from 1 to 5 July 1963 at the International Hotel, Prague, with the assistance and co-operation of the Government of the Czechoslovak Socialist Republic. A total of 151 scientists attended, from 23 countries and 4 international organizations. The participants heard and discussed more than 60 scientific papers. The Agency wishes to thank the scientists who attended this Conference for their papers and for many spirited discussions that truly mark a successful meeting. The Agency wishes also to record its gratitude for the assistance and generous hospitality accorded the Conference, the participants and the Agency's staff by the Government of the Czechoslovak Socialist Republic and by the people of Prague. The scientific information contained in these Proceedings should help to quicken the pace of progress in the fabrication of new and m ore economical fuels, and it is hoped that these proceedings will be found useful to all workers in this and related fields

  15. Fuels and Materials Division biennial report, fiscal years 1979 and 1980

    International Nuclear Information System (INIS)

    Brown, H.L.

    1980-10-01

    The Fuels and Materials Division of EG and G Idaho, Inc., reports its technical activities for the fiscal years 1979 to 1980. This Division supports technical programs at the Idaho National Engineering Laboratory with materials science, materials engineering, nondestructive evaluation engineering, and remote handling of radioactive and other hazardous materials and conducts several direct-funded materials-related programs for the US Department of Energy. This report details programmatic and other support activities, direct-funded programs, organization and facility status, and EG and G Idaho-funded technical development programs

  16. Analysis of Underground Storage Tanks System Materials to Increased Leak Potential Associated with E15 Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kass, Michael D [ORNL; Theiss, Timothy J [ORNL; Janke, Christopher James [ORNL; Pawel, Steven J [ORNL

    2012-07-01

    include model year 2001 light-duty vehicles, but specifically prohibited use in motorcycles and off-road vehicles and equipment. UST stakeholders generally consider fueling infrastructure materials designed for use with E0 to be adequate for use with E10, and there are no known instances of major leaks or failures directly attributable to ethanol use. It is conceivable that many compatibility issues, including accelerated corrosion, do arise and are corrected onsite and, therefore do not lead to a release. However, there is some concern that higher ethanol concentrations, such as E15 or E20, may be incompatible with current materials used in standard gasoline fueling hardware. In the summer of 2008, DOE recognized the need to assess the impact of intermediate blends of ethanol on the fueling infrastructure, specifically located at the fueling station. This includes the dispenser and hanging hardware, the underground storage tank, and associated piping. The DOE program has been co-led and funded by the Office of the Biomass Program and Vehicle Technologies Program with technical expertise from the Oak Ridge National Laboratory (ORNL) and the National Renewable Energy Laboratory (NREL). The infrastructure material compatibility work has been supported through strong collaborations and testing at Underwriters Laboratories (UL). ORNL performed a compatibility study investigating the compatibility of fuel infrastructure materials to gasoline containing intermediate levels of ethanol. These results can be found in the ORNL report entitled Intermediate Ethanol Blends Infrastructure Materials Compatibility Study: Elastomers, Metals and Sealants (hereafter referred to as the ORNL intermediate blends material compatibility study). These materials included elastomers, plastics, metals and sealants typically found in fuel dispenser infrastructure. The test fuels evaluated in the ORNL study were SAE standard test fuel formulations used to assess material-fuel compatibility within a

  17. Estimation of elastic properties of nuclear fuel material using longitudinal ultrasonic velocity - A new approach

    International Nuclear Information System (INIS)

    Phani, K.K.; Sanyal, Dipayan; Sengupta, A.K.

    2007-01-01

    A novel methodology has been suggested for estimation of elastic properties of nuclear fuel materials based on porosity and longitudinal ultrasonic velocity. New correlations have been proposed between the elastic moduli, namely, the Young's modulus, shear modulus, bulk modulus and Poisson's ratio, and the longitudinal ultrasonic velocity. The theoretical predictions agreed extremely well with data reported for uranium dioxide and uranium nitride, respectively. The proposed method promises to be extremely useful for non-destructive evaluation of elastic properties of irradiated fuel materials which are not amenable to shear velocity measurements due to extreme fragility

  18. Mining of Radioactive Raw Materials as an Origin of the Nuclear Fuel Chain

    Directory of Open Access Journals (Sweden)

    Bedřich Michálek

    2007-01-01

    Full Text Available The mining of radioactive raw materials may be considered as an origin of the nuclear fuel chain and thus determines the amount of radioactive wastes which have to be stored safety in the final stage of the fuel chain. The paper informs about the existing trends in mining of radioactive raw materials in the world, provides an overview of development in mining in the Czech Republic and of possibilities of future exploiting some uranium deposits. It points a possibility of non-traditional obtaining uranium from mine waters from underground uranium mines closed and flooded earlier.

  19. U.S. FUEL CYCLE TECHNOLOGIES R&D PROGRAM FOR NEXT GENERATION NUCLEAR MATERIALS MANAGEMENT

    Directory of Open Access Journals (Sweden)

    M.C. MILLER

    2013-11-01

    Full Text Available The U.S. Department of Energy's Fuel Cycle Technologies R&D program under the Office of Nuclear Energy is working to advance technologies to enhance both the existing and future fuel cycles. One thrust area is in developing enabling technologies for next generation nuclear materials management under the Materials Protection, Accounting and Control Technologies (MPACT Campaign where advanced instrumentation, analysis and assessment methods, and security approaches are being developed under a framework of Safeguards and Security by Design. An overview of the MPACT campaign's activities and recent accomplishments is presented along with future plans.

  20. Evaluation of the Total Cost of Ownership of Fuel Cell-Powered Material Handling Equipment

    Energy Technology Data Exchange (ETDEWEB)

    Ramsden, T.

    2013-04-01

    This report discusses an analysis of the total cost of ownership of fuel cell-powered and traditional battery-powered material handling equipment (MHE, or more typically 'forklifts'). A number of fuel cell MHE deployments have received funding support from the federal government. Using data from these government co-funded deployments, DOE's National Renewable Energy Laboratory (NREL) has been evaluating the performance of fuel cells in material handling applications. NREL has assessed the total cost of ownership of fuel cell MHE and compared it to the cost of ownership of traditional battery-powered MHE. As part of its cost of ownership assessment, NREL looked at a range of costs associated with MHE operation, including the capital costs of battery and fuel cell systems, the cost of supporting infrastructure, maintenance costs, warehouse space costs, and labor costs. Considering all these costs, NREL found that fuel cell MHE can have a lower overall cost of ownership than comparable battery-powered MHE.

  1. Summary of NRC LWR safety research programs on fuel behavior, metallurgy/materials and operational safety

    International Nuclear Information System (INIS)

    Bennett, G.L.

    1979-09-01

    The NRC light-water reactor safety-research program is part of the NRC regulatory program for ensuring the safety of nuclear power plants. This paper summarizes the results of NRC-sponsored research into fuel behavior, metallurgy and materials, and operational safety. The fuel behavior research program provides a detailed understanding of the response of nuclear fuel assemblies to postulated off-normal or accident conditions. Fuel behavior research includes studies of basic fuel rod properties, in-reactor tests, computer code development, fission product release and fuel meltdown. The metallurgy and materials research program provides independent confirmation of the safe design of reactor vessels and piping. This program includes studies on fracture mechanics, irradiation embrittlement, stress corrosion, crack growth, and nondestructive examination. The operational safety research provides direct assistance to NRC officials concerned with the operational and operational-safety aspects of nuclear power plants. The topics currently being addressed include qualification testing evaluation, fire protection, human factors, and noise diagnostics

  2. Direct conversion of surplus fissile materials, spent nuclear fuel, and other materials to high-level-waste glass

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Elam, K.R.

    1995-01-01

    With the end of the cold war the United States, Russia, and other countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. The United States Academy of Sciences (NAS) has recommended that these surplus fissile materials (SFMs) be processed so they are no more accessible than plutonium in spent nuclear fuel (SNF). This spent fuel standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. The NAS recommended investigation of three sets of options for disposition of SFMs while meeting the spent fuel standard: (1) incorporate SFMs with highly radioactive materials and dispose of as waste, (2) partly burn the SFMs in reactors with conversion of the SFMs to SNF for disposal, and (3) dispose of the SFMs in deep boreholes. The US Government is investigating these options for SFM disposition. A new method for the disposition of SFMs is described herein: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptinium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal

  3. Nondestructive assay of special nuclear material for uranium fuel-fabrication facilities

    International Nuclear Information System (INIS)

    Smith, H.A. Jr.; Schillebeeckx, P.

    1997-01-01

    A high-quality materials accounting system and effective international inspections in uranium fuel-fabrication facilities depend heavily upon accurate nondestructive assay measurements of the facility's nuclear materials. While item accounting can monitor a large portion of the facility inventory (fuel rods, assemblies, storage items), the contents of all such items and mass values for all bulk materials must be based on quantitative measurements. Weight measurements, combined with destructive analysis of process samples, can provide highly accurate quantitative information on well-characterized and uniform product materials. However, to cover the full range of process materials and to provide timely accountancy data on hard-to-measure items and rapid verification of previous measurements, radiation-based nondestructive assay (NDA) techniques play an important role. NDA for uranium fuel fabrication facilities relies on passive gamma spectroscopy for enrichment and U isotope mass values of medium-to-low-density samples and holdup deposits; it relies on active neutron techniques for U-235 mass values of high-density and heterogeneous samples. This paper will describe the basic radiation-based nondestructive assay techniques used to perform these measurements. The authors will also discuss the NDA measurement applications for international inspections of European fuel-fabrication facilities

  4. Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-01-26

    In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.

  5. In-core instrumentation and in-situ measurement in connection with fuel behaviour. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    The subject of this meeting has been touched on briefly in most of the Specialist's and topical meetings related to fuel behaviour. On the basis of the conclusions and recommendations of these meetings the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended the Agency to organize a dedicated Specialist's Meeting on the subject. The twenty one papers covered the instrumentation, sensors, methods and computer codes currently used in Material Test Reactor (MTR) and power reactors as well as improved instrumentation and methods. The meeting acknowledged the fast development of fuel modelling and therefore the growing need of dedicated high burnup fuel experiments carried out in MTR reactors on refabricated rods from power reactors. In order to reduce safety margins in power reactors, thus improving economics, the necessity to develop more sophisticated on-line calculations, based on improved sensors, was recognized, although this development is limited by insufficient knowledge of the mechanisms involved. Refs, figs, tabs

  6. Health assessment of gasoline and fuel oxygenate vapors: generation and characterization of test materials.

    Science.gov (United States)

    Henley, Michael; Letinski, Daniel J; Carr, John; Caro, Mario L; Daughtrey, Wayne; White, Russell

    2014-11-01

    In compliance with the Clean Air Act regulations for fuel and fuel additive registration, the petroleum industry, additive manufacturers, and oxygenate manufacturers have conducted comparative toxicology testing on evaporative emissions of gasoline alone and gasoline containing fuel oxygenates. To mimic real world exposures, a generation method was developed that produced test material similar in composition to the re-fueling vapor from an automotive fuel tank at near maximum in-use temperatures. Gasoline vapor was generated by a single-step distillation from a 1000-gallon glass-lined kettle wherein approximately 15-23% of the starting material was slowly vaporized, separated, condensed and recovered as test article. This fraction was termed vapor condensate (VC) and was prepared for each of the seven test materials, namely: baseline gasoline alone (BGVC), or gasoline plus an ether (G/MTBE, G/ETBE, G/TAME, or G/DIPE), or gasoline plus an alcohol (G/EtOH or G/TBA). The VC test articles were used for the inhalation toxicology studies described in the accompanying series of papers in this journal. These studies included evaluations of subchronic toxicity, neurotoxicity, immunotoxicity, genotoxicity, reproductive and developmental toxicity. Results of these studies will be used for comparative risk assessments of gasoline and gasoline/oxygenate blends by the US Environmental Protection Agency. Copyright © 2014 Elsevier Inc. All rights reserved.

  7. Nuclear-Fuel-Cycle Research Program: availability of geotoxic material

    Energy Technology Data Exchange (ETDEWEB)

    Wachter, B.G.; Kresan, P.L.

    1982-09-01

    This report represents an analog approach to the characterization of the environmental behavior of geotoxic waste materials (toxic material emplaced in the earth's crust) as drawn from literature on the Oklo natural fission reactors and uranium ore deposits relative to radioactive wastes, and hydrothermal metal ore deposits relative to stable toxic wastes. The natural analog data were examined in terms of mobility and immobility of selected radioactive or stable waste elements and are presented in matrix relationship with their prime geochemical variables. A numerical system of ranking those relationships for purposes of hazard-indexing is proposed. Geochemical parameters (especially oxidation/reduction potential) are apparently more potent mobilizers/immobilizers than geological or hydrological conditions in many, if not most, geologic environments for most radioactive waste elements. Heavy metal wastes, by analogy to hydrothermal ore systems and geothermal systems, are less clear in their behavior but similar geochemical patterns do apply. Depth relationships between geochemical variables and waste element behavior show some surprises. It is significantly indicated that for waste isolation, deeper is not necessarily better geochemically. Relatively shallow isolation in host rocks such as shale could offer maximum immobility. This paper provides a geochemical outline for examining analog models as well as a departure point for improved quantification of geological and geochemical indexing of toxic waste hazards.

  8. Nuclear-Fuel-Cycle Research Program: availability of geotoxic material

    International Nuclear Information System (INIS)

    Wachter, B.G.; Kresan, P.L.

    1982-09-01

    This report represents an analog approach to the characterization of the environmental behavior of geotoxic waste materials (toxic material emplaced in the earth's crust) as drawn from literature on the Oklo natural fission reactors and uranium ore deposits relative to radioactive wastes, and hydrothermal metal ore deposits relative to stable toxic wastes. The natural analog data were examined in terms of mobility and immobility of selected radioactive or stable waste elements and are presented in matrix relationship with their prime geochemical variables. A numerical system of ranking those relationships for purposes of hazard-indexing is proposed. Geochemical parameters (especially oxidation/reduction potential) are apparently more potent mobilizers/immobilizers than geological or hydrological conditions in many, if not most, geologic environments for most radioactive waste elements. Heavy metal wastes, by analogy to hydrothermal ore systems and geothermal systems, are less clear in their behavior but similar geochemical patterns do apply. Depth relationships between geochemical variables and waste element behavior show some surprises. It is significantly indicated that for waste isolation, deeper is not necessarily better geochemically. Relatively shallow isolation in host rocks such as shale could offer maximum immobility. This paper provides a geochemical outline for examining analog models as well as a departure point for improved quantification of geological and geochemical indexing of toxic waste hazards

  9. Advanced fuel pellet materials and designs for water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2004-10-01

    with the technological advances attempted in doping of fuel pellets with the primary objective of obtaining larger grains. While most of the papers gave an account of the experimental studies on addition of various dopants in different fuel materials, some of them outlined the behaviour of such pellets at sintering process. Papers dealing with 'Fission gas release from fuel pellets under high burnup conditions were presented in Session 3. Session 4 was devoted to the evolution of fuel pellet structure and thermal properties at high burnup. Session 5 was dealing with fuel pellet-cladding interaction (PCI) being a complex phenomenon that may lead to cladding failure and subsequent release of fission products into the reactor coolant. Research efforts to understand better the PCI phenomenon and minimize it with design solutions are considered necessary

  10. Evaluation of source term parameters for spent fuel disposal in foreign countries. (1) Instant release fraction from spent fuel matrices and composition materials for fuel assemblies

    International Nuclear Information System (INIS)

    Nagata, Masanobu; Chikazawa, Takahiro; Kitamura, Akira; Tachi, Yukio; Akahori, Kuniaki

    2016-01-01

    Although spent nuclear fuel is planned to be disposed after reprocessing and vitrification of high-level radioactive waste (HLW), feasibility study on direct disposal of spent nuclear fuel (SF) has been started as one of the alternative disposal options to flexibly apply change of future energy situation in Japan. Radionuclide inventories and their release behavior after breaching spent fuel container should be assessed to confirm safety of the SF disposal. Especially, instant release fractions (IRFs), which are fractions of radionuclide released relatively faster than those released with congruent dissolution with SF and construction materials after breaching spent fuel container, may have an impact on safety assessment of the direct disposal of SF. However, detailed studies on evaluation / estimation of IRF have not been performed in Japan. Therefore, we investigated some foreign safety assessment reports on direct disposal of SF by focusing on IRF for the safety assessment of Japanese SF disposal system. As a result of comparison between the safety assessment reports in foreign countries, although some fundamental data have been referred to the reports in common, the final source term dataset was seen differences between countries in the result of taking into account the national circumstances (reactor types and burnups, etc.). We also found the difference of assignment of uncertainties among the investigated reports; a report selected pessimistic values and another report selected mean values and their deviations. It is expected that these findings are useful as fundamental information for the safety assessment of Japanese SF disposal system. (author)

  11. FUNDAMENTAL STUDIES OF THE DURABILITY OF MATERIALS FOR INTERCONNECTS IN SOLID OXIDE FUEL CELLS

    Energy Technology Data Exchange (ETDEWEB)

    Frederick S. Pettit; Gerald H. Meier

    2003-06-30

    This report describes the result of the first eight months of effort on a project directed at improving metallic interconnect materials for solid oxide fuel cells (SOFCs). The results include cyclic oxidation studies of a group of ferritic alloys, which are candidate interconnect materials. The exposures have been carried out in simulated fuel cell atmospheres. The oxidation morphologies have been characterized and the ASR has been measured for the oxide scales. The effect of fuel cell electric current density on chromia growth rates has been considered The thermomechanical behavior of the scales has been investigated by stress measurements using x-ray diffraction and interfacial fracture toughness measurements using indentation. The ultimate goal of this thrust is to use knowledge of changes in oxide thickness, stress and adhesion to develop accelerated testing methods for evaluating SOFC interconnect alloys. Finally a theoretical assessment of the potential for use of ''new'' metallic materials as interconnect materials has been conducted and is presented in this report. Alloys being considered include materials based on pure nickel, materials based on the ''Invar'' concept, and coated materials to optimize properties in both the anode and cathode gases.

  12. Order for execution of the law concerning regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1978-01-01

    Under the above mentioned law this order prescribes the procedures of controls given to the persons who wish to conduct refining and fabricating businesses, to construct and operate reactors, and to use nuclear source materials, nuclear fuel materials and internationally controlled materials. The common controlling principle prescribed is that the permission or authorization necessary for above listed businesses should be applied for at each factory or each place of business. Based on the principle, the order prescribes: the procedures to apply for the authorization of the refining business, the permission of the change thereof, and the permission of the fabricating business and the change, thereof (the 1st chapter); the procedures to apply for the permission of the construction of reactors and of the change of the construction, as well as the procedure to do periodic inspections of reactor facilities (the 2nd chapter); the procedures to apply for the permission to use nuclear fuel materials and to change the use thereof, the submission of the report to use nuclear source materials, as well as the procedure to apply for the permission to use internationally controlled materials. In the 4th chapter the order lists up the items on which the competent Ministers may require reports from the person who carries on the relevant business. (Matsushima, A.)

  13. Carbon dioxide: a raw material and a future chemical fuel for a sustainable energy industry

    Science.gov (United States)

    Amouroux, J.; Siffert, P.

    2011-03-01

    Carbon dioxide is a major raw material of the future, for the capture plants which use amines, aminoacids, ammonia or zeolites. This very high purity raw material (99.9 %) opens the way of a new industrial revolution in agreement with the proposal of Nobel Prize laureates and the DOE strategy. Our goal is to explain the large advantages and the main routes for CO2 valorization, which are starting around the world. The most promising ways for this valorization are methanol synthesis as fuel for transportation and methane formation for electricity network regulation. The first way allows the use of liquid fuels, as distribution infrastructure already exists; instead of gaseous fuels (H2), for which there is storage, distribution problems and no infrastructure exist. The second way is methane synthesis during off-peak hours and burning of this methane during peak hours in order to regulate the electric network.

  14. An investigation into carbon nanostructured materials as catalyst support in proton exchange membrane fuel cells

    DEFF Research Database (Denmark)

    Veltzé, Sune

    than carbon blacks. Even then the possible durability of the platinum containing catalyst is a major concern for fuel cell degradation during operation. In order to evaluate platinum containing electrocatalysts for proton exchange membrane fuel cells (PEMFC), the rotating disc electrode (RDE......Polymer electrolyte fuel cells (PEFCs) are among the key research areas concerning clean cost-effective energy. Carbon nano fibres (CNF), single walled carbon nano tubes (SWCNT), multi walled carbon nano tubes (MWCNT) and other related materials are among the possible successors to standard carbon...... black support materials for low platinum containing electrocatalyst. This is partly due to their high electronic conductivity. Partly due to their high surface area needed for the dispersion of nanoparticulate metal-clusters. In addition carbon nano-structures (CNF, SWCNT, MWCNT etc.) are more durable...

  15. Taking into account the risk of radioactive material dissemination in French fuel cycle plants

    International Nuclear Information System (INIS)

    Ruiz, J.

    1996-01-01

    This paper presents the design principles for handling the radioactive material dissemination risk in French fuel cycle plants (fuel enrichment and fabrication plants, reprocessing plants, processing and storing facilities). Consideration of the dissemination risk in this type of facilities is made through a deep protection procedure involving the settlement of successive containment systems. These containment systems are dimensioned according to the radioactive material dissemination risk and the external exposure risk. Regulations, standards and guides are first reviewed, followed by a presentation of the various components related to the containment realization in the fuel cycle plants and the way they are realized, the evolutions of methods and designs, the supervision and inspection, and the experience gained

  16. Analysis of radwaste material management options for experimental DUPIC fuel fabrication process

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. H.; Park, J. J.; Yang, M. S.; Kim, K. H.; Shin, J. M.; Lee, H. S.; Ko, W. I.; Lee, J. W.; Yim, S. P.; Hong, D. H.; Lee, J. Y.; Baik, S. Y.; Song, W. S.; Yoo, B. O.; Lee, E. P.; Kang, I. S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    This report is desirable to review management options in advance for radioactive waste generated from manufacturing experiment of DUPIC nuclear fuel as well as residual nuclear material and dismantled equipment. This report was written for helping researchers working in related facilities to DUPIC project understanding management of DUPIC radioactive waste as well as fellows in DUPIC project. Also, it will be used as basic material to prove transparency and safeguardability of DUPIC fuel cycle. In order to meet these purposes, this report includes basic experiment plan for manufacturing DUPIC nuclear fuel, outlines for DUPIC manufacturing facility and equipment, arising source and estimated amount of radioactive waste, waste classification and packing, transport cask, transport procedures. 15 refs., 31 figs., 11 tabs. (Author)

  17. Zr-alloys, the nuclear material for water reactor fuel. A survey and update with focus on fuel for pressurized water reactor systems

    International Nuclear Information System (INIS)

    Weidinger, H.

    2008-01-01

    This paper is intended to provide a solid overview on the development of the requirements and the respective answers found as far as water cooled fuel rods and assemblies are concerned. It shall be a help as well for designers and manufacturers as also for users of this fuel, because only a broad and consistent knowledge on all aspects of the application of this material in nuclear fuel can guarantee a successful operation under the still increasing requirements in water cooled reactor cores

  18. Instrumentation Technologies for Improving an Irradiation Testing of Nuclear Fuels and Materials at the HANARO

    International Nuclear Information System (INIS)

    Kim, Bong Goo; Park, Sung Jae; Choo, Ki Nam

    2011-01-01

    Over 50 years of nuclear fuels and materials irradiation testing has led to many countries developing significant improvements in instrumentation to monitor physical parameters and to control the test conditions in Materials Test Reactors (MTRs) or research reactors. Recent effort to deploy new fuels and materials in existing and advanced reactors has increased the demand for well-instrumented irradiation tests. Specifically, demand has increased for tests with sensors capable of providing real-time measurement of key parameters, such as temperature, geometry changes, thermal conductivity, fission gas release, cracking, coating buildup, thermal and fast flux, etc. This review paper documents the current state of instrumentation technologies in MTRs in the world and summarizes on-going research efforts to deploy new sensors. There is increased interest to irradiate new materials and reactor fuels for advanced PWRs and the Gen-IV reactor systems, such as SFRs (Sodium-cooled Fast Reactors), VHTRs (Very-High-Temperature Reactors), SCWRs (Supercritical-Water-cooled Reactors) and GFRs (Gas-cooled Fast Reactor). This review documents the current state of instrumentation technologies in MTRs in the world, identifies challenges faced by previous testing methods and how these challenges were overcome. A wide range of sensors are available to measure key parameters of interest during fuels and materials irradiations in MTRs. Such sensors must be reliable, small size, highly accurate, and able to withstand harsh conditions. On-going development efforts are focusing on providing MTR users a wider range of parameter measurements with increased accuracy. In addition, development efforts are focusing on reducing the impact of sensor on measurements by reducing sensor size. This report includes not only status of instrumentation using research reactors in the world to irradiate nuclear fuels and materials but also future directions relating to instrumentation technologies for

  19. Zirconium carbonitride pellets by internal sol gel and spark plasma sintering as inert matrix fuel material

    Science.gov (United States)

    Hedberg, Marcus; Cologna, Marco; Cambriani, Andrea; Somers, Joseph; Ekberg, Christian

    2016-10-01

    Inert matrix fuel is a fuel type where the fissile material is blended with a solid diluent material. In this work zirconium carbonitride microspheres have been produced by internal sol gel technique, followed by carbothermal reduction. Material nitride purities in the produced materials ranged from Zr(N0.45C0.55) to Zr(N0.74C0.26) as determined by X-ray diffraction and application of Vegard's law. The zirconium carbonitride microspheres have been pelletized by spark plasma sintering (SPS) and by conventional cold pressing and sintering. In all SPS experiments cohesive pellets were formed. Maximum final density reached by SPS at 1700 °C was 87% theoretical density (TD) compared to 53% TD in conventional sintering at 1700 °C. Pore sizes in all the produced pellets were in the μm scale and no density gradients could be observed by computer tomography.

  20. Experience of air transport of nuclear fuel material as type A package

    International Nuclear Information System (INIS)

    Kawasaki, Masashi; Kageyama, Tomio; Suzuki, Toru

    2004-01-01

    Special law on nuclear disaster countermeasures (hereafter called as to nuclear disaster countermeasures low) that is domestic law for dealing with measures for nuclear disaster, was enforced in June, 2000. Therefore, nuclear enterprise was obliged to report accidents as required by nuclear disaster countermeasures law, besides meeting the technical requirement of existent transport regulation. For overseas procurement of plutonium reference materials that are needed for material accountability, A Type package must be transported by air. Therefore, concept of air transport of nuclear fuel materials according to the nuclear disaster countermeasures law was discussed, and the manual including measures against accident in air transport was prepared for the oversea procurement. In this presentation, the concept of air transport of A Type package containing nuclear fuel materials according to the nuclear disaster countermeasures law, and the experience of a transportation of plutonium solution from France are shown. (author)

  1. Materials, coolants, and components for perspective transmutation power units of VVER spent fuel

    International Nuclear Information System (INIS)

    Matal, O.; Hosnedl, P.

    2004-01-01

    The aim of the contribution is provide information on the selection and development of materials and molten salt coolants for high temperature applications for the new nuclear power units and discuss briefly design problems of selected components such as reactor and salt-salt and salt-gas heat exchangers. The following concepts and topics are discussed : Liquid metal reactor concepts (liquid sodium coolant concepts, lead-bismuth coolant concepts, lead coolant concept), CO 2 -cooled reactor concepts, molten-salt fueled and cooled reactor concepts, existing and prospective molten-salt fuels and coolants and structural materials, structural material development for MSR concepts, experimental studies on material corrosion in fluoride molten-salts, and R and D of some components for MSR. (P.A.)

  2. Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P.O. 1236909 Final report

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

    1978-12-01

    The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel

  3. Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P. O. 1236909 Final report

    Energy Technology Data Exchange (ETDEWEB)

    Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

    1978-12-01

    The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel.

  4. Imidazolium-Based Polymeric Materials as Alkaline Anion-Exchange Fuel Cell Membranes

    Science.gov (United States)

    Narayan, Sri R.; Yen, Shiao-Ping S.; Reddy, Prakash V.; Nair, Nanditha

    2012-01-01

    Polymer electrolyte membranes that conduct hydroxide ions have potential use in fuel cells. A variety of polystyrene-based quaternary ammonium hydroxides have been reported as anion exchange fuel cell membranes. However, the hydrolytic stability and conductivity of the commercially available membranes are not adequate to meet the requirements of fuel cell applications. When compared with commercially available membranes, polystyrene-imidazolium alkaline membrane electrolytes are more stable and more highly conducting. At the time of this reporting, this has been the first such usage for imidazolium-based polymeric materials for fuel cells. Imidazolium salts are known to be electrochemically stable over wide potential ranges. By controlling the relative ratio of imidazolium groups in polystyrene-imidazolium salts, their physiochemical properties could be modulated. Alkaline anion exchange membranes based on polystyrene-imidazolium hydroxide materials have been developed. The first step was to synthesize the poly(styrene-co-(1-((4-vinyl)methyl)-3- methylimidazolium) chloride through a free-radical polymerization. Casting of this material followed by in situ treatment of the membranes with sodium hydroxide solutions provided the corresponding hydroxide salts. Various ratios of the monomers 4-chloromoethylvinylbenzine (CMVB) and vinylbenzine (VB) provided various compositions of the polymer. The preferred material, due to the relative ease of casting the film, and its relatively low hygroscopic nature, was a 2:1 ratio of CMVB to VB. Testing confirmed that at room temperature, the new membranes outperformed commercially available membranes by a large margin. With fuel cells now in use at NASA and in transportation, and with defense potential, any improvement to fuel cell efficiency is a significant development.

  5. Not just lumber--Using wood in the sustainable future of materials, chemicals, and fuels

    Science.gov (United States)

    Joseph E. Jakes; Xavier Arzola; Richard Bergman; Peter Ciesielski; Christopher G. Hunt; Nima Rahbar; Mandla Tshabalala; Alex C. Wiedenhoeft; Samuel L. Zelinka

    2016-01-01

    Forest-derived biomaterials can play an integral role in a sustainable and renewable future. Research across a range of disciplines is required to develop the knowledge necessary to overcome the challenges of incorporating more renewable forest resources in materials, chemicals, and fuels. We focus on wood specifically because in our view, better characterization of...

  6. Ab initio Defect Energetics in LaBO3 Perovskite Solid Oxide Fuel Cell Materials

    DEFF Research Database (Denmark)

    Lee, Yueh-Lin; Morgan, Dane; Kleis, Jesper

    2009-01-01

    Perovskite materials of the form ABO3 are a promising family of compounds for use in solid oxide fuel cell (SOFC) cathodes. Study of the physics of these compounds under SOFC conditions with ab initio methods is particularly challenging due to high temperatures, exchange of oxygen with O2 gas...

  7. Status of the nondestructive examination equipment for the fuels and materials examination facility

    International Nuclear Information System (INIS)

    Frandsen, G.B.

    1980-01-01

    The present status of Nondestructive Examination (NDE) Equipment proposed for the Fuels and Materials Examination Facility (FMEF) now under construction at the Hanford Engineering Development Laboratory is discussed. Items discussed include the NDE cell receiving machine, the dismantling machine, the standard examination stage, profilometry, eddy current, wire wrap removal machine, surface examination, gamma scan and general NDE equipment

  8. The reprocessing of irradiated MTR fuel and the nuclear material accountancy - Dounreay, UKAEA

    International Nuclear Information System (INIS)

    Barrett, T.R.; Harrison, R.

    1997-01-01

    The reprocessing of irradiated HEU MTR fuel is a sensible part of a safeguards regime. It brings together fuel otherwise scattered around the world into a concerted accountancy and protection arrangement. From a nuclear material accountants view the overall accountancy performance has been excellent. While investigations have been required for a few individual MUFs or trends, very little effort has required to be expended by the Nuclear Materials Control Department. That is a definition of a 'good plant'; it operates, measures and records input and output streams, and then the accountancy falls into place. As identified in this paper, the accountancy of the nuclear material processed in the plant is well founded and sound. The accountancy results over several decades confirm the adequacy of the safeguards arrangements at Dounreay. The processing makes good commercial sense and meets the current philosophy of recycling valuable resource materials. The risk of operating the full fuel cycle are less than those of extended storage of irradiated fuel at disparate diverse locations. The reprocessing at Dounreay accords with all of these philosophies. The assessed risk is at a very low level, well within published UK HSE 'tolerability of risk' regulatory guidelines. The impact of the operations are similarly low within the guidelines, for the operators and for the general public. (author)

  9. Corrosion resistance of metallic materials for use in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Legry, J.P.; Pelras, M.; Turluer, G.

    1989-01-01

    This paper reviews the corrosion resistance properties required from metallic materials to be used in the various developments of the PUREX process for nuclear fuel reprocessing. Stainless steels, zirconium or titanium base alloys are considered for the various plant components, where nitric acid is the main electrolyte with differing acid and nitrate concentrations, temperature and oxidizing species. (author)

  10. Energy Frontier Research Center, Center for Materials Science of Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Todd R. Allen

    2011-12-01

    This is a document required by Basic Energy Sciences as part of a mid-term review, in the third year of the five-year award period and is intended to provide a critical assessment of the Center for Materials Science of Nuclear Fuels (strategic vision, scientific plans and progress, and technical accomplishments).

  11. Standard guide for qualification of laboratory analysts for the analysis of nuclear fuel cycle materials

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This guide covers the qualification of analysts to perform chemical analysis or physical measurements of nuclear fuel cycle materials. The guidance is general in that it is applicable to all analytical methods, but must be applied method by method. Also, the guidance is general in that it may be applied to initial qualification or requalification.

  12. Corrosion of materials in Molten Carbonate Fuel Cells (MCFC): Literature review

    Science.gov (United States)

    1987-01-01

    Molten carbonate fuel cell (MCFC) separator plates are subjected to corrosive environments. This review focuses on the most comprehensive compiled available data from the major government contractors like ERC and UTC, together with a basic review of alloys resistant to oxidation, sulfidation, carburization, and hot corrosion. Results of this review are used in selecting test materials for the current program. A number of metal oxides were examined. This review discusses the tests performed by UTC and ERC on separator plate materials.

  13. Coordinated safeguards for materials management in chemical separation, conversion, and fuel fabrication facilities

    International Nuclear Information System (INIS)

    Dayem, H.A.; Cobb, D.; Dietz, R.J.; Hakkila, E.A.; Shipley, J.P.; Smith, D.B.

    1978-01-01

    The benefits of dynamic materials accounting are demonstrated by comparing the timeliness, spatial specificity, and sensitivity with that of current materials accounting requirements. Examples of conceptual dynamic accountability systems are given for proposed domestic fuel-cycle facilities. Modeling and simulation results indicate that such systems can detect diversion in hours or days with improved diversion sensitivity and can localize diversion to a single unit process area. 17 refs

  14. Transfer tunnel transporter system for the Fuels and Materials Examination Facility

    International Nuclear Information System (INIS)

    Petty, J.A.; Miller, S.C.; Richards, J.T.

    1981-01-01

    The detail design is complete and fabrication is approximately 75% complete on the Transfer Tunnel Transporter System. This system provides material handling capability for large, bulky equipment between two hot cells in a new Breeder Reactor Program support facility, the Fuels and Materials Examination Facility. One hot cell has an air atmosphere, the other a high purity inert gas atmosphere which must be maintained during transfer operations. System design features, operational capabilities and remote recovery provisions are described

  15. Mass, energy and material balances of SRF production process. Part 3: solid recovered fuel produced from municipal solid waste.

    Science.gov (United States)

    Nasrullah, Muhammad; Vainikka, Pasi; Hannula, Janne; Hurme, Markku; Kärki, Janne

    2015-02-01

    This is the third and final part of the three-part article written to describe the mass, energy and material balances of the solid recovered fuel production process produced from various types of waste streams through mechanical treatment. This article focused the production of solid recovered fuel from municipal solid waste. The stream of municipal solid waste used here as an input waste material to produce solid recovered fuel is energy waste collected from households of municipality. This article presents the mass, energy and material balances of the solid recovered fuel production process. These balances are based on the proximate as well as the ultimate analysis and the composition determination of various streams of material produced in a solid recovered fuel production plant. All the process streams are sampled and treated according to CEN standard methods for solid recovered fuel. The results of the mass balance of the solid recovered fuel production process showed that 72% of the input waste material was recovered in the form of solid recovered fuel; 2.6% as ferrous metal, 0.4% as non-ferrous metal, 11% was sorted as rejects material, 12% as fine faction and 2% as heavy fraction. The energy balance of the solid recovered fuel production process showed that 86% of the total input energy content of input waste material was recovered in the form of solid recovered fuel. The remaining percentage (14%) of the input energy was split into the streams of reject material, fine fraction and heavy fraction. The material balances of this process showed that mass fraction of paper and cardboard, plastic (soft) and wood recovered in the solid recovered fuel stream was 88%, 85% and 90%, respectively, of their input mass. A high mass fraction of rubber material, plastic (PVC-plastic) and inert (stone/rock and glass particles) was found in the reject material stream. © The Author(s) 2014.

  16. Analysis of difficulties accounting and evaluating nuclear material of PWR fuel plant

    International Nuclear Information System (INIS)

    Zhang Min; Jue Ji; Liu Tianshu

    2013-01-01

    Background: Nuclear materials accountancy must be developed for nuclear facilities, which is required by regulatory in China. Currently, there are some unresolved problems for nuclear materials accountancy of bulk nuclear facilities. Purpose: The retention values and measurement errors are analyzed in nuclear materials accountancy of Power Water Reactor (PWR) fuel plant to meet the regulatory requirements. Methods: On the basis of nuclear material accounting and evaluation data of PWR fuel plant, a deep analysis research including ratio among random error variance, long-term systematic error variance, short-term systematic error variance and total error involving Material Unaccounted For (MUF) evaluation is developed by the retention value measure in equipment and pipeline. Results: In the equipment pipeline, the holdup estimation error and its total proportion are not more than 5% and 1.5%, respectively. And the holdup estimation can be regraded as a constant in the PWR nuclear material accountancy. Random error variance, long-term systematic error variance, short-term systematic error variance of overall measurement, and analytical and sampling methods are also obtained. A valuable reference is provided for nuclear material accountancy. Conclusion: In nuclear material accountancy, the retention value can be considered as a constant. The long-term systematic error is a main factor in all errors, especially in overall measurement error and sampling error: The long-term systematic errors of overall measurement and sampling are considered important in the PWR nuclear material accountancy. The proposals and measures are applied to the nuclear materials accountancy of PWR fuel plant, and the capacity of nuclear materials accountancy is improved. (authors)

  17. Study on the standard establishment for the integrity assessment of nuclear fuel cladding Materials

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S. S.; Kim, S. H.; Jung, Y. K.; Yang, C. Y.; Kim, I. G.; Choi, Y. H.; Kim, H. J.; Kim, M. W.; Rho, B. H. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2007-02-15

    Fuel cladding material plays important role as a primary structure under the high temperature, high pressure and neutron environment of nuclear power plant. According to this environment, cladding material can be experienced several type aging phenomena including the neutron irradiation embrittlement. On the other hand, although the early nuclear power plant was designed to fitting into the 40MWd/KgU burn-up, the currently power plant intends to go to the high burn-up range. In this case, the safety criteria which was established at low burn-up needs to conform the applicability at the high burn-up. In this study, the safety criteria of fuel cladding material was reviewed to assess the cladding material integrity, and the material characteristics of cladding were reviewed. The current LOCA criterial was also reviewed, and the basic study for re-establishment of LOCA criteria was performed. The time concept safety criteria was also discussed to prevent the breakaway oxidation. Through the this study, safety issues will be produced and be helpful for integrity insurance of nuclear fuel cladding material. This report is 2nd term report.

  18. Study on the Standard Establishment for the Integrity Assessment of Nuclear Fuel Cladding Materials

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S-S; Kim, S-H; Jung, Y-K; Yang, C-Y; Kim, I-G; Choi, Y-H; Kim, H-J; Kim, M-W; Rho, B-H [KINS, Daejeon (Korea, Republic of)

    2008-02-15

    Fuel cladding material plays important role as a primary structure under the high temperature, high pressure and neutron environment of nuclear power plant. According to this environment, cladding material can be experienced several type aging phenomena including the neutron irradiation embrittlement. On the other hand, although the early nuclear power plant was designed to fitting into the 40MWd/KgU burn-up, the currently power plant intends to go to the high burn-up range. In this case, the safety criteria which was established at low burn-up needs to conform the applicability at the high burn-up. In this study, the safety criteria of fuel cladding material was reviewed to assess the cladding material integrity, and the material characteristics of cladding were reviewed. The current LOCA criterial was also reviewed, and the basic study for re-establishment of LOCA criteria was performed. The time concept safety criteria was also discussed to prevent the breakaway oxidation. Through the this study, safety issues will be produced and be helpful for integrity insurance of nuclear fuel cladding material. This report is the final report.

  19. Phase Formation and Transformations in Transmutation Fuel Materials for the LIFE Engine Part I - Path Forward

    Energy Technology Data Exchange (ETDEWEB)

    Turchi, P E; Kaufman, L; Fluss, M J

    2008-11-10

    The current specifications of the LLNL fusion-fission hybrid proposal, namely LIFE, impose severe constraints on materials, and in particular on the nuclear fissile or fertile nuclear fuel and its immediate environment. This constitutes the focus of the present report with special emphasis on phase formation and phase transformations of the transmutation fuel and their consequences on particle and pebble thermal, chemical and mechanical integrities. We first review the work that has been done in recent years to improve materials properties under the Gen-IV project, and with in particular applications to HTGR and MSR, and also under GNEP and AFCI in the USA. Our goal is to assess the nuclear fuel options that currently exist together with their issues. Among the options, it is worth mentioning TRISO, IMF, and molten salts. The later option will not be discussed in details since an entire report is dedicated to it. Then, in a second part, with the specific LIFE specifications in mind, the various fuel options with their most critical issues are revisited with a path forward for each of them in terms of research, both experimental and theoretical. Since LIFE is applicable to very high burn-up of various fuels, distinctions will be made depending on the mission, i.e., energy production or incineration. Finally a few conclusions are drawn in terms of the specific needs for integrated materials modeling and the in depth knowledge on time-evolution thermochemistry that controls and drastically affects the performance of the nuclear materials and their immediate environment. Although LIFE demands materials that very likely have not yet been fully optimized, the challenge are not insurmountable and a well concerted experimental-modeling effort should lead to dramatic advances that should well serve other fission programs such as Gen-IV, GNEP, AFCI as well as the international fusion program, ITER.

  20. Lead-Cooled Fast Reactor Systems and the Fuels and Materials Challenges

    Directory of Open Access Journals (Sweden)

    T. R. Allen

    2007-01-01

    Full Text Available Anticipated developments in the consumer energy market have led developers of nuclear energy concepts to consider how innovations in energy technology can be adapted to meet consumer needs. Properties of molten lead or lead-bismuth alloy coolants in lead-cooled fast reactor (LFR systems offer potential advantages for reactors with passive safety characteristics, modular deployment, and fuel cycle flexibility. In addition to realizing those engineering objectives, the feasibility of such systems will rest on development or selection of fuels and materials suitable for use with corrosive lead or lead-bismuth. Three proposed LFR systems, with varying levels of concept maturity, are described to illustrate their associated fuels and materials challenges. Nitride fuels are generally favored for LFR use over metal or oxide fuels due to their compatibility with molten lead and lead-bismuth, in addition to their high atomic density and thermal conductivity. Ferritic/martensitic stainless steels, perhaps with silicon and/or oxide-dispersion additions for enhanced coolant compatibility and improved high-temperature strength, might prove sufficient for low-to-moderate-temperature LFRs, but it appears that ceramics or refractory metal alloys will be necessary for higher-temperature LFR systems intended for production of hydrogen energy carriers.

  1. Material Flow Analysis of Fossil Fuels in China during 2000–2010

    Directory of Open Access Journals (Sweden)

    Sheng Wang

    2012-01-01

    Full Text Available Since the relationship between the supply and demand of fossil fuels is on edge in the long run, the contradiction between the economic growth and limited resources will hinder the sustainable development of the Chinese society. This paper aims to analyze the input of fossil fuels in China during 2000–2010 via the material flow analysis (MFA that takes hidden flows into account. With coal, oil, and natural gas quantified by MFA, three indexes, consumption and supply ratio (C/S ratio, resource consumption intensity (RCI, and fossil fuels productivity (FFP, are proposed to reflect the interactions between population, GDP, and fossil fuels. The results indicated that in the past 11 years, China’s requirement for fossil fuels has been increasing continuously because of the growing mine productivity in domestic areas, which also leads to a single energy consumption structure as well as excessive dependence on the domestic exploitation. It is advisable to control the fossil fuels consumption by energy recycling and new energy facilities’ popularization in order to lead a sustainable access to nonrenewable resources and decrease the soaring carbon emissions.

  2. Material Flow Analysis of Fossil Fuels in China during 2000–2010

    Science.gov (United States)

    Wang, Sheng; Dai, Jing; Su, Meirong

    2012-01-01

    Since the relationship between the supply and demand of fossil fuels is on edge in the long run, the contradiction between the economic growth and limited resources will hinder the sustainable development of the Chinese society. This paper aims to analyze the input of fossil fuels in China during 2000–2010 via the material flow analysis (MFA) that takes hidden flows into account. With coal, oil, and natural gas quantified by MFA, three indexes, consumption and supply ratio (C/S ratio), resource consumption intensity (RCI), and fossil fuels productivity (FFP), are proposed to reflect the interactions between population, GDP, and fossil fuels. The results indicated that in the past 11 years, China's requirement for fossil fuels has been increasing continuously because of the growing mine productivity in domestic areas, which also leads to a single energy consumption structure as well as excessive dependence on the domestic exploitation. It is advisable to control the fossil fuels consumption by energy recycling and new energy facilities' popularization in order to lead a sustainable access to nonrenewable resources and decrease the soaring carbon emissions. PMID:23365525

  3. Material flow analysis of fossil fuels in China during 2000-2010.

    Science.gov (United States)

    Wang, Sheng; Dai, Jing; Su, Meirong

    2012-01-01

    Since the relationship between the supply and demand of fossil fuels is on edge in the long run, the contradiction between the economic growth and limited resources will hinder the sustainable development of the Chinese society. This paper aims to analyze the input of fossil fuels in China during 2000-2010 via the material flow analysis (MFA) that takes hidden flows into account. With coal, oil, and natural gas quantified by MFA, three indexes, consumption and supply ratio (C/S ratio), resource consumption intensity (RCI), and fossil fuels productivity (FFP), are proposed to reflect the interactions between population, GDP, and fossil fuels. The results indicated that in the past 11 years, China's requirement for fossil fuels has been increasing continuously because of the growing mine productivity in domestic areas, which also leads to a single energy consumption structure as well as excessive dependence on the domestic exploitation. It is advisable to control the fossil fuels consumption by energy recycling and new energy facilities' popularization in order to lead a sustainable access to nonrenewable resources and decrease the soaring carbon emissions.

  4. Risk associated with the transport of radioactive materials in the fuel cycle

    International Nuclear Information System (INIS)

    Lange, F.; Mairs, J.; Niel, C.

    1997-01-01

    This paper sets out the regulatory framework within which nuclear fuel cycle materials are transported. It establishes the basic principles of those safety regulations and explains the graded approach to satisfying those requirements depending on the hazard of the radioactive contents. The paper outlines the minimum performance standards required by the Regulations. It covers the performance standards for Type C packages in a little more detail because these are new to the 1996 Edition of the IAEA's Regulations for the Safe Transport of Radioactive Material and are less well reported elsewhere at present. The paper then gives approximate data on the number of shipments of radioactive materials that service the nuclear fuel cycles in France, Germany and the UK. The quantities are expressed as average annual quantities per GW el installed capacity. There is also a short discussion of the general performance standards required of Type B packages in comparison with tests that have simulated specific accident conditions involving particular packages. There follows a discussion on the probability of packages experiencing accident conditions that are comparable with the tests that Type B packages are required to withstand. Finally there is a summary of the implementation of the Regulations for sea and air transport and a description of ongoing work that may have a bearing on the future development of mode related Regulations. Nuclear fuel cycle materials are transported in accordance with strict and internationally agreed safety regulations which are the result of a permanent and progressive process based on social concern and on the advancement of knowledge provided by research and development. Transport operations take place in the public domain and some become high profile events in the management of these materials, attracting a lot of public, political and media attention. The risks associated with the transport of radioactive materials are low and it is important

  5. Backfilling techniques and materials in underground excavations: Potential alternative backfill materials in use in Posiva's spent fuel repository concept

    International Nuclear Information System (INIS)

    Dixon, D.A.; Keto, P.

    2009-05-01

    A variety of geologic media options have been proposed as being suitable for safely and permanently disposing of spent nuclear fuel or fuel reprocessing wastes. In Finland the concept selected is construction of a deep repository in crystalline rock (Posiva 1999, 2006; SKB 1999), likely at the Olkiluoto site (Posiva 2006). Should that site prove suitable, excavation of tunnels and several vertical shafts will be necessary. These excavations will need to be backfilled and sealed as emplacement operations are completed and eventually all of the openings will need to be backfilled and sealed. Clay-based materials were selected after extensive review of materials options and the potential for practical implementation in a repository and work over a 30+ year period has led to the development of a number of workable clay-based backfilling options, although discussion persists as to the most suitable clay materials and placement technologies to use. As part of the continuous process of re-evaluating backfilling options in order to provide the best options possible, placement methods and materials that have been given less attention have been revisited. Primary among options that were and continue to be evaluated as a potential backfill are cementitious materials. These materials were included in the list of candidate materials initially screened in the late 1970's for use in repository backfilling. Conventional cement-based materials were quickly identified as having some serious technical limitations with respect their ability to fulfil the identified requirements of backfill. Concerns related to their ability to achieve the performance criteria defined for backfill resulted in their exclusion from large-scale use as backfill in a repository. Development of new, less chemically aggressive cementitious materials and installation technologies has resulted in their re-evaluation. Concrete and cementitious materials have and are being developed that have chemical, durability

  6. Evaluation of improved materials for stationary diesel engines operating on residual and coal based fuels. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    Experimental results to date from an on-going research program on improved materials for stationary diesel engines using residual or coal-based fuels are presented with little discussion of conclusions about these results. Information is included on ring and liner wear, fuel oil qualities, ceramic materials, coatings, test procedures and equipment, and tribology test results. (LCL)

  7. Permission for the alteration in fabrication business of nuclear fuel materials in Tokai Works of Japan Nuclear Fuel Conversion Company, Ltd. (report)

    International Nuclear Information System (INIS)

    1984-01-01

    A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry, concerning the permission of the alteration in the fabrication business of nuclear fuel materials in the Tokai Works Japan Nuclear Fuel Conversion Co. was presented; The technical capability of Japan Nuclear Fuel Conversion Co. and the safety after the alteration were confirmed by the NSC. The examination made by the Committee on Examination of Nuclear Fuel Safety of the NSC for the safety after the alteration is described: the aseismatic design and the countermeasures to fire in the conversion testing building, criticality management, waste treatment, radiation control, the exposure dose of general people. (Mori, K.)

  8. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumntation, and measurement techniques in fuel fabrication facilities, P.O.1236909. Final report

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-12-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. Some of the material included has appeared elswhere and it has been summarized. An extensive bibliography is included. A spcific example of application of the accountability methods to a model fuel fabrication facility which is based on the Westinghouse Anderson design

  9. Improvements in materials reliability in the back end of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    1993-02-01

    Many of the problems that occur in the day-to-day operation of nuclear fuel cycle facilities can be traced to the degradation of the physical and chemical properties of the materials involved. In order to provide an international forum for the scientists and engineers working in this field, the IAEA convened a Technical Committee Meeting (TCM) on ''Materials Reliability in the Back End of the Nuclear Fuel Cycle'' in September 1986. The proceedings were subsequently published as IAEA-TECDOC-421. One of the major conclusions of the meeting was to hold a second TCM in 1989 to update progress in the field. This meeting was held in Vienna from 28 November to 1 December 1989, with the participation of 23 scientists and engineers from 13 countries who presented the 14 papers contained in this Technical Document. The papers cover the resistance of materials to corrosion in chemicals and/or under irradiation in three important segments of the back-end of the nuclear fuel cycle: during the interim storage of spent fuel, in reprocessing plants and during the final disposal of high level waste. In addition, a panel discussion was held on each of these three topics. Refs, figs and tabs

  10. Social-economic processes of fuel and raw materials region: diversification, mathematical modeling and regularities

    Science.gov (United States)

    Novichikhin, A. V.; Fryanov, V. N.; Petrova, T. V.; Pavlova, L. D.; Temlyantsev, M. V.

    2017-09-01

    The paper proposes the structure of production diversification in the fuel and raw material region. A distinctive feature of diversification is the provision of regional reproduction of natural resources - reserves of mineral deposits, as well as fuel and energy deposits - traditionally consumed energy products. The algorithm for determining the regularities of socio-economic processes in the region is developed and the results of its concretization are presented. The difference of the proposed algorithm is the joint accounting of quantitative and qualitative information, the emergence properties of social and economic systems under favorable and the following problem situations: crisis, resource scarcity, uncertainty of the prediction of changes in external conditions, changes in legislative bases. It is suggested to implement identification of possible negative processes at the initial stage and when eliminating them step by step: the formation of a hypothesis about the existence of regularity, cognitive modeling, formulation of statements. Based on the results of cognitive modeling of development scenarios for fuel and raw material regions, a set of statements that form the methodological basis for correcting hypotheses about regularities is justified. Four regularities of social and economic processes of the fuel and raw materials region are revealed.

  11. Modelling of Physical, Chemical, and Material Properties of Solid Oxide Fuel Cells

    Directory of Open Access Journals (Sweden)

    Jakub Kupecki

    2015-01-01

    Full Text Available This paper provides a review of modelling techniques applicable for system-level studies to account for physical, chemical, and material properties of solid oxide fuel cells. Functionality of 0D to 3D models is discussed and selected examples are given. Author provides information on typical length scales in evaluation of power systems with solid oxide fuel cells. In each section, proper examples of previous studies done in the field of 0D–3D modelling are recalled and discussed.

  12. Considerations of material transport and blockage formation in deformed fuel element clusters of LMFBRS

    International Nuclear Information System (INIS)

    Schultheiss, G.F.

    1984-01-01

    Local coolant blockages have been considered so far only for intact fuel element geometry. Operational experience of subassembly behaviour in sodium cooled fast breeder reactors however show that deformation of bundles have to be included in further considerations. Material sources for particulate and blockage forming matter as well as the transport behaviour of granulate are discussed in this report. In principle no safety relevant worse consequences are expected in the case of deformed fuel rods and spacers compared to the intact geometry with respect to formation and growth of local blockages in the high power core region. (orig.) [de

  13. Fuel, structural material and coolant for an advanced fast micro-reactor

    International Nuclear Information System (INIS)

    Nascimento, Jamil A. do; Guimaraes, Lamartine N.F.; Ono, Shizuca

    2011-01-01

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials. (author)

  14. Irradiation experiments on materials for core internals, pressure vessel and fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Tsukada, Takashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Materials degradation due to the aging phenomena is one of the key issues for the life assessment and extension of the light water reactors (LWRs). This presentation introduces JAERI`s activities in the field of LWR material researches which utilize the research and testing reactors for irradiation experiments. The activities are including the material studies for the core internals, pressure vessel and fuel cladding. These materials are exposed to the neutron/gamma radiation and high temperature water environments so that it is worth reviewing their degradation phenomena as the continuum. Three topics are presented; For the core internal materials, the irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steels is the present major concern. At JAERI the effects of alloying elements on IASCC have been investigated through the post-irradiation stress corrosion cracking tests in high-temperature water. The radiation embrittlement of pressure vessel steels is still a significant issue for LWR safety, and at JAERI some factors affecting the embrittlement behavior such as a dose rate have been investigated. Waterside corrosion of Zircaloy fuel cladding is one of the limiting factors in fuel rod performance and an in-situ measurement of the corrosion rate in high-temperature water was performed in JMTR. To improve the reliability of experiments and to extent the applicability of experimental techniques, a mutual utilization of the technical achievements in those irradiation experiments is desired. (author)

  15. Measurement of properties of sealant materials for solid oxide fuel cell systems

    International Nuclear Information System (INIS)

    Boersma, R.J.; Sammes, N.M.; Zhang, Y.

    1998-01-01

    Thermal expansion of ceramic materials, ceramic cements and steels was studied, and their compatibility with materials used in the SOFC system was evaluated. A number of ceramic cements, a glass paste and a glass ceramic were examined to identify their potential as a sealing material in a tubular fuel cell system. Thermal expansion coefficients of these materials were compared for thermal matching with materials employed in the fuel cell construction, such as yttria stabilised with 8 mole% zirconia (8YSZ) and stainless steel. A reasonable match was found for Macor, a glass ceramic, with one of the steel samples studied and with 8YSZ. Candidate sealants were tested for bonding quality, which made it evident that the ceramic cements had to be discarded. Good bonds between Macor and one of the steel samples and Macor and 8YSZ were formed, resulting in a gas tight seal. Uncertainty remains regarding the long term effect of the interface reactions between steel and Macor. Exposure of the Macor and a steel-Macor-8YSZ assembly to the different fuel cell gas atmospheres revealed that the Macor reacts with hydrogen, the long term effect of which is also unknown. Copyright (1998) Australasian Ceramic Society

  16. Contribution to the study of nuclear fuel materials with a metallic uranium base

    International Nuclear Information System (INIS)

    Englander, M.

    1957-11-01

    In a power reactor destined to supply industrially recoverable thermal energy, the most economical source of heat still consists of natural metallic uranium. However, the nuclear fuel material, most often employed in the form of rods of 20 to 40 mm diameter, is subjected to a series of stresses which lead to irreversible distortions usually incompatible with the substructure of the reactor. As a result the fuel material must possess at the outset a certain number of qualities which must be determined. Investigations have therefore been carried out, first on the technological characters peculiar to each of the three allotropic phases of pure uranium metal, and on their interactions on the stabilisation of the material which consists of either cast uranium or uranium pile-treated in the γ phase. (author) [fr

  17. THE COMPLEX USE OF LOCAL TYPES OF FUEL IN THE POROUS CONSTRUCTION MATERIALS PRODUCTION

    Directory of Open Access Journals (Sweden)

    N. P. Voronova

    2014-01-01

    Full Text Available The article presents a comprehensive low-waste technology is the use of local fuels, which can be used in the technology of some porous building materials. Also provides new methods of preparation of porous building materials based on aggloporite using local fuels and waste energy on the basis of milled peat, fuel briquettes and wood chips allow to replace expensive imported components that comprise the raw mixtures (coal, anthracite.On the basis of mathematical modeling of cooling in reheat furnaces pusher drive developed a method of engineering calculation mode batch hardening in agglomeration. Submitted constructive solution for the development of the cooling charge with thermophysical rational justification cooling modes. A study of the temperature distribution within the charge depending on the different speeds of the belt sintering machine, and hence on the cooling time.The characteristics of the raw material deposits "Fanipol" and the optimal composition of the charge which includes loam, coal, milled peat. In industrial research obtained aggloporite this formulation has shown positive results in strength and density. Established that by decreasing the particle size of the fuel increases the redox potential of the combustion products, which reduces the height of the oxidizing zone and the speed of the sintering raw mix. These processes increase the productivity of sinter machine.Technology is implemented on the "Minsk factory of building materials". The tests analyzed production technology porous construction materials using milled peat with the addition of sawdust. The study results recommend further use of sapropel, which cost significantly lower raw material mixture of submissions and in their physical and mechanical properties much closer to the properties of milled peat.

  18. Order for execution of the law concerning regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1985-01-01

    This ordinance is stipulated under the law concerning the regulation of nuclear raw materials, nuclear fuel materials and reactors. The designation for refining and processing businesses under the law shall be obtained for each works or enterprise where these operations are to be practiced. Persons who intend to accept the designation shall file applications attaching business plans and the other documents specified by the ordinances of the Prime Minister's Office and other ministry orders. The permission for the installation of nuclear reactors under the law shall be received for each works or enterprise where reactors are to be set up. Persons who intend to get the permission shall file applications attaching the financing plans required for the installation of reactors and the other documents designated by the orders of the competent ministry. The permission concerning the reactors installed on foreign ships shall be obtained for each ship which is going to enter into the Japanese waters. Persons who ask for the permission shall file applications attaching the documents which explain the safety of reactor facilities and the other documents defined by the orders of the Ministry of Transportation. The designation for reprocessing business and the application for it are provided for, respectively. The usage of nuclear fuel materials, nuclear raw materials and internationally regulated goods is ruled in detail. (Kubozone, M.)

  19. Order for execution of the law concerning regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1981-01-01

    This ordinance is stipulated under the law concerning the regulation of nuclear raw materials, nuclear fuel materials and reactors. The designation for refining and processing businesses under the law shall be obtained for each works or enterprise where these operations are to be practiced. Persons who intend to accept the designation shall file applications attaching business plans and the other documents specified by the ordinances of the Prime Minister's Office and other ministry orders. The permission for the installation of nuclear reactors under the law shall be received for each works or enterprise where reactors are to be set up. Persons who intend to get the permission shall file applications attaching the financing plans required for the installation of reactors and the other documents designated by the orders of the competent ministry. The permission concerning the reactors installed on foreign ships shall be obtained for each ship which is going to enter into the Japanese waters. Persons who ask for the permission shall file applications attaching the documents which explain the safety of reactor facilities and the other documents defined by the orders of the Ministry of Transportation. The designation for reprocessing business and the application for it are provided for, respectively. The usage of nuclear fuel materials, nuclear raw materials and internationally regulated goods is ruled in detail.(Okada, K.)

  20. Method and device of producing spherical particles of nuclear fuel or nuclear source material

    International Nuclear Information System (INIS)

    Yamagishi, Shigeru; Takahashi, Yoshihisa; Shiba, Koreyuki.

    1984-01-01

    Purpose: To obtain nuclear fuel particles with satisfactory sphericalness. Method: In the crust-gelling process for producing nuclear fuel particles, the following three kinds of solution are used (1) organic solvent (carbon tetrachloride) as a medium for forming droplets and crust gelification, (2) starting gelifying solution containing nuclear fuel material (for example, a sol of solution mixture of thorium nitrate - uranyl nitrate adjusted to pH 3.0 by the addition of ammonia), (3) an aqueous ammonia solution for the completion of gelification and recovery of gelled spheres, with the density of the solutions being increased in the recited order. The apparatus for practicing the method comprises an intermediate portion connected to a starting liquid container and filled with an ammoniac organic solvent for to gelification, a gelification column for completion of gelification, gelled sphere-organic solvent separating portion and a gelified sphere recovery containeer. (Seki, T.)

  1. Urania-gadolinia nuclear fuel physical and irradiation characteristics and material properties

    International Nuclear Information System (INIS)

    Potts, G.A.

    1977-01-01

    The concept of the incorporation of burnable nuclear poisons in BWR fuel for supplementary reactivity control and power shaping has received extensive investigation. It has been determined that the most effective burnable poison system is one provided by a material having a very large neutron absorption cross section loaded in the nuclear fuel in concentrations that make the absorber highly self-shielding. Through the use of a burnable self-shielded absorber, the reactivity control can be approximately linearized and can theoretically provide a matching compensation to the reactivity loss rate due to U-235 depletion and fission product poisoning. The nuclear and material properties of the rare earth oxide Gd 2 O 3 have made it the most promising material compatible with the requirements and objectives of a burnable poison. The purpose of the document presented is to provide a summary of the information presently available on the material properties and characteristics of UO 2 to Gd 2 O 3 nuclear fuel

  2. Change in nuclear fuel material processing operation at Tokai Plant of Mitsubishi Atomic Fuel Co., Ltd. (report)

    International Nuclear Information System (INIS)

    1987-01-01

    This report, compiled by the Nuclear Safety Commission to be submitted to the Prime Minister, deals with studies on a proposed change in the operation of processing nuclear fuel substances at the Tokai Plant of Mitsubishi Atomic Fuel Co., Ltd. The conclusions of and principles for the examination and evaluation are described. It is concluded that part of the proposed change is appropriate with respect to required technical capability and that part of the change will not have adverse effects on the safety of the plant. The studies carried out are focused on the safety of the facilities. The study on the earthquake resistance reveals that anti-earthquake design for the new buildings is properly developed. The buildings are of fireproof construction and the systems and equipment to be installed are made of incomustible materials to ensure the prevention of fire and explosion. It is confirmed that criticality control (for each unit and for the group of units) will be performed appropriately and that the waste (gaseous waste, liquid waste, solid waste) treatment systems are designed appropriately. A study is also made on the radiation control methods (working condition control, individual exposure control, surrounding environment control). In addition accident evaluation is carried out to confirm the safety of the residents around the plant. (Nogami, K.)

  3. Change in nuclear fuel material processing operation at Tokai Plant of Mitsubishi Atomic Fuel Co. , Ltd. (report)

    Energy Technology Data Exchange (ETDEWEB)

    1987-09-01

    This report, compiled by the Nuclear Safety Commission to be submitted to the Prime Minister, deals with studies on a proposed change in the operation of processing nuclear fuel substances at the Tokai Plant of Mitsubishi Atomic Fuel Co., Ltd. The conclusions of and principles for the examination and evaluation are described. It is concluded that part of the proposed change is appropriate with respect to required technical capability and that part of the change will not have adverse effects on the safety of the plant. The studies carried out are focused on the safety of the facilities. The study on the earthquake resistance reveals that anti-earthquake design for the new buildings is properly developed. The buildings are of fireproof construction and the systems and equipment to be installed are made of incomustible materials to ensure the prevention of fire and explosion. It is confirmed that criticality control (for each unit and for the group of units) will be performed appropriately and that the waste (gaseous waste, liquid waste, solid waste) treatment systems are designed appropriately. A study is also made on the radiation control methods (working condition control, individual exposure control, surrounding environment control). In addition accident evaluation is carried out to confirm the safety of the residents around the plant. (Nogami, K.).

  4. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    In recent years, engineering oriented work, rather than basic research and development (R&D), has led to significant progress in improving the economics of innovative fast reactors and associated fuel cycle facilities, while maintaining and even enhancing the safety features of these systems. Optimization of plant size and layout, more compact designs, reduction of the amount of plant materials and the building volumes, higher operating temperatures to attain higher generating efficiencies, improvement of load factor, extended core lifetimes, high fuel burnup, etc. are good examples of achievements to date that have improved the economics of fast neutron systems. The IAEA, through its Technical Working Group on Fast Reactors (TWG-FR) and Technical Working Group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), devotes many of its initiatives to encouraging technical cooperation and promoting common research and technology development projects among Member States with fast reactor and advanced fuel cycle development programmes, with the general aim of catalysing and accelerating technology advances in these fields. In particular the theme of fast reactor deployment, scenarios and economics has been largely debated during the recent IAEA International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios, held in Paris in March 2013. Several papers presented at this conference discussed the economics of fast reactors from different national and regional perspectives, including business cases, investment scenarios, funding mechanisms and design options that offer significant capital and energy production cost reductions. This Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics addresses Member States’ expressed need for information exchange in the field, with the aim of identifying the main open issues and launching possible initiatives to help and

  5. Simulations of the Thermodynamic and Diffusion Properties of Actinide Oxide Fuel Materials

    International Nuclear Information System (INIS)

    Becker, Udo

    2013-01-01

    Spent nuclear fuel from commercial reactors is comprised of 95-99 percent UO 2 and 1-5 percent fission products and transuranic elements. Certain actinides and fission products are of particular interest in terms of fuel stability, which affects reprocessing and waste materials. The transuranics found in spent nuclear fuels are Np, Pu, Am, and Cm, some of which have long half- lives (e.g., 2.1 million years for 237 Np). These actinides can be separated and recycled into new fuel matrices, thereby reducing the nuclear waste inventory. Oxides of these actinides are isostructural with UO 2 , and are expected to form solid solutions. This project will use computational techniques to conduct a comprehensive study on thermodynamic properties of actinide-oxide solid solutions. The goals of this project are to: Determine the temperature-dependent mixing properties of actinide-oxide fuels; Validate computational methods by comparing results with experimental results; Expand research scope to complex (ternary and quaternary) mixed actinide oxide fuels. After deriving phase diagrams and the stability of solid solutions as a function of temperature and pressure, the project team will determine whether potential phase separations or ordered phases can actually occur by studying diffusion of cations and the kinetics of potential phase separations or ordered phases. In addition, the team will investigate the diffusion of fission product gases that can also have a significant influence on fuel stability. Once the system has been established for binary solid solutions of Th, U, Np, and Pu oxides, the methodology can be quickly applied to new compositions that apply to ternaries and quaternaries, higher actinides (Am, Cm), burnable poisons (B, Gd, Hf), and fission products (Cs, Sr, Tc) to improve reactivity

  6. WaterTransport in PEM Fuel Cells: Advanced Modeling, Material Selection, Testing and Design Optimization

    Energy Technology Data Exchange (ETDEWEB)

    J. Vernon Cole; Abhra Roy; Ashok Damle; Hari Dahr; Sanjiv Kumar; Kunal Jain; Ned Djilai

    2012-10-02

    Water management in Proton Exchange Membrane, PEM, Fuel Cells is challenging because of the inherent conflicts between the requirements for efficient low and high power operation. Particularly at low powers, adequate water must be supplied to sufficiently humidify the membrane or protons will not move through it adequately and resistance losses will decrease the cell efficiency. At high power density operation, more water is produced at the cathode than is necessary for membrane hydration. This excess water must be removed effectively or it will accumulate in the Gas Diffusion Layers, GDLs, between the gas channels and catalysts, blocking diffusion paths for reactants to reach the catalysts and potentially flooding the electrode. As power density of the cells is increased, the challenges arising from water management are expected to become more difficult to overcome simply due to the increased rate of liquid water generation relative to fuel cell volume. Thus, effectively addressing water management based issues is a key challenge in successful application of PEMFC systems. In this project, CFDRC and our partners used a combination of experimental characterization, controlled experimental studies of important processes governing how water moves through the fuel cell materials, and detailed models and simulations to improve understanding of water management in operating hydrogen PEM fuel cells. The characterization studies provided key data that is used as inputs to all state-of-the-art models for commercially important GDL materials. Experimental studies and microscopic scale models of how water moves through the GDLs showed that the water follows preferential paths, not branching like a river, as it moves toward the surface of the material. Experimental studies and detailed models of water and airflow in fuel cells channels demonstrated that such models can be used as an effective design tool to reduce operating pressure drop in the channels and the associated

  7. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    International Nuclear Information System (INIS)

    Shropshire, D.E.; Herring, J.S.

    2004-01-01

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

  8. Energy Frontier Research Center, Center for Materials Science of Nuclear Fuels

    International Nuclear Information System (INIS)

    Allen, Todd R.

    2011-01-01

    The Office of Science, Basic Energy Sciences, has funded the INL as one of the Energy Frontier Research Centers in the area of material science of nuclear fuels. This document is the required annual report to the Office of Science that outlines the accomplishments for the period of May 2010 through April 2011. The aim of the Center for Material Science of Nuclear Fuels (CMSNF) is to establish the foundation for predictive understanding of the effects of irradiation-induced defects on thermal transport in oxide nuclear fuels. The science driver of the center's investigation is to understand how complex defect and microstructures affect phonon mediated thermal transport in UO2, and achieve this understanding for the particular case of irradiation-induced defects and microstructures. The center's research thus includes modeling and measurement of thermal transport in oxide fuels with different levels of impurities, lattice disorder and irradiation-induced microstructure, as well as theoretical and experimental investigation of the evolution of disorder, stoichiometry and microstructure in nuclear fuel under irradiation. With the premise that thermal transport in irradiated UO2 is a phonon-mediated energy transport process in a crystalline material with defects and microstructure, a step-by-step approach will be utilized to understand the effects of types of defects and microstructures on the collective phonon dynamics in irradiated UO2. Our efforts under the thermal transport thrust involved both measurement of diffusive phonon transport (an approach that integrates over the entire phonon spectrum) and spectroscopic measurements of phonon attenuation/lifetime and phonon dispersion. Our distinct experimental efforts dovetail with our modeling effort involving atomistic simulation of phonon transport and prediction of lattice thermal conductivity using the Boltzmann transport framework.

  9. Energy Frontier Research Center, Center for Materials Science of Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Todd R. Allen, Director

    2011-04-01

    The Office of Science, Basic Energy Sciences, has funded the INL as one of the Energy Frontier Research Centers in the area of material science of nuclear fuels. This document is the required annual report to the Office of Science that outlines the accomplishments for the period of May 2010 through April 2011. The aim of the Center for Material Science of Nuclear Fuels (CMSNF) is to establish the foundation for predictive understanding of the effects of irradiation-induced defects on thermal transport in oxide nuclear fuels. The science driver of the center’s investigation is to understand how complex defect and microstructures affect phonon mediated thermal transport in UO2, and achieve this understanding for the particular case of irradiation-induced defects and microstructures. The center’s research thus includes modeling and measurement of thermal transport in oxide fuels with different levels of impurities, lattice disorder and irradiation-induced microstructure, as well as theoretical and experimental investigation of the evolution of disorder, stoichiometry and microstructure in nuclear fuel under irradiation. With the premise that thermal transport in irradiated UO2 is a phonon-mediated energy transport process in a crystalline material with defects and microstructure, a step-by-step approach will be utilized to understand the effects of types of defects and microstructures on the collective phonon dynamics in irradiated UO2. Our efforts under the thermal transport thrust involved both measurement of diffusive phonon transport (an approach that integrates over the entire phonon spectrum) and spectroscopic measurements of phonon attenuation/lifetime and phonon dispersion. Our distinct experimental efforts dovetail with our modeling effort involving atomistic simulation of phonon transport and prediction of lattice thermal conductivity using the Boltzmann transport framework.

  10. A PEM fuel cell based on electrocatalyst and membrane materials modified by PANAM dendrimers

    Energy Technology Data Exchange (ETDEWEB)

    Ledesma-Garcia, J.; Chapman, T.W.; Godinez, L.A. [Centro de Investigacion y Desarrollo Tecnologico en Electroquimica, Queretaro (Mexico)

    2008-10-15

    Due to its high energy conversion efficiency and low emission of pollutants, fuel-cell technology has been generally recognized as a key twenty-first century energy source. For polymer electrolyte membrane fuel cells (PEMFC), it has been found that platinum and its alloys exhibit the best electrocatalytic activity for oxygen reduction. The highest electrocatalytic activity of platinum and its alloys can be achieved when the particles are produced in the nanometer range. In this context, organic molecules have been adopted as templates to control the size of metal nanoparticles. Dendrimers, in particular, have shown promising properties for this application, and strategies that include direct adsorption, electrostatic attachment and covalent bonding have been developed for connecting metal-bearing dendrimers to conducting substrates. This paper reported on the preliminary results of a study that involved the construction and testing of a hydrogen-oxygen PEM fuel cell based on carbon-fiber-paper electrodes coated with hydroxyl-terminated dendrimers that encapsulated nanoparticles of platinum. This prototype cell also employed an ion exchange membrane comprising a cellulose acetate filter functionalized with proton-exchanging dendrimers. A proton-exchange membrane was prepared by binding duplex amine-carboxylate dendrimers to a cellulose-acetate support. With these dendrimer-based materials, a hydrogen-oxygen fuel cell was assembled and the performance compared with cells prepared with Nafion-based membranes. The voltage-current profiles and the power-density curves from the new cell provide encouragement to continue work with these dendrimer-modified materials. The paper discussed the experimental methods, with particular reference to materials; electrode preparation and characterization; proton-exchange membrane preparation; and PEM fuel-cell assembly and testing. It was concluded that the use of the dendritic macromolecules as supports for the nanoparticulate

  11. Perovskites for energy applications. From cathode material for fuel cells to a gas separation membrane

    Energy Technology Data Exchange (ETDEWEB)

    Meulenberg, W.A.; Baumann, S.; Betz, M.; Buchkremer, H.P.; Stoever, D. [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energieforschung (IEF); Serra, J.M.; Vert, V.B. [Universidad Politecnica de Valencia (Spain). Inst. de Tecnologia Quimica

    2010-07-01

    Oxyfuel power plants are one possibility for Carbon Capture and Storage (CCS) using pure oxygen instead of air to combust a carbon containing fuel. This oxygen can be produced by ceramic membranes, which consist of a Mixed Ionic Electronic Conductor (MIEC). Appropriate materials for oxygen separation from air are perovskites transporting oxygen ions through oxygen vacancies in the crystal lattice. Perovskites show highest permeability in particular Ba{sub 0.5}Sr{sub 0.5}Co{sub 0.8}Fe{sub 0.2}O{sub 3-{delta}} (BSCF) and offer a theoretical selectivity of 100%. However, perovskites with high permeability show in principle poor chemical stability e.g. in atmosphere containing CO{sub 2}, SO{sub 2}, or H{sub 2}O and particularly reducing conditions. Moreover the thermal and chemical expansion coefficient is very high, which makes the manufacturing of a gas-tight thin film on or joining to a material different from BSCF nearly impossible. Solid oxide fuel cells (SOFCs) are becoming promising candidates for highly efficient energy generation from conventional and biomass-derived fuels due to different reasons: (i) electricity can be obtained directly from a fuel; (ii) the sub-product is a high quality heat, usable in (micro) turbines and for building central heating (CHP) units; (iii) zero-emission operation is achieved when hydrogen is fuelled; (iv) SOFCs can operate besides H{sub 2} with hydrocarbons without extensive fuel purification and reforming; and (v) SOFCs are noiseless and modular. However, conventional SOFCs need to operate in the 800-1000 C temperature range. The reduction of the operating temperature below 700 C implies that the electrode polarization resistance of classical cathodes limits the whole cell operation, and consequently the performance is significantly reduced. Therefore, it is needed the development of new cathode materials with sufficient chemical stability and electrochemical activity to enable the operation at lower temperatures with

  12. Evaluation of Ni/SDC as anode material for dry CH4 fueled Solid Oxide Fuel Cells

    Science.gov (United States)

    Wang, Zhiming; Li, Yongdan; Schwank, Johannes W.

    2014-02-01

    A Ni/Sm0.2Ce0.8O1.9 (SDC) composite was employed as anode material for direct electrochemical oxidation (DEO) of dry CH4 in a solid oxide fuel cell. The anodic performance was investigated at temperatures between 600 °C and 700 °C using SDC as electrolyte material and Ba0.5Sr0.5Co0.8Fe0.2O3-δ (BSCF) as cathode material. The single cell exhibited maximum power densities of 671 mW cm-2, 494 mW cm-2 and 305 mW cm-2 in dry CH4 at 700 °C, 650 °C and 600 °C, respectively. Remarkably, at 700 °C the power density in CH4 was higher than in H2, thanks to the carbon tolerance of the anode. Durability tests under constant 300 mA output current showed only 3.7% performance loss after 72 h operation. The results demonstrate that Ni/SDC can be used as anode for the DEO of dry CH4 even at temperatures as low as 600 °C.

  13. Storage tank materials for biodiesel blends; the analysis of fuel property changes

    Directory of Open Access Journals (Sweden)

    Nurul Komariah Leily

    2017-01-01

    Full Text Available Fuel stability is one of major problem in biodiesel application. Some of the physical properties of biodiesel are commonly changed during storage. The change in physico-chemical properties is strongly correlated to the stability of the fuel. This study is objected to observe the potential materials for biodiesel storage. The test was conducted in three kinds of tank materials, such as glass, HDPE, and stainless steel. The fuel properties are monitored in 12 weeks, while the sample was analyzed every week. Biodiesel used is palm oil based. The storage tanks were placed in a confined indoor space with range of temperature 27–34 °C. The relative humidity and sunshine duration on the location was also evaluated. The observed properties of the fuel blends were density, viscosity and water content. During 12 weeks of storage, the average density of B20 was changed very slightly in all tanks, while the viscosity was tend to increase sharply, especially in polimerics tank. Water content of B20 was increased by the increase of storage time especially in HDPE tank. In short period of storage, the biodiesel blends is found more stable in glass tank due to its versatility to prohibit oxidation, degradation, and its chemical resistance.

  14. Investigation of altenative carbon materials for fuel-cell catalyst support

    DEFF Research Database (Denmark)

    Larsen, Mikkel Juul

    samples. Generally, the measurements using this method suffered from poor electrode homogeneity due to the problem of dispersing the CNT and CNF material on the electrode disc. For the carbon samples the peroxide percentage was 25–75 % at 0.066 V vs. the dynamic hydrogen electrode (DHE), but no ORR......In order to ensure high utilization of the catalyst material in a polymer electrolyte membrane fuel cell (PEMFC) it is usually fixed in the form of nanoparticles on a supporting material. The catalyst is platinum or a platinum alloy, and the commonly used support is carbon black (CB). Although...... structured carbon forms such as graphitized CBs, carbon nanotubes (CNTs), and carbon nanofibres (CNFs). This thesis concerns the investigation of an array of different materials which may prospec-tively replace the conventional materials used in the catalyst. The study comprised 13 carbon samples which...

  15. Functional mesoporous materials for energy applications: solar cells, fuel cells, and batteries.

    Science.gov (United States)

    Ye, Youngjin; Jo, Changshin; Jeong, Inyoung; Lee, Jinwoo

    2013-06-07

    This feature article presents recent progress made in the synthesis of functional ordered mesoporous materials and their application as high performance electrodes in dye-sensitized solar cells (DSCs) and quantum dot-sensitized solar cells (QDSCs), fuel cells, and Li-ion batteries. Ordered mesoporous materials have been mainly synthesized using two representative synthetic methods: the soft template and hard template methods. To overcome the limitations of these two methods, a new method called CASH was suggested. The CASH method combines the advantages of the soft and hard template methods by employing a diblock copolymer, PI-b-PEO, which contains a hydrophilic block and an sp(2)-hybridized-carbon-containing hydrophobic block as a structure-directing agent. After discussing general techniques used in the synthesis of mesoporous materials, this article presents recent applications of mesoporous materials as electrodes in DSCs and QDSCs, fuel cells, and Li-ion batteries. The role of material properties and mesostructures in device performance is discussed in each case. The developed soft and hard template methods, along with the CASH method, allow control of the pore size, wall composition, and pore structure, providing insight into material design and optimization for better electrode performances in these types of energy conversion devices. This paper concludes with an outlook on future research directions to enable breakthroughs and overcome current limitations in this field.

  16. Novel Processing of Unique Ceramic-Based Nuclear Materials and Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hui Zhang; Raman P. Singh

    2008-11-30

    Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These include refractory alloys base on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as those based on silicon carbide (SiCf-SiC); carbon-carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor componets is necessary for improved efficiency. Improving thermal conductivity of the materials used in nuclear fuels and other temperature critical components can lower the center-line fuel temperature and thereby enhance durability and reduce the risk of premature failure.

  17. MATERIALS COMPATIBILITY OF SNAP FUEL COMPONENTS DURING SHIPMENT IN 9975 PACKAGING

    Energy Technology Data Exchange (ETDEWEB)

    Vormelker, P

    2006-11-14

    Materials Science and Technology has evaluated materials compatibility for the SNAP (Systems for Nuclear Auxiliary Power) fuel for containment within a 9975 packaging assembly for a shipping period of one year. The evaluation included consideration for potential for water within the convenience can, corrosion from water, galvanic corrosion, tape degradation, and thermal expansion risk. Based on a review of existing literature and assumed conditions, corrosion and/or degradation of the 304 stainless steel (SS) Primary Containment Vessel (PCV) and the 304 stainless steel convenience cans containing the SNAP fuel is not significant to cause failure during the 1 year time shipping period in the 9975 packaging assembly. However, storage beyond the 1 year shipping period has not been validated.

  18. Novel Processing of Unique Ceramic-Based Nuclear Materials and Fuels

    International Nuclear Information System (INIS)

    Zhang, Hui; Singh, Raman P.

    2008-01-01

    Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These include refractory alloys base on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as those based on silicon carbide (SiCf-SiC); carbon-carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor components is necessary for improved efficiency. Improving thermal conductivity of the materials used in nuclear fuels and other temperature critical components can lower the center-line fuel temperature and thereby enhance durability and reduce the risk of premature failure.

  19. Neutron Absorbing Ability Variation in Neutron Absorbing Material Caused by the Neutron Irradiation in Spent Fuel Storage Facility

    International Nuclear Information System (INIS)

    Sohn, Hee Dong; Han, Seul Gi; Lee, Sang Dong; Kim, Ki Hong; Ryu, Eag Hyang; Park, Hwa Gyu

    2014-01-01

    In spent fuel storage facility like high density spent fuel storage racks and dry storage casks, spent fuels are stored with neutron absorbing materials installed as a part of those facilities, and they are used for absorbing neutrons emitted from spent fuels. Usually structural material with neutron absorbing material of racks and casks are located around spent fuels, so it is irradiated by neutrons for long time. Neutron absorbing ability could be changed by the variation of nuclide composition in neutron absorbing material caused by the irradiation of neutrons. So, neutron absorbing materials are continuously faced with spent fuels with boric acid solution or inert gas environment. Major nuclides in neutron absorbing material are Al 27 , C 12 , B 11 , B 10 and they are changed to numerous other ones as radioactive decay or neutron absorption reaction. The B 10 content in neutron absorbing material dominates the neutron absorbing ability, so, the variation of nuclide composition including the decrease of B 10 content is the critical factor on neutron absorbing ability. In this study, neutron flux in spent fuel, the activation of neutron absorbing material and the variation of nuclide composition are calculated. And, the minimum neutron flux causing the decrease of B 10 content is calculated in spent fuel storage facility. Finally, the variation of neutron multiplication factor is identified according to the one of B 10 content in neutron absorbing material. The minimum neutron flux to impact the neutron absorbing ability is 10 10 order, however, usual neutron flux from spent fuel is 10 8 order. Therefore, even though neutron absorbing material is irradiated for over 40 years, B 10 content is little decreased, so, initial neutron absorbing ability could be kept continuously

  20. Effect of Fuel Structure Materials on Radiation Source Term in Reactor Core Meltdown

    International Nuclear Information System (INIS)

    Jeong, Hae Sun; Ha, Kwang Soon

    2014-01-01

    The fission product (Radiation Source) releases from the reactor core into the containment is obligatorily evaluated to guarantee the safety of Nuclear Power Plant (NPP) under the hypothetical accident involving a core meltdown. The initial core inventory is used as a starting point of all radiological consequences and effects on the subsequent results of accident assessment. Hence, a proper evaluation for the inventory can be regarded as one of the most important part over the entire procedure of accident analysis. The inventory of fission products is typically evaluated on the basis of the uranium material (e.g., UO2 and USi2) loaded in nuclear fuel assembly, except for the structure materials such as the end fittings, grids, and some kinds of springs. However, the structure materials are continually activated by the neutrons generated from the nuclear fission, and some nuclides of them (e.g., 14 C and 60 Co) can significantly influence on accident assessment. During the severe core accident, the structure components can be also melted with the melting points of temperature relatively lower than uranium material. A series of the calculation were performed by using ORIGEN-S module in SCALE 6.1 package code system. The total activity in each part of structure materials was specifically analyzed from these calculations. The fission product inventory is generally evaluated based on the uranium materials of fuel only, even though the structure components of the assembly are continually activated by the neutrons generated from the nuclear fission. In this study, the activation calculation of the fuel structure materials was performed for the initial source term assessment in the accident of reactor core meltdown. As a result, the lower end fitting and the upper plenum greatly contribute to the total activity except for the cladding material. The nuclides of 56 Mn, '5 1 Cr, 55 Fe, 58 Co, 54 Mn, and 60 Co are analyzed to mainly effect on the activity. This result

  1. Systematic approach to remote maintenance in the fuels and materials examination facility

    International Nuclear Information System (INIS)

    Frandsen, G.B.; Nash, C.R.; Divona, C.J.; May, R.F.

    1979-01-01

    The Fuels and Materials Examination Facility (FMEF) is systematically analyzed from a remote maintenance standpoint using functional analysis methods. From the analysis the remote maintainability of equipment is ascertained, required tooling lists are formed, and maintenance downtimes are established. These techniques identify deficiencies or inefficiencies in the early design stage where changes have a minimum impact on cost. Special tooling and fixture requirements are minimized by standardizing remote maintenance design features

  2. Coordinated safeguards for materials management in a mixed-oxide fuel facility

    Energy Technology Data Exchange (ETDEWEB)

    Shipley, J.P.; Cobb, D.D.; Dietz, R.J.; Evans, M.L.; Schelonka, E.P.; Smith, D.B.; Walton, R.B.

    1977-02-01

    A coordinated safeguards system is described for safeguarding strategic quantities of special nuclear materials in mixed-oxide recycle fuel fabrication facilities. The safeguards system is compatible with industrial process requirements and combines maximum effectiveness consistent with modest cost and minimal process interference. It is based on unit process accounting using a combination of conventional and state-of-the-art NDA measurement techniques. The effectiveness of the system against single and multiple thefts is evaluated using computer modeling and simulation techniques.

  3. Coordinated safeguards for materials management in a mixed-oxide fuel facility

    International Nuclear Information System (INIS)

    Shipley, J.P.; Cobb, D.D.; Dietz, R.J.; Evans, M.L.; Schelonka, E.P.; Smith, D.B.; Walton, R.B.

    1977-02-01

    A coordinated safeguards system is described for safeguarding strategic quantities of special nuclear materials in mixed-oxide recycle fuel fabrication facilities. The safeguards system is compatible with industrial process requirements and combines maximum effectiveness consistent with modest cost and minimal process interference. It is based on unit process accounting using a combination of conventional and state-of-the-art NDA measurement techniques. The effectiveness of the system against single and multiple thefts is evaluated using computer modeling and simulation techniques

  4. Structural analysis of hatch cover plates on Fuels and Materials Examination Facility high bay mezzanine

    International Nuclear Information System (INIS)

    Dixson, G.E.

    1997-01-01

    In order to move the Idaho National Engineering Laboratory (INEL) Light Duty Utility Arm (LDUA) trailer into position for testing on the Fuels and Materials Examination Facility (FMEF) 42 ft level mezzanine one of the trailer's wheels will have to sit on a circular hatch cover fabricated from one-inch thick steel plate. The attached calculations verify that the hatch cover plate is strong enough to support the weight of the INEL LDUA trailer's wheel

  5. Order for execution of the law concerning regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1987-01-01

    Chapeter 1 specifies regulations concerning business management for refining and processing, which cover application for designation of refining operation, application for permission for processing operation, and approval of personnel responsible for handling nuclear fuel. Chapter 2 specifies regulations concerning construction and operation of nuclear reactors, which cover application for construction of nuclear reactors, reactors in a research and development stage, application for permission concerning nuclear reactors mounted on foreign nuclear powered ships, application for permission for alteration concerning construction of nuclear reactors, application for permission for alteration concerning nuclear reactors mounted on foreign nuclear powered ships, nuclear reactor facilities to be subjected to regular inspection, nuclear reactor for which submission of operation plan is not required, and application for permission for transfer of nuclear reactor. Chapter 2 also specifies regulations concerning business management for reprocessing and waste disposal. Chapter 3 stipulates regulations concerning use of nuclear fuel substances, nuclear material substances and other substances covered by international regulations, which include rules for application for permission for use of nuclear fuel substances, etc. Supplementary provisions are provided in Chapter 4. (Nogami, K.)

  6. Reports of the 8th new type nuclear fuel materials studying meeting. Present status of the plutonium mixed oxide fuel application

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-05-01

    This was the reports of the 8th New Type Nuclear Fuel Materials Studying Meeting, as a circle of Yayoi Studying Group meeting held on March 17, 1997. This meeting was added to a subtitle of `Present status and problems of plutonium mixed oxide application`, which had 12 lectures. In this meeting, for the MOX fuels putting the most attention in the field of nuclear fuel development at present, many specialists introduced faithfully on present status and problems of its nuclear features, reactor core design, and application to light water reactor and fast reactor. And, following reports were executed: (A) On feature of plutonium and reactor core design; (1) nuclear feature of plutonium, (2) nuclear design of BWR, (3) nuclear design of PWR, (4) nuclear design of FBR, and (5) and (6) properties of the MOX fuel; (B) On application of plutonium to the light water reactor; (1) preparation of the MOX fuel for light water reactor, (2) radiation behavior and using result of the MOX fuel for BWR, and (3) radiation behavior and using result of the MOX fuel for PWR; and (C) On application of plutonium to the fast reactor; (1) fuel preparation, (2) radiation behavior, and (3) reprocessing of the fast reactor fuel. (G.K.)

  7. Synthesis of the Novel MAX Phases for the Future Nuclear Fuel Cladding and Structural Materials

    International Nuclear Information System (INIS)

    Chung, Seung Hyeok; Kim, Taehee; Lee, Taegyu; Ryu, H. J.

    2016-01-01

    With these properties, the MAX phases are expected to be used for the Accident Tolerant Fuel (ATF) cladding and oxidation/corrosion resistance materials. Especially, the MAX phase can be used for the Gen-IV, SFR and HTGR, component materials which have to possess the thermal and corrosion resistance. The zirconium has been used to the nuclear industry for fuel cladding because of the small thermal neutron cross-section. Zr-based MAX phase was discovered by group Lapauw et al. They observed the Zr 2 AlC and Zr 3 AlC 2 with the X-ray diffraction (XRD) patterns and backscattered electron detector. Fabrication of the Zr-containing MAX phase was investigated for nuclear fuel cladding and structural materials applications. A MAX phase with the Zr 3 AlC 2 structure was synthesized by spark plasma sintering of a powder mixture targeting (Zr 0.5 Cr 0.5 ) 4 AlC 3 . The formation of MAX phases was confirmed by XRD and EDS of sintered samples. In the future work, the electron probe micro analyzer (EPMA) and transmission electron microscopy (TEM) are required to certain analyze the elements composition and formation of the MAX phase

  8. The HTR modular power reactor system. Qualification of fuel elements and materials

    International Nuclear Information System (INIS)

    Heidenreich, U.; Breitling, H.; Nieder, R.; Ohly, W.; Mittenkuehler, A.; Ragoss, H.; Seehafer, H.J.; Wirtz, K.; Serafin, N.

    1989-01-01

    For further development of the HTR modular power reactor system (HTR-M-KW), the project activities for 'Qualification of fuel elements and materials' reported here cover the work for specifying the qualifications to be met by metallic and ceramic materials, taking into account the design-based requirements and the engineered safety requirements. The fission product retention data determined for the HTR modular reactor fuel elements could be better confirmed by evaluation of the experiments, and have been verified by various calculation methods for different operating conditions. The qualification of components was verified by strength analyses including a benchmark calculation for specified normal operation and emergencies; the results show a convenient behaviour of the components and their materials. In addition, a fuel element burnup measuring system was designed that applies Cs-137 gamma spectroscopy; its feasibility was checked by appropriate analyses, and qualification work is in progress. The installation of a prototype measurement system is the task for project No. 03 IAT 211. (orig.) [de

  9. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    International Nuclear Information System (INIS)

    Rebak, Raul B.

    2014-01-01

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  10. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, Raul B. [General Electric Global Research, Schnectady, NY (United States)

    2014-09-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  11. Fundamental studies for life prediction of materials used in spent fuel reprocessing plant

    International Nuclear Information System (INIS)

    Kiuchi, Kiyoshi; Hayashi, Masanori; Hayakawa, Hitoshi; Kikuchi, Masahiko

    1992-01-01

    Fundamental corrosion study and experimental technology development have been made with respect to the life prediction of device materials used in purex type spent fuel reprocessing. Solution chemistry and corrosiveness of nitric acid environments were examined minutely by means of ICP, IC, NO x gas analysis and electrochemical measurements. New testing methods for evaluating corrosion and environmental cracking in nitric acid environments were also developed. The dominant corrosion failure of each candidate material of austenitic stainless steels, zirconium and titanium alloys was examined quantitatively as functions of environmental and metallurgical controlling parameters. (author)

  12. Optimization of a fuel converter for the MERLIN materials testing facility

    International Nuclear Information System (INIS)

    Pouleur, Y.; Raedt, Ch. de; Malambu, E.; Minsart, G.; Vermunt, J.

    1998-01-01

    SCK-CEN is at the present time developing the MERLIN materials testing facility, to be placed in the pool at the BR2 high flux materials testing reactor. It aims at irradiating large amounts of steel samples that are subsequently to be analyzed in the framework of reactor vessel embrittlement research programmes. In order to fulfill the required fast neutron flux conditions, a converter made of highly enriched uranium fuel plates, has to be inserted between the reactor vessel and the samples. The converter will transform the BR2 thermal neutron outflow into the required fast neutron flux. This converter has to be optimised. (author)

  13. Novel Catalysts and Processing Technologies for Production of Aerospace Fuels from Non-Petroleum Raw Materials

    Science.gov (United States)

    Hepp, Aloysius F.; Kulis, Michael J.; Psarras, Peter C.; Ball, David W.; Timko, Michael T.; Wong, Hsi-Wu; Peck, Jay; Chianelli, Russell R.

    2014-01-01

    Transportation fuels production (including aerospace propellants) from non-traditional sources (gases, waste materials, and biomass) has been an active area of research and development for decades. Reducing terrestrial waste streams simultaneous with energy conversion, plentiful biomass, new low-cost methane sources, and/or extra-terrestrial resource harvesting and utilization present significant technological and business opportunities being realized by a new generation of visionary entrepreneurs. We examine several new approaches to catalyst fabrication and new processing technologies to enable utilization of these nontraditional raw materials. Two basic processing architectures are considered: a single-stage pyrolysis approach that seeks to basically re-cycle hydrocarbons with minimal net chemistry or a two-step paradigm that involves production of supply or synthesis gas (mainly carbon oxides and H2) followed by production of fuel(s) via Sabatier or methanation reactions and/or Fischer-Tröpsch synthesis. Optimizing the fraction of product stream relevant to targeted aerospace (and other transportation) fuels via modeling, catalyst fabrication and novel reactor design are described. Energy utilization is a concern for production of fuels for either terrestrial or space operations; renewable sources based on solar energy and/or energy efficient processes may be mission enabling. Another important issue is minimizing impurities in the product stream(s), especially those potentially posing risks to personnel or operations through (catalyst) poisoning or (equipment) damage. Technologies being developed to remove (and/or recycle) heteroatom impurities are briefly discussed as well as the development of chemically robust catalysts whose activities are not diminished during operation. The potential impacts on future missions by such new approaches as well as balance of system issues are addressed.

  14. In-Pile Testing and Instrumentation for Development of Generation-IV Fuels and Materials. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2013-12-01

    For many years, the increase in efficiency in the production of nuclear electricity has been an economic challenge in many countries which have developed this kind of energy. The increase in fuel burnup and fuel residence time leads to a reduction in the volume of fresh fuel loaded and spent fuel discharged, respectively. More demanding nuclear fuel cycle parameters are combined with a need to operate nuclear power plants with maximal availability and load factors, in load-follow mode and with longer fuel cycles. In meeting these requirements, fuel has to operate in a demanding environment of high radiation fields, high temperatures, high mechanical stresses and high coolant flow. Requirements of increased fuel reliability and minimal fuel failures also remain in force. Under such circumstances, continuous development of more radiation resistant fuel materials, especially advanced cladding, careful and incremental examinations, and improved understanding and modelling of high burnup fuel behaviour are required. Following a recommendation of the IAEA Technical Working Group on Fuel Performance and Technology, the Technical Meeting on In-pile Testing and Instrumentation for Development of Generation-IV Fuels and Materials was held in Halden, Norway, on 21-24 August 2012. The purpose of the meeting was to review the current status and the progress in methods and technologies used for the in-pile testing of nuclear fuel achieved since the previous IAEA meeting on In-core Instrumentation and Reactor Core Assessment, also held in Halden in 2007. Emphasis was placed on advanced techniques applied for the understanding of high burnup fuel behaviour of water cooled power reactors that represent the vast majority of the current nuclear reactor fleet. However, the meeting also included papers and discussion on testing techniques applied or developed specifically for new fuel and structural materials considered for Generation-IV systems. The meeting was attended by 43

  15. Current status of materials development of nuclear fuel cladding tubes for light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Duan, Zhengang, E-mail: duan_zg@imr.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Yang, Huilong [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Satoh, Yuhki [Institute for Materials Research, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Abe, Hiroaki, E-mail: abe.hiroaki@n.t.u-tokyo.ac.jp [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan)

    2017-05-15

    Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

  16. Bioavailability and biodegradation of weathered diesel fuel in aquifer material under denitrifying conditions

    International Nuclear Information System (INIS)

    Bregnard, T.P.A.; Hoehener, P.; Zeyer, J.

    1998-01-01

    During the in situ bioremediation of a diesel fuel-contaminated aquifer in Menziken, Switzerland, aquifer material containing weathered diesel fuel (WDF) and indigenous microorganisms was excavated. This material was used to identify factors limiting WDF biodegradation under denitrifying conditions. Incubations of this material for 360 to 390 d under denitrifying conditions resulted in degradation of 23% of the WDF with concomitant consumption of NO 3 - and production of inorganic carbon. The biodegradation of WDF and the rate of NO 3 - consumption was stimulated by agitation of the microcosms. Biodegradation was not stimulated by the addition of a biosurfactant (rhamnolipids) or a synthetic surfactant (Triton X-100) at concentrations above their critical micelle concentrations. The rhamnolipids were biodegraded preferentially to WDF, whereas Triton X-100 was not degraded. Both surfactants reduced the surface tension of the growth medium from 72 to <35 dynes/cm and enhanced the apparent aqueous solubility of the model hydrocarbon n-hexadecane by four orders of magnitude. Solvent-extracted WDF, added at a concentration equal to that already present in the aquifer material, was also biodegraded by the microcosms, but not at a higher rate than the WDF already present in the material. The results show that the denitrifying biodegradation of WDF is not necessarily limited by bioavailability but rather by the inherent recalcitrance of WDF

  17. Certification for copper concentration in reference material for fuel anhydro ethylic alcohol; Certificacao da concentracao de cobre em material de referencia para alcool etilico anidro combustivel (AEC)

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Lindomar Augusto dos; Rocha, Marcia Silva da; Mesko, Marcia Foster; Silva, Fagner Francisco da; Quaresma, Maria Cristina Baptista; Araujo, Thiago Oliveira [Instituto Nacional de Metrologia, Normalizacao e Qualidade Industrial (DIMCI/INMETRO), Duque de Caxias, RJ (Brazil). Diretoria de Metrologia Cientifica e Industrial], E-mail: lareis@inmetro.gov.br

    2009-07-01

    This work aiming to obtain the first certified reference material for fuel anhydro ethylic alcohol relative to the copper concentration, which has his maximum limit determined by the in force legislation providing traceability and reliability for the measurement results.

  18. Down Select Report of Chemical Hydrogen Storage Materials, Catalysts, and Spent Fuel Regeneration Processes

    Energy Technology Data Exchange (ETDEWEB)

    Ott, Kevin; Linehan, Sue; Lipiecki, Frank; Aardahl, Christopher L.

    2008-08-24

    The DOE Hydrogen Storage Program is focused on identifying and developing viable hydrogen storage systems for onboard vehicular applications. The program funds exploratory research directed at identifying new materials and concepts for storage of hydrogen having high gravimetric and volumetric capacities that have the potential to meet long term technical targets for onboard storage. Approaches currently being examined are reversible metal hydride storage materials, reversible hydrogen sorption systems, and chemical hydrogen storage systems. The latter approach concerns materials that release hydrogen in endothermic or exothermic chemical bond-breaking processes. To regenerate the spent fuels arising from hydrogen release from such materials, chemical processes must be employed. These chemical regeneration processes are envisioned to occur offboard the vehicle.

  19. Material accountancy measurement techniques in dry-powdered processing of nuclear spent fuels

    International Nuclear Information System (INIS)

    Wolf, S. F.

    1999-01-01

    The paper addresses the development of inductively coupled plasma-mass spectrometry (ICPMS), thermal ionization-mass spectrometry (TIMS), alpha-spectrometry, and gamma spectrometry techniques for in-line analysis of highly irradiated (18 to 64 GWD/T) PWR spent fuels in a dry-powdered processing cycle. The dry-powdered technique for direct elemental and isotopic accountancy assay measurements was implemented without the need for separation of the plutonium, uranium and fission product elements in the bulk powdered process. The analyses allow the determination of fuel burn-up based on the isotopic composition of neodymium and/or cesium. An objective of the program is to develop the ICPMS method for direct fissile nuclear materials accountancy in the dry-powdered processing of spent fuel. The ICPMS measurement system may be applied to the KAERI DUPIC (direct use of spent PWR fuel in CANDU reactors) experiment, and in a near-real-time mode for international safeguards verification and non-proliferation policy concerns

  20. Final Scientific Report, New Proton Conductive Composite Materials for PEM Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Lvov, Serguei

    2010-11-08

    This project covered one of the main challenges in present-day PEM fuel cell technology: to design a membrane capable of maintaining high conductivity and mechanical integrity when temperature is elevated and water vapor pressure is severely reduced. The DOE conductivity milestone of 0.1 S cm-1 at 120 degrees C and 50 % relative humidity (RH) for designed membranes addressed the target for the project. Our approach presumed to develop a composite membrane with hydrophilic proton-conductive inorganic material and the proton conductive polymeric matrix that is able to “bridge” the conduction paths in the membrane. The unique aspect of our approach was the use of highly functionalized inorganic additives to benefit from their water retention properties and high conductivity as well. A promising result turns out that highly hydrophilic phosphorsilicate gels added in Nafion matrix improved PEM fuel cell performance by over 50% compared with bare Nafion membrane at 120 degrees C and 50 % RH. This achievement realizes that the fuel cell operating pressure can be kept low, which would make the PEM fuel cell much more cost efficient and adaptable to practical operating conditions and facilitate its faster commercialization particularly in automotive and stationary applications.

  1. Proton exchange membrane materials for the advancement of direct methanol fuel-cell technology

    Science.gov (United States)

    Cornelius, Christopher J [Albuquerque, NM

    2006-04-04

    A new class of hybrid organic-inorganic materials, and methods of synthesis, that can be used as a proton exchange membrane in a direct methanol fuel cell. In contrast with Nafion.RTM. PEM materials, which have random sulfonation, the new class of materials have ordered sulfonation achieved through self-assembly of alternating polyimide segments of different molecular weights comprising, for example, highly sulfonated hydrophilic PDA-DASA polyimide segment alternating with an unsulfonated hydrophobic 6FDA-DAS polyimide segment. An inorganic phase, e.g., 0.5 5 wt % TEOS, can be incorporated in the sulfonated polyimide copolymer to further improve its properties. The new materials exhibit reduced swelling when exposed to water, increased thermal stability, and decreased O.sub.2 and H.sub.2 gas permeability, while retaining proton conductivities similar to Nafion.RTM.. These improved properties may allow direct methanol fuel cells to operate at higher temperatures and with higher efficiencies due to reduced methanol crossover.

  2. Order for execution of the law concerning regulation of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1977-01-01

    The designations according to Item 1, Article 3 and Item 1, Article 13 of the Law must be obtained for each factory or business place where refining and fabrication of nuclear material are to be performed. One who wants to obtain such designation should file an application attached with a business plan and other documents via the director of a regional bureau of international trade and industry having jurisdiction over such factory or business place. When nuclear material refiners and nuclear material fabricators wish to obtain the approval for change stipulated in Item 1, Article 6 and Item 1, Article 16 of the Law, they must file applications to the Prime Minister and the Minister of International Trade and Industry via said directors. Chief handlers of nuclear fuel materials shall be approved among those meeting the strict requirements. One who wishes to install reactors must obtain the approval for each factory or business place where the reactors are to be installed. The permission must be obtained for each nuclear ship entering Japanese waters. The reactors proper and several facilities are subject to periodic inspection. (Rikitake, Y.)

  3. Investigation and basic evaluation for ultra-high burnup fuel cladding material

    International Nuclear Information System (INIS)

    Ioka, Ikuo; Nagase, Fumihisa; Futakawa, Masatoshi; Kiuchi, Kiyoshi

    2001-03-01

    In ultra-high burnup of the power reactor, it is an essential problem to develop the cladding with excellent durability. First, development history and approach of the safety assessment of Zircaloy for the high burnup fuel were summarized in the report. Second, the basic evaluation and investigation were carried out on the material with high practicability in order to select the candidate materials for the ultra-high burnup fuel. In addition, the basic research on modification technology of the cladding surface was carried out from the viewpoint of the addition of safety margin as a cladding. From the development history of the zirconium alloy including the Zircaloy, it is hard to estimate the results of in-pile test from those of the conventional corrosion test (out-pile test). Therefore, the development of the new testing technology that can simulate the actual environment and the elucidation of the corrosion-controlling factor of the cladding are desired. In cases of RIA (Reactivity Initiated Accident) and LOCA (Loss of Coolant Accident), it seems that the loss of ductility in zirconium alloys under heavy irradiation and boiling of high temperature water restricts the extension of fuel burnup. From preliminary evaluation on the high corrosion-resistance materials (austenitic stainless steel, iron or nickel base superalloys, titanium alloy, niobium alloy, vanadium alloy and ferritic stainless steel), stabilized austenitic stainless steels with a capability of future improvement and high-purity niobium alloys with a expectation of the good corrosion resistance were selected as candidate materials of ultra-high burnup cladding. (author)

  4. Frost induced damages within porous materials - from concrete technology to fuel cells technique

    Science.gov (United States)

    Palecki, Susanne; Gorelkov, Stanislav; Wartmann, Jens; Heinzel, Angelika

    2017-12-01

    Porous media like concrete or layers of membrane electrode assemblies (MEA) within fuel cells are affected by a cyclic frost exposure due to different damage mechanisms which could lead to essential degradation of the material. In general, frost damages can only occur in case of a specific material moisture content. In fuel cells, residual water is generally available after shut down inside the membrane i.e. the gas diffusion layer (GDL). During subsequent freezing, this could cause various damage phenomena such as frost heaves and delamination effects of the membrane electrode assembly, which depends on the location of pore water and on the pore structure itself. Porous materials possess a pore structure that could range over several orders of magnitudes with different properties and freezing behaviour of the pore water. Latter can be divided into macroscopic, structured and pre-structured water, influenced by surface interactions. Therefore below 0 °C different water modifications can coexist in a wide temperature range, so that during frost exposure a high amount of unfrozen and moveable water inside the pore system is still available. This induces transport mechanisms and shrinkage effects. The physical basics are similar for porous media. While the freezing behaviour of concrete has been studied over decades of years, in order to enhance the durability, the know-how about the influence of a frost attack on fuel cell systems is not fully understood to date. On the basis of frost damage models for concrete structures, an approach to describe the impact of cyclic freezing and thawing on membrane electrode assemblies has been developed within this research work. Major aim is beyond a better understanding of the frost induced mechanisms, the standardization of a suitable test procedure for the assessment of different MEA materials under such kind of attack. Within this contribution first results will be introduced.

  5. Method for assessing the performance of a material control and accounting system at an operating nuclear fuel processing facility

    International Nuclear Information System (INIS)

    Ellwein, L.B.; Harris, L.; Altman, W.D.; Gramann, R.H.

    1981-01-01

    This paper discusses a method for assessing the performance of a material control and accounting (MCandA) system in an operating nuclear fuel processing facility. The performance criteria inherent in the assessment are 16 key goals established by NRC's 1978 Material Control and Material Accounting Task Force. 7 refs

  6. Bipolar plate materials in molten carbonate fuel cells. Final CRADA report.

    Energy Technology Data Exchange (ETDEWEB)

    Krumpelt, M.

    2004-06-01

    Advantages of implementation of power plants based on electrochemical reactions are successfully demonstrated in the USA and Japan. One of the msot promising types of fuel cells (FC) is a type of high temperature fuel cells. At present, thanks to the efforts of the leading countries that develop fuel cell technologies power plants on the basis of molten carbonate fuel cells (MCFC) and solid oxide fuel cells (SOFC) are really close to commercialization. One of the problems that are to be solved for practical implementation of MCFC and SOFC is a problem of corrosion of metal components of stacks that are assembled of a number of fuel cells. One of the major components of MCFC and SOFC stacks is a bipolar separator plate (BSP) that performs several functions - it is separation of reactant gas flows sealing of the joints between fuel cells, and current collection from the surface of electrodes. The goal of Task 1 of the project is to develop new cost-effective nickel coatings for the Russian 20X23H18 steel for an MCFC bipolar separator plate using technological processes usually implemented to apply corrosion stable coatings onto the metal parts for products in the defense. There was planned the research on production of nickel coatings using different methods, first of all the galvanic one and the explosion cladding one. As a result of the works, 0.4 x 712 x 1296 mm plates coated with nickel on one side were to be made and passed to ANL. A line of 4 galvanic baths 600 liters was to be built for the galvanic coating applications. The goal of Task 2 of the project is the development of a new material of an MCFC bipolar separator plate with an upgraded corrosion stability, and development of a technology to produce cold roll sheets of this material the sizes of which will be 0.8 x 712x 1296 mm. As a result of these works, a pilot batch of the rolled material in sheets 0.8 x 712 x 1296 mm in size is to be made (in accordance with the norms and standards of the Russian

  7. The Attractiveness Of Materials In Advanced Nuclear Fuel Cycles For Various Proliferation And Theft Scenarios

    International Nuclear Information System (INIS)

    Bathke, C.G.; Ebbinghaus, Bartley B.; Collins, Brian A.; Sleaford, Brad W.; Hase, Kevin R.; Robel, Martin; Wallace, R.K.; Bradley, Keith S.; Ireland, J.R.; Jarvinen, G.D.; Johnson, M.W.; Prichard, Andrew W.; Smith, Brian W.

    2012-01-01

    We must anticipate that the day is approaching when details of nuclear weapons design and fabrication will become common knowledge. On that day we must be particularly certain that all special nuclear materials (SNM) are adequately accounted for and protected and that we have a clear understanding of the utility of nuclear materials to potential adversaries. To this end, this paper examines the attractiveness of materials mixtures containing SNM and alternate nuclear materials associated with the plutonium-uranium reduction extraction (Purex), uranium extraction (UREX), coextraction (COEX), thorium extraction (THOREX), and PYROX (an electrochemical refining method) reprocessing schemes. This paper provides a set of figures of merit for evaluating material attractiveness that covers a broad range of proliferant state and subnational group capabilities. The primary conclusion of this paper is that all fissile material must be rigorously safeguarded to detect diversion by a state and must be provided the highest levels of physical protection to prevent theft by subnational groups; no 'silver bullet' fuel cycle has been found that will permit the relaxation of current international safeguards or national physical security protection levels. The work reported herein has been performed at the request of the U.S. Department of Energy (DOE) and is based on the calculation of 'attractiveness levels' that are expressed in terms consistent with, but normally reserved for, the nuclear materials in DOE nuclear facilities. The methodology and findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security is discussed.

  8. New innovative materials for advanced electrochemical applications in battery and fuel cell systems

    Science.gov (United States)

    Voß, S.; Kollmann, H.; Kollmann, W.

    The advanced material POLYMET is an innovative high tech polymer with a three-dimensional polymeric structure metallized with an enclosing coating of different kinds of metals or alloys. The result is a range of tailor-made, microporous structures on a designable scale. By varying the metals and alloys, it is possible to draw upon extremely diverse areas of applications such as battery systems, fuel cells, filters or efficient catalysts as well as air regeneration systems, e.g. in aerospace. The three-dimensional structure of metallized high tech woven or non-woven materials or foams causes a lot of advantages such as high conductivity, high corrosion resistance, flexibility or mechanical strength. Therefore, POLYMET is suitable for, e.g. current collectors or gas diffusion layers in energy storage systems. They supply an economic and environmental alternative material to improve functional electrochemical systems.

  9. TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

    International Nuclear Information System (INIS)

    Grover, S.B.

    2004-01-01

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

  10. A Review of Metallic Bipolar Plates for Proton Exchange Membrane Fuel Cells: Materials and Fabrication Methods

    Directory of Open Access Journals (Sweden)

    Shahram Karimi

    2012-01-01

    Full Text Available The proton exchange membrane fuel cell offers an exceptional potential for a clean, efficient, and reliable power source. The bipolar plate is a key component in this device, as it connects each cell electrically, supplies reactant gases to both anode and cathode, and removes reaction products from the cell. Bipolar plates have been fabricated primarily from high-density graphite, but in recent years, much attention has been paid to developing cost-effective and feasible alternative materials. Two different classes of materials have attracted attention: metals and composites. This paper offers a comprehensive review of the current research being carried out on metallic bipolar plates, covering materials and fabrication methods.

  11. Carbon superfine materials as a promising material for Gluconobacter oxydans based microbial fuel cells

    Science.gov (United States)

    Tenchurin, Timur K.; Reshetilov, Anatoly N.; Plekhanova, Yuliya V.; Tarasov, Sergey E.; Bykov, Aleksandr G.; Gutorov, Michail A.; Alferov, Sergey V.; Chvalun, Sergei N.; Orekhov, Anton S.; Shepelev, Alexey D.; Gotovtsev, Pavel M.; Vasilov, Raif G.

    2018-02-01

    We have investigated the properties of a several bioelectrodes based on the immobilization of Gluconobacter oxydans bacterial cells on carbon superfine materials (CFMs). We use three types of CFMs (as adopted by the working classification CFM 1-3). All bioelectrodes was formed by covering the surface of the CFM via suspension of bacteria in a chitosan gel. The properties of samples are evaluated by measuring the physiological state of the bacteria immobilized: (a) recording the intensity of cellular respiration, (b) for measuring the charge transport characteristics of electrode (bioelectrocatalysis), and (c) by measuring the electrode impedance. Measurements (b) and (c) are made on two and three-electrode circuits in the oxidation of ethanol in the presence of 2,6-dichlorophenol electron transport mediator. For CFMs 1 and 2 the electron transport by the oxidation of the substrate is not registered, while for CFM 3 the current generation occurs. The resistance of CFM 3 bioelectrode is below the resistance of CFMs 1 and 2 both before (39.6 kΩ/cm2 for CFM 3, 630 Ω/cm2 for CFM 2, and 1329 Ω/cm2 for CFM 1) and after the addition of the substrate (2.9 kΩ/cm2 for CFM 3, 45 kΩ/cm2 for CFM 2, and 58 kΩ/cm2 for CFM 1). The bioelectrode made of CFM 3 has a capacitance of 196 μF/cm2—greater than two orders of magnitude of the bioelectrode capacity of CFMs 1 and 2 (0.51 and 0.58 μF/cm2, respectively). It is important to further study the properties of the CFM class of materials, which are promising as the basis of mechanically flexible electrodes with controlled parameters.

  12. Oxygen transport in perovskite-type solid oxide fuel cell materials: insights from quantum mechanics.

    Science.gov (United States)

    Muñoz-García, Ana B; Ritzmann, Andrew M; Pavone, Michele; Keith, John A; Carter, Emily A

    2014-11-18

    CONSPECTUS: Global advances in industrialization are precipitating increasingly rapid consumption of fossil fuel resources and heightened levels of atmospheric CO2. World sustainability requires viable sources of renewable energy and its efficient use. First-principles quantum mechanics (QM) studies can help guide developments in energy technologies by characterizing complex material properties and predicting reaction mechanisms at the atomic scale. QM can provide unbiased, qualitative guidelines for experimentally tailoring materials for energy applications. This Account primarily reviews our recent QM studies of electrode materials for solid oxide fuel cells (SOFCs), a promising technology for clean, efficient power generation. SOFCs presently must operate at very high temperatures to allow transport of oxygen ions and electrons through solid-state electrolytes and electrodes. High temperatures, however, engender slow startup times and accelerate material degradation. SOFC technologies need cathode and anode materials that function well at lower temperatures, which have been realized with mixed ion-electron conductor (MIEC) materials. Unfortunately, the complexity of MIECs has inhibited the rational tailoring of improved SOFC materials. Here, we gather theoretically obtained insights into oxygen ion conductivity in two classes of perovskite-type materials for SOFC applications: the conventional La1-xSrxMO3 family (M = Cr, Mn, Fe, Co) and the new, promising class of Sr2Fe2-xMoxO6 materials. Using density functional theory + U (DFT+U) with U-J values obtained from ab initio theory, we have characterized the accompanying electronic structures for the two processes that govern ionic diffusion in these materials: (i) oxygen vacancy formation and (ii) vacancy-mediated oxygen migration. We show how the corresponding macroscopic oxygen diffusion coefficient can be accurately obtained in terms of microscopic quantities calculated with first-principles QM. We find that the

  13. Behavior or Nonmetallic Materials in Shale Oil Derived Jet Fuels and in High Aromatic and High Sulfur Petroleum Fuels

    Science.gov (United States)

    1978-07-01

    These developments, coupled with the expectation of deriving future aviation turbine fuels from non-petroleum sources such as oil shale and coal liquid...707, 36 flights -70 -34 KC-135, 8 hours -73 -40 e. Thermal Stability It is required that the fuel not form lacquer or tar type deposits or form...turbine fuels. The requirement influences the degree of finishing processing such as hydrotreating and filtration required to meet the specification. Engine

  14. Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Lippek, H.E.; Schuller, C.R.

    1979-03-01

    A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light most of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel.

  15. Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

    International Nuclear Information System (INIS)

    Lippek, H.E.; Schuller, C.R.

    1979-03-01

    A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light most of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel

  16. Management of radioisotope, radiation generator and fuel materials for independent administrative corporations of national university

    International Nuclear Information System (INIS)

    2003-03-01

    This report states the situation, problems and proposal of management of radioisotope, radiation generator and fuel materials by independent administrative corporations of national universities. Four proposals are stated as followings; 1) in order to improve management of radioisotope, radiation generator, fuel materials and X-ray in the universities, organization and definition of the control department in each university and accident measures have to be decided. The middle object and plan should be needed. An appropriate management for proceeding researches should be discussed by closer connection of universities in the country. 2) The budget for safety control has to be identified at distribution of budget of each national university corporations. The insurance method is needed to be discussed. 3) The department in the MEXT (Ministry of Education, Culture, Sports, Science and Technology) should be enriched to support researches and safety control of the staff and students. 4) The system, which carries out treatment and disposal of disuse materials and keeps them under the responsibility of the nation, is necessary. (S.Y.)

  17. A regulatory analysis on emergency preparedness for fuel cycle and other radioactive material licensees: Final report

    International Nuclear Information System (INIS)

    McGuire, S.A.

    1988-01-01

    The question this Regulatory Analysis sought to answer is: should the NRC impose additional emergency preparedness requirements on certain fuel cycle and other radioactive material licensees for dealing with accidents that might have offsite releases of radioactive material. To answer the question, we analyzed potential accidents for 15 types of fuel cycle and other radioactive material licensees. An appropriate plan would: (1) identify accidents for which protective actions should be taken by people offsite; (2) list the licensee's responsibilities for each type of accident, including notification of local authorities (fire and police generally); and (3) give sample messages for local authorities including protective action recommendations. This approach more closely follows the approach used for research reactors than for power reactors. The low potential offsite doses (acute fatalities and injuries not possible except possibly for UF 6 releases), the small areas where actions would be warranted, the small number of people involved, and the fact that the local police and fire departments would be doing essentially the same things they normally do, are all factors that tend to make a simple plan adequate. This report discusses the potentially hazardous accidents, and the likely effects of these accidents in terms of personnel danger

  18. Research Reactors for the Development of Materials and Fuels for Innovative Nuclear Energy Systems

    International Nuclear Information System (INIS)

    2017-01-01

    This publication presents an overview of research reactor capabilities and capacities in the development of fuels and materials for innovative nuclear reactors, such as GenIV reactors. The compendium provides comprehensive information on the potential for materials and fuel testing research of 30 research reactors, both operational and in development. This information includes their power levels, mode of operation, current status, availability and historical overview of their utilization. A summary of these capabilities and capacities is presented in the overview tables of section 6. Papers providing a technical description of the research reactors, including their specific features for utilization are collected as profiles on a CD-ROM and represent an integral part of this publication. The publication is intended to foster wider access to information on existing research reactors with capacity for advanced material testing research and thus ensure their increased utilization in this particular domain. It is expected that it can also serve as a supporting tool for the establishment of regional and international networking through research reactor coalitions and IAEA designated international centres based on research reactors.

  19. Fuel cell studies of perovskite-type materials for IT-SOFC

    Science.gov (United States)

    Peña-Martínez, J.; Marrero-López, D.; Ruiz-Morales, J. C.; Buergler, B. E.; Núñez, P.; Gauckler, L. J.

    The electrochemical performance of solid oxide fuel cells (SOFCs) based on perovskite-type materials (ABO 3) was investigated. La 0.9Sr 0.1Ga 0.8Mg 0.2O 3- δ (LSGM) ceramics were used as electrolyte and a composite containing La 0.8Sr 0.2MnO 3 (LSM) as cathode. Ba 0.5Sr 0.5Co 0.8Fe 0.2O 3- δ (BSCF) was also used as cathode and La 0.75Sr 0.25Cr 0.5Mn 0.5O 3- δ (LSCM) as anode materials. Furthermore, fluorite-type Sm 0.15Ce 0.85O 2- δ (SDC) material was used as buffer layer between the electrolyte and the anode to avoid possible interfacial reactions. The maximum power density value of BSCF/LSGM/LSCM with 1.5 mm thick electrolyte supported cell was 160 mW cm -2 at 1073 K, using moist H 2 diluted with N 2 as fuel and air as oxidant.

  20. Suitability of granular carbon as an anode material for sediment microbial fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Arends, Jan B.A.; Blondeel, Evelyne; Boon, Nico; Verstraete, Willy [Ghent Univ. (Belgium). Faculty of Bioscience Engineering; Tennison, Steve R. [Mast Carbon International Ltd., Basingstoke, Hampshire (United Kingdom)

    2012-08-15

    Purpose: Sediment microbial fuel cells (S-MFCs) are bio-electrochemical devices that are able to oxidize organic matter directly into harvestable electrical power. The flux of organic matter into the sediment is rather low; therefore, other researchers have introduced plants for a continuous supply of organic matter to the anode electrode. Until now only interconnected materials have been considered as anode materials in S-MFCs. Here, granular carbon materials were investigated for their suitability as an anode material in S-MFCs. Materials and methods: Laboratory microcosms with eight different electrode materials (granules, felts and cloths) were examined with controlled organic matter addition under brackish conditions. Current density, organic matter removal and microbial community composition were monitored using 16S rRNA gene PCR followed by denaturing gradient gel electrophoresis (DGGE). The main parameters investigated were the influence of the amount of electrode material applied to the sediment, the size of the granular material and the electrode configuration. Results and discussion: Felt material had an overall superior performance in terms of current density per amount of applied electrode material; felt and granular anode obtained similar current densities (approx. 50-60 mA m{sup -2}), but felt materials required 29 % less material to be applied. Yet, when growing plants, granular carbon is more suited because it is considered to restore, upon disturbance, the electrical connectivity within the anode compartment. Small granules (0.25-0.5 mm) gave the highest current density compared to larger granules (1-5 mm) of the same material. Granules with a rough surface had a better performance compared to smooth granules of the same size. The different granular materials lead to a selection of distinct microbial communities for each material, as shown by DGGE. Conclusions: Granular carbon is suitable as an anode material for S-MFCs. This opens the possibility

  1. Advanced in-situ characterisation of corrosion properties of LWR fuel cladding materials

    International Nuclear Information System (INIS)

    Arilahti, E.; Bojinov, M.; Beverskog, B.

    1999-01-01

    The trend towards higher fuel burnups imposes a demand for better corrosion and hydriding resistance of cladding materials. Development of new and improved cladding materials is a long process. There is a lack of fast and reliable in-situ techniques to investigate zirconium alloys in simulated or in-core LWR coolant conditions. This paper describes a Thin Layer Electrode (TLE) arrangement suitable for in-situ characterization of oxide films formed on fuel cladding materials. This arrangement enables us to carry out: Versatile Thin Layer Electrochemical measurements, including: (i) Thin Layer Electrochemical impedance Spectroscopic (TLEIS) measurements to characterize the oxidation kinetics and mechanisms of metals and the properties of their oxide films in aqueous environments. These measurements can also be performed in low conductivity electrolytes. (ii) Thin-Layer Wall-Jet (TLWJ) measurements, which give the possibility to detect soluble reaction products and to evaluate the influence of novel water chemistry additions on their release. Solid Contact measurements: (i) Contact Electric Resistance (CER) measurements to investigate the electronic properties of surface films on the basis of d.c. resistance measurements. (i) Contact Electric impedance (CEI) measurements to study the electronic properties of surface films using a.c. perturbation. All the above listed measurements can be performed using one single measurement device developed at VTT. This device can be conveniently inserted into an autoclave. Its geometry is currently being optimized in cooperation with the OECD Halden Reactor Project. In addition, the applicability of the device for in-core measurements has been investigated in a joint feasibility study performed by VTT and JRC Petten. Results of some autoclave studies of the effect of LiOH concentration on the stability of fuel cladding oxide films are presented in this paper. (author)

  2. Notes on the nuclear safety regulations concerning research reactors and use of nuclear fuel material

    International Nuclear Information System (INIS)

    Tanzawa, Tomio

    2014-01-01

    Nuclear regulation system of Japan has drastically reconsidered after the nuclear disaster caused by the reactor accident of Fukushima Daiichi Nuclear Power Plant. As a result a law was enacted to establish a nuclear regulation authority in June 27, 2012, and the laws relating the atomic power were largely revised. In this article, we show major revisions in the laws concerning research reactor facilities and those facilities which treat nuclear fuel materials. It should be noted that the law is only a formality and that the application of the law by persons enables to produces effectiveness. Persons hold the key to the safety. (J.P.N.)

  3. Idaho National Laboratory Materials and Fuels Complex Natural Phenomena Hazards Flood Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Gerald Sehlke; Paul Wichlacz

    2010-12-01

    This report presents the results of flood hazards analyses performed for the Materials and Fuels Complex (MFC) and the adjacent Transient Reactor Experiment and Test Facility (TREAT) located at Idaho National Laboratory. The requirements of these analyses are provided in the U.S. Department of Energy Order 420.1B and supporting Department of Energy (DOE) Natural Phenomenon Hazard standards. The flood hazards analyses were performed by Battelle Energy Alliance and Pacific Northwest National Laboratory. The analyses addressed the following: • Determination of the design basis flood (DBFL) • Evaluation of the DBFL versus the Critical Flood Elevations (CFEs) for critical existing structures, systems, and components (SSCs).

  4. Gas cooled fast reactor materials: compatibility and reaction kinetics of fuel/matrices couples

    International Nuclear Information System (INIS)

    Lechelle, J.; Aufore, L.; Basini, V.; Belin, R.; Vaudez, S.

    2004-01-01

    Fourth Generation Gas cooled Fast Reactor concept implies a fast neutron spectrum and aims to lead to an iso-generation of minor actinides. Criteria have been defined for these fuels such as: high core filling factor, efficient fuel cooling, low operation temperature, i.e. 400-850 deg C, good fission product retention, burn-ups in the range of 5-8 atom%, Pu content in the range of 15-25%. Materials matching this demand are considered: mixed uranium - plutonium nitrides and carbides as fuels, whereas TiN, TiC, ZrN, ZrC, SiC are investigated as inert matrices. Thermo-chemical compatibility studies have been carried out, mostly for (U,Pu)N/SiC and (U,Pu)N/TiN couples. They have been associated to matching diffusional studies. For the first studies, accidental reactor conditions have been chosen (1600 deg C) so as to select a couple. Results are presented in terms of nature and quantity of resulting phases identified by XRD and SEM for thermodynamical equilibrium experiments. (authors)

  5. Demonstration of improved vehicle fuel efficiency through innovative tire design, materials, and weight reduction technologies

    Energy Technology Data Exchange (ETDEWEB)

    Donley, Tim [Cooper Tire & Rubber Company Incorporated, Findlay, OH (United States)

    2014-12-31

    Cooper completed an investigation into new tire technology using a novel approach to develop and demonstrate a new class of fuel efficient tires using innovative materials technology and tire design concepts. The objective of this work was to develop a new class of fuel efficient tires, focused on the “replacement market” that would improve overall passenger vehicle fuel efficiency by 3% while lowering the overall tire weight by 20%. A further goal of this project was to accomplish the objectives while maintaining the traction and wear performance of the control tire. This program was designed to build on what has already been accomplished in the tire industry for rolling resistance based on the knowledge and general principles developed over the past decades. Cooper’s CS4 (Figure #1) premium broadline tire was chosen as the control tire for this program. For Cooper to achieve the goals of this project, the development of multiple technologies was necessary. Six technologies were chosen that are not currently being used in the tire industry at any significant level, but that showed excellent prospects in preliminary research. This development was divided into two phases. Phase I investigated six different technologies as individual components. Phase II then took a holistic approach by combining all the technologies that showed positive results during phase one development.

  6. Material control and accountability aspects of safeguards for the USA 233U/Th fuel recycle plant

    International Nuclear Information System (INIS)

    Carpenter, J.A. Jr.; McNeany, S.R.; Angelini, P.; Holder, N.D.; Abraham, L.

    1978-01-01

    The materials control and accountability aspects of the reprocessing and refabrication of a conceptual large-scale HTGR fuel recycle plant have been discussed. Two fuel cycles were considered. The traditional highly enriched uranium cycle uses an initial or makeup fuel element with a fissile enrichment of 93% 235 U. The more recent medium enriched uranium cycle uses initial or makeup fuel elements with a fissile enrichment less than 20% 235 U. In both cases, 233 U bred from the fertile thorium is recycled. Materials control and accountability in the plant will be by means of a real-time accountability method. Accountability data will be derived from monitoring of total material mass through the processes and a system of numerous assays, both destructive and nondestructive

  7. Novel materials for more robust solid oxide fuel cells in small scale applications

    DEFF Research Database (Denmark)

    Holtappels, Peter

    Solid oxide fuel cells can offer supply of electrical energy with a high efficiency and based on a wide range of fuels. While natural gas and/or bio methane is a commonly used fuel for combined heat and power supply, liquid fuels such as gasoline, Diesel and alcohols are interesting fuels...

  8. 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

  9. 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.

  10. Toward new fuel cell support materials: a theoretical and experimental study of nitrogen-doped graphene.

    Science.gov (United States)

    Seo, Min Ho; Choi, Sung Mook; Lim, Eun Ja; Kwon, In Hye; Seo, Joon Kyo; Noh, Seung Hyo; Kim, Won Bae; Han, Byungchan

    2014-09-01

    Nano-scale Pt particles are often reported to be more electrochemically active and stable in a fuel cell if properly displaced on support materials; however, the factors that affect their activity and stability are not well understood. We applied first-principles calculations and experimental measurements to well-defined model systems of N-doped graphene supports (N-GNS) to reveal the fundamental mechanisms that control the catalytic properties and structural integrity of nano-scale Pt particles. DFT calculations predict thermodynamic and electrochemical interactions between N-GNS and Pt nanoparticles in the methanol oxidation reaction (MOR) and oxygen reduction reaction (ORR). Moreover, the dissolution potentials of the Pt nanoparticles supported on GNS and N-GNS catalysts are calculated under acidic conditions. Our results provide insight into the design of new support materials for enhanced catalytic efficiency and long-term stability. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  11. Mechanism of chromium poisoning the conventional cathode material for solid oxide fuel cells

    Science.gov (United States)

    Zhang, Xiaoqiang; Yu, Guangsen; Zeng, Shumao; Parbey, Joseph; Xiao, Shuhao; Li, Baihai; Li, Tingshuai; Andersson, Martin

    2018-03-01

    Chromium poisoning the La0.875Sr0.125MnO3 (LSM) cathode for solid oxide fuel cells is a critical issue that can strongly affect the stability. In this study, we evaluate the temperature distribution in a SOFC based on a 3D model and then combine conductivity test and material computation to reveal the effects of chromium in SUS430 stainless steels on LSM conductivities. The starch concentration in LSM pellets and the applied pressure on the contact with interconnect materials show close relationships with the chromium poisoning behavior. The density functional theory (DFT) computing results indicate that chromium atoms preferably adsorb on the MnO2-terminated and La (Sr)-O-terminated (001) surfaces. The resulting conclusions are expected to deeply understand mechanism of chromium deactivating conventional cathodes at some typical operational conditions, and offer crucial information to optimize the structure to avoid the poisoning effect.

  12. Engineered Graphene Materials: Synthesis and Applications for Polymer Electrolyte Membrane Fuel Cells.

    Science.gov (United States)

    He, Daping; Tang, Haolin; Kou, Zongkui; Pan, Mu; Sun, Xueliang; Zhang, Jiujun; Mu, Shichun

    2017-05-01

    Engineered graphene materials (EGMs) with unique structures and properties have been incorporated into various components of polymer electrolyte membrane fuel cells (PEMFCs) such as electrode, membrane, and bipolar plates to achieve enhanced performances in terms of electrical conductivity, mechanical durability, corrosion resistance, and electrochemical surface area. This research news article provides an overview of the recent development in EGMs and EGM-based PEMFCs with a focus on the effects of EGMs on PEMFC performance when they are incorporated into different components of PEMFCs. The challenges of EGMs for practical PEMFC applications in terms of production scale, stability, conductivity, and coupling capability with other materials are also discussed and the corresponding measures and future research trends to overcome such challenges are proposed. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  13. Experimental Measurement of Relative Permeability Functions for Fuel Cell GDL Materials

    KAUST Repository

    Hussaini, Irfan

    2009-01-01

    Gas diffusion layer in PEM fuel cells plays a pivotal role in water management. Modeling of liquid water transport through the GDL relies on knowledge of relative permeability functions in the in-plane and through-plane directions. In the present work, air and water relative permeabilities are experimentally determined as functions of saturation for typical GDL materials such as Toray-060, -090, -120 carbon paper and E-Tek carbon cloth materials in their plain, untreated forms. Saturation is measured using an ex-situ gravimetric method. Absolute and relative permeability functions in the two directions of interest are presented. Significant departure from the generally assumed cubic function of saturation is observed. ©The Electrochemical Society.

  14. Mesoporous nitrogen-rich carbon materials as cathode catalysts in microbial fuel cells

    KAUST Repository

    Ahn, Yongtae

    2014-12-01

    The high cost of the catalyst material used for the oxygen reduction reaction in microbial fuel cell (MFC) cathodes is one of the factors limiting practical applications of this technology. Mesoporous nitrogen-rich carbon (MNC), prepared at different temperatures, was examined as an oxygen reduction catalyst, and compared in performance to Pt in MFCs and electrochemical cells. MNC calcined at 800 °C produced a maximum power density of 979 ± 131 mW m-2 in MFCs, which was 37% higher than that produced using MNC calined at 600 °C (715 ± 152 mW m-2), and only 14% lower than that obtained with Pt (1143 ± 54 mW m-2). The extent of COD removal and coulombic efficiencies were the same for all cathode materials. These results show that MNC could be used as an alternative to Pt in MFCs. © 2014 Elsevier B.V. All rights reserved.

  15. Electrode materials: a challenge for the exploitation of protonic solid oxide fuel cells

    International Nuclear Information System (INIS)

    Fabbri, Emiliana; Pergolesi, Daniele; Traversa, Enrico

    2010-01-01

    High temperature proton conductor (HTPC) oxides are attracting extensive attention as electrolyte materials alternative to oxygen-ion conductors for use in solid oxide fuel cells (SOFCs) operating at intermediate temperatures (400-700 0 C). The need to lower the operating temperature is dictated by cost reduction for SOFC pervasive use. The major stake for the deployment of this technology is the availability of electrodes able to limit polarization losses at the reduced operation temperature. This review aims to comprehensively describe the state-of-the-art anode and cathode materials that have so far been tested with HTPC oxide electrolytes, offering guidelines and possible strategies to speed up the development of protonic SOFCs. (topical review)

  16. Electrode materials: a challenge for the exploitation of protonic solid oxide fuel cells

    Directory of Open Access Journals (Sweden)

    Emiliana Fabbri, Daniele Pergolesi and Enrico Traversa

    2010-01-01

    Full Text Available High temperature proton conductor (HTPC oxides are attracting extensive attention as electrolyte materials alternative to oxygen-ion conductors for use in solid oxide fuel cells (SOFCs operating at intermediate temperatures (400–700 °C. The need to lower the operating temperature is dictated by cost reduction for SOFC pervasive use. The major stake for the deployment of this technology is the availability of electrodes able to limit polarization losses at the reduced operation temperature. This review aims to comprehensively describe the state-of-the-art anode and cathode materials that have so far been tested with HTPC oxide electrolytes, offering guidelines and possible strategies to speed up the development of protonic SOFCs.

  17. A Viable Electrode Material for Use in Microbial Fuel Cells for Tropical Regions

    Directory of Open Access Journals (Sweden)

    Felix Offei

    2016-01-01

    Full Text Available Electrode materials are critical for microbial fuel cells (MFC since they influence the construction and operational costs. This study introduces a simple and efficient electrode material in the form of palm kernel shell activated carbon (AC obtained in tropical regions. The novel introduction of this material is also targeted at introducing an inexpensive and durable electrode material, which can be produced in rural communities to improve the viability of MFCs. The maximum voltage and power density obtained (under 1000 Ω load using an H-shaped MFC with AC as both anode and cathode electrode material was 0.66 V and 1.74 W/m3, respectively. The power generated by AC was as high as 86% of the value obtained with the extensively used carbon paper. Scanning electron microscopy and Denaturing Gradient Gel Electrophoresis (DGGE analysis of AC anode biofilms confirmed that electrogenic bacteria were present on the electrode surface for substrate oxidation and the formation of nanowires.

  18. Study utilization of extractable petroleum hydrocarbons biodegradation waste as the main material for making solid fuels

    Science.gov (United States)

    Hendrianie, Nuniek; Juliastuti, Sri Rachmania; Ar-rosyidah, Fanny Husna; Rochman, Hilal Abdur

    2017-05-01

    Nowadays the existence of energy sources of oil and was limited. Therefore, it was important to searching for new innovations of renewable energy sources by utilizing the waste into a source of energy. On the other hand, the process of extractable petroleum hydrocarbons biodegradation generated sludge that had calorific value and untapped. Because of the need for alternative sources of energy innovation with the concept of zero waste and the fuel potential from extractable petroleum hydrocarbons biodegradation waste, so it was necessary to study the use of extractable petroleum hydrocarbons biodegradation waste as the main material for making solid fuel. In addition, sawdust is a waste that had a great quantities and also had a high calorific value to be mixed with extractable petroleum hydrocarbons biodegradation waste. The purpose of this study was to determine the characteristics of the extractable petroleum hydrocarbons biodegradation waste and to determine the potential and a combination of a mixture of extractable petroleum hydrocarbons biodegradation waste and sawdust which has the best calorific value. The variables of this study was the composition of the waste and sawdust as follows 1:1; 1:3; and 3:1 (mass of sawdust : mass of waste) and time of sawdust carbonization was 10, 15 and 20 minutes. Sawdust was carbonized to get the high heating value. The characteristic of main material and fuel analysis performed with proximate analysis. While the calorific value analysis was performed with a bomb calorimeter. From the research, it was known that extractable petroleum hydrocarbons biodegradation waste had a moisture content of 3.06%; volatile matter 19.98%; ash content of 0.56%; fixed carbon content of 76.4% and a calorific value of 717 cal/gram. And a mixture that had the highest calorific value (4286.5 cal/gram) achieved in comparison sawdust : waste (3:1) by carbonization of sawdust for 20 minutes.

  19. Preliminary Compatibility Assessment of Metallic Dispenser Materials for Service in Ethanol Fuel Blends

    Energy Technology Data Exchange (ETDEWEB)

    Pawel, Steven J [ORNL; Kass, Michael D [ORNL; Janke, Christopher James [ORNL

    2009-11-01

    The compatibility of selected metals representative of those commonly used in dispensing systems was evaluated in an aggressive E20 formulation (CE20a) and in synthetic gasoline (Reference Fuel C) in identical testing to facilitate comparison of results. The testing was performed at modestly elevated temperature (nominally 60 C) and with constant fluid flow in an effort to accelerate potential interactions in the screening test. Based on weight change, the general corrosion of all individual coupons exposed in the vapor phase above Reference Fuel C and CE20a as well as all coupons immersed in Reference Fuel C was essentially nil (<0.3 {micro}m/y), with no evidence of localized corrosion such as pitting/crevice corrosion or selective leaching at any location. Modest discoloration was observed on the copper-based alloys (cartridge brass and phosphor bronze), but the associated corrosion films were quite thin and apparently protective. For coupons immersed in CE20a, four different materials exhibited net weight loss over the entire course of the experiment: cartridge brass, phosphor bronze, galvanized steel, and terne-plated steel. None of these exhibited substantial incompatibility with the test fluid, with the largest general corrosion rate calculated from coupon weight loss to be approximately 4 {micro}m/y for the cartridge brass specimens. Selective leaching of zinc (from brass) and tin (from bronze) was observed, as well as the presence of sulfide surface films rich in these elements, suggesting the importance of the role of sulfuric acid in the CE20a formulation. Analysis of weight loss data for the slightly corroded metals indicated that the corrosivity of the test environment decreased with exposure time for brass and bronze and increased for galvanized and terne-plated steel. Other materials immersed in CE20a - type 1020 mild steel, type 1100 aluminum, type 201 nickel, and type 304 stainless steel - each appeared essentially immune to corrosion at the test

  20. Evaluation of materials for bipolar plates in simulated PEM fuel-cell cathodic environments

    Energy Technology Data Exchange (ETDEWEB)

    Rivas, S.V.; Belmonte, M.R.; Moron, L.E.; Torres, J.; Orozco, G. [Centro de Investigacion y Desarrollo Technologico en Electroquimica S.C. Parcque Sanfandila, Queretaro (Mexico); Perez-Quiroz, J.T. [Mexican Transport Inst., Queretaro (Mexico); Cortes, M. A. [Mexican Petroleum Inst., Mexico City (Mexico)

    2008-04-15

    The bipolar plates in proton exchange membrane fuel cells (PEMFC) are exposed to an oxidizing environment on the cathodic side, and therefore are susceptible to corrosion. Corrosion resistant materials are needed for the bipolar plates in order to improve the lifespan of fuel cells. This article described a study in which a molybdenum (Mo) coating was deposited over austenitic stainless steel 316 and carbon steel as substrates in order to evaluate the resulting surfaces with respect to their corrosion resistance in simulated anodic and cathodic PEMFC environments. The molybdenum oxide films were characterized by scanning electron microscopy (SEM) and Raman spectroscopy. The article presented the experiment and discussed the results of the corrosion behaviour of coated stainless steel. In general, the electrochemical characterization of bare materials and coated steel consisted of slow potentiodynamic polarization curves followed by a constant potential polarization test. The test medium was 0.5M sulfuric acid with additional introduction of oxygen to simulate the cathodic environment. All tests were performed at ambient temperature and at 50 degrees Celsius. The potentiostat used was a Gamry instrument. It was concluded that it is possible to deposit Mo-oxides on steel without using another alloying metal. The preferred substrate for corrosion prevention was found to be an alloy with high chromium content. 24 refs., 4 figs.

  1. Lanthanum chromite materials as potential symmetrical electrodes for Solid Oxide Fuel Cells

    Directory of Open Access Journals (Sweden)

    Ruiz-Morales, J. C.

    2007-08-01

    Full Text Available A commonly used interconnector material has been tested as electrode for a new concept of Solid Oxide Fuel Cell, where the same material could be used, simultaneously, as interconnector, anode and cathode. We have found that a typical substituted chromite, such as La0.7Ca0.3CrO3-δ (LCC can be considered a good candidate for such configuration, due to its high electronic conductivity in both reducing and oxidising conditions, and moderate catalytic properties for oxygen reduction and hydrogen oxidation. The symmetrical design renders performances of 100 mWcm-2 at 950ºC, using O2 and H2 as oxidant and fuel respectively. Performances exceeding 300 mWcm-2 can be predicted for a 100μm-thick YSZ electrolyte.

    Un material comúnmente utilizado como interconector ha sido probado como electrodo para un nuevo concepto de Pila de Combustible de Óxidos Sólido, en el cual el mismo material se utiliza, simultáneamente, como interconector, ánodo y cátodo. Hemos encontrado que una cromita típica como La0.7Ca0.3CrO3-δ (LCC puede ser considerada una buena candidata para dicha configuración, debido a sus altas conductividades eléctricas tanto en condiciones reductoras como oxidantes y una aceptable actividad catalítica para la reducción del oxígeno y la oxidación del hidrógeno. El diseño simétrico permite obtener rendimientos del orden de 100mWcm-2 a 950ºC, utilizando O2 e H2 como oxidante y combustible, respectivamente. Rendimientos que superan los 300mWcm-2 pueden predecirse para pilas con electrolitos de YSZ de 100 μm de grosor.

  2. Order for execution of the law concerning regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The order is enacted under the law for the regulations of nuclear source materials, nuclear fuel materials and reactors. Any person who engages in refining business shall get designation for each works or place of enterprise. The application shall be filed through the director of International Trade and Industry Office in charge of the location of the works or the enterprise with a business program and other specified documents attached. Any person who undertakes processing business shall get permission for each works or place of enterprise. The application shall be submitted with a business program and other documents defined by the Ordinance of the Prime Minister's Office. Any person who sets up reactor shall get permission for each works or place of enterprise. The application shall be presented with a financial project and other documents stipulated by the ordinance. Fast breeding reactor, heavy-water moderated boiling water reactor and light-water moderated pressurized water reactor are designated as reactor in the phase of research and development. Each foreign nuclear ship equipped with reactor which enters into Japanese waters shall get permission of the Minister of Transport. The application shall be presented with the papers explaining safety of reactor facilities and other documents provided by the ordinance of the ministry concerned. (Okada, K.)

  3. Fuels and materials research under the high neutron fluence using a fast reactor Joyo and post-irradiation examination facilities

    International Nuclear Information System (INIS)

    Soga, Tomonori; Ito, Chikara; Aoyama, Takafumi; Suzuki, Soju

    2009-01-01

    The experimental fast reactor Joyo at Oarai Research and Development Center (ORDC) of Japan Atomic Energy Agency (JAEA) is Japan's sodium-cooled fast reactor (FR). In 2003, this reactor's upgrade to the 140MWt MK-III core was completed to increase the irradiation testing capability. The MK-III core provides the fast neutron flux of 4.0x10 15 n/cm 2 s as an irradiation test bed for improving the fuels and material of FR in Japan. Three post-irradiation examination (PIE) facilities named FMF, MMF and AGF related to Joyo are in ORDC. Irradiated subassemblies and core components are carried into the FMF (Fuel Monitoring Facility) and conducted nondestructive examinations. Each subassembly is disassembled to conduct some destructive examinations and to prepare the fuel and material samples for further detailed examinations. Fuel samples are sent to the AGF (Alpha-Gamma Facility), and material samples are sent to the MMF (Materials Monitoring Facility). These overall and elaborate data provided by PIE contribute to investigate the irradiation effect and behavior of fuels and materials. This facility complex is indispensable to promote the R and D of FR in Japan. And, the function and technology of irradiation test and PIE enable to contribute to the R and D of innovative fission or fusion reactor material which will be required to use under the high neutron exposure. (author)

  4. Evaluation of source term parameters for spent fuel disposal in foreign countries. (2) Dissolution rates of spent fuel matrices and construction materials for fuel assemblies

    International Nuclear Information System (INIS)

    Kitamura, Akira; Chikazawa, Takahiro; Tachi, Yukio; Akahori, Kuniaki

    2016-01-01

    The Japanese geological disposal program has started researching disposal of spent nuclear fuel (SF) in deep geological strata (hereafter 'direct disposal of SF') as an alternative management option other reprocessing followed by vitrification and geological disposal of high-level radioactive waste. We conducted literature survey of dissolution rate of SF matrix and constructing materials (e.g. zircaloy cladding and control rods) selected in safety assessment reports for direct disposal of SF in Europe and United States. We also investigated basis of release rate determination and assignment of uncertainties in the safety assessment reports. Furthermore, we summarized major conclusions proposed by some European projects governed by European Commission. It was found that determined release rates are fairly similar to each other due to use of similar literature data in all countries of interest. It was also found that the determined release rates were including conservativeness because it was difficult to assign uncertainties quantitatively. It is expected that these findings are useful as fundamental information for determination of the release rates for the safety assessment of Japanese SF disposal system. (author)

  5. Use of natural materials from Northern Russia for the isolation of radioactive wastes and spent nuclear fuel

    International Nuclear Information System (INIS)

    Komlev, V.N.

    1998-01-01

    The application of natural materials to the isolation of radioactive waste and spent nuclear fuel is being assessed, together with possible isolation technologies. The operational requirements for such materials are identified and a proposal for an inter-regional ecological and technological project is discussed. 39 refs

  6. 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.)

  7. Development of Melting Crucible Materials of Metallic Fuel Slug for SFR

    International Nuclear Information System (INIS)

    Kim, K. H.; Lee, C. T.; Oh, S. J.; Kim, S. K.; Lee, C. B.; Ko, Y. M.; Woo, W. M.

    2010-01-01

    The fabrication process of metallic fuel for SFR(sodium fast reactor) of Generation-IV candidate reactors is composed of the fabrication of fuel pin, fuel rod, and fuel assembly. The key technology of the fabrication process for SFR can be referred to the fabrication technology of fuel pin. As SFR fuel contains MA(minor actinide) elements proceeding the recycling of actinide elements, it is so important to extinguish MA during irradiation in SFR, included in nuclear fuel through collection of volatile MA elements during fabrication of fuel pin. Hence, it is an imminent circumstance to develop the fabrication process of fuel pin. This report is an state-of art report related to the characteristics of irradiation performance for U-Zr-Pu metallic fuel, and the apparatus and the technology of conventional injection casting process. In addition, to overcome the drawbacks of the conventional injection casting and the U-Zr-Pu fuel, new fabrication technologies such as the gravity casting process, the casting of fuel pin to metal-barrier mold, the fabrication of particulate metallic fuel utilizing centrifugal atomization is surveyed and summarized. The development of new U-10Mo-X metallic fuel as nuclear fuel having a single phase in the temperature range between 550 and 950 .deg. C, reducing the re-distribution of the fuel elements and improving the compatibility between fuel and cladding, is also surveyed and summarized

  8. Material control and accounting in the Department of Energy's nuclear fuel complex

    Energy Technology Data Exchange (ETDEWEB)

    None

    1989-01-01

    Material control and accounting takes place within an envelope of activities related to safeguards and security, as well as to safety, health, and environment, all of which need to be managed to assure that the entire nuclear fuel complex can operate in a societally accepted manner. Within this envelope the committee was directed to carry out the following scope of work: (1) Review the MCandA systems in use at selected DOE facilities that are processing special nuclear material (SNM) in various physical and chemical forms. (2) Design and convene a workshop for senior representatives from each of DOE's facilities on the flows and inventories of nuclear materials. (3) Plan and conduct a series of site visits to each of the facilities to observe first hand the processing operations and the related MCandA systems. (4) Review the potential improvement in overall safeguard systems effectiveness, as measured by expected reduction in inventory difference control limits and inventory differences for materials balance accounts and facilities, or other criteria as appropriate. Indicate how this affects the relative degree of uncertainty in the system. (5) Review the efficiency of operating the MCandA system with and without the upgrading options and assess whether upgrading will contribute further efficiencies in operation, which may reduce many of the current operations costs. Determine if the current system is cost-effective. (6) Recommend the most promising technical approaches for further development by DOE and further study as warranted.

  9. Sorbent Material Property Requirements for On-Board Hydrogen Storage for Automotive Fuel Cell Systems.

    Energy Technology Data Exchange (ETDEWEB)

    Ahluwalia, R. K.; Peng, J-K; Hua, T. Q.

    2015-05-25

    Material properties required for on-board hydrogen storage in cryogenic sorbents for use with automotive polymer electrolyte membrane (PEM) fuel cell systems are discussed. Models are formulated for physical, thermodynamic and transport properties, and for the dynamics of H-2 refueling and discharge from a sorbent bed. A conceptual storage configuration with in-bed heat exchanger tubes, a Type-3 containment vessel, vacuum insulation and requisite balance-of-plant components is developed to determine the peak excess sorption capacity and differential enthalpy of adsorption for 5.5 wt% system gravimetric capacity and 55% well-to-tank (WTT) efficiency. The analysis also determines the bulk density to which the material must be compacted for the storage system to reach 40 g.L-1 volumetric capacity. Thermal transport properties and heat transfer enhancement methods are analyzed to estimate the material thermal conductivity needed to achieve 1.5 kg.min(-1) H-2 refueling rate. Operating temperatures and pressures are determined for 55% WTT efficiency and 95% usable H-2. Needs for further improvements in material properties are analyzed that would allow reduction of storage pressure to 50 bar from 100 bar, elevation of storage temperature to 175-200 K from 150 K, and increase of WTT efficiency to 57.5% or higher.

  10. Mesoporous TiO2 : an alternative material for PEM fuel cells catalyst support

    Energy Technology Data Exchange (ETDEWEB)

    Do, T.B. [Michigan Univ., Ann Arbor, MI (United States). Dept. of Materials Science; Ruthkosky, M.; Cai, M. [General Motors, Warren, MI (United States). Research and Development Center

    2008-07-01

    This paper discussed the feasibility of using an alternative catalyst support material to replace carbon in proton exchange membrane (PEM) fuel cells. The alternative catalyst support material requires a high surface area with a large porosity but must have comparable conductivity with carbon. A mesoporous titanium oxide (TiO2) material produced by coprecipitation was introduced. The conductivity of the material is about one order of that of carbon. The 8 mole per cent Nb-doped TiO2 was formed and deposited on the surface of a nano polystyrene (PS) template via the hydrolysis of a co-solution of Ti(OC4H9)4 and Nb(OC2H5)5. The removal of PS by heat treatment produced porous structure of TiO2 with the appearance of 3 different pore types, notably open pore, ink-pot pores and closed pores. TiO2 formed from the rutile phase, allowing a lower activation temperature at 850 degrees C in a hydrogen atmosphere. The pore structures were retained after this heat treatment. The BET surface area was 116 m{sup 2}/g, porosity was 22 per cent and the average pore size was 159 angstrom. The conductivity improved considerably from almost non-conductive to one order of that of carbon.

  11. Nanostructured Carbon Materials as Supports in the Preparation of Direct Methanol Fuel Cell Electrocatalysts

    Directory of Open Access Journals (Sweden)

    María Jesús Lázaro

    2013-08-01

    Full Text Available Different advanced nanostructured carbon materials, such as carbon nanocoils, carbon nanofibers, graphitized ordered mesoporous carbons and carbon xerogels, presenting interesting features such as high electrical conductivity and extensively developed porous structure were synthesized and used as supports in the preparation of electrocatalysts for direct methanol fuel cells (DMFCs. The main advantage of these supports is that their physical properties and surface chemistry can be tailored to adapt the carbonaceous material to the catalytic requirements. Moreover, all of them present a highly mesoporous structure, diminishing diffusion problems, and both graphitic character and surface area can be conveniently modified. In the present work, the influence of the particular features of each material on the catalytic activity and stability was analyzed. Results have been compared with those obtained for commercial catalysts supported on Vulcan XC-72R, Pt/C and PtRu/C (ETEK. Both a highly ordered graphitic and mesopore-enriched structure of these advanced nanostructured materials resulted in an improved electrochemical performance in comparison to the commercial catalysts assayed, both towards CO and alcohol oxidation.

  12. FMEF profilometry and visual examination feasibility and conceptual design. [Fuels and Materials Examination Facility; LMFBR and GCFR

    Energy Technology Data Exchange (ETDEWEB)

    Philipp, L.D.; Dilbeck, R.A.; Hartman, J.S.; Hildebrand, B.P.; Reich, F.R.; Swinth, K.L.

    1976-07-01

    The Fuels and Materials Examination Facility (FMEF) is being scoped to provide postirradiation examination capabilities for FFTF, LMFBR and GCFR fuels and materials. The Hanford Engineering Development Laboratory has requested that the Battelle Pacific Northwest Laboratory (PNL) complete a feasibility study for the development of equipment to meet the FMEF Measurement Requirements for irradiated fuel pin and absorber rod bow, length, profile and visual examination stations. The purpose of the report is to provide a conceptual design for development of the examination equipment. The design analysis assumes that fuel pins and absorber rods to be examined are in the main cell. The cell's environment will be argon or nitrogen gas at a pressure between --1 and --4 in. of water and at a temperature between 70 and 100/sup 0/F. Oxygen content of the cell gas will normally be controlled between 25 and 50 ppM. Water content will be controlled within the same limits.

  13. Fabrication of uranium-americium mixed oxide fuels: thermodynamical modeling and materials properties

    International Nuclear Information System (INIS)

    Prieur, D.

    2011-01-01

    Fuel irradiation in pressurized water reactors lead to the formation of fission products and minor actinides (Np, Am, Cm) which can be transmuted in fast neutrons reactors. In this context, the aim of this work was to study the fabrication conditions of the U 1-y Am y O 2+x fuels which exhibit particular thermodynamical properties requiring an accurate monitoring of the oxygen potential during the sintering step. For this reason, a thermodynamical model was developed to assess the optimum sintering conditions for these materials. From these calculations, U 1-y Am y O 2+x (y=0.10; 0.15; 0.20; 0.30) were sintered in two range of atmosphere. In hyper-stoichiometric conditions at low temperature, porous and multiphasic compounds are obtained whereas in reducing conditions at high temperature materials are dense and monophasic. XAFS analyses were performed in order to obtain additional experimental data for the thermodynamical modeling refinement. These characterizations also showed the reduction of Am(+IV) to Am(+III) and the partial oxidation of U(+IV) to U(+V) due to a charge compensation mechanism occurring during the sintering. Finally, taking into account the high - activity of Am, self-irradiation effects were studied for two types of microstructures and two Am contents (10 and 15%). For each composition, a lattice parameter increase was observed without structural change coupled with a macroscopic swelling of the pellet diameter up to 1.2% for the dense compounds and 0.6% for the tailored porosity materials. (author) [fr

  14. Materials and characterization techniques for high-temperature polymer electrolyte membrane fuel cells.

    Science.gov (United States)

    Zeis, Roswitha

    2015-01-01

    The performance of high-temperature polymer electrolyte membrane fuel cells (HT-PEMFC) is critically dependent on the selection of materials and optimization of individual components. A conventional high-temperature membrane electrode assembly (HT-MEA) primarily consists of a polybenzimidazole (PBI)-type membrane containing phosphoric acid and two gas diffusion electrodes (GDE), the anode and the cathode, attached to the two surfaces of the membrane. This review article provides a survey on the materials implemented in state-of-the-art HT-MEAs. These materials must meet extremely demanding requirements because of the severe operating conditions of HT-PEMFCs. They need to be electrochemically and thermally stable in highly acidic environment. The polymer membranes should exhibit high proton conductivity in low-hydration and even anhydrous states. Of special concern for phosphoric-acid-doped PBI-type membranes is the acid loss and management during operation. The slow oxygen reduction reaction in HT-PEMFCs remains a challenge. Phosphoric acid tends to adsorb onto the surface of the platinum catalyst and therefore hampers the reaction kinetics. Additionally, the binder material plays a key role in regulating the hydrophobicity and hydrophilicity of the catalyst layer. Subsequently, the binder controls the electrode-membrane interface that establishes the triple phase boundary between proton conductive electrolyte, electron conductive catalyst, and reactant gases. Moreover, the elevated operating temperatures promote carbon corrosion and therefore degrade the integrity of the catalyst support. These are only some examples how materials properties affect the stability and performance of HT-PEMFCs. For this reason, materials characterization techniques for HT-PEMFCs, either in situ or ex situ, are highly beneficial. Significant progress has recently been made in this field, which enables us to gain a better understanding of underlying processes occurring during fuel cell

  15. Summary report on transportation of nuclear fuel materials in Japan : transportation infrastructure, threats identified in open literature, and physical protection regulations.

    Energy Technology Data Exchange (ETDEWEB)

    Cochran, John Russell; Ouchi, Yuichiro (Japan Atomic Energy Agency, Japan); Furaus, James Phillip; Marincel, Michelle K.

    2008-03-01

    This report summarizes the results of three detailed studies of the physical protection systems for the protection of nuclear materials transport in Japan, with an emphasis on the transportation of mixed oxide fuel materials1. The Japanese infrastructure for transporting nuclear fuel materials is addressed in the first section. The second section of this report presents a summary of baseline data from the open literature on the threats of sabotage and theft during the transport of nuclear fuel materials in Japan. The third section summarizes a review of current International Atomic Energy Agency, Japanese and United States guidelines and regulations concerning the physical protection for the transportation of nuclear fuel materials.

  16. Spent Fuel Working Group report on inventory and storage of the Department's spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities

    International Nuclear Information System (INIS)

    1993-11-01

    The Secretary of Energy's memorandum of August 19, 1993, established an initiative for a Department-wide assessment of the vulnerabilities of stored spent nuclear fuel and other reactor irradiated nuclear materials. A Project Plan to accomplish this study was issued on September 20, 1993 by US Department of Energy, Office of Environment, Health and Safety (EH) which established responsibilities for personnel essential to the study. The DOE Spent Fuel Working Group, which was formed for this purpose and produced the Project Plan, will manage the assessment and produce a report for the Secretary by November 20, 1993. This report was prepared by the Working Group Assessment Team assigned to the Hanford Site facilities. Results contained in this report will be reviewed, along with similar reports from all other selected DOE storage sites, by a working group review panel which will assemble the final summary report to the Secretary on spent nuclear fuel storage inventory and vulnerability

  17. Characterizing automotive fuel cell materials by soft x-ray scanning transmission x-ray microscopy

    International Nuclear Information System (INIS)

    Hitchcock, A. P.; Lee, V.; Wu, J.; Cooper, G.; West, M. M.; Berejnov, V.; Soboleva, T.; Susac, D.; Stumper, J.

    2016-01-01

    Proton-Exchange Membrane Fuel Cell (PEM-FC) based engines are being developed rapidly for near-term implementation in hydrogen fueled, mass production, personal automobiles. Research is focused on understanding and controlling various degradation processes (carbon corrosion, Pt migration, cold start), and reducing cost by reducing or eliminating Pt catalyst. We are using soft X-ray scanning transmission X-ray microscopy (STXM) at the S 2p, C 1s, O 1s and F 1s edges to study a variety of issues related to optimization of PEM-FC materials for automotive applications. A method to efficiently and accurately measure perfluorosulfonic acid distributions was developed and is being used to better understand how different loadings and preparation methods affect the ionomer distribution in the cathode. Progress towards an environmental cell capable of controlling the temperature and humidity of a PEM-FC sample in the STXM is described. Methods for studying the 3D chemical structure of PEM-FC are outlined

  18. Preliminary analysis of the creep behaviour of nuclear fuel-waste container materials

    International Nuclear Information System (INIS)

    Dutton, R.; Leitch, B.W.; Crosthwaite, J.L.; Kasprick, G.R.

    1996-12-01

    In the Canadian Nuclear Fuel Waste Management Program, it is proposed that nuclear fuel waste be placed in a durable container and disposed of in a deep underground vault. Consideration of various disposal-container designs has identified either titanium or copper as the material suitable for constructing the container shell. As part of the R and D program to examine the structural integrity of the container, creep tests are being conducted on commercially pure titanium and oxygen-free copper. This report presents the preliminary data obtained. It also describes the evaluation of various constitutive equations to represent the creep curves, thus providing the basis for extrapolation of the creep behaviour over the design lifetime of the container. In this regard, a specific focus is placed on equations derived from the 0-Projection Concept. Recognizing that the container lifetime will be determined by the onset of tertiary creep leading to creep rupture, we present the results of the metallographic examination of creep damage. This shows that the tertiary stage in titanium is associated with the formation of transgranular cavities within the region of localized necking of the creep specimens. In contrast, creep damage in copper is in the form of intergranular cavities uniformly distributed throughout the gauge length. These results are analyzed within the context of the extant literature, and their implications for future container design are discussed. (author)

  19. Advanced materials for solid oxide fuel cells: Hafnium-Praseodymium-Indium Oxide System

    Energy Technology Data Exchange (ETDEWEB)

    Bates, J.L.; Griffin, C.W.; Weber, W.J.

    1988-06-01

    The HfO/sub 2/-PrO/sub 1.83/-In/sub 2/O/sub 3/ system has been studied at the Pacific Northwest Laboratory to develop alternative, highly electrically conducting oxides as electrode and interconnection materials for solid oxide fuel cells. A coprecipitation process was developed for synthesizing single-phase, mixed oxide powders necessary to fabricate powders and dense oxides. A ternary phase diagram was developed, and the phases and structures were related to electrical transport properties. Two new phases, an orthorhombic PrInO/sub 3/ and a rhombohedral Hf/sub 2/In/sub 2/O/sub 7/ phase, were identified. The highest electronic conductivity is related to the presence of a bcc, In/sub 2/O/sub 3/ solid solution (ss) containing HfO/sub 2/ and PrO/sub 1.83/. Compositions containing more than 35 mol % of the In/sub 2/O/sub 3/ ss have electrical conductivities greater than 10/sup /minus/1/ (ohm-cm)/sup /minus/1/, and the two or three phase structures that contain this phase appear to exhibit mixed electronic-ionic conduction. The high electrical conductivities and structures similar to the Y/sub 2/O/sub 3/-stabilized ZrO/sub 2/(HfO/sub 2/) electrolyte give these oxides potential for use as cathodes in solid oxide fuel cells. 21 refs.

  20. Designed Materials for Enhanced Oxygen Reduction Electrocatalysis in PEM Fuel Cells: Novel Materials and Next Generation Synchrotron Based in Situ

    Science.gov (United States)

    2005-11-15

    fonnation at the electrocatalyst- ionomer interface in a PEM fuel cell context is especially important from the perspective ofattempts to develop alternative...Reduction in a Fully Hydrated Proton Exchange Membrane Interface : Problems and Prospects’, S. Mukerjee. Keynote speaker at International Fuel Cells , May 15th...Synchrotron XAS Investigation’S. Mukerjee, invited speaker at the Workshop on Theory and Surface Measurement of Fuel Cell Catalysts , Magleas Conference

  1. Optimal measurement uncertainties for materials accounting in a fast breeder reactor spent-fuel reprocessing plant

    International Nuclear Information System (INIS)

    Dayem, H.A.; Kern, E.A.; Markin, J.T.

    1982-01-01

    Optimization techniques are used to calculate measurement uncertainties for materials accountability instruments in a fast breeder reactor spent-fuel reprocessing plant. Optimal measurement uncertainties are calculated so that performance goals for detecting materials loss are achieved while minimizing the total instrument development cost. Improved materials accounting in the chemical separations process (111 kg Pu/day) to meet 8-kg plutonium abrupt (1 day) and 40-kg plutonium protracted (6 months) loss-detection goals requires: process tank volume and concentration measurements having precisions less than or equal to 1%; accountability and plutonium sample tank volume measurements having precisions less than or equal to 0.3%, short-term correlated errors less than or equal to 0.04%, and long-term correlated errors less than or equal to 0.04%; and accountability and plutonium sample tank concentration measurements having precisions less than or equal to 0.4%, short-term correlated errors less than or equal to 0.1%, and long-term correlated errors less than or equal to 0.05%

  2. Neutron environmental characterization requirements for reactor fuels and materials development and surveillance programs

    International Nuclear Information System (INIS)

    McElroy, W.N.; Bennett, R.A.; Johnson, D.L.; Dudey, N.D.

    1975-01-01

    Neutron environmental characterization requirements for reactor fuels and materials development and surveillance programs for Light Water Reactors (LWRs), High Temperature Gas-Cooled Reactors (HTGRs), Fast Breeder Reactors (FBRs), and Controlled Thermonuclear Reactors (CTRs) are varied. Presently, the most demanding requirements are associated with the development of FBRs where goal accuracies in the range of 1 to 3 percent (1sigma) have been requested for the determination of fission rates, burnups, and neutron fluxes and fluences. Total fluence associated with a measured material property change in a fast test reactor can presently be determined in the 5 to 10 percent (1sigma) range, and application of improved dosimetry techniques is expected to reduce this to the 2 to 5 percent (1sigma) range. Without direct dosimetry measurements, however, uncertainties of 25 percent and more are not uncommon. International standardization, development, and application of improved dosimetry methods for reactor materials development and surveillance programs for LWRs, HTGRs, FBRs, and CTRs are essential. The discussion of requirements for neutron environmental characterization for these different reactor concepts is an important aspect of this conference. Here, these requirements are reviewed in light of currently known design, development, testing, and operation considerations for U.S. LMFBR and CTR programs. 97 references. (auth)

  3. FABRICATION AND MATERIAL ISSUES FOR THE APPLICATION OF SiC COMPOSITES TO LWR FUEL CLADDING

    Directory of Open Access Journals (Sweden)

    WEON-JU KIM

    2013-08-01

    Full Text Available The fabrication methods and requirements of the fiber, interphase, and matrix of nuclear grade SiCf/SiC composites are briefly reviewed. A CVI-processed SiCf/SiC composite with a PyC or (PyC-SiCn interphase utilizing Hi-Nicalon Type S or Tyranno SA3 fiber is currently the best combination in terms of the irradiation performance. We also describe important material issues for the application of SiC composites to LWR fuel cladding. The kinetics of the SiC corrosion under LWR conditions needs to be clarified to confirm the possibility of a burn-up extension and the cost-benefit effect of the SiC composite cladding. In addition, the development of end-plug joining technology and fission products retention capability of the ceramic composite tube would be key challenges for the successful application of SiC composite cladding.

  4. Development of fuel cladding material for the next-generation nuclear power reactor

    International Nuclear Information System (INIS)

    Kawaharada, Yoshiyuki; Kano, Fumihisa; Kitano, Koji; Okonogi, Kazunari; Higuchi, Shinichi

    2009-01-01

    The capability for the fuel cladding made of silicon carbide (SiC) fiber-bonded ceramic composite, which consists of a close-packed structure of silicon carbide fiber, was studied. The microstructure was composed of the close-packed structure of the fiber with the interfacial carbon layer. For the thermomechanical properties, SiC composite indicated higher thermal conductivity than zirconium alloy and a relatively high tensile strength. Moreover, fracture toughness of SiC composite was higher than that of other monolithic ceramics owing to fiber layer debonding. The observed characteristics were favorable for SiC composite to be a candidate material applicable to the next-generation reactor. (author)

  5. Methane steam reforming kinetics over Ni-YSZ anode materials for Solid Oxide Fuel Cells

    DEFF Research Database (Denmark)

    Mogensen, David

    of internal reforming has to be carefully controlled. The objective of this thesis is to make such a careful control possible by examining the rate of internal steam reforming in SOFCs. The catalytic steam reforming activity of Ni-YSZ anode material was tested both in a packed bed reactor to determine...... energy. The overall efficiency of a fuel cell system operating on natural gas can be significantly improved by having part of the steam reforming take place inside the SOFC stack. In order to avoid large temperature gradients as a result of the highly endothermal steam reforming reaction, the amount...... accurately predict the steam reforming rate in a stack from the rate expression obtained from the packed bed experiments. During the experiments a previously unreported long term dynamic behavior of the catalyst was observed. After startup, the initial high reactivity was slowly reduced by a factor 5-10 over...

  6. Electrodeposited gold nanoparticles on carbon nanotube-textile: Anode material for glucose alkaline fuel cells

    KAUST Repository

    Pasta, Mauro

    2012-06-01

    In the present paper we propose a new anode material for glucose-gluconate direct oxidation fuel cells prepared by electrodepositing gold nanoparticles onto a conductive textile made by conformally coating single walled carbon nanotubes (SWNT) on a polyester textile substrate. The electrodeposition conditions were optimized in order to achieve a uniform distribution of gold nanoparticles in the 3D porous structure of the textile. On the basis of previously reported studies, the reaction conditions (pH, electrolyte composition and glucose concentration) were tuned in order to achieve the highest oxidation rate, selectively oxidizing glucose to gluconate. The electrochemical characterization was carried out by means of cyclic voltammetry. © 2012 Elsevier B.V. All rights reserved.

  7. Process Knowledge Summary Report for Materials and Fuels Complex Contact-Handled Transuranic Debris Waste

    Energy Technology Data Exchange (ETDEWEB)

    R. P. Grant; P. J. Crane; S. Butler; M. A. Henry

    2010-02-01

    This Process Knowledge Summary Report summarizes the information collected to satisfy the transportation and waste acceptance requirements for the transfer of transuranic (TRU) waste between the Materials and Fuels Complex (MFC) and the Advanced Mixed Waste Treatment Project (AMWTP). The information collected includes documentation that addresses the requirements for AMWTP and the applicable portion of their Resource Conservation and Recovery Act permits for receipt and treatment of TRU debris waste in AMWTP. This report has been prepared for contact-handled TRU debris waste generated by the Idaho National Laboratory at MFC. The TRU debris waste will be shipped to AMWTP for purposes of supercompaction. This Process Knowledge Summary Report includes information regarding, but not limited to, the generation process, the physical form, radiological characteristics, and chemical contaminants of the TRU debris waste, prohibited items, and packaging configuration. This report, along with the referenced supporting documents, will create a defensible and auditable record for waste originating from MFC.

  8. Advanced Ultrafast Spectroscopy for Chemical Detection of Nuclear Fuel Cycle Materials

    Energy Technology Data Exchange (ETDEWEB)

    Villa-Aleman, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Houk, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Spencer, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-09-29

    The development of new signatures and observables from processes related to proliferation activities are often related to the development of technologies. In our physical world, the intensity of observables is linearly related to the input drivers (light, current, voltage, etc.). Ultrafast lasers with high peak energies, opens the door to a new regime where the intensity of the observables is not necessarily linear with the laser energy. Potential nonlinear spectroscopic applications include chemical detection via remote sensing through filament generation, material characterization and processing, chemical reaction specificity, surface phenomena modifications, X-ray production, nuclear fusion, etc. The National Security Directorate laser laboratory is currently working to develop new tools for nonproliferation research with femtosecond and picosecond lasers. Prior to this project, we could only achieve laser energies in the 5 nano-Joule range, preventing the study of nonlinear phenomena. To advance our nonproliferation research into the nonlinear regime we require laser pulses in the milli-Joule (mJ) energy range. We have procured and installed a 35 fs-7 mJ laser, operating at one-kilohertz repetition rate, to investigate elemental and molecular detection of materials in the laboratory with potential applications in remote sensing. Advanced, nonlinear Raman techniques will be used to study materials of interest that are in a matrix of many materials and currently with these nonlinear techniques we can achieve greater than three orders of magnitude signal enhancement. This work studying nuclear fuel cycle materials with nonlinear spectroscopies will advance SRNL research capabilities and grow a core capability within the DOE complex.

  9. 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

  10. High Temperature Steam Oxidation Testing of Candidate Accident Tolerant Fuel Cladding Materials

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Nelson, Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parker, Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parkison, Adam [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2013-12-23

    The Fuel Cycle Research and Development (FCRD) program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels in order to overcome the inherent shortcomings of light water reactor (LWR) fuels when exposed to beyond design basis accident conditions. The campaign has invested in development of experimental infrastructure within the Department of Energy complex capable of chronicling the performance of a wide range of concepts under prototypic accident conditions. This report summarizes progress made at Oak Ridge National Laboratory (ORNL) and Los Alamos National Laboratory (LANL) in FY13 toward these goals. Alternative fuel cladding materials to Zircaloy for accident tolerance and a significantly extended safety margin requires oxidation resistance to steam or steam-H2 environments at ≥1200°C for short times. At ORNL, prior work focused attention on SiC, FeCr and FeCrAl as the most promising candidates for further development. Also, it was observed that elevated pressure and H2 additions had minor effects on alloy steam oxidation resistance, thus, 1 bar steam was adequate for screening potential candidates. Commercial Fe-20Cr-5Al alloys remain protective up to 1475°C in steam and CVD SiC up to 1700°C in steam. Alloy development has focused on Fe-Cr-Mn-Si-Y and Fe-Cr-Al-Y alloys with the aluminaforming alloys showing more promise. At 1200°C, ferritic binary Fe-Cr alloys required ≥25% Cr to be protective for this application. With minor alloy additions to Fe-Cr, more than 20%Cr was still required, which makes the alloy susceptible to α’ embrittlement. Based on current results, a Fe-15Cr-5Al-Y composition was selected for initial tube fabrication and welding for irradiation experiments in FY14. Evaluations of chemical vapor deposited (CVD) SiC were conducted up to 1700°C in steam. The reaction of H2O with the alumina reaction tube at 1700°C resulted in Al(OH)3

  11. Intelligent Materials Used in Hydraulic, Fuel, and Rudder Control Systems of Aircrafts

    Directory of Open Access Journals (Sweden)

    D. B. Chernov

    2017-01-01

    Full Text Available The device is really intelligent, only if it is capable to respond to changing external conditions. The devices, which "feel" the external environment and can change their characteristics, have many advantages compared to the conventional devices: they are more efficient, wear out more slowly, and have lower operating costs.The scope of smart products is truly infinite. Alloys with memory effect also apply to intellectual content. Natural piezoelectric crystals such as silicon dioxide (intellectual material have been known for over a hundred years. They have greater stiffness and can be used at high operating frequencies. Due to the direct piezoelectric effect, they have been successfully used as a strain gage. Later came artificial ceramic piezoelectric materials; they are used as mechanical transducers. Thus, an inverse piezoelectric effect is usually used. It consists in the change of dimensions when an electric field is applied. Control of intellectual structure can be provided by heat fluxes, electromagnetic, hydraulic or piezoelectric forces and through application of electro-rheological, and magneto-rheological fluids. The article examines the intellectual materials and technologies that are already in place or will find its application in aviation hydraulic and fuel systems and control systems of rudders (CSR of aircrafts in the near future.The paper considers in detail the shape memory effect alloys (SMEA as "intelligent" materials. Actuators made from SMEA have a number of advantages: high working power; large recoverable deformation; different types of strain (tensile, compressive, bending and torsional; most specific value of the work per unit mass. All the SMEA advantages may be well used for the so-called thermo-mechanical connections (TMС of pipelines where SMEA drawbacks in this application, practically, do not affect the quality of TMC. In aircraft engineering the TMC were first used in hydraulic systems of the aircraft TU204

  12. Nuclear fuel particles with a core of fissible or fertile material and method of production of such nuclear fuel particles

    International Nuclear Information System (INIS)

    Chin, J.; Luby, C.S.; Mills, R.G.

    1975-01-01

    The method deals with the coating of 4 layers on a fuel core: a first layer of pyrolytic carbon, the second layer of non-penetrable carbon, a third of dense silicon or zirconium carbide, as well as a fourth layer of dense pyrolytic carbon. The precipitation takes place in an atmosphere of hydrocarbons. (UA/LH) [de

  13. Canadian power reactor fuel

    International Nuclear Information System (INIS)

    Page, R.D.

    1976-03-01

    The following subjects are covered: the basic CANDU fuel design, the history of the bundle design, the significant differences between CANDU and LWR fuel, bundle manufacture, fissile and structural materials and coolants used in the CANDU fuel program, fuel and material behaviour, and performance under irradiation, fuel physics and management, booster rods and reactivity mechanisms, fuel procurement, organization and industry, and fuel costs. (author)

  14. Heat Modeling and Material Development of Mg-Based Nanomaterials Combined with Solid Oxide Fuel Cell for Stationary Energy Storage

    Directory of Open Access Journals (Sweden)

    Huaiyu Shao

    2017-11-01

    Full Text Available Mg-based materials have been investigated as hydrogen storage materials, especially for possible onboard storage in fuel cell vehicles for decades. Recently, with the development of large-scale fuel cell technologies, the development of Mg-based materials as stationary storage to supply hydrogen to fuel-cell components and provide electricity and heat is becoming increasingly promising. In this work, numerical analysis of heat balance management for stationary solid oxide fuel cell (SOFC systems combined with MgH2 materials based on a carbon-neutral design concept was performed. Waste heat from the SOFC is supplied to hydrogen desorption as endothermic heat for the MgH2 materials. The net efficiency of this model achieves 82% lower heating value (LHV, and the efficiency of electrical power output becomes 68.6% in minimizing heat output per total energy output when all available heat of waste gas and system is supplied to warm up the storage. For the development of Mg-based hydrogen storage materials, various nano-processing techniques have been widely applied to synthesize Mg-based materials with small particle and crystallite sizes, resulting in good hydrogen storage kinetics, but poor thermal conductivity. Here, three kinds of Mg-based materials were investigated and compared: 325 mesh Mg powers, 300 nm Mg nanoparticles synthesized by hydrogen plasma metal reaction, and Mg50Co50 metastable alloy with body-centered cubic structure. Based on the overall performances of hydrogen capacity, absorption kinetics and thermal conductivity of the materials, the Mg nanoparticle sample by plasma synthesis is the most promising material for this potential application. The findings in this paper may shed light on a new energy conversion and utilization technology on MgH2-SOFC combined concept.

  15. Material control and accountability aspects of safeguards for the USA 233U/TH fuel recycle plant

    International Nuclear Information System (INIS)

    Carpenter, J.A. Jr.; McNeany, S.R.; Angelini, P.; Holder, N.D.; Abraham, L.

    1979-01-01

    Two fuel cycles are considered. The highly enriched uranium (HEU) cycle uses uranium enriched 93% in 235 U as the initial fuel. The medium enriched uranium (MEU) cycle uses uranium with a 235 U enrichment less than 20% as its initial fuel. In both, 233 U is bred from thorium. The HEU 235 U and the 233 U of both cycles are recycled. The MEU 235 U is retired to waste after one reactor cycle. Typical heavy metal contents of spent fuel elements from both cycles are presented. The main functional areas of the recycle plant are Shipping, Receiving, and Storage; Reprocessing; Refabrication; and Waste Treatment. A real-time materials accountability system will manage the data provided by measurements from all four areas. Simulations of material flow used in the HTGR development program are forerunners of such a system. The material control and accountability aspects of Reprocessing and Refabrication only are discussed. The proposed accountability areas are identified and the measurement techniques appropriate to various streams crossing the boundaries of the areas are identified. Special emphasis is placed on novel nondestructive methods developed for assaying solid materials containing 233 U-Th. The material form, total uranium and plutonium, and activity of selected reprocessing streams are listed. The isotopics and activity of the uranium input into Refabrication are also presented

  16. Determination of technical details concerning measures for transportation of nuclear fuel materials in the works or the enterprise

    International Nuclear Information System (INIS)

    1979-01-01

    The determination is defined under the regulations concerning the fabricating business of nuclear fuel materials, the regulation concerning installation and operation of test reactor, the regulations concerning the reprocessing business of spent fuel and the regulations concerning the uses of nuclear source materials. The notification determining technical details concerning measures for transportation of nuclear fuel materials is abolished. Measures for prevention of hazard designated by the Director General of Science and Technology Agency include such ones not to let radioactive materials easily fly about or leak in regular transport, not to let rain water easily penetrate or make each exterior side of a cubic load more than 10 centi-meters. The application for permission shall be filed for transportation of things highly difficult to be sealed in a vessel, listing name and address of the applicant, kind, quantity, form and nature of the load contaminated by nuclear fuel materials, date and route of transfer and measures taken for prevention of hazard in transport. Radiation doses of load and transporting apparatus are stipulated by the Director for an hour as 200 mili-rem on the surface of load, 10 mili-rem at the distance of 1 meter from the surface of load, and 200 mili-rem on the surface of the vehicle, etc. Dangerous things, signals and radiation dose of particular loads are specified respectively. (Okada, K.)

  17. Quantitative determination of uranium distribution homogeneity in material test reactor (MTR) fuel type plates

    International Nuclear Information System (INIS)

    Ferrufino, F.B.J.; Durazzo, M.

    2010-01-01

    IPEN-CNEN/SP produces the fuel to supply its nuclear research reactor IEA-R1. The fuel is assembled from fuel plates containing a core of U 3 Si 2 -Al composite. A good homogeneity in the distribution of uranium in the core of the fuel plate is important from the standpoint of performance under irradiation. Considering the low power reactor IEA-R1, currently the distribution of uranium in the fuel plate is evaluated only by visual inspection of radiographs. However, in view of the possibility of IPEN manufacture the fuel for the new Brazilian Multipurpose Reactor, which has high power, it became urgent to develop a methodology to determine quantitatively the distribution of uranium into the fuel. This paper presents a methodology based on X-ray attenuation, quantifying the distribution of the uranium concentration in the core of the fuel plate by using optical densities of radiographs and comparison with standards. (author)

  18. Characterization of stored defense production spent nulcear fuel and associated materials at Hanford Site, Richland Washington: Environmental assessment

    International Nuclear Information System (INIS)

    1995-03-01

    There are about 2,100 tonnes (2,300 tons) of defense production spent nuclear fuel stored in the 100-K Area Basins located along the south shore of the Columbia River in the northern part of the Hanford Site. Some of the fuel which has been in storage for a number of years is in poor condition and continues to deteriorate. The basins also contain fuel fragments and radioactively contaminated sludge. The DOE needs to characterize defense production spent nuclear fuel and associated materials stored on the Hanford Site. In order to satisfy that need, the Department of Energy (DOE) proposes to select, collect and transport samples of spent nuclear fuel and associated materials to the 327 Building for characterization. As a result of that characterization, modes of interim storage can be determined that would be compatible with the material in its present state and alternative treatment processes could be developed to permit a broader selection of storage modes. Environmental impacts of the proposed action were determined to be limited principally to radiation exposure of workers, which, however, were found to be small. No health effects among workers or the general public would be expected under routine operations. Implementation of the proposed action would not result in any impacts on cultural resources, threatened, endangered and candidate species, air or water quality, socioeconomic conditions, or waste management

  19. Characterization of stored defense production spent nulcear fuel and associated materials at Hanford Site, Richland Washington: Environmental assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

    There are about 2,100 tonnes (2,300 tons) of defense production spent nuclear fuel stored in the 100-K Area Basins located along the south shore of the Columbia River in the northern part of the Hanford Site. Some of the fuel which has been in storage for a number of years is in poor condition and continues to deteriorate. The basins also contain fuel fragments and radioactively contaminated sludge. The DOE needs to characterize defense production spent nuclear fuel and associated materials stored on the Hanford Site. In order to satisfy that need, the Department of Energy (DOE) proposes to select, collect and transport samples of spent nuclear fuel and associated materials to the 327 Building for characterization. As a result of that characterization, modes of interim storage can be determined that would be compatible with the material in its present state and alternative treatment processes could be developed to permit a broader selection of storage modes. Environmental impacts of the proposed action were determined to be limited principally to radiation exposure of workers, which, however, were found to be small. No health effects among workers or the general public would be expected under routine operations. Implementation of the proposed action would not result in any impacts on cultural resources, threatened, endangered and candidate species, air or water quality, socioeconomic conditions, or waste management.

  20. A Brief Description of High Temperature Solid Oxide Fuel Cell’s Operation, Materials, Design, Fabrication Technologies and Performance

    Directory of Open Access Journals (Sweden)

    Muneeb Irshad

    2016-03-01

    Full Text Available Today’s world needs highly efficient systems that can fulfill the growing demand for energy. One of the promising solutions is the fuel cell. Solid oxide fuel cell (SOFC is considered by many developed countries as an alternative solution of energy in near future. A lot of efforts have been made during last decade to make it commercial by reducing its cost and increasing its durability. Different materials, designs and fabrication technologies have been developed and tested to make it more cost effective and stable. This article is focused on the advancements made in the field of high temperature SOFC. High temperature SOFC does not need any precious catalyst for its operation, unlike in other types of fuel cell. Different conventional and innovative materials have been discussed along with properties and effects on the performance of SOFC’s components (electrolyte anode, cathode, interconnect and sealing materials. Advancements made in the field of cell and stack design are also explored along with hurdles coming in their fabrication and performance. This article also gives an overview of methods required for the fabrication of different components of SOFC. The flexibility of SOFC in terms fuel has also been discussed. Performance of the SOFC with varying combination of electrolyte, anode, cathode and fuel is also described in this article.

  1. Designing New Materials for Converting Solar Energy to Fuels via Quantum Mechanics

    Science.gov (United States)

    2014-07-11

    methanol, and formic acid . Methanol can be converted by known means – zeolite catalysis – to energy-dense, higher hydrocarbon fuels. The position of...electrodes that can convert sunlight, carbon dioxide, and water into fuels. We focused on the physics and chemistry of transition metal oxide...largely all hydrogen today is produced from fossil fuels with carbon dioxide as a byproduct. While hydrogen can be used directly as a fuel, a more

  2. Novel approaches to the creation of alternative motor fuels from renewable raw materials

    Directory of Open Access Journals (Sweden)

    Ольга Олександрівна Гайдай

    2016-06-01

    Full Text Available The paper considers the method of obtaining aliphatic alcohols as components of alternative fuels by catalytic processing of synthesis gas under the conditions of mechanochemical activation of the catalyst without using high pressure.It is established that the introduction of hydrocarbon spherical clusters (onions in the alternative fuel changes physical, chemical and chemmotological characteristics of fuel due to the effect of structure formation. The results of comparative studies of the performance properties of hydrocarbon and alternative fuels are displayed

  3. Assessment of radioactive material released from a fuel fabrication plant under accidental conditions

    International Nuclear Information System (INIS)

    1981-01-01

    This report evaluates the amounts of fissile material released both inside and outside a mixed oxide fuel fabrication plant (MOFFP) for light water reactors. The first section begins with a descriptive study of fissile material containment systems, and the methods available for quantifying accident occurrence probabilities. In addition to accidents common to all industrial facilities, other much rarer accidents were considered, such as aircraft crashes. The minimum occurrence probability limit for consideration in this study was set at 10 -6 per annum. The second part of this report attempts to assess the consequences of the accidents considered (i.e. with occurrence probabilities exceeding 10 -6 per annum) by determining maximum values for such accidents. Acts of sabotage and other accidents of this type are beyond the scope of this study and were not taken into consideration. The most serious potential accident would be a fire involving all of the glove boxes in the PuO 2 powder calcination and preparation cell, which could release 76.5 mg of PuO 2 powder into the atmosphere; the occurrence probability of such an accident, however, is slight (less than 10 -5 per annum). The second possibility, is a specially nuclear hazard that would release fission products into the atmosphere. The occurrence probability of such an accident is currently evaluated at 10 -3 per annum

  4. Leaching studies of heavy concrete material for nuclear fuel waste immobilization containers

    International Nuclear Information System (INIS)

    Onofrei, M.; Raine, D.; Brown, L.; Hooton, R.D.

    1989-08-01

    The leaching behaviour of a high-density concrete was studied as part of a program to evaluate its potential use as a container material for nuclear fuel waste under conditions of deep geologic disposal. Samples of concrete material were leached in deionized distilled water, Standard Canadian Shield Saline Solution (SCSSS), SCSSS plus 20% Na-bentonite, and SCSSS plus granite and 20% Na-bentonite under static conditions at 100 degrees celsius for periods up to 365 days. The results of these leaching experiments suggest that the stability of concrete depends on the possible internal structural changes due to hydration reactions of unhydrated components, leading to the formation of C-S-H gel plus portlandite (Ca(OH) 2 ). The factors controlling the concrete leaching process were the composition of the leachant and the concentration of elements in solution capable of forming precipitates on the concrete surface, e.g., silicon, Mg 2+ and Ca 2+ . The main effect observed during leaching was an increase in groundwater pH (from 7 to 9). However, the addition of Na-bentonite suppressed the normal tendency of the pH of the groundwater in contact with concrete to rise rapidly. It was shown that the solution concentration of elements released from the concrete, particularly potassium, increased in the presence of Na-bentonite

  5. Automation of potentiometric titration for the determination of uranium in nuclear fuel materials

    International Nuclear Information System (INIS)

    Kelkar, Anoop; Pandey, Ashish; Kapoor, Y.S.; Kumar, Manish; Singh, Mamta; Fulzele, Ajeet; Prakash, Amrit; Afzal, Mohd; Panakkal, J.P.

    2010-01-01

    Advanced Fuel Fabrication Facility is fabricating various types of mixed oxide fuels, namely for PHWR, BWR, FBTR and PFBR. Precise determination of uranium in MOX fuel sample is important to get desired burn up in the reactor. The modified Davies and Gray method is routinely used for the potentiometric titration of uranium

  6. Use of Friction Stir Welding and Friction Stir Processing for Advanced Nuclear Fuels and Materials Joining Applications

    International Nuclear Information System (INIS)

    J. I. Cole; J. F. Jue

    2006-01-01

    Application of the latest developments in materials technology may greatly aid in the successful pursuit of next generation reactor and transmutation technologies. One such area where significant progress is needed is joining of advanced fuels and materials. Rotary friction welding, also referred to as friction stir welding (FSW), has shown great promise as a method for joining traditionally difficult to join materials such as aluminum alloys. This relatively new technology, first developed in 1991, has more recently been applied to higher melting temperature alloys such as steels, nickel-based and titanium alloys. An overview of the FSW technology is provided and two specific nuclear fuels and materials applications where the technique may be used to overcome limitations of conventional joining technologies are highlighted

  7. Material synthesis and fabrication method development for intermediate temperature solid oxide fuel cells

    Science.gov (United States)

    Ding, Hanping

    Solid oxide fuel cells (SOFCs) are operated in high temperature conditions (750-1000 °C). The high operating temperature in turn may lead to very complicated material degradation issues, significantly increasing the cost and reducing the durability of SOFC material systems. In order to widen material selections, reduce cost, and increase durability of SOFCs, there is a growing interest to develop intermediate temperature SOFCs (500-750 °C). However, lowering operating temperature will cause substantial increases of ohmic resistance of electrolyte and polarization resistance of electrodes. This dissertation aimed at developing high-performance intermediate-temperature SOFCs through the employment of a series of layered perovskite oxides as novel cathode materials to minimize the potential electrode polarization on oxygen reduction reaction resulting from the unique crystal structure. The high performance of such perovskites under lower temperatures lies in the fact that a simple cubic perovskite with randomly occupied A-sites transforming into a layered compound with ordered lanthanide and alkali-earth cations may reduce the oxygen bonding strength and provide disorder-free channels for oxygen ion migrations. In order to compromise the cell performance and chemical and mechanical stability, the substitution of Fe in B site was comprehensively investigated to explore the effects of Fe doping on the crystal structure, thermal and electrical properties, as well as electrochemical performance. Furthermore, a platinum nanowire network was successfully developed as an ultrathin electrochemically efficient current collector for SOFCs. The unique platinum network on cathode surface can connect the oxygen reduction reaction (ORR) sites at the nano-scale to the external circuit while being able to substantially avoid blocking the open pores of the cathode. The superior electrochemical performance was exhibited, including the highly reduced electrode polarization resistance

  8. Implementation and evaluation of fuel creep using advanced light-water reactor materials in FRAPCON 3.5

    Science.gov (United States)

    Carroll, Spencer

    As current reactors approach the end of their operable lifetime, new reactors are needed if nuclear power is to continue being generated in the United States. Some utilities have already began construction on newer, more advanced LWR reactors, which use the same fuel as current reactors and have a similar but updated design. Others are researching next generation (GEN-IV) reactors which have new designs that utilize alternative fuel, coolants and other reactor materials. Many of these alternative fuels are capable of achieving higher burnups and are designed to be more accident tolerant than the currently used UO2 fuel. However, before these new materials can be used, extensive research must be done in order to obtain a detailed understanding of how the new fuels and other materials will interact. New fuels, such as uranium nitride (UN) and uranium carbide (UC) have several advantages over UO2, such as increased burnup capabilities and higher thermal conductivities. However, there are issues with each that prevent UC and UN from being used as direct replacements for UO2. Both UC and UN swell at a significantly higher rate than UO2 and neither fuel reacts favorably when exposed to water. Due to this, UC and UN are being considered more for GEN-IV reactors that use alternative coolant rather than for current LWRs. In an effort to increase accident tolerance, silicon carbide (SiC) is being considered for use as an alternative cladding. The high strength, high melting point and low oxidation of SiC make it an attractive cladding choice, especially in an accident scenario. However, as a ceramic, SiC is not ductile and will not creep outwards upon pellet-clad mechanical interaction (PCMI) which can cause a large build up in interfacial pressure. In order to understand the interaction between the high swelling fuels and unyielding SiC cladding, data on the properties and behaviors of these materials must be gathered and incorporated into FRAPCON. FRAPCON is a fuel

  9. MASSAHAKE whole tree harvesting method for pulp raw-material and fuel -- R&D in 1993--1998

    Energy Technology Data Exchange (ETDEWEB)

    Asplund, D.A.; Ahonen, M.A. [Technical Research Centre of Finland, Jyvaeskylae (Finland)

    1993-12-31

    In Finland biofuels and hydropower are the only indigenous fuels available. Peat, wood and wood derived fuels form about 18% of total primary energy requirement. The largest wood and wood fuel user in Finland is wood processing industry, paper, pulp, sawmills. Due to silvicultural activities the growth of forests has developed an instant need for first thinnings. This need is about 12% of total stem wood growth. With conventional harvesting methods this would produce about 8 mill. m{sup 3} pulp raw material and 2 mill. m{sup 3} wood fuel. By using integrated harvesting methods about 12 mill. m{sup 3} pulp raw material and 8 mill. m{sup 3} (about 1, 3 mill. toe) fuel could be produced. At the moment, there is no economically profitable method for harvesting first thinning trees for industrial use or energy production. Hence, there are a few ongoing research projects aiming at solving the question of integrated harvesting. MASSAHAKE chip purification method has been under R&D since 1987. Research with continuous experimental line (capacity 5--10 loose-m{sup 3}) has been done in 1991 and 1992. The research has concentrated on pine whole tree chip treatment, but preliminary tests with birch whole tree chips has been done. The experiment line will be modified for birth whole tree chips during 1993. Based on the research results more than 60% of the whole tree chips can be separated to pulp raw material with < 1% bark content. This amount is 1.5--2 times more than with present technology. The yield of fuel fraction is 2--4 times higher compared to present methods. Fuel fraction is homogeneous and could be used in most furnaces for energy production. By replacing fossil fuels with wood fuel in energy production it is possible to reduce CO{sub 2}-emissions significantly. This paper presents the wood fuel research areas in Finland and technical potential of MASSAHAKE-method including the plant for building a demonstration plant based on this technology.

  10. Standard format and content for emergency plans for fuel-cycle and materials facilities: Draft report for comment

    International Nuclear Information System (INIS)

    1987-11-01

    This report is issued as guidance to those fuel-cycle and major materials licensees who are required by the NRC to prepare and submit an emergency plan. This Standard Format has been prepared to help ensure uniformity and completeness in the preparation of those plans

  11. Analysis of material effect in molten fuel-coolant interaction, comparison of thermodynamic calculations and experimental observations

    Czech Academy of Sciences Publication Activity Database

    Tyrpekl, Václav; Piluso, P.

    2012-01-01

    Roč. 46, AUGUST (2012), s. 197-203 ISSN 0306-4549 Institutional support: RVO:61388980 Keywords : Nuclear reactor severe accident * Fuel -Coolant Interaction * Material effect * Steam explosion Subject RIV: CA - Inorganic Chemistry Impact factor: 0.800, year: 2012

  12. Applying distance-to-target weighing methodology to evaluate the environmental performance of bio-based energy, fuels, and materials

    NARCIS (Netherlands)

    Weiss, M.; Patel, M.K.; Heilmeier, H.; Bringezu, S.

    2007-01-01

    The enhanced use of biomass for the production of energy, fuels, and materials is one of the key strategies towards sustainable production and consumption. Various life cycle assessment (LCA) studies demonstrate the great potential of bio-based products to reduce both the consumption of

  13. Analysis of alternative transportation methods for radioactive materials shipments including the use of special trains for spent fuel and wastes

    International Nuclear Information System (INIS)

    Smith, D.R.; Luna, R.E.; Taylor, J.M.

    1978-01-01

    Two studies were completed which evaluate the environmental impact of radioactive material transport. The first was a generic study which evaluated all radioactive materials and all transportation modes; the second addressed spent fuel and fuel-cycle wastes shipped by truck, rail and barge. A portion of each of those studies dealing with the change in impact resulting from alternative shipping methods is presented in this paper. Alternatives evaluated in each study were mode shifts, operational constraints, and, in generic case, changes in material properties and package capabilities. Data for the analyses were obtained from a shipper survey and from projections of shipments that would occur in an equilibrium fuel cycle supporting one hundred 1000-MW(e) reactors. Population exposures were deduced from point source radiation formulae using separation distances derived for scenarios appropriate to each shipping mode and to each exposed population group. Fourteen alternatives were investigated for the generic impact case. All showed relatively minor changes in the overall radiological impact. Since the radioactive material transport is estimated to be fewer than 3 latent cancer fatalities (LCF) for each shipment year (compared to some 300,000 yearly cancer fatalities or 5000 LCF's calculated for background radiation using the same radiological effects model), a 15% decrease caused by shifting from passenger air to cargo air is a relatively small effect. Eleven alternatives were considered for the fuel cycle/special train study, but only one produced a reduction in total special train baseline LCF's (.047) that was larger than 5%

  14. Scientific Assessment in support of the Materials Roadmap enabling Low Carbon Energy Technologies: Hydrogen and Fuel Cells

    DEFF Research Database (Denmark)

    Cerri, I.; Lefebvre-Joud, F.; Holtappels, Peter

    A group of experts from European research organisations and industry have assessed the state of the art and future needs for materials' R&D for hydrogen and fuel cell technologies. The work was performed as input to the European Commission's roadmapping exercise on materials for the European...... Strategic Energy Technology Plan. The report summarises the results, including key targets identified for medium term (2020/2030) and long term (2050) timescales....

  15. Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2008-01-01

    The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U-Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease

  16. Functionalisation of mesoporous materials for application as additives in high temperature PEM fuel cell membranes

    Energy Technology Data Exchange (ETDEWEB)

    Sharifi, Monir

    2012-03-06

    The presented thesis contains six original research articles dedicated to the preparation and characterization of organic-inorganic mesoporous materials as additives for polymer electroly1e membrane fuel cells (PEMFCs). The mesoporous materials Si-MCM-41 and benzene-PMO (periodic mesoporous organosilica) were chosen for the investigations. These materials were modified with functional groups for enhanced proton conductivity and water-keeping properties. In order to improve these materials Broenstedt acidic groups were introduced in the framework of mesoporous Si-MCM-41. Therefore, some silicium atoms in the framework were substituted by aluminium using different aluminium sources. Here NaAlO{sub 2} exhibits clearly the best results because the entire aluminium incorporated within the framework is tetragonally coordinated as observed by {sup 2}7AI MAS NMR. The increase of the proton conductivities results from an improved hydrophilicity, a decreased particle size, and newly introduced Broenstedt acidity in the mesoporous Al-MCM-41. However, mesoporous Si-MCM-41 materials functionalised by co-condensation with sulphonic acid groups exhibit the best results concerning proton conductivity, compared to those prepared by grafting. Hence, these materials where characterized in more detail by SANS and by MAS NMR measurements. The first one indicated that by co-condensation the entire inner pore surface is altered by functional groups which are, thus, distributed much more homogeneously than samples functionalised by grafting. This result explains the improved proton conductivities. Additionally, {sup 2}9Si NMR spectra proved that samples prepared by co-condensation lead to a successful and almost complete incorporation of mercaptopropyltrimethoxysilan (MPMS) into the mesoporous framework. Furthermore, it was shown by {sup 1}3C MAS NMR spectroscopy that the majority of the organic functional groups remained intact after H{sub 2}0{sub 2}-oxidation. However, proton

  17. Foreign materials in a deep repository for spent nuclear fuels; Fraemmande material i ett djupfoervar foer anvaent kaernbraensle

    Energy Technology Data Exchange (ETDEWEB)

    Jones, C.; Christiansson, Aa.; Wiborgh, M. [Kemakta Konsult AB, Stockholm (Sweden)

    1999-12-01

    The effects of foreign substances introduced into a spent-fuel repository are reviewed. Possible impacts on processes and barrier-functions are examined, and the following areas are identified: Corrosion of the spent-fuel canister through the presence of sulfur and substances that favor microbial growth; impacts on the bentonite properties through the presence of cations as calcium, potassium and iron; radionuclide transport through the presence of complex-formers and surface-active substances.

  18. Laboratory Investigations of the High Temperature Corrosion of Various Materials in Simulated oxy-fuel and Conventional Coal Firing

    International Nuclear Information System (INIS)

    Folkeson, N.; Pettersson, J.; Svensson, J.E.; Hjornhede, A.; Montgomery, M.; Bjurman, M.

    2009-01-01

    Laboratory exposures in horizontal tube furnaces were conducted to test various materials for corrosion resistance in simulated oxy-fuel firing and conventional coal firing environments. Two different exposures were done at 630 C for 672 hours. The reaction atmosphere, consisting of CO 2 , H 2 O, O 2 , N 2 and SO 2 , was mixed to resemble that of oxy-fuel firing in the first exposure and that of conventional coal firing in the second exposure (N 2 was added during the second exposure only). Four different materials were tested in the first exposure; Sanicro 63, Alloy 800HT, 304L and 304HCu. In the second exposure four different materials were tested; 304L, Alloy 800HT, Kanthal APMT and NiCrAl. Apart from cleaned sample coupons, some samples pre-exposed in a test rig under oxy-fuel conditions with lignite as fuel and some pre-exposed with bituminous coal as fuel were investigated in the first exposure. In the second exposure some samples were pre-exposed in a rig under conventional firing conditions with lignite as fuel. The corrosion attack on the investigated samples was analysed by gravimetry, x-ray diffraction (XRD) and scanning electron microscopy (SEM) with energy dispersive x-ray (EDX). The SEM/EDX analysis was made on both the sample envelope and metallographic cross sections of the samples. The results show that there is small difference in the corrosion attack between the two environments. There was also little difference in oxide morphology and composition between cleaned samples and pre-exposed samples of the same material. The austenitic chromia former 304HCu suffered the most extensive corrosion attack in the oxy-fuel environment. In the conventional air firing environment 304L showed the highest mass gain. Chromia formers with higher chromium concentrations performed better, especially the super austenitic Alloy 800HT, with its high chromium concentration, formed a thin and protective corundum type oxide. The nickel based Sanicro 63 showed very low

  19. Spent Fuel Working Group report on inventory and storage of the Department's spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities

    International Nuclear Information System (INIS)

    1993-11-01

    A self assessment was conducted of those Hanford facilities that are utilized to store Reactor Irradiated Nuclear Material, (RINM). The objective of the assessment is to identify the Hanford inventories of RINM and the ES ampersand H concerns associated with such storage. The assessment was performed as proscribed by the Project Plan issued by the DOE Spent Fuel Working Group. The Project Plan is the plan of execution intended to complete the Secretary's request for information relevant to the inventories and vulnerabilities of DOE storage of spent nuclear fuel. The Hanford RINM inventory, the facilities involved and the nature of the fuel stored are summarized. This table succinctly reveals the variety of the Hanford facilities involved, the variety of the types of RINM involved, and the wide range of the quantities of material involved in Hanford's RINM storage circumstances. ES ampersand H concerns are defined as those circumstances that have the potential, now or in the future, to lead to a criticality event, to a worker radiation exposure event, to an environmental release event, or to public announcements of such circumstances and the sensationalized reporting of the inherent risks

  20. Modeled salt density for nuclear material estimation in the treatment of spent nuclear fuel

    Science.gov (United States)

    Mariani, Robert D.; Vaden, DeeEarl

    2010-09-01

    Spent metallic nuclear fuel is being treated in a pyrometallurgical process that includes electrorefining the uranium metal in molten eutectic LiCl-KCl as the supporting electrolyte. We report a model for determining the density of the molten salt. Material balances account for the net mass of salt and for the mass of actinides present. It was necessary to know the molten salt density, but difficult to measure. It was also decided to model the salt density for the initial treatment operations. The model assumes that volumes are additive for the ideal molten salt solution as a starting point; subsequently, a correction factor for the lanthanides and actinides was developed. After applying the correction factor, the percent difference between the net salt mass in the electrorefiner and the resulting modeled salt mass decreased from more than 4.0% to approximately 0.1%. As a result, there is no need to measure the salt density at 500 °C for inventory operations; the model for the salt density is found to be accurate.

  1. Thermal expansion coefficient prediction of fuel-cell seal materials from silica sand

    Science.gov (United States)

    Hidayat, Nurul; Triwikantoro, Baqiya, Malik A.; Pratapa, Suminar

    2013-09-01

    This study is focused on the prediction of coefficient of thermal expansion (CTE) of silica-sand-based fuel-cell seal materials (FcSMs) which in principle require a CTE value in the range of 9.5-12 ppm/°C. A semi-quantitative theoretical method to predict the CTE value is proposed by applying the analyzed phase compositions from XRD data and characterized density-porosity behavior. A typical silica sand was milled at 150 rpm for 1 hour followed by heating at 1000 °C for another hour. The sand and heated samples were characterized by means of XRD to perceive the phase composition correlation between them. Rietveld refinement was executed to investigate the weight fraction of the phase contained in the samples, and then converted to volume fraction for composite CTE calculations. The result was applied to predict their potential physical properties for FcSM. Porosity was taken into account in the calculation after which it was directly measured by the Archimedes method.

  2. Distribution of 60Co and 54Mn in graphite material of irradiated HTGR fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Kikuchi, Teruo; Kobayashi, Fumiaki; Minato, Kazuo; Fukuda, Kousaku; Ikawa, Katsuichi; Iwamoto, Kazumi

    1984-05-01

    Distribution of 60 Co and 54 Mn was measured in the graphite sleeves and blocks of the third and fourth HTGR fuel assemblies irradiated in the Oarai Gas Loop-1 (OGL-1), which is a high temperature inpile gas loop installed in the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Research Institute (JAERI). Axial and circumferential profiles were obtained by gamma spectrometry, and radial profiles by lathe sectioning with gamma spectrometry. Distribution of 60 Co is in good agreement with that of thermal neutron flux, and the Co content in the graphite is estimated to be -- 1 x 10 -9 in weight fraction. Concentration of 54 Mn decreases toward the axial center in its axial profile, and radially is almost uniform inside and appreciably higher at free surfaces. An estimated Fe content of --10 -8 in wight fraction is smaller by two orders of magnitude than that from chemical analysis. Higher concentraion of 60 Co and 54 Mn at the free surfaces suggests the importance of transportation process of these nuclides in the coolant loop. (author)

  3. Routing and scheduling of hazardous materials shipments: algorithmic approaches to managing spent nuclear fuel transport

    International Nuclear Information System (INIS)

    Cox, R.G.

    1984-01-01

    Much controversy surrounds government regulation of routing and scheduling of Hazardous Materials Transportation (HMT). Increases in operating costs must be balanced against expected benefits from local HMT bans and curfews when promulgating or preempting HMT regulations. Algorithmic approaches for evaluating HMT routing and scheduling regulatory policy are described. A review of current US HMT regulatory policy is presented to provide a context for the analysis. Next, a multiobjective shortest path algorithm to find the set of efficient routes under conflicting objectives is presented. This algorithm generates all efficient routes under any partial ordering in a single pass through the network. Also, scheduling algorithms are presented to estimate the travel time delay due to HMT curfews along a route. Algorithms are presented assuming either deterministic or stochastic travel times between curfew cities and also possible rerouting to avoid such cities. These algorithms are applied to the case study of US highway transport of spent nuclear fuel from reactors to permanent repositories. Two data sets were used. One data set included the US Interstate Highway System (IHS) network with reactor locations, possible repository sites, and 150 heavily populated areas (HPAs). The other data set contained estimates of the population residing with 0.5 miles of the IHS and the Eastern US. Curfew delay is dramatically reduced by optimally scheduling departure times unless inter-HPA travel times are highly uncertain. Rerouting shipments to avoid HPAs is a less efficient approach to reducing delay

  4. Boiling performance and material robustness of modified surfaces with multi scale structures for fuel cladding development

    International Nuclear Information System (INIS)

    Jo, HangJin; Kim, Jin Man; Yeom, Hwasung; Lee, Gi Cheol; Park, Hyun Sun; Kiyofumi, Moriyama; Kim, Moo Hwan; Sridharan, Kumar; Corradini, Michael

    2015-01-01

    Highlights: • We improved boiling performance and material robustness using surface modification. • We combined micro/millimeter post structures and nanoparticles with heat treatments. • Compactly-arranged micrometer posts had improved boiling performance. • CHF increased significantly due to capillary pumping by the deposited NP layers. • Sintering procedure increased mechanical strength of the NP coating surface. - Abstract: By regulating the geometrical characteristics of multi-scale structures and by adopting heat treatment for protective layer of nanoparticles (NPs), we improved critical heat flux (CHF), boiling heat transfer (BHT), and mechanical robustness of the modified surface. We fabricated 1-mm and 100-μm post structures and deposited NPs on the structured surface as a nano-scale structured layer and protective layer at the same time, then evaluated the CHF and BHT and material robustness of the modified surfaces. On the structured surfaces without NPs, the surface with compactly-arranged micrometer posts had improved CHF (118%) and BHT (41%). On the surface with structures on which NPs had been deposited, CHF increased significantly (172%) due to capillary pumping by the deposited NP layers. The heat treatment improved robustness of coating layer in comparison to the one of before heat treatment. In particular, low-temperature sintering increased the hardness of the modified surface by 140%. The increased mechanical strength of the NP coating is attributed to reduction in coating porosity during sintering. The combination of micrometer posts structures and sintered NP coating can increase the safety, efficiency and reliability of advanced nuclear fuel cladding

  5. Determination of the details of technical standard concerning the transport of nuclear fuel materials outside the works or the enterprise

    International Nuclear Information System (INIS)

    1979-01-01

    The determination is defined under the regulations concerning the transport of nuclear fuel materials outside the works or the enterprise. The least dangerous nuclear fuel materials designated by the Director General of Science and Technology Agency include, besides such materials listed and corresponding to the particular terms, natural or depleted uranium, natural thorium and their compounds whose surface is covered by corrosion resistance metal. The permissible density of surface stipulated by the Director is 1/100,000 micro-curie per centi-meter 2 for radioactive substances not emitting alpha rays and 1/10,000 micro-curie per centi-meter 2 for radioactive substances emitting alpha rays. General test conditions of A, BM and BU type of loads are in detail provided for. The nuclear fuel materials defined by the Director not as fission materials are uranium 233 and 235, plutonium 238 or 241, whose quantity or combined quantity of such substances do not exceed 15 grams and natural or depleted uranium. The permissible accumulative dose specified for employees is the figure (unit rem) calculated by the formula D=5 (N-18), when D means the permissible accumulative dose and N number of the age. The permissible exposure dose shall be 3 rem for 3 months for employees and 1.5 rem for a year for persons other than employees. (Okada, K.)

  6. Thermal Properties of Green Fuel Briquettes from Residue Corncobs Materials Mixed Macadamia Shell Charcoal Powder

    Science.gov (United States)

    Teeta, Suminya; Nachaisin, Mali; Wanish, Suchana

    2017-09-01

    The objective of this research was to produce green fuel briquettes from corncobs by adding macadamia shell charcoal powder. The study was sectioned into 3 parts: 1) Quality improvement of green fuel briquettes by adding macadamia; 2) Fuel property analysis based on ASTM standards and thermal fuel efficiency; and 3) Economics appropriateness in producing green fuel briquettes. This research produced green fuel briquettes using the ratio of corncobs weight and macadamia shell charcoal powder in 100:0 90:10 80:20 70:30 60:40 and 50:50 and pressing in the cold briquette machine. Fuel property analysis showed that green fuel briquettes at the ratio 50:50 produced maximum heating values at 21.06 Megajoule per kilogram and briquette density of 725.18 kilograms per cubic meter, but the percent of moisture content, volatile matter, ash, and fixed carbon were 10.09, 83.02, 2.17 and 4.72 respectively. The thermal efficiency of green fuel briquettes averaged 20.22%. Economics appropriateness was most effective where the ratio of corncobs weight to macadamia shell charcoal powder was at 50:50 which accounted for the cost per kilogram at 5.75 Baht. The net present value was at 1,791.25 Baht. Internal rate of return was at 8.62 and durations for a payback period of investment was at 1.9 years which was suitable for investment.

  7. 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.

  8. Effects of radiation and environmental factors on the durability of materials in spent fuel storage and disposal

    International Nuclear Information System (INIS)

    2002-12-01

    This is the second report that addresses results from the Coordinated Research Project (CRP) on Irradiation Enhanced Degradation of Materials in Spent Fuel Storage Facilities. This second report addresses results of topical studies that are relevant to issues important to materials behaviour in wet storage technology, but also involves topics on materials behaviour in dry storage and repository environments, including effects of radiation. The material is in seven separate papers contributed by the participants in the CRP and contains details of research studies started within the framework of the CRP and in several cases completed well after the CRP was finished. The seven contributions fall into three broad subject areas: Effects of temperature and radiation on aqueous and moist air corrosion of stainless steels; Studies of materials behaviour in wet and dry storage; Effects of gamma radiation on the durability of candidate canister materials for repository applications: carbon steel, titanium, and copper. Each of the papers has been indexed separately

  9. Hazardous Materials Verification and Limited Characterization Report on Sodium and Caustic Residuals in Materials and Fuel Complex Facilities MFC-799/799A

    Energy Technology Data Exchange (ETDEWEB)

    Gary Mecham

    2010-08-01

    This report is a companion to the Facilities Condition and Hazard Assessment for Materials and Fuel Complex Sodium Processing Facilities MFC-799/799A and Nuclear Calibration Laboratory MFC-770C (referred to as the Facilities Condition and Hazards Assessment). This report specifically responds to the requirement of Section 9.2, Item 6, of the Facilities Condition and Hazards Assessment to provide an updated assessment and verification of the residual hazardous materials remaining in the Sodium Processing Facilities processing system. The hazardous materials of concern are sodium and sodium hydroxide (caustic). The information supplied in this report supports the end-point objectives identified in the Transition Plan for Multiple Facilities at the Materials and Fuels Complex, Advanced Test Reactor, Central Facilities Area, and Power Burst Facility, as well as the deactivation and decommissioning critical decision milestone 1, as specified in U.S. Department of Energy Guide 413.3-8, “Environmental Management Cleanup Projects.” Using a tailored approach and based on information obtained through a combination of process knowledge, emergency management hazardous assessment documentation, and visual inspection, this report provides sufficient detail regarding the quantity of hazardous materials for the purposes of facility transfer; it also provides that further characterization/verification of these materials is unnecessary.

  10. Compatibility Study for Plastic, Elastomeric, and Metallic Fueling Infrastructure Materials Exposed to Aggressive Formulations of Ethanol-blended Gasoline

    Energy Technology Data Exchange (ETDEWEB)

    Kass, Michael D [ORNL; Pawel, Steven J [ORNL; Theiss, Timothy J [ORNL; Janke, Christopher James [ORNL

    2012-07-01

    In 2008 Oak Ridge National Laboratory began a series of experiments to evaluate the compatibility of fueling infrastructure materials with intermediate levels of ethanol-blended gasoline. Initially, the focus was elastomers, metals, and sealants, and the test fuels were Fuel C, CE10a, CE17a and CE25a. The results of these studies were published in 2010. Follow-on studies were performed with an emphasis on plastic (thermoplastic and thermoset) materials used in underground storage and dispenser systems. These materials were exposed to test fuels of Fuel C and CE25a. Upon completion of this effort, it was felt that additional compatibility data with higher ethanol blends was needed and another round of experimentation was performed on elastomers, metals, and plastics with CE50a and CE85a test fuels. Compatibility of polymers typically relates to the solubility of the solid polymer with a solvent. It can also mean susceptibility to chemical attack, but the polymers and test fuels evaluated in this study are not considered to be chemically reactive with each other. Solubility in polymers is typically assessed by measuring the volume swell of the polymer exposed to the solvent of interest. Elastomers are a class of polymers that are predominantly used as seals, and most o-ring and seal manufacturers provide compatibility tables of their products with various solvents including ethanol, toluene, and isooctane, which are components of aggressive oxygenated gasoline as described by the Society of Automotive Engineers (SAE) J1681. These tables include a ranking based on the level of volume swell in the elastomer associated with exposure to a particular solvent. Swell is usually accompanied by a decrease in hardness (softening) that also affects performance. For seal applications, shrinkage of the elastomer upon drying is also a critical parameter since a contraction of volume can conceivably enable leakage to occur. Shrinkage is also indicative of the removal of one or more

  11. Study of the corrosion of AA 6061 in spent fuel materials

    International Nuclear Information System (INIS)

    Rodriguez, Sebastian; Haddad, Roberto; Lanzani, Liliana A.

    2003-01-01

    Localized attack induced by dust or other particles deposited on alloy AA 6061 surface under water has been addressed as a matter of concern after completion of an IAEA Coordinated Research Program (CRP) on the corrosion of aluminum clad spent fuel during storage in water basins. This attack has been observed in all kinds of waters, although it is more pronounced in those of higher conductivity. In these cases a strong attack (similar to pitting corrosion) up to several hundred microns in depth and about a millimeter in length has been found beneath the aluminum hydroxide blister formed in those places where specks had seated on. As this problem could seriously affect the fission product containing capacity of stored spent fuel even in well maintained high quality water, it is important to learn about the involved mechanism of attack and find out about the influence of particle composition, in order to establish the convenience or disapproval of the use of materials and procedures in storage basins. With this objective, an experimental approach has been developed to study the mechanism of corrosion attack linked with the deposition of particles of different composition on aluminum surfaces; this include two kinds of iron flakes, concrete powder, and sand particles. Immersion tests of up to 40 days have been conducted in nuclear grade demineralized water and sodium citrate solutions of several conductivities. The position of sediments was marked and followed through the process and the final state of the aluminum surfaces was assessed by optical and electron microscopy and other microanalysis techniques. Other complementary activities carried on in relation with this work are: through characterization of intermetallic particles in AA 6061, and the study of the electrochemical behavior of precipitates in high purity water. Mg 2 Si particles perform very actively, dissolving even in high pure water at open circuit potential, leaving a small hole on the surface. Iron

  12. Safety of some fuel cladding materials, alternative to Zr-alloys

    International Nuclear Information System (INIS)

    Hache, Georges; Clement, Bernard; Barrachin, Marc

    2013-01-01

    The Fukushima accident underlined the impact of hydrogen production on LWR core melt accident behaviour. New fuel cladding and structural materials are under development by the industry. IRSN performed a bibliographic study on the behaviour of these materials during LWR core melt accidents. Method This presentation is focused on cladding oxidation by steam and more precisely on: - number of H 2 moles produced per cladding length unit at thermochemical equilibrium; - oxidation kinetics; - heat of reaction; - physic-chemical interactions between material or oxidation products and fuel. Silicon carbide (SiC) - During SiC oxidation by steam, nearly 3 times more explosive gases (CO+H 2 ) moles are produced per cladding length unit at thermochemical equilibrium than for Zr-alloys. - SiC oxidation kinetics below 1700 deg. C: According to early tests performed by NASA and ORNL, the oxidation is linear but slow, there is an effective protection by a thin vitreous SiO 2 layer; these tests underlined the importance of the steam pressure and flow rate. Recently, published MIT and ORNL tests confirm that under large break LOCA conditions (∼5 bars) and up to 1200 deg. C, SiC recession is much slower than for Zr-alloys. Tests under small break conditions (3 inches LOCA: ∼40 bars) were not performed or not published. - SiC oxidation kinetics above 1700 deg. C (melting point of SiO 2 ): Molten SiO 2 loses its protective effect; this is known in the literature as 'catastrophic oxidation by molten oxides'. There will be a cliff-edge effect. For un-inerted containments, H 2 recombiners will be saturated, leading to a risk of CO+H 2 explosion in these containments. - During SiC oxidation by steam, the heat of reaction produced per cladding length unit at thermochemical equilibrium is of the same order of magnitude as for Zr alloys. Molten SiO 2 will interact with UO 2 to form molten mixtures at temperatures well below UO 2 melting temperature. - Calculations were

  13. Organic-inorganic hybrid nanostructured materials for photovoltaics and solar fuels

    NARCIS (Netherlands)

    Lai, Lai-Hung

    2016-01-01

    The hydrogen economy aiming to use hydrogen as a new potential fuel for motive power has been proposed as a promising model for this century. However, until now most of the H2 in use still comes from steam reforming which produces H2 via steam reaction at high temperature with fossil fuel. Solar

  14. Homogeneous forming technology of composite materials and its application to dispersion nuclear fuel

    International Nuclear Information System (INIS)

    Hong, Soon Hyun; Ryu, Ho Jin; Sohn, Woong Hee; Kim, Chang Kyu

    1997-01-01

    Powder metallurgy processing technique of metal matrix composites is reviewed and its application to process homogeneous dispersion nuclear fuel is considered. The homogeneous mixing of reinforcement with matrix powders is very important step to process metal matrix composites. The reinforcement with matrix powders is very important step to process metal matrix composites. The reinforcement can be ceramic particles, whiskers or chopped fibers having high strength and high modulus. The blended powders are consolidated into billets and followed by various deformation processing, such as extrusion, forging, rolling or spinning into final usable shapes. Dispersion nuclear fuel is a class of metal matrix composite consisted of dispersed U-compound fuel particles and metallic matrix. Dispersion nuclear fuel is fabricated by powder metallurgy process such as hot pressing followed by hot extrusion, which is similar to that of SiC/Al metal matrix composite. The fabrication of homogeneous dispersion nuclear fuel is very difficult mainly due to the inhomogeneous mixing characteristics of the powders from quite different densities between uranium alloy powders and aluminum powders. In order to develop homogeneous dispersion nuclear fuel, it is important to investigate the effect of powder characteristics and mixing techniques on homogeneity of dispersion nuclear fuel. An new quantitative analysis technique of homogeneity is needed to be developed for more accurate analysis of homogeneity in dispersion nuclear fuel. (author). 28 refs., 7 figs., 1tab

  15. Solar fuel production by using PV/PEC junctions based on earth-abundant materials

    NARCIS (Netherlands)

    Perez Rodriguez, P.; Digdaya, I.A.; Mangel Raventos, A.; Falkenberg, M.; Vasudevan, R.A.; Zeman, M.; Smith, W.A.; Smets, A.H.M.

    2016-01-01

    One of the main problems of renewable energies is storage of the energy carrier. For long-term storage, solar fuels seem to be a good option. Direct solar water splitting could play an important role in the production of these solar fuels. One of the main challenges of this process is the charge

  16. Fundamental Processes of Coupled Radiation Damage and Mechanical Behavior in Nuclear Fuel Materials for High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Phillpot, Simon; Tulenko, James

    2011-09-08

    The objective of this work has been to elucidate the relationship among microstructure, radiation damage and mechanical properties for nuclear fuel materials. As representative nuclear materials, we have taken an hcp metal (Mg as a generic metal, and Ti alloys for fast reactors) and UO2 (representing fuel). The degradation of the thermo-mechanical behavior of nuclear fuels under irradiation, both the fissionable material itself and its cladding, is a longstanding issue of critical importance to the nuclear industry. There are experimental indications that nanocrystalline metals and ceramics may be more resistant to radiation damage than their coarse-grained counterparts. The objective of this project look at the effect of microstructure on radiation damage and mechanical behavior in these materials. The approach to be taken was state-of-the-art, large-scale atomic-level simulation. This systematic simulation program of the effects of irradiation on the structure and mechanical properties of polycrystalline Ti and UO2 identified radiation damage mechanisms. Moreover, it will provided important insights into behavior that can be expected in nanocrystalline microstructures and, by extension, nanocomposites. The fundamental insights from this work can be expected to help in the design microstructures that are less susceptible to radiation damage and thermomechanical degradation.

  17. Technical Division quarterly progress report, October 1--December 31, 1977. [Fuel cycle research and development; special materials production

    Energy Technology Data Exchange (ETDEWEB)

    Slansky, C.M. (ed.)

    1978-02-01

    Results are presented on the fluidized-bed calcination of simulated radioactive waste from the reprocessing of spent commercial nuclear fuel and on the removal of actinide elements from the waste prior to calcination. Other programs include the development of storage technology for /sup 85/Kr waste; and the behavior of volatile radionuclides during the combustion of HTGR graphite-based fuel. The long-term management of defense waste from the ICPP covers post-calcination treatment of ICPP calcined waste; the removal of actinide elements from first-cycle raffinate; the retrieval and handling of calcined waste from ICPP storage vaults; and the preparation of the Programmatic Environmental Impact Statement on ICPP waste. Process improvements are reported on the Fluorinel headend process for Zircaloy-clad fuels and on uranium accountability measurements. Other development results cover the process for recovering spent Rover fuel, buried pipeline transfer systems, support to the Waste Management Program, Waste Calcining Facility support, New Waste Calcining Facility support, and effluent monitoring methods evaluation and development. In this category are studies on nuclear materials security, application of a liquid-solid fluidized-bed heat exchanger to the recovery of geothermal heat, inplant reactor source term measurements, burnup methods for fast breeder reactor fuels, absolute thermal fission yield measurements, analytical support to light water breeder reactor development, research on analytical methods, and the behavior of environmental species of iodine.

  18. Hydrogen as a fuel for today and tomorrow: expectations for advanced hydrogen storage materials/systems research.

    Science.gov (United States)

    Hirose, Katsuhiko

    2011-01-01

    History shows that the evolution of vehicles is promoted by several environmental restraints very similar to the evolution of life. The latest environmental strain is sustainability. Transport vehicles are now facing again the need to advance to use sustainable fuels such as hydrogen. Hydrogen fuel cell vehicles are being prepared for commercialization in 2015. Despite intensive research by the world's scientists and engineers and recent advances in our understanding of hydrogen behavior in materials, the only engineering phase technology which will be available for 2015 is high pressure storage. Thus industry has decided to implement the high pressure tank storage system. However the necessity of smart hydrogen storage is not decreasing but rather increasing because high market penetration of hydrogen fuel cell vehicles is expected from around 2025 onward. In order to bring more vehicles onto the market, cheaper and more compact hydrogen storage is inevitable. The year 2025 seems a long way away but considering the field tests and large scale preparation required, there is little time available for research. Finding smart materials within the next 5 years is very important to the success of fuel cells towards a low carbon sustainable world.

  19. UKAEA fast reactor project research and development programme on fuel element cladding and sub-assembly wrapper materials

    International Nuclear Information System (INIS)

    Harries, D.R.

    1977-01-01

    Research and development work on fuel element component (cladding, subassembly wrappers, etc.) materials for the U.K. sodium cooled fast reactor programme has been conducted at the United Kingdom Atomic Energy Authority (UKAEA) establishments at Dounreay, Harwell, Risley, and Springfields during the past fifteen years or so. This work has formed an integral part of, and has been co-ordinated by, the UKAEA Fast Reactor Project and has involved close liaison with the Nuclear Power Company (NPC) and the Central Electricity Generating Board (CEGB). The research and development were initially related to the Prototype Fast Reactor (PFR) but the scope has now been extended to cover the first Civil Fast Reactor (CFR1), which has recently been re-designated the Civil Demonstration Fast Reactor (CDFR). The paper outlines the present status of the development of sodium cooled fast reactors in the U.K. and proceeds to summarize the principal PFR and CDFR core and fuel element parameters which have determined the planning and direction of the fuel element materials programme. The current position on the fuel element cladding and wrapper research and development programme is reviewed, and the facilities and future irradiation programme to be carried out in PFR are described

  20. Optimization of material flow in the nuclear fuel cycle using a cyclic multi-stage production-to-inventory model

    International Nuclear Information System (INIS)

    DePorter, E.L.

    1977-01-01

    The nuclear fuel cycle is modelled as a cyclic, multi-stage production-to-inventory system. The objective is to meet a known deterministic demand for energy while minimizing acquisition, production, and inventory holding costs for all stages of the fuel cycle. The model allows for cyclic flow (feedback) of materials, material flow conversion factors at each stage, production lag times at each stage, and for escalating costs of uranium ore. It does not allow shortages to occur in inventories. The model is optimized by the application of the calculus of variations and specifically through recently developed theorems on the solution of functionals constrained by inequalities. The solution is a set of optimal cumulative production trajectories which define the stagewise production rates. Analysis of these production rates reveals the optimal nuclear fuel cycle costs and that inventories (stockpiles) occur in uranium fields, enriched uranium hexafluoride, and fabricated fuel assemblies. An analysis of the sensitivity of the model to variation in three important parameters is performed

  1. Methodologies to determine the Pu content of spent fuel assemblies for input nuclear material accountancy of pyroporcessing

    International Nuclear Information System (INIS)

    Lee, Taehoon; Shin, Heesung; Kim, Youngsoo; Kim, Hodong; Kwon, Taeje

    2011-01-01

    This study shows two different non-destructive approaches to determine the Pu mass of spent fuel assemblies, and the analysis results on the errors in their Pu mass. For both methods, the Cm mass of the assembly is obtained based on the neutron measurement results. The Cm ratio of the assembly is determined from the Cm mass and the Pu mass obtained by using either of the two methods. In a comparison of two methods, the second method is simpler than the first and may not need a homogeneously-mixed sample of the spent fuel assembly. On the other hand, the second approach shows larger error in the estimated Pu mass than the first one for many different spent fuel cases of various burnup, initial enrichment, and cooling times. A member state support program for the development of the IAEA safeguards approach for an engineering-scale pyroprocessing facility, which is designated as the Reference Engineering-scale Pyroprocessing Facility(REPF), has been carried out by Korea Atomic Energy Research Institute since 2008. The nuclear material accountancy of the REPF is based on the 'Cm balance' technique. The Pu content of processing materials of pyroprocessing can be determined by measuring the Cm mass of the materials and multiplying it by the Cm ratio. The spent fuel assembly is de-cladded, and the irradiated UO 2 material of the assembly is homogeneously mixed in the homogenization process in order to obtain a representative sample of the spent fuel assembly for determining the mass of Pu, U and Cm elements, as well as the Cm ratio of the campaign. The shipper-receiver difference between the nuclear power plant and HPC of REPF is determined at this point. We found that the error for the Pu mass and Cm ratio determined from the homogenized uranium oxide powder is the most critical for the determination of the material unaccounted for throughout the whole processes. This paper presents two approaches to determine the Pu mass of spent fuel assemblies using non

  2. The Thermochemical Degradation of Hot Section Materials for Gas Turbine Engines in Alternative-Fuel Combustion Environments

    Science.gov (United States)

    Montalbano, Timothy

    Gas turbine engines remain an integral part of providing the world's propulsion and power generation needs. The continued use of gas turbines requires increased temperature operation to reach higher efficiencies and the implementation of alternative fuels for a lower net-carbon footprint. This necessitates evaluation of the material coatings used to shield the hot section components of gas turbines in these new extreme environments in order to understand how material degradation mechanisms change. Recently, the US Navy has sought to reduce its use of fossil fuels by implementing a blended hydroprocessed renewable diesel (HRD) derived from algae in its fleet. To evaluate the material degradation in this alternative environment, metal alloys are exposed in a simulated combustion environment using this blended fuel or the traditional diesel-like fuel. Evaluation of the metal alloys showed the development of thick, porous scales with a large depletion of aluminum for the blend fuel test. A mechanism linking an increased solubility of the scale to the blend fuel test environment will be discussed. For power generation applications, Integrated Gasification Combined Cycle (IGCC) power plants can provide electricity with 45% efficiency and full carbon capture by using a synthetic gas (syngas) derived from coal, biomass, or another carbon feedstock. However, the combustion of syngas is known to cause high water vapor content levels in the exhaust stream with unknown material consequences. To evaluate the effect of increased humidity, air-plasma sprayed (APS), yttria-stabilized zirconia (YSZ) is thermally aged in an environment with and without humidity. An enhanced destabilization of the parent phase by humid aging is revealed by x-ray diffraction (XRD) and Raman spectroscopy. Microstructural analysis by transmission electron microscopy (TEM) and scanning-TEM (STEM) indicate an enhanced coarsening of the domain structure of the YSZ in the humid environment. The enhanced

  3. Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Lumin [Univ. of Michigan, Ann Arbor, MI (United States). Department of Nuclear Engineering and Radiological Science; Wierschke, Jonathan Brett [Univ. of Michigan, Ann Arbor, MI (United States). Department of Nuclear Engineering and Radiological Science

    2015-04-08

    The objective of this work was to understand the corrosion behavior of Boral® and Bortec® neutron absorbers over long-term deployment in a used nuclear fuel dry cask storage environment. Corrosion effects were accelerated by flowing humidified argon through an autoclave at temperatures up to 570°C. Test results show little corrosion of the aluminum matrix but that boron is leaching out of the samples. Initial tests performed at 400 and 570°C were hampered by reduced flow caused by the rapid build-up of solid deposits in the outlet lines. Analysis of the deposits by XRD shows that the deposits are comprised of boron trioxide and sassolite (H3BO3). The collection of boron- containing compounds in the outlet lines indicated that boron was being released from the samples. Observation of the exposed samples using SEM and optical microscopy show the growth of new phases in the samples. These phases were most prominent in Bortec® samples exposed at 570°C. Samples of Boral® exposed at 570°C showed minimal new phase formation but showed nearly the complete loss of boron carbide particles. Boron carbide loss was also significant in Boral samples at 400°C. However, at 400°C phases similar to those found in Bortec® were observed. The rapid loss of the boron carbide particles in the Boral® is suspected to inhibit the formation of the new secondary phases. However, Material samples in an actual dry cask environment would be exposed to temperatures closer to 300°C and less water than the lowest test. The results from this study conclude that at the temperature and humidity levels present in a dry cask environment, corrosion and boron leaching will have no effect on the performance of Boral® and Bortec® to maintain criticality control.

  4. Some issues on the Law for the Regulations of Nuclear Source Material, Nuclear Fuel Material and Reactors Amendment after JCO criticality accident

    International Nuclear Information System (INIS)

    Tanabe, Tomoyuki

    2001-01-01

    As the Amendment of the Law for the Regulation of Nuclear Material, Nuclear Fuel Material and Reactors on an opportunity of the JCO criticality accident can be almost evaluated at a viewpoint of upgrading on effectiveness of safety regulation, it is thought to remain a large problem to rely on only enforcement of regulation due to amendment of the Law at future accident. In future, it can be also said to be important subjects to further expand a philosophy on the regulation (material regulation) focussed to hazards of nuclear material itself, not only to secure effectiveness on the multi-complementary safety regulation due to the administrative agency and the Nuclear Safety Commission but also to prepare a mechanism reflexible of a new information to the safety regulation, and to prepare a mechanism to assist adequate business execution and so forth of enterprises. (G.K.)

  5. Hydrogen like energy and materials for fuel cells; Hidrogeno como energetico y materiales para celdas de combustible

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez V, S. M., E-mail: suilma.fernandez@inin.gob.m [ININ, Departamento de Quimica, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2010-07-01

    The researches on the production, storage and the use of hydrogen like fuel or energy carrying are carried out in several laboratories around the world. In the Instituto Nacional de Investigaciones Nucleares (ININ), from the year of 1993 they are carried out researches about the synthesis of electro-catalysts materials than can serve in the hydrogen production starting from the electrolysis of the water, or in fuel cells, as well as of semiconductor materials for the photo-electrolysis of the water. Recently, in collaboration with other Departments of the ININ, the hydrogen production has been approached starting from fruit and vegetable wastes, with the purpose of evaluating the possibility that this residuals can be utilized for the energy obtaining and that they are not only garbage that causes problems of environmental pollution, generate toxic gases and pollute the soil with the organic acids that take place during their fermentation. (Author)

  6. Ab initio study of perovskite type oxide materials for solid oxide fuel cells

    Science.gov (United States)

    Lee, Yueh-Lin

    2011-12-01

    Perovskite type oxides form a family of materials of significant interest for cathodes and electrolytes of solid oxide fuel cells (SOFCs). These perovskites not only are active catalysts for surface oxygen reduction (OR) reactions but also allow incorporating the spilt oxygen monomers into their bulk, an unusual and poorly understood catalytic mechanism that couples surface and bulk properties. The OR mechanisms can be influenced strongly by defects in perovskite oxides, composition, and surface defect structures. This thesis work initiates a first step in developing a general strategy based on first-principles calculations for detailed control of oxygen vacancy content, transport rates of surface and bulk oxygen species, and surface/interfacial reaction kinetics. Ab initio density functional theory methods are used to model properties relevant for the OR reactions on SOFC cathodes. Three main research thrusts, which focus on bulk defect chemistry, surface defect structures and surface energetics, and surface catalytic properties, are carried to investigate different level of material chemistry for improved understanding of key physics/factors that govern SOFC cathode OR activity. In the study of bulk defect chemistry, an ab initio based defect model is developed for modeling defect chemistry of LaMnO 3 under SOFC conditions. The model suggests an important role for defect interactions, which are typically excluded in previous defect models. In the study of surface defect structures and surface energetics, it is shown that defect energies change dramatically (1˜2 eV lower) from bulk values near surfaces. Based on the existing bulk defect model with the calculated ab initio surface defect energetics, we predict the (001) MnO 2 surface oxygen vacancy concentration of (La0.9Sr0.1 )MnO3 is about 5˜6 order magnitude higher than that of the bulk under typical SOFC conditions. Finally, for surface catalytic properties, we show that area specific resistance, oxygen

  7. Nuclear fuel element containing strips of an alloyed Zr, Ti, and Ni getter material

    International Nuclear Information System (INIS)

    Grossman, L.N.; Packard, D.R.

    1975-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. The nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of strips and preferably the strips are positioned inside a helical member in the plenum. The position of the alloy strips permits gases and liquids entering the plenum to contact and react with the alloy strips. (U.S.)

  8. New technologies in the production of motor fuels from renewable materials

    Directory of Open Access Journals (Sweden)

    Adnađević Borivoj K.

    2012-01-01

    Full Text Available This work presents resources of the Autonomous Province of Vojvodina available for bioethanol and motor fuels (gasoline and diesel fuel from sustainable resources: corn-stalks, straw, sweet sorghum, pork fat. The physicochemical basis for novel processes for motor fuel production is coupling microwave pyrolysis of oil shale and catalytic cracking of purified pyrolysis oil, hydrothermal liquefaction of algae and swine manure. The effects of the degree of purification of crude pyrolysis oil and oil shale on the degree of their conversion to gasoline and diesel fuel, as well as the product distribution are investigated. The effects of the duration and temperature of hydrothermal liquefaction of microalga, Botryoccocus braunii, and swine manure on their degrees of conversion into bio-oil and its thermal properties are investigated. The development of novel strategy of biofuel in the Autonomous Province of Vojvodina is presented.

  9. Storage, inspection and sip testing of spent nuclear fuel from the HIFAR materials test reactor

    International Nuclear Information System (INIS)

    Selwyn, H.; Finlay, R.; Bull, P.; Irwin, A.

    2002-01-01

    Aluminum clad U-Al fuel used within the HIFAR MTR has been stored both in dry (underground) and wet (pond) storage facilities at the Lucas Heights site since the 1960's. As part of ANSTO's current program to send this fuel for long term storage or reprocessing, a significant level of visual inspection and water sip testing has been performed. This data has been used to demonstrate the integrity and suitability of the fuel for transport and receipt at the re processors interim storage ponds. This paper presents the key technical background-history of HIFAR fuel and its storage at Lucas Heights, presents the data obtained to date regarding its condition and discusses some observations regarding visual corrosion indicators and actual sip test results. (author)

  10. Nuclear fuel element containing particles of an alloyed Zr, Ti, and Ni getter material

    International Nuclear Information System (INIS)

    Grossman, L.N.; Levin, H.A.

    1975-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. The nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of particles in a hollow gas permeable container having a multiplicity of openings of size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles of alloy. The container is preferably held in the spring in the plenum of the fuel element. (Official Gazette)

  11. Supply guarantee initiatives for nuclear fuel materials and services and their compatibility with the market

    International Nuclear Information System (INIS)

    Nagano, Koji

    2009-01-01

    This paper first discusses where and why those ideas for nuclear fuel supply assurance and guarantee mechanisms came out, and attempts to draw a comparison among the proposed schemes and thereby examine possible steps forward. (author)

  12. 49 CFR 173.230 - Fuel cell cartridges containing hazardous material.

    Science.gov (United States)

    2010-10-01

    ... integral vent feature) is deemed to pass the fire test if: (A) The internal pressure vents to zero gauge... required to power the equipment, plus two spares; (5) Large robust articles containing fuel cells may be...

  13. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objectives of the meeting were: - To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; - To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; - To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; - To discuss the results of studies and on-going R&D activities that address cost reduction and the future economic competitiveness of fast reactors; and - To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

  14. Proton Conductors based on Metal Phosphonate Hybrid Materials for PEM Fuel Cells

    OpenAIRE

    Cabeza-Diaz, Aurelio; Colodrero, Rosario M. P.; Bazaga-García, Montse; Hernández Alonso, Daniel; Compaña Prieto, José Manuel; Olivera-Pastor, Pascual; Losilla, Enrique R.; Aranda, Miguel A. G.; León-Reina, Laura

    2014-01-01

    The necessity of energy and power generation is constantly growing. Fossils fuels are quickly becoming unsatisfactory substrates due to both their emission of pollutants and their finite expectancy. Fuel cells are one of the best alternatives as they are clean, durable, and highly efficient 1. MOFs will be attractive candidates for this application because of their tuneable pore size functionality as well as their chemical and thermal stability and presence of acidic protons within their stru...

  15. Energy Frontier Research Center Center for Materials Science of Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Todd Allen

    2014-04-01

    boundaries due to the more dense dislocation arrays associated with large angle boundaries; this observation has potentially important ramifications for thermal transport in the high burn-up rim region of light water reactor fuel. • A variable charge interatomic potential has been developed that not only provides an accurate representation of the fluorite UO2 phase, it is further capable of describing continuous stoichiometry changes from UO2 to hyper-stoichiometric UO2+x, to U4O9 and U3O7, and possibly to orthorhombic U3O8. This is the first potential that features many-body effects in all possible interactions (U-U, U-O and O-O) combined with the variable charge. • A theoretical proof has been formulated showing that it is necessary to use the so-called model C phase field approach, consisting of Cahn-Hilliard and Allen-Cahn equations, to describe void evolution in irradiated materials. This work resolved a longstanding literature controversy regarding how to model voids at the mesoscale. • A novel cluster dynamics model has been developed for the nucleation of voids and loops in UO2 under irradiation. This model is important in understanding the defect state of UO2 after irradiation and, more importantly, reveals off-stoichiometric states of irradiated UO2 that are critical for understanding the impact of irradiation on thermal transport. Personnel Successes

  16. Estimation of the radionuclide inventory in LWR spent fuel assembly structural materials for long-term safety analysis

    Directory of Open Access Journals (Sweden)

    Caruso Stefano

    2016-01-01

    Full Text Available The radionuclide inventory of materials irradiated in a reactor depends on the initial material composition, irradiation history and on the magnitude and spectrum of the neutron flux. The material composition of a fuel assembly structure includes various alloys of Zircaloy, Inconel and stainless steel. The existing impurities in these materials are very important for accurate determination of the activation of all nuclides with a view to assessing the radiological consequences of their geological disposal. In fact, the safety assessments of geological repositories require the average and maximum (in the sense of very conservative inventories of the very long-lived nuclides as input. The purpose of the present work is to describe the methodology applied for determining the activation of these nuclides in fuel assembly structural materials by means of coupled depletion/activation calculations and also to crosscheck the results obtained from two approaches. UO2 and MOX PWR fuels have been simulated using SCALE/TRITON, simultaneously irradiating the fuel region in POWER mode and the cladding region in FLUX mode and aiming to produce binary macro cross-section libraries by applying accurate local neutron spectra in the cladding region as a function of irradiation history that are suitable for activation calculations. The developed activation libraries have been re-employed in a second run using the ORIGEN-S program for a dedicated activation calculation. The axial variation of the neutron flux along the fuel assembly length has also been considered. The SCALE calculations were performed using a 238-group cross-section library, according to the ENDF/B-VII. The results obtained with the ORIGEN-S activation calculations are compared with the results obtained from TRITON via direct irradiation of the cladding, as allowed by the FLUX mode. It is shown that an agreement on the total calculated activities can be found within 55% for MOX and within 22% for

  17. Impact on vehicle fuel economy of the soot loading on diesel particulate filters made of different substrate materials

    International Nuclear Information System (INIS)

    Millo, Federico; Andreata, Maurizio; Rafigh, Mahsa; Mercuri, Davide; Pozzi, Chiara

    2015-01-01

    Wall flow DPFs (Diesel Particulate Filters) are nowadays universally adopted for all European passenger cars. Since the properties of the filter substrate material play a fundamental role in determining the optimal soot loading level to be reached before DPF regeneration, three different filter material substrates (Silicon Carbide, Aluminum Titanate and Cordierite) were investigated in this work, considering different driving conditions, after treatment layouts and regeneration strategies. In the first step of the research, an experimental investigation on the three different substrates over the NEDC (New European Driving Cycle) was performed. The data obtained from experiments were then used for the calibration and the validation of a one dimensional fluid-dynamic engine and after treatment simulation model. Afterward, the model was used to predict the vehicle fuel consumption increments as a function of the exhaust back pressure due to the soot loading for different driving cycles. The results showed that appreciable fuel consumption increments could be noticed only in particular driving conditions, and, as a consequence, in most of the cases the optimal filter regeneration strategy corresponds to reach the highest soot loading that still ensures the component safety even in case of uncontrolled regeneration events. - Highlights: • Three different substrate materials for a Diesel Particulate Filter were investigated. • Fuel consumption increases due to DPF soot loading were generally not appreciable. • Optimal soot loading before regeneration was the highest safeguarding DPF integrity. • SiC substrate showed highest soot load limit and lowest fuel consumption penalties. • AT and Cd substrate properties lead to lower soot load limits than SiC

  18. Materiais cerâmicos para células a combustível Ceramic materials for fuel cells

    Directory of Open Access Journals (Sweden)

    D. Z. de Florio

    2004-12-01

    Full Text Available A partir da definição de células a combustível, é feita uma introdução sucinta dos tipos de células e dos materiais cerâmicos que são empregados em projeto e fabricação destes dispositivos geradores de energia elétrica. Tomando por base a ampla literatura científica disponível em publicações periódicas internacionais indexadas e arbitradas, bem como patentes, são relatados com detalhes os materiais cerâmicos com comportamento elétrico adequado para uso como eletrólitos, anodos, catodos, interconectores e selantes, que são os componentes básicos de células a combustível de óxidos sólidos. Por fim, é feita uma avaliação do estado da arte na pesquisa e desenvolvimento de materiais cerâmicos para uso em células a combustível de óxidos sólidos.Basic definitions of fuel cells and a brief introduction of different types of fuel cells are given. A review of the most important ceramic materials being considered for the design and fabrication of devices for producing electrical energy is presented. Ceramic materials with suitable electrical behavior to be used as electrolytes, anodes, cathodes, interconnectors, and sealants of solid oxide fuel cells are reported with details, taking into account the large number of available indexed and refereed scientific publications and patents. Finally, an evaluation of the state of the art of the research and development of ceramic materials for solid oxide fuel cells is presented.

  19. Assessment of chemical and material contamination in waste wood fuels--A case study ranging over nine years.

    Science.gov (United States)

    Edo, Mar; Björn, Erik; Persson, Per-Erik; Jansson, Stina

    2016-03-01

    The increased demand for waste wood (WW) as fuel in Swedish co-combustion facilities during the last years has increased the import of this material. Each country has different laws governing the use of chemicals and therefore the composition of the fuel will likely change when combining WW from different origins. To cope with this, enhanced knowledge is needed on WW composition and the performance of pre-treatment techniques for reduction of its contaminants. In this study, the chemical and physical characteristics of 500 WW samples collected at a co-combustion facility in Sweden between 2004 and 2013 were investigated to determine the variation of contaminant content over time. Multivariate data analysis was used for the interpretation of the data. The concentrations of all the studied contaminants varied widely between sampling occasions, demonstrating the highly variable composition of WW fuels. The efficiency of sieving as a pre-treatment measure to reduce the levels of contaminants was not sufficient, revealing that sieving should be used in combination with other pre-treatment methods. The results from this case study provide knowledge on waste wood composition that may benefit its management. This knowledge can be applied for selection of the most suitable pre-treatments to obtain high quality sustainable WW fuels. Copyright © 2015 The Authors. Published by Elsevier Ltd.. All rights reserved.

  20. Radioactive material releases in the nuclear fuel cycle - Recent experience and improvements

    International Nuclear Information System (INIS)

    Allan, C.J.; Allsop, P.J.; Anderson, R.W.; Boss, C.R.; Frost, S.E.

    1997-01-01

    The nuclear fuel cycle involves a wide range of activities and technologies from the mining of uranium, to the production of electricity and radioisotopes for medical and industrial applications, to the reprocessing and recycling of used fuel, to decommissioning and waste disposal. Worker exposures and releases to the environment are carefully controlled in: (a) all stages of uranium mining, refining and fuel fabrication, where occupational exposures and releases have decreased while production has increased; (b) the operation of nuclear power plants, where occupational exposures and releases have decreased as reactor vendors evolve their products and reactor operators optimize their procedures; (c) fuel reprocessing facilities in the U.K. and France, where occupational exposures and releases have decreased while the amount of fuel processed has increased; and in (d) decommissioning nuclear facilities and waste management activities. The nuclear industry's recent record of achievement in controlling its releases and ensuring the radiological protection of its employees has been excellent. It is clear that releases and occupational exposures from modem nuclear facilities of all types contribute negligibly to the radiation environment to which all biota are exposed. But the general public seems not to appreciate the low environmental impact of nuclear activities. The future of nuclear power and of other applications of nuclear technology applications in medicine, in agriculture and in industry will depend on maintaining a high standard of performance so that the public and decision makers can be assured that the industry is safe. (author)

  1. Fossil Fuels.

    Science.gov (United States)

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  2. Standard guide for evaluation of materials used in extended service of interim spent nuclear fuel dry storage systems

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 Part of the total inventory of commercial spent nuclear fuel (SNF) is stored in dry cask storage systems (DCSS) under licenses granted by the U.S. Nuclear Regulatory Commission (NRC). The purpose of this guide is to provide information to assist in supporting the renewal of these licenses, safely and without removal of the SNF from its licensed confinement, for periods beyond those governed by the term of the original license. This guide provides information on materials behavior under conditions that may be important to safety evaluations for the extended service of the renewal period. This guide is written for DCSS containing light water reactor (LWR) fuel that is clad in zirconium alloy material and stored in accordance with the Code of Federal Regulations (CFR), at an independent spent-fuel storage installation (ISFSI). The components of an ISFSI, addressed in this document, include the commercial SNF, canister, cask, and all parts of the storage installation including the ISFSI pad. The language of t...

  3. Novel Mg-Doped SrMoO3 Perovskites Designed as Anode Materials for Solid Oxide Fuel Cells

    Directory of Open Access Journals (Sweden)

    Vanessa Cascos

    2016-07-01

    Full Text Available SrMo1−xMxO3−δ (M = Fe and Cr, x = 0.1 and 0.2 oxides have been recently described as excellent anode materials for solid oxide fuel cells at intermediate temperatures (IT-SOFC with LSGM as the electrolyte. In this work, we have improved their properties by doping with aliovalent Mg ions at the B-site of the parent SrMoO3 perovskite. SrMo1−xMgxO3−δ (x = 0.1, 0.2 oxides have been prepared, characterized and tested as anode materials in single solid-oxide fuel cells, yielding output powers near 900 mW/cm−2 at 850 °C using pure H2 as fuel. We have studied its crystal structure with an “in situ” neutron power diffraction (NPD experiment at temperatures as high as 800 °C, emulating the working conditions of an SOFC. Adequately high oxygen deficiencies, observed by NPD, together with elevated disk-shaped anisotropic displacement factors suggest a high ionic conductivity at the working temperatures. Furthermore, thermal expansion measurements, chemical compatibility with the LSGM electrolyte, electronic conductivity and reversibility upon cycling in oxidizing-reducing atmospheres have been carried out to find out the correlation between the excellent performance as an anode and the structural features.

  4. Recent Progress on the Key Materials and Components for Proton Exchange Membrane Fuel Cells in Vehicle Applications

    Directory of Open Access Journals (Sweden)

    Cheng Wang

    2016-07-01

    Full Text Available Fuel cells are the most clean and efficient power source for vehicles. In particular, proton exchange membrane fuel cells (PEMFCs are the most promising candidate for automobile applications due to their rapid start-up and low-temperature operation. Through extensive global research efforts in the latest decade, the performance of PEMFCs, including energy efficiency, volumetric and mass power density, and low temperature startup ability, have achieved significant breakthroughs. In 2014, fuel cell powered vehicles were introduced into the market by several prominent vehicle companies. However, the low durability and high cost of PEMFC systems are still the main obstacles for large-scale industrialization of this technology. The key materials and components used in PEMFCs greatly affect their durability and cost. In this review, the technical progress of key materials and components for PEMFCs has been summarized and critically discussed, including topics such as the membrane, catalyst layer, gas diffusion layer, and bipolar plate. The development of high-durability processing technologies is also introduced. Finally, this review is concluded with personal perspectives on the future research directions of this area.

  5. Lightweighting Automotive Materials for Increased Fuel Efficiency and Delivering Advanced Modeling and Simulation Capabilities to U.S. Manufacturers

    Energy Technology Data Exchange (ETDEWEB)

    Hale, Steve

    2013-09-11

    Abstract The National Center for Manufacturing Sciences (NCMS) worked with the U.S. Department of Energy (DOE), National Energy Technology Laboratory (NETL), to bring together research and development (R&D) collaborations to develop and accelerate the knowledgebase and infrastructure for lightweighting materials and manufacturing processes for their use in structural and applications in the automotive sector. The purpose/importance of this DOE program: • 2016 CAFÉ standards. • Automotive industry technology that shall adopt the insertion of lightweighting material concepts towards manufacturing of production vehicles. • Development and manufacture of advanced research tools for modeling and simulation (M&S) applications to reduce manufacturing and material costs. • U.S. competitiveness that will help drive the development and manufacture of the next generation of materials. NCMS established a focused portfolio of applied R&D projects utilizing lightweighting materials for manufacture into automotive structures and components. Areas that were targeted in this program: • Functionality of new lightweighting materials to meet present safety requirements. • Manufacturability using new lightweighting materials. • Cost reduction for the development and use of new lightweighting materials. The automotive industry’s future continuously evolves through innovation, and lightweight materials are key in achieving a new era of lighter, more efficient vehicles. Lightweight materials are among the technical advances needed to achieve fuel/energy efficiency and reduce carbon dioxide (CO2) emissions: • Establish design criteria methodology to identify the best materials for lightweighting. • Employ state-of-the-art design tools for optimum material development for their specific applications. • Match new manufacturing technology to production volume. • Address new process variability with new production-ready processes.

  6. Concerning change in nuclear fuel material processing business at Tokai plant of Japan Nuclear Fuel Conversion Co., Ltd. Report to Prime Minister

    International Nuclear Information System (INIS)

    1988-01-01

    The Nuclear Safety Committee of Japan on April 7, 1988, directed the Nuclear Safety Expert Group to make a study concerning the proposed changes in the nuclear fuel material processing business at the Tokai plant of Japan Nuclear Fuel Conversion Co., Ltd., and after receiving and reviewing the report from the Group, concluded that the proposed changes should be approved. The conclusions together with results of the study were reported to the Prime Minister on June 9. 1988. The proposed plan included changes in the maximum processing capacity of the No.2 processing facilities; construction of a new powder warehouse and changes in the maximum capacity of the No.3 powder storage room and No.2 powder warehouse; reuse of No.1 powder warehouse as No.3 solid waste warehouse; and abolition of UF 6 dispensing equipment installed at the No.1 processing facilities and changes in procedures for criticality control of the hydrolysis facilities. The safety of these facilities were studied in terms of resistance to earthquakes, prevention of fire and explosion, criticality control, operations of waste processing, and radiation management. Exposure doses expected during normal operations were also examined to confirm that the possible exposure doses to the public would be sufficiently small. (N.K.)

  7. Electrocatalysts and their Supporting Materials for Proton Exchange Membrane Fuel Cells: Activity and Durability Studies

    DEFF Research Database (Denmark)

    Permyakova, Anastasia Aleksandrovna

    other methods. The thesis begins with an introduction in Chapter 1 providing an overview of fuel cells, their associated reaction mechanisms, catalysts and catalysts supports. Chapter 2 presents the theoretical background to the study including equipment and the techniques used to analyse the catalysts......This thesis describes investigations conducted exploring the activity, stability and durability of supported nano-particulate, bulk and thin film electrocatalysts used in proton exchange membrane fuel cells (PEMFCs). The effects of different factors and conditions on the reactions involved...

  8. Applying distance-to-target weighing methodology to evaluate the environmental performance of bio-based energy, fuels, and materials

    International Nuclear Information System (INIS)

    Weiss, Martin; Patel, Martin; Heilmeier, Hermann; Bringezu, Stefan

    2007-01-01

    The enhanced use of biomass for the production of energy, fuels, and materials is one of the key strategies towards sustainable production and consumption. Various life cycle assessment (LCA) studies demonstrate the great potential of bio-based products to reduce both the consumption of non-renewable energy resources and greenhouse gas emissions. However, the production of biomass requires agricultural land and is often associated with adverse environmental effects such as eutrophication of surface and ground water. Decision making in favor of or against bio-based and conventional fossil product alternatives therefore often requires weighing of environmental impacts. In this article, we apply distance-to-target weighing methodology to aggregate LCA results obtained in four different environmental impact categories (i.e., non-renewable energy consumption, global warming potential, eutrophication potential, and acidification potential) to one environmental index. We include 45 bio- and fossil-based product pairs in our analysis, which we conduct for Germany. The resulting environmental indices for all product pairs analyzed range from -19.7 to +0.2 with negative values indicating overall environmental benefits of bio-based products. Except for three options of packaging materials made from wheat and cornstarch, all bio-based products (including energy, fuels, and materials) score better than their fossil counterparts. Comparing the median values for the three options of biomass utilization reveals that bio-energy (-1.2) and bio-materials (-1.0) offer significantly higher environmental benefits than bio-fuels (-0.3). The results of this study reflect, however, subjective value judgments due to the weighing methodology applied. Given the uncertainties and controversies associated not only with distance-to-target methodologies in particular but also with weighing approaches in general, the authors strongly recommend using weighing for decision finding only as a

  9. Spent fuel working group report on inventory and storage of the Department's spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities

    International Nuclear Information System (INIS)

    1993-11-01

    In a memo dated 19 August 1993, Secretary O'Leary assigned the Office of Environment, Safety and Health the primary responsibility to identify, characterize, and assess the safety, health, and environmental vulnerabilities of the DOE's existing storage conditions and facilities for the storage of irradiated reactor fuel and other reactor irradiated nuclear materials. This volume is divided into three major sections. Section 1 contains the Working Group Assessment Team reports on the following facilities: Hanford Site, INEL, SRS, Oak Ridge Site, West Valley Site, LANL, BNL, Sandia, General Atomics (San Diego), Babcock ampersand Wilcox (Lynchburg Technology Center), and ANL. Section 2 contains the Vulnerability Development Forms from most of these sites. Section 3 contains the documents used by the Working Group in implementing this initiative

  10. The future supply of and demand for candidate materials for the fabrication of nuclear fuel waste disposal containers

    International Nuclear Information System (INIS)

    Grover, L.K.

    1990-01-01

    This report summarizes the findings of a literature survey carried out to assess the future world supply of and demand for titanium, copper and lead. These metals are candidate materials for the fabrication of containers for the immobilization and disposal of Canada's nuclear used-fuel waste for a reference Used-fuel Disposal Centre. Such a facility may begin operation by approximately 2020, and continue for about 40 years. The survey shows that the world has abundant supplies of titanium minerals (mostly in the form of ilmenite), which are expected to last up to at least 2110. However, for copper and lead the balance between supply and demand may warrant increased monitoring beyond the year 2000. A number of factors that can influence future supply and demand are discussed in the report

  11. 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

    , Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO 2 ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO 2 and MOX ceramics (Chromium oxide-doped UO 2 fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The nature of spent nuclear

  12. A cost/benefit analysis of methods for controlling the release of radioactive materials in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Blanco, R.E.; Dahlman, R.C.; Davis, W. Jr.; Finney, B.C.; Groenier, W.S.; Hill, G.S.; Kibbey, A.H.; Kitts, F.G.; Lindauer, R.B.; Moore, R.E.; Pechin, W.H.; Roddy, J.W.; Ryon, A.D.; Seagren, R.D.; Sears, M.B.; Witherspoon, J.P.

    1977-01-01

    Cost/benefit surveys were made to determine the cost (in dollars) and effectiveness of radwaste treatment systems for decreasing the release of radioactive materials from model fuel cycle facilities, and to determine the benefits in terms of reduction in radiological dose commitment to individuals and populations in the surrounding areas. The studies include milling of uranium ores, conversion of virgin uranium and recycle uranium to UF 6 , fabrication of light-water reactor (LWR) fuels containing enriched uranium or enriched uranium and plutonium, fabrication of high-temperature gas-cooled reactor (HTGR) fuels containing 233 U and thorium, and reprocessing of LWR and HTGR fuels. Conceptual flowsheets were prepared for each model facility illustrating the treatment methods for gaseous and liquid effluents. The ''base'' case represents the lowest treatment cost, current treatment technology, and highest radiological dose. In succeeding cases, increasingly efficient radwaste treatment equipment is added to the ''base'' plant to reduce the amounts of radioactive materials released. The technology ranges from that currently available to that which may be developed over the next 30 years. The status of development for these technologies is discussed. The dose estimates are for maximum individual total body and organ doses at the plant boundary and for population total-body and organ doses out to 89 km. Comparisons of the doses vs annual costs in dollars are presented. In summary, they indicate that (1) the annual doses can be reduced to very low fractions of the natural background dose by the successful development and application of the radwaste treatment methods; and (2) excluding mills, the capital costs for the treatment methods vary from 0.02 to 8% of the capital cost of the base plants and the total annual operating costs (fixed charges plus operating costs) vary from 0.009 to 7.0% of the capital costs for the plant

  13. Fuel pellet

    International Nuclear Information System (INIS)

    Hayashi, K.

    1980-01-01

    Fuel pellet for insertion into a cladding tube in order to form a fuel element or a fuel rod. The fuel pellet has got a belt-like projection around its essentially cylindrical lateral circumferential surface. The upper and lower edges in vertical direction of this belt-like projection are wave-shaped. The projection is made of the same material as the bulk pellet. Both are made in one piece. (orig.) [de

  14. Development of metal uranium fuel and testing of construction materials (I-VI); Part I

    International Nuclear Information System (INIS)

    Mihajlovic, A.

    1965-11-01

    This project includes the following tasks: Study of crystallisation of metal melt and beta-alpha transforms in uranium and uranium alloys; Study of the thermal treatment influence on phase transformations and texture in uranium alloys; Radiation damage of metal uranium; Project related to irradiation of metal uranium in the reactor; Development of fuel element for nuclear reactors

  15. Materials and Concepts for Full Ceramic SOFCs with Focus on Carbon Containing Fuels

    DEFF Research Database (Denmark)

    Holtappels, Peter; Sudireddy, Bhaskar Reddy; Veltzé, Sune

    stimulated the development for full ceramic anodes based on strontium titanates. Furthermore, the Ni-cermet is primarily a hydrogen oxidation electrode and efficiency losses might occur when operating on carbon containing fuels. In a recent European project full ceramic cells comprising CGO/Ni infiltrated Nb...

  16. Effects of catalyst-support materials on the performance of fuel cells

    CSIR Research Space (South Africa)

    Ejikeme, PM

    2016-07-01

    Full Text Available of the historically significant work on various catalyst supports for Direct Alcohol Fuel Cells (DAFCs), as far as possible, the most recent developments are accentuated. References are made to other reports that have reviewed similar subject matter for specific cases...

  17. Taking into account the dissemination risk of radioactive materials in the French fuel cycle factories

    International Nuclear Information System (INIS)

    Ruiz, J.

    1994-01-01

    In this text the conception principles retained for treating the dissemination risk of radioactive matters in the French fuel cycle factories are presented. For taking into account this risk successives containment systems are used with respects to the ventilation regulations and fire protection

  18. Materials and Fuels Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

    International Nuclear Information System (INIS)

    Harvego, Lisa; Bennett, Brion

    2011-01-01

    Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Materials and Fuels Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool for developing the radioactive waste management basis.

  19. Selection of the reference concept for the surface examination stations in the fuels and materials examination facility

    International Nuclear Information System (INIS)

    Frandsen, G.B.; Nash, C.R.

    1978-01-01

    The prototype surface examination station for the Fuels and Materials Examination Facility (FMEF) will use closed circuit television (CCTV) for routine modes of operation along with a nuclear periscope for special examination needs. The CCTV and the nuclear periscope were evaluated against prescribed station requirements and compared in a side-by-side demonstration. A quantitative evaluation of their outputs showed that both systems were capable of meeting surface anomaly detection requirements. The CCTV system was superior in its ability to collect, suppress and present data into a more useful form for the experimenters

  20. Fuel-containing materials of 'Shelter' object today: actual physical properties and facilities for their state prognosis creation

    International Nuclear Information System (INIS)

    Zhidkov, A.V.

    2001-01-01

    The review of main available data on the physical properties of fuel-containing materials (FCM) of the 'Shelter' object is represented. The FCM properties with time changeability is connected with the radiation damages originated from FCM internal self-irradiation. The final stage of FCM degradation will be their total conversion into the submicronic radioactive dust, which already provides 90 % of inhalation dose rate when working in 'Shelter'. The preliminary prognostic estimation for such a situation development is represented as well; the possible scheme for FCM properties monitoring is also suggested