Advances in carbide fuel element development for fast reactor application

International Nuclear Information System (INIS)

Dienst, W.; Kleykamp, H.; Muehling, G.; Reiser, H.; Steiner, H.; Thuemmler, F.; Wedermeyer, H.; Weimar, P.

1977-01-01

The features of the carbide fuel development programme are reviewed and evaluated. Single pin and bundle irradiations are carried out under thermal, epithermal and fast flux conditions, the latter in the DFR and KNK-II reactors. Several fuel concepts in the region of representative SNR clad temperatures are compared by parameter and performance tests. A conservative concept is based on He-bonded 8 mm pins with (U,Pu)C pellets and a smear density of 75% TD, operating at 800 W/cm rod power and burnup to 70 MWd/kg. The preparation of mixed carbide fuels is carried out by carbothermic reduction of the oxides in different methods supported by equivalent carbon content, grain size and phase distribution analysis. The fuel for subassembly performance tests is produced in a pilot plant of 0,5 t/year capacity. Compatibility studies reveal that cladding carburization is the only chemical interaction with carbide fuels. This effect leads to a reduction in ductility of the stainless steel. Fission products apparently play no role in the compatibility behaviour. Comprehensive studies lead to reliable information on the chemical and thermodynamic state of the fuel under irradiation. The swelling of carbide fuels and the fission gas release are examined and analysed. Cladding plastic strain by fuel swelling occurs during steady-state operation because the irradiation creep is rather slow compared to oxide fuels. The cladding strain observed depends on the fuel porosity and the cladding strength. The development of carbide fuel pins is complemented by the application of comprehensive computer models. In addition to the steady-state tests power cycling and safety tests are under performance. Up to 1980 the results are summarized for the final design and specification. The development target of the present program is to fabricate several subassemblies for test operation in the SNR 300 by 1981

Synthesis of carbides of refractory metals in salt melts

International Nuclear Information System (INIS)

Ilyushchenko, N.G.; Anfinogenov, A.I.; Chebykin, V.V.; Chernov, Ya.B.; Shurov, N.I.; Ryaposov, Yu.A.; Dobrynin, A.I.; Gorshkov, A.V.; Chub, A.V.

2003-01-01

The ion-electron melts, obtained through dissolving the alkali and alkali-earth metals in the molten chlorides above the chloride melting temperature, were used for manufacturing the high-melting metal carbides as the transport melt. The lithium, calcium and magnesium chlorides and the mixture of the lithium chloride with the potassium or calcium chloride were used from the alkali or alkali-earth metals. The metallic lithium, calcium, magnesium or the calcium-magnesium mixtures were used as the alkali or alkali-earth metals. The carbon black or sugar was used as carbon. It is shown, that lithium, magnesium or calcium in the molten salts transfer the carbon on the niobium, tantalum, titanium, forming the carbides of the above metals. The high-melting metal carbides are obtained both from the metal pure powders and from the oxides and chlorides [ru

A study on the formation of uranium carbide in an induction furnace

International Nuclear Information System (INIS)

Song, In Young; Lee, Yoon Sang; Kim, Eung Soo; Lee, Don Bae; Kim, Chang Kyu

2005-01-01

Uranium is a typical carbide-forming element. Three carbides, UC, U 2 C 3 and UC 2 , are formed in the uranium-carbon system. The most important of these as fuel is uranium monocarbide UC. It is well known that Uranium carbides can be obtained by three basic methods: 1) by reaction of uranium metal with carbon; 2) by reaction of uranium metal powder with gaseous hydrocarbons; 3) by reaction of uranium oxides with carbon. The use of uranium monocarbide, or materials based on it, has great prospects as fuel for nuclear reactors. It is quite possible that uranium dicarbide UC 2 may also acquire great importance as a fuel, particularly in dispersion fuel elements with graphite matrix. In the present study, uranium carbides are obtained by direct reaction of uranium metal with graphite in a high frequency induction furnace

Friction and metal transfer for single-crystal silicon carbide in contact with various metals in vacuum

International Nuclear Information System (INIS)

Miyoshi, K.; Buckley, D.H.

1978-04-01

Sliding friction experiments were conducted with single-crystal silicon carbide in contact with transition metals (tungsten, iron, rhodium, nickel, titanium, and cobalt), copper, and aluminum. Results indicate the coefficient of friction for a silicon carbide-metal system is related to the d bond character and relative chemical activity of the metal. The more active the metal, the higher the coefficient of friction. All the metals examined transferred to the surface of silicon carbide in sliding. The chemical activity of metal to silicon and carbon and shear modulus of the metal may play important roles in metal transfer and the form of the wear debris. The less active metal is, and the greater resistance to shear it has, with the exception of rhodium and tungsten, the less transfer to silicon carbide

Stable carbides in transition metal alloys

International Nuclear Information System (INIS)

Piotrkowski, R.

1991-01-01

In the present work different techniques were employed for the identification of stable carbides in two sets of transition metal alloys of wide technological application: a set of three high alloy M2 type steels in which W and/or Mo were total or partially replaced by Nb, and a Zr-2.5 Nb alloy. The M2 steel is a high speed steel worldwide used and the Zr-2.5 Nb alloy is the base material for the pressure tubes in the CANDU type nuclear reactors. The stability of carbide was studied in the frame of Goldschmidt's theory of interstitial alloys. The identification of stable carbides in steels was performed by determining their metallic composition with an energy analyzer attached to the scanning electron microscope (SEM). By these means typical carbides of the M2 steel, MC and M 6 C, were found. Moreover, the spatial and size distribution of carbide particles were determined after different heat treatments, and both microstructure and microhardness were correlated with the appearance of the secondary hardening phenomenon. In the Zr-Nb alloy a study of the α and β phases present after different heat treatments was performed with optical and SEM metallographic techniques, with the guide of Abriata and Bolcich phase diagram. The α-β interphase boundaries were characterized as short circuits for diffusion with radiotracer techniques and applying Fisher-Bondy-Martin model. The precipitation of carbides was promoted by heat treatments that produced first the C diffusion into the samples at high temperatures (β phase), and then the precipitation of carbide particles at lower temperature (α phase or (α+β)) two phase field. The precipitated carbides were identified as (Zr, Nb)C 1-x with SEM, electron microprobe and X-ray diffraction techniques. (Author) [es

Fabrication of metallic channel-containing UO2 fuels

International Nuclear Information System (INIS)

Yang, Jae Ho; Song, Kun Woo; Kim, Keon Sik; Jung, Youn Ho

2004-01-01

The uranium dioxide is widely used as a fuel material in the nuclear industry, owing to many advantages. But it has a disadvantage of having the lowest thermal conductivity of all kinds of nuclear fuels; metal, carbide, nitride. It is well known that the thermal conductivity of UO 2 fuel is enhanced by making, so called, the CERMET (ceramic-metal) composite which consists of both continuous body of highly thermal-conducting metal and UO 2 islands. The CERMET fuel fabrication technique needs metal phase of at least 30%, mostly more than 50%, of the volume of the pellet in order to keep the metal phase interconnected. This high volume fraction of metal requires such a high enrichment of U that the parasitic effect of metal should be compensated. Therefore, it is attractive to develop an innovative composite fuel that can form continuous metal phase with a small amount of metal. In this investigation, a feasibility study was made on how to make such an innovative fuel. Candidate metals (W, Mo, Cr) were selected, and fabrication process was conceptually designed from thermodynamic calculations. We have experimentally found that a metal phase envelops perfectly UO 2 grains, forming continuous channel throughout the pellet, and improving the thermal conductivity of pellet

Boron-carbide-aluminum and boron-carbide-reactive metal cermets. [B/sub 4/C-Al

Science.gov (United States)

Halverson, D.C.; Pyzik, A.J.; Aksay, I.A.

1985-05-06

Hard, tough, lighweight boron-carbide-reactive metal composites, particularly boron-carbide-aluminum composites, are produced. These composites have compositions with a plurality of phases. A method is provided, including the steps of wetting and reacting the starting materials, by which the microstructures in the resulting composites can be controllably selected. Starting compositions, reaction temperatures, reaction times, and reaction atmospheres are parameters for controlling the process and resulting compositions. The ceramic phases are homogeneously distributed in the metal phases and adhesive forces at ceramic-metal interfaces are maximized. An initial consolidated step is used to achieve fully dense composites. Microstructures of boron-carbide-aluminum cermets have been produced with modules of rupture exceeding 110 ksi and fracture toughness exceeding 12 ksi..sqrt..in. These composites and methods can be used to form a variety of structural elements.

Survey of post-irradiation examinations made of mixed carbide fuels

International Nuclear Information System (INIS)

Coquerelle, M.

1997-01-01

Post-irradiation examinations on mixed carbide, nitride and carbonitride fuels irradiated in fast flux reactors Rapsodie and DFR were carried out during the seventies and early eighties. In this report, emphasis was put on the fission gas release, cladding carburization and head-end gaseous oxidation process of these fuels, in particular, of mixed carbides. (author). 8 refs, 16 figs, 3 tabs

Multi-criteria methodology to design a sodium-cooled carbide-fueled Gen-IV reactor

International Nuclear Information System (INIS)

Stauff, N.

2011-01-01

Compared with earlier plant designs (Phenix, Super-Phenix, EFR), Gen IV Sodium-cooled Fast Reactor requires improved economics while meeting safety and non-proliferation criteria. Mixed Oxide (U-Pu)O 2 fuels are considered as the reference fuels due to their important and satisfactory feedback experience. However, innovative carbide (U-Pu)C fuels can be considered as serious competitors for a prospective SFR fleet since carbide-fueled SFRs can offer another type of optimization which might overtake on some aspects the oxide fuel technology. The goal of this thesis is to reveal the potentials of carbide by designing an optimum carbide-fueled SFR with competitive features and a naturally safe behavior during transients. For a French nuclear fleet, a 1500 MW(e) break-even core is considered. To do so, a multi-physic approach was developed taking into account neutronics, fuel thermo-mechanics and thermal-hydraulic at a pre-design stage. Simplified modeling with the calculation of global neutronic feedback coefficients and a quasi-static evaluation was developed to estimate the behavior of a core during overpower transients, loss of flow and/or loss of heat removal transients. The breakthrough of this approach is to provide the designer with an overall view of the iterative process, emphasizing the well-suited innovations and the most efficient directions that can improve the SFR design project.This methodology was used to design a core that benefits from the favorable features of carbide fuels. The core developed is a large carbide-fueled SFR with high power density, low fissile inventory, break-even capability and forgiving behaviors during the un-scrammed transients studied that should prevent using expensive mitigate systems. However, the core-peak burnup is unlikely to significantly exceed 100 MWd/kg because of the large swelling of the carbide fuel leading to quick pellet-clad mechanical interaction and the low creep capacity of carbide. Moderate linear power fuel

Process for the preparation of fine grain metal carbide powders

International Nuclear Information System (INIS)

Gortsema, F.P.

1976-01-01

Fine grain metal carbide powders are conveniently prepared from the corresponding metal oxide by heating in an atmosphere of methane in hydrogen. Sintered articles having a density approaching the theoretical density of the metal carbide itself can be fabricated from the powders by cold pressing, hot pressing or other techniques. 8 claims, no drawings

Interaction of noble-metal fission products with pyrolytic silicon carbide

International Nuclear Information System (INIS)

Lauf, R.J.; Braski, D.N.

1982-01-01

Fuel particles for the High-Temperature Gas-Cooled Reactor (HTGR) contain layers of pyrolytic carbon and silicon carbide, which act as a miniature pressure vessel and form the primary fission product barrier. Of the many fission products formed during irradiation, the noble metals are of particular interest because they interact significantly with the SiC layer and their concentrations are somewhat higher in the low-enriched uranium fuels currently under consideration. To study fission product-SiC interactions, particles of UO 2 or UC 2 are doped with fission product elements before coating and are then held in a thermal gradient up to several thousand hours. Examination of the SiC coatings by TEM-AEM after annealing shows that silver behaves differently from the palladium group

Silver diffusion through silicon carbide in microencapsulated nuclear fuels TRISO

International Nuclear Information System (INIS)

Cancino T, F.; Lopez H, E.

2013-10-01

The silver diffusion through silicon carbide is a challenge that has persisted in the development of microencapsulated fuels TRISO (Tri structural Isotropic) for more than four decades. The silver is known as a strong emitter of gamma radiation, for what is able to diffuse through the ceramic coatings of pyrolytic coal and silicon carbide and to be deposited in the heat exchangers. In this work we carry out a recount about the art state in the topic of the diffusion of Ag through silicon carbide in microencapsulated fuels and we propose the role that the complexities in the grain limit can have this problem. (Author)

Square lattice honeycomb tri-carbide fuels for 50 to 250 KN variable thrust NTP design

International Nuclear Information System (INIS)

Anghaie, Samim; Knight, Travis; Gouw, Reza; Furman, Eric

2001-01-01

Ultrahigh temperature solid solution of tri-carbide fuels are used to design an ultracompact nuclear thermal rocket generating 950 seconds of specific impulse with scalable thrust level in range of 50 to 250 kilo Newtons. Solid solutions of tri-carbide nuclear fuels such as uranium-zirconium-niobium carbide. UZrNbC, are processed to contain certain mixing ratio between uranium carbide and two stabilizing carbides. Zirconium or niobium in the tri-carbide could be replaced by tantalum or hafnium to provide higher chemical stability in hot hydrogen environment or to provide different nuclear design characteristics. Recent studies have demonstrated the chemical compatibility of tri-carbide fuels with hydrogen propellant for a few to tens of hours of operation at temperatures ranging from 2800 K to 3300 K, respectively. Fuel elements are fabricated from thin tri-carbide wafers that are grooved and locked into a square-lattice honeycomb (SLHC) shape. The hockey puck shaped SLHC fuel elements are stacked up in a grooved graphite tube to form a SLHC fuel assembly. A total of 18 fuel assemblies are arranged circumferentially to form two concentric rings of fuel assemblies with zirconium hydride filling the space between assemblies. For 50 to 250 kilo Newtons thrust operations, the reactor diameter and length including reflectors are 57 cm and 60 cm, respectively. Results of the nuclear design and thermal fluid analyses of the SLHC nuclear thermal propulsion system are presented

Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

International Nuclear Information System (INIS)

Radulescu, H.

2001-01-01

There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report

Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

Energy Technology Data Exchange (ETDEWEB)

H. radulescu

2001-09-28

There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report.

UK irradiation experience relevant to advanced carbide fuel concepts for LMFBR's

International Nuclear Information System (INIS)

Bagley, K.Q.; Batey, W.; Paris, R.; Sloss, W.M.; Snape, G.P.

1977-01-01

Despite discouraging prognoses of fabrication and reprocessing problems, it is recognized that the quest for a carbide fuel pin design which fully exploits the favourable density and thermal conductivity of (U,Pu) monocarbide must be maintained. Studies in aid of carbide fuel development have, therefore, continued in the UK in parallel with those on oxide, albeit at a substantially lower level of effort, and a sufficient body of irradiation experience has been accumulated to allow discrimination of realistic fuel pin designs

Failed fuel identification techniques for liquid-metal cooled reactors

International Nuclear Information System (INIS)

Lambert, J.D.B.; Gross, K.C.; Mikaili, R.; Frank, S.M.; Cutforth, D.C.; Angelo, P.L.

1995-01-01

The Experimental Breeder Reactor II (EBR-II), located in Idaho and operated for the US Department of Energy by Argonne National Laboratory, has been used as an irradiation testbed for LMR fuels and components for thirty years. During this time many endurance tests have been carried out with experimental LMR metal, oxide, carbide and nitride fuel elements, in which cladding failures were intentionally allowed to occur. This paper describes methods that have been developed for the detection, identification and verification of fuel failures

Properties of cemented carbides alloyed by metal melt treatment

International Nuclear Information System (INIS)

Lisovsky, A.F.

2001-01-01

The paper presents the results of investigations into the influence of alloying elements introduced by metal melt treatment (MMT-process) on properties of WC-Co and WC-Ni cemented carbides. Transition metals of the IV - VIll groups (Ti, Zr, Ta, Cr, Re, Ni) and silicon were used as alloying elements. It is shown that the MMT-process allows cemented carbides to be produced whose physico-mechanical properties (bending strength, fracture toughness, total deformation, total work of deformation and fatigue fracture toughness) are superior to those of cemented carbides produced following a traditional powder metallurgy (PM) process. The main mechanism and peculiarities of the influence of alloying elements added by the MMT-process on properties of cemented carbides have been first established. The effect of alloying elements on structure and substructure of phases has been analyzed. (author)

Irradiation performance of helium-bonded uranium--plutonium carbide fuel elements

International Nuclear Information System (INIS)

Latimer, T.W.; Petty, R.L.; Kerrisk, J.F.; DeMuth, N.S.; Levine, P.J.; Boltax, A.

1979-01-01

The current irradiation program of helium-bonded uranium--plutonium carbide elements is achieving its original goals. By August 1978, 15 of the original 171 helium-bonded elements had reached their goal burnups including one that had reached the highest burnup of any uranium--plutonium carbide element in the U.S.--12.4 at.%. A total of 66 elements had attained burnups over 8 at.%. Only one cladding breach had been identified at that time. In addition, the systematic and coordinated approach to the current steady-state irradiation tests is yielding much needed information on the behavior of helium-bonded carbide fuel elements that was not available from the screening tests (1965 to 1974). The use of hyperstoichiometric (U,Pu)C containing approx. 10 vol% (U,Pu) 2 C 3 appears to combine lower swelling with only a slightly greater tendency to carburize the cladding than single-phase (U,Pu)C. The selected designs are providing data on the relationship between the experimental parameters of fuel density, fuel-cladding gap size, and cladding type and various fuel-cladding mechanical interaction mechanisms

Hydrogen evolution activity and electrochemical stability of selected transition metal carbides in concentrated phosphoric acid

DEFF Research Database (Denmark)

Tomás García, Antonio Luis; Jensen, Jens Oluf; Bjerrum, Niels J.

2014-01-01

phosphoric acid were investigated in a temperature range from 80 to 170°C. A significant dependence of the activities on temperature was observed for all five carbide samples. Through the entire temperature range Group 6 metal carbides showed higher activity than that of the Group 5 metal carbides......Alternative catalysts based on carbides of Group 5 (niobium and tantalum) and 6 (chromium, molybdenum and tungsten) metals were prepared as films on the metallic substrates. The electrochemical activities of these carbide electrodes towards the hydrogen evolution reaction (HER) in concentrated...

Metal Immiscibility Route to Synthesis of Ultrathin Carbides, Borides, and Nitrides.

Science.gov (United States)

Wang, Zixing; Kochat, Vidya; Pandey, Prafull; Kashyap, Sanjay; Chattopadhyay, Soham; Samanta, Atanu; Sarkar, Suman; Manimunda, Praveena; Zhang, Xiang; Asif, Syed; Singh, Abhisek K; Chattopadhyay, Kamanio; Tiwary, Chandra Sekhar; Ajayan, Pulickel M

2017-08-01

Ultrathin ceramic coatings are of high interest as protective coatings from aviation to biomedical applications. Here, a generic approach of making scalable ultrathin transition metal-carbide/boride/nitride using immiscibility of two metals is demonstrated. Ultrathin tantalum carbide, nitride, and boride are grown using chemical vapor deposition by heating a tantalum-copper bilayer with corresponding precursor (C 2 H 2 , B powder, and NH 3 ). The ultrathin crystals are found on the copper surface (opposite of the metal-metal junction). A detailed microscopy analysis followed by density functional theory based calculation demonstrates the migration mechanism, where Ta atoms prefer to stay in clusters in the Cu matrix. These ultrathin materials have good interface attachment with Cu, improving the scratch resistance and oxidation resistance of Cu. This metal-metal immiscibility system can be extended to other metals to synthesize metal carbide, boride, and nitride coatings. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Safety research needs for carbide and nitride fueled LMFBR's. Final report

International Nuclear Information System (INIS)

Kastenberg, W.E.

1975-01-01

The results of a study initiated at UCLA during the academic year 1974--1975 to evaluate and review the potential safety related research needs for carbide and nitride fueled LMFBR's are presented. The tasks included the following: (1) Review Core and primary system designs for any significant differences from oxide fueled reactors, (2) Review carbide (and nitride) fuel element irradiation behavior, (3) Review reactor behavior in postulated accidents, (4) Examine analytical methods of accident analysis to identify major gaps in models and data, and (5) Examine post accident heat removal. (TSS)

Irradiation of a 19 pin subassembly with mixed carbide fuel in KNK II

Science.gov (United States)

Geithoff, D.; Mühling, G.; Richter, K.

1992-06-01

The presentation deals with the fabrication, irradiation and nondestructive postirradiation examinations of LMR fuel pins with mixed (U, Pu)-carbide fuels. The mixed carbide fuel was fabricated by the European Institute of Transuranium Elements using various fabrication procedures. Fuel composition varied therefore in a wide range of tolerances with respect to oxygen and phase content and microstructure. The 19 carbide pins were irradiated in the fast neutron flux of the KNK II reactor to a burn-up of about 7 at% without any failure in the centre of a KNK "carrier element" at a maximum linear rating of 800 W/cm. After dismantling in the Hot Cells of KfK nondestructive examinations were carried out comprising dimensional controls, radiography, γ-scanning and eddy-current testing. The results indicate differences in fuel behaviour with respect to composition of the fuel.

A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

International Nuclear Information System (INIS)

Vasudevamurthy, G.; Radecka, A.; Massey, C.

2015-01-01

High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

Energy Technology Data Exchange (ETDEWEB)

Vasudevamurthy, G.; Radecka, A.; Massey, C. [Virginia Commonwealth Univ., Richmond, VA (United States). High Temperature Materials Lab.

2015-07-01

High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

Catalytic activity of metall-like carbides in carbon oxide oxidation reaction

International Nuclear Information System (INIS)

Kharlamov, A.I.; Kosolapova, T.Ya.; Rafal, A.N.; Kirillova, N.V.

1980-01-01

Kinetics of carbon oxide oxidation upon carbides of hafnium, niobium, tantalum, molybdenum, zirconium and chromium is studied. Probable mechanism of the catalysts action is suggested. The established character of the change of the carbide catalytic activity is explained by the change of d-electron contribution to the metal-metal interaction

Features of order-disorder phase transformation in nonstoichiometric transition metals carbides

International Nuclear Information System (INIS)

Emel'yanov, A.N.

1996-01-01

Measurements of temperature and electric conductivity of nonstoichiometric transition metals carbides TiC χ and NbC χ in the area of order-disorder phase transformation are carried out. There are certain peculiarities on the temperature and electric conductivity curves of the carbides, connected with the carbon sublattice disordering. On the basis of the anomalies observed on the curves of the temperature conductivity of nonstoichiometric carbides of transition metals above the temperature of the order-disorder transition the existence of the second structural transition is supposed

Separation of Nuclear Fuel Surrogates from Silicon Carbide Inert Matrix

International Nuclear Information System (INIS)

Baney, Ronald

2008-01-01

The objective of this project has been to identify a process for separating transuranic species from silicon carbide (SiC). Silicon carbide has become one of the prime candidates for the matrix in inert matrix fuels, (IMF) being designed to reduce plutonium inventories and the long half-lives actinides through transmutation since complete reaction is not practical it become necessary to separate the non-transmuted materials from the silicon carbide matrix for ultimate reprocessing. This work reports a method for that required process

Point defects and transport properties in carbides

International Nuclear Information System (INIS)

Matzke, Hj.

1984-01-01

Carbides of transition metals and of actinides are interesting and technologically important. The transition-metal carbides (or carbonitrides) are extensively being used as hard materials and some of them are of great interest because of the high transition temperature for superconductivity, e.g. 17 K for Nb(C,N). Actinide carbides and carbonitrides, (U,Pu)C and (U,Pu)(C,N) are being considered as promising advanced fuels for liquid metal cooled fast breeder nuclear reactors. Basic interest exists in all these materials because of their high melting points (e.g. 4250 K for TaC) and the unusually broad range of homogeneity of nonstoichiometric compositions (e.g. from UCsub(0.9) to UCsub(1.9) at 2500 K). Interaction of point defects to clusters and short-range ordering have recently been studied with elastic neutron diffraction and diffuse scattering techniques, and calculations of energies of formation and interaction of point defects became available for selected carbides. Diffusion measurements also exist for a number of carbides, in particular for the actinide carbides. The existing knowledge is discussed and summarized with emphasis on informative examples of particular technological relevance. (Auth.)

Status of steady-state irradiation testing of mixed-carbide fuel designs

International Nuclear Information System (INIS)

Harry, G.R.

1983-01-01

The steady-state irradiation program of mixed-carbide fuels has demonstrated clearly the ability of carbide fuel pins to attain peak burnup greater than 12 at.% and peak fluences of 1.4 x 10 23 n/cm 2 (E > 0.1 MeV). Helium-bonded fuel pins in 316SS cladding have achieved peak burnups of 20.7 at.% (192 MWd/kg), and no breaches have occurred in pins of this design. Sodium-bonded fuel pins in 316SS cladding have achieved peak burnups of 15.8 at.% (146 MWd/kg). Breaches have occurred in helium-bonded fuel pins in PE-16 cladding (approx. 5 at.% burnup) and in D21 cladding (approx. 4 at.% burnup). Sodium-bonded fuel pins achieved burnups over 11 at.% in PE-16 cladding and over 6 at.% in D9 and D21 cladding

GEN IV: Carbide Fuel Elaboration for the 'Futurix Concepts' experiment

International Nuclear Information System (INIS)

Vaudez, Stephane; Riglet-Martial, Chantal; Paret, Laurent; Abonneau, Eric

2008-01-01

In order to collect information on the behaviour of the future GFR (Gas Fast Reactor) fuel under fast neutron irradiation, an experimental irradiation program, called 'Futurix-concepts' has been launched at the CEA. The considered concept is a composite material made of a fissile fuel embedded in an inert ceramic matrix. Fissile fuel pellets are made of UPuN or UPuC while ceramics are SiC for the carbide fuel and TiN for the nitride fuel. This paper focuses on the description of the carbide composite fabrication. The UPuC pellets are manufactured using a metallurgical powder process. Fabrication and handling of the fuels are carried out in glove boxes under a nitrogen atmosphere. Carbide fuel is synthesized by carbo-thermic reduction under vacuum of a mixture of actinide oxide and graphitic carbon up to 1550 deg. C. After ball milling, the UPuC powder is pressed to create hexagonal or spherical compacts. They are then sintered up to 1750 deg. C in order to obtain a density of 85 % of the theoretical one. The sintered pellets are inserted into an inert and tight capsule of SiC. In order to control the gap between the fuel and the matrix precisely, the pellets are abraded. The inert matrix is then filled with the pellets and the whole system is sealed by a BRASiC R process at high temperature under a helium atmosphere. Fabrication of the sample to be irradiated was done in 2006 and the irradiation began in May 2007 in the Phenix reactor. This presentation will detail and discuss the results obtained during this fabrication phase. (authors)

Molten fuel-coolant interaction behaviours of various fast reactor fuels (Paper No. HMT-45-87)

International Nuclear Information System (INIS)

Doshi, J.B.

1987-01-01

A parametric computational model of molten fuel-coolant interaction (MFCI) including a particle size distribution is developed and employed to analyse behaviours of various possible reactor fuels, such as oxide, carbide and metal in MFCI scenario. It is observed that while higher thermal conductivity and lower specific heat of carbide compared to oxide is responsible for higher peak pressure and work done per unit mass, the trend is not observed in the metal fuel. The reason for this is the lower operation temperature and latent heat of metallic fuel. (author). 9 refs., 1 fig

Development of a Robust Tri-Carbide Fueled Reactor for Multi-Megawatt Space Power and Propulsion Applications

International Nuclear Information System (INIS)

Samim Anghaie; Knight, Travis W.; Plancher, Johann; Gouw, Reza

2004-01-01

An innovative reactor core design based on advanced, mixed carbide fuels was analyzed for nuclear space power applications. Solid solution, mixed carbide fuels such as (U,Zr,Nb)c and (U,Zr, Ta)C offer great promise as an advanced high temperature fuel for space power reactors

Characterization of fuel swelling in helium-bonded carbide fuel pins

International Nuclear Information System (INIS)

Louie, D.L.Y.

1987-08-01

This work is not only the first attempt at characterizing the swelling of (U,Pu)C fuel pellets, but it also represents the only detailed examinations on carbide fuel swelling at high fuel burnups (4 to 16 at. %). This characterization includes the contributions of fission gases, cracks and solid fission products to fuel swelling. Significantly, the contributions of fission gases and cracks were determined by using the image analysis technique (IAT) which allows researchers to take areal measurements of the irradiated fuel porosity and cracks from the photographs of metallographic fuel samples. However, because areal measurements for varying depths in the fuel pellet could not be obtained, the crack areal measurements could not be converted into volumetric quantities. Consequently, in this situation, an areal fuel swelling analysis was used. The macroscopic fission-gas induced fuel swelling (MAS) caused by fission-gas bubbles and pores > 1 μm was determined using the measured irradiated fuel porosity because the measuring range of IAT is limited to bubbles and pores >1 μm. Conversely, for fuel swelling induced by fission-gas bubbles < 1 μm, the microscopic fission-gas induced fuel swelling (MIS) was estimated using an areal fuel swelling model

High 240Pu FTR/EMC experiments and analysis: Carbide fuel and UO2 blanket subassembly worths

International Nuclear Information System (INIS)

Ombrellaro, P.A.

1977-06-01

Carbide-plutonium fuel and UO 2 blanket subassembly worth measurements performed at ANL in the EMC/LWR were analyzed. Composition exchange worth calculations were performed for: (a) the replacement of high- 240 Pu fuel composition for low- 240 Pu fuel composition and carbide-plutonium fuel composition, successively, in the center subassembly of the core; (b) the replacement of low- 240 Pu fuel composition for carbide--plutonium fuel composition in one outer driver subassembly; and (c) the replacement of the radial reflector composition with UO 2 blanket composition in one subassembly of the radial reflector. The composition exchange worth calculations were performed in two-dimensional x,y geometry, using diffusion theory and perturbation theory. Each method produces about the same calculated-to-experimental bias factors

Characterization of Transition Metal Carbide Layers Synthesized by Thermo-reactive Diffusion Processes

DEFF Research Database (Denmark)

Laursen, Mads Brink; Fernandes, Frederico Augusto Pires; Christiansen, Thomas Lundin

2015-01-01

. In this study halide-activated pack cementation techniques were used on tool steel Vanadis 6 and martensitic stainless steel AISI 420 in order to produce hard layers of titanium carbide (TiC), vanadium carbide (V8C7) and chromium carbides (Cr23C6 and Cr7C3). Surface layers were characterized by scanning......Hard wear resistant surface layers of transition metal carbides can be produced by thermo-reactive diffusion processes where interstitial elements from a steel substrate together with external sources of transition metals (Ti, V, Cr etc.) form hard carbide and/or nitride layers at the steel surface...... electron microscopy, X-ray diffraction and Vickers hardness testing. The study shows that porosityfree, homogenous and very hard surface layers can be produced by thermo-reactive diffusion processes. The carbon availability of the substrate influences thickness of obtained layers, as Vanadis 6 tool steel...

The compatibility of stainless steels with particles and powders of uranium carbide and low-sulphur UCS fuels

International Nuclear Information System (INIS)

Venter, S.

1978-05-01

Slightly hyperstoichiometric (U,Pu)C is a potential nuclear fuel for fast breeder reactors. The excess carbon above the stoichiometric amount results in a higher carbon activity in the fuel, and carbon is transferred to the stainless steel cladding, resulting in embrittlement of the cladding. It is with this problem of carbon transfer from the fuel to the cladding that this thesis is concerned. For practical reasons, UC and not (U,Pu)C was used as the fuel. The theory of decarburisation of carbide fuel and the carburisation of stainless steel, the facilities constructed for the project at the Atomic Energy Board, and the experimental techniques used, including preparation of the fuels, are discussed. The effect of a number of variables of uranium carbide fuel on its compatibility behaviour with stainless steels was investigated, as well as the effect om microstructure and type of stainless steel (304, 304 L and 316) on the rate of carburisation. These studies can be briefly summarised under the following headings: powder-particle size; surface oxidation of uranium carbide; preparation temperature of uranium carbide; low sulfur UCS fuels; uranium sulfide and the microstructure and type of steel. The author concludes that: the effect of surface oxidation and particle size must be taken into account when evaluating out-of-pile tests; the possible effects of surface oxidation must be taken into account when considering vibro-compacted carbide fuels; there is no advantage in replacing a fraction of the carbon atoms by sulphur atoms in slightly hyperstoichiometric carbide fuels, and the type and thermo-mechanical treatment of the stainless steel used as cladding material in a fuel pin is not important as far as the rate of carburisation by the fuel is concerned

Corrosion resistant cemented carbide

International Nuclear Information System (INIS)

Hong, J.

1990-01-01

This paper describes a corrosion resistant cemented carbide composite. It comprises: a granular tungsten carbide phase, a semi-continuous solid solution carbide phase extending closely adjacent at least a portion of the grains of tungsten carbide for enhancing corrosion resistance, and a substantially continuous metal binder phase. The cemented carbide composite consisting essentially of an effective amount of an anti-corrosion additive, from about 4 to about 16 percent by weight metal binder phase, and with the remaining portion being from about 84 to about 96 percent by weight metal carbide wherein the metal carbide consists essentially of from about 4 to about 30 percent by weight of a transition metal carbide or mixtures thereof selected from Group IVB and of the Periodic Table of Elements and from about 70 to about 96 percent tungsten carbide. The metal binder phase consists essentially of nickel and from about 10 to about 25 percent by weight chromium, the effective amount of an anti-corrosion additive being selected from the group consisting essentially of copper, silver, tine and combinations thereof

Electronic specific heat of transition metal carbides

International Nuclear Information System (INIS)

Conte, R.

1964-07-01

The experimental results that make it possible to define the band structure of transition metal carbides having an NaCI structure are still very few. We have measured the electronic specific heat of some of these carbides of varying electronic concentration (TiC, either stoichiometric or non-stoichiometric, TaC and mixed (Ti, Ta) - C). We give the main characteristics (metallography, resistivity, X-rays) of our samples and we describe the low temperature specific heat apparatus which has been built. In one of these we use helium as the exchange gas. The other is set up with a mechanical contact. The two use a germanium probe for thermometer. The measurement of the temperature using this probe is described, as well as the various measurement devices. The results are presented in the form of a rigid band model and show that the density of the states at the Fermi level has a minimum in the neighbourhood of the group IV carbides. (author) [fr

Failure analysis of carbide fuels under transient overpower (TOP) conditions

International Nuclear Information System (INIS)

Nguyen, D.H.

1980-06-01

The failure of carbide fuels in the Fast Test Reactor (FTR) under Transient Overpower (TOP) conditions has been examined. The Beginning-of-Cycle Four (BOC-4) all-oxide base case, at $.50/sec ramp rate was selected as the reference case. A coupling between the advanced fuel performance code UNCLE-T and HCDA Code MELT-IIIA was necessary for the analysis. UNCLE-T was used to determine cladding failure and fuel preconditioning which served as initial conditions for MELT-III calculations. MELT-IIIA determined the time of molten fuel ejection from fuel pin

Dissolution of nuclear fuel samples for analytical purposes. I

International Nuclear Information System (INIS)

Krtil, J.

1983-01-01

Main attention is devoted to procedures for dissolving fuels based on uranium metal and its alloys, uranium oxides and carbides, plutonium metal, plutonium dioxide, plutonium carbides, mixed PuC-UC carbides and mixed oxides (PuU)O 2 . Data from the literature and experience gained with the dissolution of nuclear fuel samples at the Central Control Laboratory of the Nuclear Research Institute at Rez are given. (B.S.)

Design and fuel fabrication processes for the AC-3 mixed-carbide irradiation test

International Nuclear Information System (INIS)

Latimer, T.W.; Chidester, K.M.; Stratton, R.W.; Ledergerber, G.; Ingold, F.

1992-01-01

The AC-3 test was a cooperative U.S./Swiss irradiation test of 91 wire-wrapped helium-bonded U-20% Pu carbide fuel pins irradiated to 8.3 at % peak burnup in the Fast Flux Test Facility. The test consisted of 25 pins that contained spherepac fuel fabricated by the Paul Scherrer Institute (PSI) and 66 pins that contained pelletized fuel fabricated by the Los Alamos National Laboratory. Design of AC-3 by LANL and PSI was begun in 1981, the fuel pins were fabricated from 1983 to 1985, and the test was irradiated from 1986 to 1988. The principal objective of the AC-3 test was to compare the irradiation performance of mixed-carbide fuel pins that contained either pelletized or sphere-pac fuel at prototypic fluence and burnup levels for a fast breeder reactor

GEN IV: Carbide Fuel Elaboration for the 'Futurix Concepts' experiment

Energy Technology Data Exchange (ETDEWEB)

Vaudez, Stephane; Riglet-Martial, Chantal; Paret, Laurent; Abonneau, Eric [Commissariat a l' Energie Atomique (C.E.A.), Direction de l' Energie Nucleaire, Centre d' Etudes de Cadarache, 13108 Saint Paul lez Durance Cedex (France)

2008-07-01

In order to collect information on the behaviour of the future GFR (Gas Fast Reactor) fuel under fast neutron irradiation, an experimental irradiation program, called 'Futurix-concepts' has been launched at the CEA. The considered concept is a composite material made of a fissile fuel embedded in an inert ceramic matrix. Fissile fuel pellets are made of UPuN or UPuC while ceramics are SiC for the carbide fuel and TiN for the nitride fuel. This paper focuses on the description of the carbide composite fabrication. The UPuC pellets are manufactured using a metallurgical powder process. Fabrication and handling of the fuels are carried out in glove boxes under a nitrogen atmosphere. Carbide fuel is synthesized by carbo-thermic reduction under vacuum of a mixture of actinide oxide and graphitic carbon up to 1550 deg. C. After ball milling, the UPuC powder is pressed to create hexagonal or spherical compacts. They are then sintered up to 1750 deg. C in order to obtain a density of 85 % of the theoretical one. The sintered pellets are inserted into an inert and tight capsule of SiC. In order to control the gap between the fuel and the matrix precisely, the pellets are abraded. The inert matrix is then filled with the pellets and the whole system is sealed by a BRASiC{sup R} process at high temperature under a helium atmosphere. Fabrication of the sample to be irradiated was done in 2006 and the irradiation began in May 2007 in the Phenix reactor. This presentation will detail and discuss the results obtained during this fabrication phase. (authors)

Platinum group metal nitrides and carbides: synthesis, properties and simulation

International Nuclear Information System (INIS)

Ivanovskii, Alexander L

2009-01-01

Experimental and theoretical data on new compounds, nitrides and carbides of the platinum group 4d and 5d metals (ruthenium, rhodium, palladium, osmium, iridium, platinum), published over the past five years are summarized. The extreme mechanical properties of platinoid nitrides and carbides, i.e., their high strength and low compressibility, are noted. The prospects of further studies and the scope of application of these compounds are discussed.

Analysis of refabricated fuel: determination of carbon in uranium plutonium mixed carbide

International Nuclear Information System (INIS)

Huwyler, S.

1977-09-01

In developing uranium plutonium mixed carbide which represents an advanced fuel for breeder reactors carbon analysis is an important means of determining the stoichiometry. Methods of carbon determination are briefly reviewed. The carbon determination using a LECO WR-12 Carbon Determinator is treated in detail and experience of three years operation communicated. Problems arising from operating the LECO-apparatus in a glove box are discussed. It is pointed out that carbon determination with the LECO-apparatus is a very fast method with good precision and well suited for the routine analysis of mixed carbide fuel. The accuracy of the method is checked by means of a standard. (Auth.)

Neutron irradiation damage in transition metal carbides

International Nuclear Information System (INIS)

Matsui, Hisayuki; Nesaki, Kouji; Kiritani, Michio

1991-01-01

Effects of neutron irradiation on the physical properties of light transition metal carbides, TiC x , VC x and NbC x , were examined, emphasizing the characterization of irradiation induced defects in the nonstoichiometric composition. TiC x irradiated with 14 MeV (fusion) neutrons showed higher damage rates with increasing C/Ti (x) ratio. A brief discussion is made on 'cascade damage' in TiC x irradiated with fusion neutrons. Two other carbides (VC x and NbC x ) were irradiated with fission reactor neutrons. The irradiation effects on VC x were not so simple, because of the complex irradiation behavior of 'ordered' phases. For instance, complete disordering was revealed in an ordered phase, 'V 8 C 7 ', after an irradiation dose of 10 25 n/m 2 . (orig.)

Generation and Characteristics of IV-VI transition Metal Nitride and Carbide Nanoparticles using a Reactive Mesoporous Carbon Nitride

KAUST Repository

Alhajri, Nawal Saad

2016-02-22

Interstitial nitrides and carbides of early transition metals in groups IV–VI exhibit platinum-like electronic structures, which make them promising candidates to replace noble metals in various catalytic reactions. Herein, we present the preparation and characterization of nano-sized transition metal nitries and carbides of groups IV–VI (Ti, V, Nb, Ta, Cr, Mo, and W) using mesoporous graphitic carbon nitride (mpg-C3N4), which not only provides confined spaces for restricting primary particle size but also acts as a chemical source of nitrogen and carbon. We studied the reactivity of the metals with the template under N2 flow at 1023 K while keeping the weight ratio of metal to template constant at unity. The produced nanoparticles were characterized by powder X-ray diffraction, CHN elemental analysis, nitrogen sorption, X-ray photoelectron spectroscopy, and transmission electron microscopy. The results show that Ti, V, Nb, Ta, and Cr form nitride phases with face centered cubic structure, whereas Mo and W forme carbides with hexagonal structures. The tendency to form nitride or carbide obeys the free formation energy of the transition metal nitrides and carbides. This method offers the potential to prepare the desired size, shape and phase of transition metal nitrides and carbides that are suitable for a specific reaction, which is the chief objective of materials chemistry.

A Review of Metal Injection Molding- Process, Optimization, Defects and Microwave Sintering on WC-Co Cemented Carbide

Science.gov (United States)

Shahbudin, S. N. A.; Othman, M. H.; Amin, Sri Yulis M.; Ibrahim, M. H. I.

2017-08-01

This article is about a review of optimization of metal injection molding and microwave sintering process on tungsten cemented carbide produce by metal injection molding process. In this study, the process parameters for the metal injection molding were optimized using Taguchi method. Taguchi methods have been used widely in engineering analysis to optimize the performance characteristics through the setting of design parameters. Microwave sintering is a process generally being used in powder metallurgy over the conventional method. It has typical characteristics such as accelerated heating rate, shortened processing cycle, high energy efficiency, fine and homogeneous microstructure, and enhanced mechanical performance, which is beneficial to prepare nanostructured cemented carbides in metal injection molding. Besides that, with an advanced and promising technology, metal injection molding has proven that can produce cemented carbides. Cemented tungsten carbide hard metal has been used widely in various applications due to its desirable combination of mechanical, physical, and chemical properties. Moreover, areas of study include common defects in metal injection molding and application of microwave sintering itself has been discussed in this paper.

Transition metal carbide nanocomposite and amorphous thin films

OpenAIRE

Tengstrand, Olof

2014-01-01

This thesis explores thin films of binary and ternary transition metal carbides, in the Nb-C, Ti-Si-C, Nb-Si-C, Zr-Si-C, and Nb-Ge-C systems. The electrical and mechanical properties of these systems are affected by their structure and here both nanocomposite and amorphous thin films are thus investigated. By appropriate choice of transition metal and composition the films can be designed to be multifunctional with a combination of properties, such as low electric resistivity, low contact res...

Properties of zirconium carbide for nuclear fuel applications

Energy Technology Data Exchange (ETDEWEB)

Katoh, Yutai; Vasudevamurthy, Gokul, E-mail: gvasudev@vcu.edu; Nozawa, Takashi; Snead, Lance L.

2013-10-15

Zirconium carbide (ZrC) is a potential coating, oxygen-gettering, or inert matrix material for advanced high temperature reactor fuels. ZrC has demonstrated attractive properties for these fuel applications including excellent resistance against fission product corrosion and fission product retention capabilities. However, fabrication of ZrC results in a range of stable sub-stoichiometric and carbon-rich compositions with or without substantial microstructural inhomogeneity, textural anisotropy, and a phase separation, leading to variations in physical, chemical, thermal, and mechanical properties. The effects of neutron irradiation at elevated temperatures, currently only poorly understood, are believed to be substantially influenced by those compositional and microstructural features further adding complexity to understanding the key ZrC properties. This article provides a survey of properties data for ZrC, as required by the United States Department of Energy’s advanced fuel programs in support of the current efforts toward fuel performance modeling and providing guidance for future research on ZrC for fuel applications.

Change of Composition in Metallic Fuel Slug of U-Zr Alloy from High-Temperature Annealing

Energy Technology Data Exchange (ETDEWEB)

Youn, Young Sang; Lee, Jeong Mook; Kim, Jong Yun; Kim, Jong Hwan; Song, Hoon [KAERI, Daejeon (Korea, Republic of)

2016-09-15

The U–Zr alloy is a candidate for fuel to be used as metallic fuel in sodium-cooled fast reactors (SFRs). Its chemical composition before and after annealing at the operational temperature of SFRs (610 .deg. C) was investigated using X-ray photoelectron spectroscopy, Raman spectroscopy, and X-ray diffraction. The original alloy surface contained uranium oxides with the U(IV) and U(VI) oxidation states, Zr{sub 2}O{sub 3}, and a low amount of uranium metal. After annealing at 610 .deg. C, the alloy was composed of uranium metal, uranium carbide, uranium oxide with the U(V) valence state, zirconium metal, and amorphous carbon. Meanwhile, X-ray diffraction data indicate that the bulk composition of the alloy remained unchanged.

Change of Composition in Metallic Fuel Slug of U-Zr Alloy from High-Temperature Annealing

International Nuclear Information System (INIS)

Youn, Young Sang; Lee, Jeong Mook; Kim, Jong Yun; Kim, Jong Hwan; Song, Hoon

2016-01-01

The U–Zr alloy is a candidate for fuel to be used as metallic fuel in sodium-cooled fast reactors (SFRs). Its chemical composition before and after annealing at the operational temperature of SFRs (610 .deg. C) was investigated using X-ray photoelectron spectroscopy, Raman spectroscopy, and X-ray diffraction. The original alloy surface contained uranium oxides with the U(IV) and U(VI) oxidation states, Zr 2 O 3 , and a low amount of uranium metal. After annealing at 610 .deg. C, the alloy was composed of uranium metal, uranium carbide, uranium oxide with the U(V) valence state, zirconium metal, and amorphous carbon. Meanwhile, X-ray diffraction data indicate that the bulk composition of the alloy remained unchanged

The oxidative corrosion of carbide inclusions at the surface of uranium metal during exposure to water vapour

International Nuclear Information System (INIS)

Scott, T.B.; Petherbridge, J.R.; Harker, N.J.; Ball, R.J.; Heard, P.J.; Glascott, J.; Allen, G.C.

2011-01-01

Highlights: → High resolution imagery (FIB, SEM and SIMS) of carbide inclusions in uranium metal. → Real time images following the reaction of the carbide inclusions with water vapour. → Shown preferential consumption of carbide over that of the bulk metal. → Quantity of impurities in the metal therefore seriously influence reaction rate. → Metal purity must be considered when storing uranium in air or moist conditions. - Abstract: The reaction between uranium and water vapour has been well investigated, however discrepancies exist between the described kinetic laws, pressure dependence of the reaction rate constant and activation energies. Here this problem is looked at by examining the influence of impurities in the form of carbide inclusions on the reaction. Samples of uranium containing 600 ppm carbon were analysed during and after exposure to water vapour at 19 mbar pressure, in an environmental scanning electron microscope (ESEM) system. After water exposure, samples were analysed using secondary ion mass spectrometry (SIMS), focused ion beam (FIB) imaging and sectioning and transmission electron microscopy (TEM) with X-ray diffraction (micro-XRD). The results of the current study indicate that carbide particles on the surface of uranium readily react with water vapour to form voluminous UO 3 .xH 2 O growths at rates significantly faster than that of the metal. The observation may also have implications for previous experimental studies of uranium-water interactions, where the presence of differing levels of undetected carbide may partly account for the discrepancies observed between datasets.

Review of experimental studies of zirconium carbide coated fuel particles for high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Minato, Kazuo; Ogawa, Toru; Fukuda, Kousaku

1995-03-01

Experimental studies of zirconium carbide(ZrC) coated fuel particles were reviewed from the viewpoints of fuel particle designs, fabrication, characterization, fuel performance, and fission product retentiveness. ZrC is known as a refractory and chemically stable compound, so ZrC is a candidate to replace the silicon carbide(SiC) coating layer of the Triso-coated fuel particles. The irradiation experiments, the postirradiation heating tests, and the out-of-reactor experiments showed that the ZrC layer was less susceptible than the SiC layer to chemical attack by fission products and fuel kernels, and that the ZrC-coated fuel particles performed better than the standard Triso-coated fuel particles at high temperatures, especially above 1600degC. The ZrC-coated fuel particles demonstrated better cesium retention than the standard Triso-coated fuel particles though the ZrC layer showed a less effective barrier to ruthenium than the SiC layer. (author) 51 refs

Effect of deposition conditions on the properties of pyrolytic silicon carbide coatings for high-temperature gas-cooled reactor fuel particles

International Nuclear Information System (INIS)

Stinton, D.P.; Lackey, W.J.

1977-10-01

Silicon carbide coatings on HTGR microsphere fuel act as the barrier to contain metallic fission products. Silicon carbide coatings were applied by the decomposition of CH 3 SiCl 3 in a 13-cm-diam (5-in.) fluidized-bed coating furnace. The effects of temperature, CH 3 SiCl 3 supply rate and the H 2 :CH 3 SiCl 3 ratio on coating properties were studied. Deposition temperature was found to control coating density, whole particle crushing strength, coating efficiency, and microstructure. Coating density and microstructure were also partially determined by the H 2 :CH 3 SiCl 3 ratio. From this work, it appears that the rate at which high quality SiC can be deposited can be increased from 0.2 to 0.5 μm/min

Friction and Wear of Metals With a Single-Crystal Abrasive Grit of Silicon Carbide - Effect of Shear Strength of Metal

National Research Council Canada - National Science Library

Miyoshi, Kazuhisa

1978-01-01

An investigation was conducted to examine the removal and plastic deformation of metal as a function of the metal properties when the metal is in sliding contact with a single-crystal abrasive grit of silicon carbide...

Novel fabrication of silicon carbide based ceramics for nuclear applications

Science.gov (United States)

Singh, Abhishek Kumar

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 materials include refractory alloys based on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as SiC--SiCf; 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 fuel can lower the center-line temperature and, thereby, enhance power production capabilities and reduce the risk of premature fuel pellet failure. Crystalline silicon carbide has superior characteristics as a structural material from the viewpoint of its thermal and mechanical properties, thermal shock resistance, chemical stability, and low radioactivation. Therefore, there have been many efforts to develop SiC based composites in various forms for use in advanced energy systems. In recent years, with the development of high yield preceramic precursors, the polymer infiltration and pyrolysis (PIP) method has aroused interest for the fabrication of ceramic based materials, for various applications ranging from disc brakes to nuclear reactor fuels. The pyrolysis of preceramic polymers allow new types of ceramic materials to be processed at relatively low temperatures. The raw materials are element-organic polymers whose composition and architecture can be tailored and varied. The primary focus of this study is to use a pyrolysis based process to fabricate a host of novel silicon carbide-metal carbide or oxide composites, and to synthesize new materials based on mixed-metal silicocarbides that cannot be processed using conventional techniques. Allylhydridopolycarbosilane (AHPCS), which is an organometal polymer, was used as the precursor for silicon carbide. Inert gas pyrolysis of AHPCS produces near-stoichiometric amorphous

Ceramics as nuclear reactor fuels

International Nuclear Information System (INIS)

Reeve, K.D.

1975-01-01

Ceramics are widely accepted as nuclear reactor fuel materials, for both metal clad ceramic and all-ceramic fuel designs. Metal clad UO 2 is used commercially in large tonnages in five different power reactor designs. UO 2 pellets are made by familiar ceramic techniques but in a reactor they undergo complex thermal and chemical changes which must be thoroughly understood. Metal clad uranium-plutonium dioxide is used in present day fast breeder reactors, but may eventually be replaced by uranium-plutonium carbide or nitride. All-ceramic fuels, which are necessary for reactors operating above about 750 0 C, must incorporate one or more fission product retentive ceramic coatings. BeO-coated BeO matrix dispersion fuels and silicate glaze coated UO 2 -SiO 2 have been studied for specialised applications, but the only commercial high temperature fuel is based on graphite in which small fuel particles, each coated with vapour deposited carbon and silicon carbide, are dispersed. Ceramists have much to contribute to many aspects of fuel science and technology. (author)

Dependence of silicon carbide coating properties on deposition parameters: preliminary report

International Nuclear Information System (INIS)

Lauf, R.J.; Braski, D.N.

1980-05-01

Fuel particles for the High-Temperature Gas-Cooled Reactor (HTGR) contain a layer of pyrolytic silicon carbide, which acts as a pressure vessel and provides containment of metallic fission products. The silicon carbide (SiC) is deposited by the thermal decomposition of methyltrichlorosilane (CH 3 SiCl 3 or MTS) in an excess of hydrogen. The purpose of the current study is to determine how the deposition variables affect the structure and properties of the SiC layer

Evaluation of titanium carbide metal matrix composites deposited via laser cladding

Science.gov (United States)

Cavanaugh, Daniel Thomas

Metal matrix composites have been widely studied in terms of abrasion resistance, but a particular material system may behave differently as particle size, morphology, composition, and distribution of the hardening phase varies. The purpose of this thesis was to understand the mechanical and microstructural effects of combining titanium carbide with 431 series stainless steel to create a unique composite via laser cladding, particularly regarding wear properties. The most predominant effect in increasing abrasion resistance, measured via ASTM G65, was confirmed to be volume fraction of titanium carbide addition. Macrohardness was directly proportional to the amount of carbide, though there was an overall reduction in individual particle microhardness after cladding. The reduction in particle hardness was obscured by the effect of volume fraction carbide and did not substantially contribute to the wear resistance changes. A model evaluating effective mean free path of the titanium carbide particles was created and correlated to the measured data. The model proved successful in linking theoretical mean free path to overall abrasion resistance. The effects of the titanium carbide particle distributions were limited, while differences in particle size were noticeable. The mean free path model did not correlate well with the particle size, but it was shown that the fine carbides were completely removed by the coarse abrasive particles in the ASTM G65 test. The particle morphology showed indications of influencing the wear mode, but no statistical reduction was observed in the volume loss figures. Future studies may more specifically focus on particle morphology or compositional effects of the carbide particles.

Post irradiation examinations of uranium-plutonium mixed carbide fuels irradiated at low linear power rate

International Nuclear Information System (INIS)

Maeda, Atsushi; Sasayama, Tatsuo; Iwai, Takashi; Aizawa, Sakuei; Ohwada, Isao; Aizawa, Masao; Ohmichi, Toshihiko; Handa, Muneo

1988-11-01

Two pins containing uranium-plutonium carbide fuels which are different in stoichiometry, i.e. (U,Pu)C 1.0 and (U,Pu)C 1.1 , were constructed into a capsule, ICF-37H, and were irradiated in JRR-2 up to 1.0 at % burnup at the linear heat rate of 420 W/cm. After being cooled for about one year, the irradiated capsule was transferred to the Reactor Fuel Examination Facility where the non-destructive examinations of the fuel pins in the β-γ cells and the destructive ones in two α-γ inert gas atmosphere cells were carried out. The release rates of fission gas were low enough, 0.44 % from (U,Pu)C 1.0 fuel pin and 0.09% from (U,Pu)C 1.1 fuel pin, which is reasonable because of the low central temperature of fuel pellets, about 1000 deg C and is estimated that the release is mainly governed by recoil and knock-out mechanisms. Volume swelling of the fuels was observed to be in the range of 1.3 ∼ 1.6 % for carbide fuels below 1000 deg C. Respective open porosities of (U,Pu)C 1.0 and (U,Pu)C 1.1 fuel were 1.3 % and 0.45 %, being in accordance with the release behavior of fission gas. Metallographic observation of the radial sections of pellets showed the increase of pore size and crystal grain size in the center and middle region of (U,Pu)C 1.0 pellets. The chemical interaction between fuel pellets and claddings in the carbide fuels is the penetration of carbon in the fuels to stainless steel tubes. The depth of corrosion layer in inner sides of cladding tubes ranged 10 ∼ 15 μm in the (U,Pu)C 1.0 fuel and 15 #approx #25 μm in the (U,Pu)C 1.1 fuel, which is correlative with the carbon potential of fuels posibly affecting the amount of carbon penetration. (author)

Chemical Engineering Division Fuel Cycle Programs. Quarterly progress report, April-June 1978. [Advanced solvent extraction; accidents; pyrochemical; radwaste in metal matrix; waste migration

Energy Technology Data Exchange (ETDEWEB)

Steindler, M. J.; Ader, M.; Barletta, R. E.

1979-12-01

Fuel cycle studies reported include development of centrifugal contactors for Purex processes. Tricaprylmethyl-ammonium nitrate and di-n-amyl-n-amylphosphonate are being evaluated as Thorex extractants. Dispersion of uranium and plutonium by fires, and mechanisms for subdividing and dispersing liquids and solids were reviewed. In the pyrochemical and dry processing program, a facility for testing containment materials is under construction; a flowsheet for carbide fuel processing has been designed and studies of carbide reactions in bismuth are underway; salt transport processes are being studied; process-size refractory metal vessels are being fabricated; the feasibility of AIROX reprocessing is being determined; the solubility of UO/sub 2/, UO/sub 2/ + fission products, and PuO/sub 2/ in molten alkali metal nitrates, has been investigated; a flowsheet was developed for reprocessing actinide oxides in molten salts; preparation of Th-U carbide from the oxide is being studied; new flowsheets based on the Dow Aluminum Pyrometallurgical process for reprocessing of spent uranium metal fuel have been prepared; the chloride volitility processing of thorium-based fuels is being studied; the reprocessing of (Th,U)O/sub 2/ solid solution in KCl-LiCl-ThCl/sub 4/-Th is being studied; and a flowsheet for processing spent nuclear fuel in molten tin has been constructed. Leach rates of simulated encapsulated waste forms in a metal matrix were studied. Nine criteria for handling waste cladding hulls were established. Strontium and tin migration in glauconite columns was measured. Radioactive Sr in a stream of water moved through oolitic limestone as rapidly as water, but in a stream of water equilibrated with the limestone, Sr moved through the limestone one-tenth as fast. Migration of trace quantities of Cs and I through kaolinite was studied. 88 figures, 53 tables.

High reliability fuel in the US

International Nuclear Information System (INIS)

Neuhold, R.J.; Leggett, R.D.; Walters, L.C.; Matthews, R.B.

1986-05-01

The fuels development program of the United States is described for liquid metal reactors (LMR's). The experience base, status and future potential are discussed for the three systems - oxide, metal and carbide - that have proved to have high reliability. Information is presented showing burnup capability of the oxide fuel system in a large core, e.g., FFTF, to be 150 MWd/kgM with today's technology with the potential for a capability as high as 300 MWd/kgM. Data provided for the metal fuel system show 8 at. % being routinely achieved as the EBR-II driver fuel with good potential for extending this to 15 at. % since special test pins have already exceeded this burnup level. The data included for the carbide fuel system are from pin and assembly irradiations in EBR-II and FFTF, respectively. Burnup to 12 at. % appears readily achievable with burnups to 20 at. % being demonstrated in a few pins. Efforts continue on all three systems with the bulk of the activity on metal and oxide

Evaluation of mechanical properties of aluminium alloy–alumina–boron carbide metal matrix composites

International Nuclear Information System (INIS)

Vijaya Ramnath, B.; Elanchezhian, C.; Jaivignesh, M.; Rajesh, S.; Parswajinan, C.; Siddique Ahmed Ghias, A.

2014-01-01

Highlights: • Fabrication of MMC with aluminium alloy–alumina–boron carbide is done. • Different proportions of reinforcements are added. • The effects of varying proportions are studied. • Investigation on mechanical properties above composites is performed. • Failure morphology analysis is done using SEM. - Abstract: This paper deals with the fabrication and mechanical investigation of aluminium alloy, alumina (Al 2 O 3 ) and boron carbide metal matrix composites. Aluminium is the matrix metal having properties like light weight, high strength and ease of machinability. Alumina which has better wear resistance, high strength, hardness and boron carbide which has excellent hardness and fracture toughness are added as reinforcements. Here, the fabrication is done by stir casting which involves mixing the required quantities of additives into stirred molten aluminium. After solidification, the samples are prepared and tested to find the various mechanical properties like tensile, flexural, impact and hardness. The internal structure of the composite is observed using Scanning Electron Microscope (SEM)

Particle fuel bed tests

International Nuclear Information System (INIS)

Horn, F.L.; Powell, J.R.; Savino, J.M.

1985-01-01

Gas-cooled reactors, using packed beds of small diameter coated fuel particles have been proposed for compact, high-power systems. The particulate fuel used in the tests was 800 microns in diameter, consisting of a thoria kernel coated with 200 microns of pyrocarbon. Typically, the bed of fuel particles was contained in a ceramic cylinder with porous metallic frits at each end. A dc voltage was applied to the metallic frits and the resulting electric current heated the bed. Heat was removed by passing coolant (helium or hydrogen) through the bed. Candidate frit materials, rhenium, nickel, zirconium carbide, and zirconium oxide were unaffected, while tungsten and tungsten-rhenium lost weight and strength. Zirconium-carbide particles were tested at 2000 K in H 2 for 12 hours with no visible reaction or weight loss

Nuclear fuel concept for the 21st century

International Nuclear Information System (INIS)

Tulenko, J.S.; Schoessow, G.

1996-01-01

In a previous paper, the author presented his rationale for the fuel cycle for the 21st century. This cycle, driven by both environmental and economic factors, required that the fuel should be able to operate in a range from 90 000 MWd/tonne of heavy metal and above. Such an operation would require the development of a cladding material that would not undergo waterside corrosion at these ultrahigh burnups. The University of Florida is proposing a new fuel arrangement that the authors feel meets the demands of high burnup and provides a safer fuel assembly. It is believed that the liquid-metal bond concept combined with a silicon carbide composite cladding and the collapsible fission gas plenum offers outstanding potential for ultrahigh burnup fuels while providing a potentially ultrasafe reactor operation. Efforts at various facilities are under way to determine the radiation stability of silicon carbide fuel and to fabricate SiC materials that will provide the radiation stability needed. Other parameters offer strong incentives to successfully develop silicon carbide as a cladding material

Advanced technologies of production of cemented carbides and composite materials based on them

International Nuclear Information System (INIS)

Bondarenko, V.; Pavlotskaya, E.; Martynova, L.; Epik, I.

2001-01-01

The paper presents new technological processes of production of W, WC and (Ti, W)C powders, cemented carbides having a controlled carbon content, high-strength nonmagnetic nickel-bonded cemented carbides, cemented carbide-based composites having a wear-resistant antifriction working layer as well as processes of regeneration of cemented carbide waste. It is shown that these technological processes permit radical changes in the production of carbide powders and products of VK, TK, VN and KKhN cemented carbides. The processes of cemented carbide production become ecologically acceptable and free of carbon black, the use of cumbersome mixers is excluded, the power expenditure is reduced and the efficiency of labor increases. It becomes possible to control precisely the carbon content within a two-phase region -carbide-metal. A high wear resistance of parts of friction couples which are lubricated with water, benzine, kerosene, diesel fuel and other low-viscosity liquids, is ensured with increased strength and shock resistance. (author)

Chapter 19: Catalysis by Metal Carbides and Nitrides

Energy Technology Data Exchange (ETDEWEB)

Schaidle, Joshua A [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Nash, Connor P [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Yung, Matthew M [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Chen, Yuan [Pacific Northwest National Laboratory; Carl, Sarah [University of Michigan; Thompson, Levi [University of Michigan

2017-08-09

Early transition metal carbides and nitrides (ETMCNs), materials in which carbon or nitrogen occupies interstitial sites within a parent metal lattice, possess unique physical and chemical properties that motivate their use as catalysts. Specifically, these materials possess multiple types of catalytic sites, including metallic, acidic, and basic sites, and as such, exhibit reactivities that differ from their parent metals. Moreover, their surfaces are dynamic under reaction conditions. This chapter reviews recent (since 2010) experimental and computational investigations into the catalytic properties of ETMCN materials for applications including biomass conversion, syngas and CO2 upgrading, petroleum and natural gas refining, and electrocatalytic energy conversion, energy storage, and chemicals production, and attempts to link catalyst performance to active site identity/surface structure in order to elucidate the present level of understanding of structure-function relationships for these materials. The chapter concludes with a perspective on leveraging the unique properties of these materials to design and develop improved catalysts through a dedicated, multidisciplinary effort.

High burnup, high power irradiation behavior of helium-bonded mixed carbide fuel pins

International Nuclear Information System (INIS)

Levine, P.J.; Nayak, U.P.; Boltax, A.

1983-01-01

Large diameter (9.4 mm) helium-bonded mixed carbide fuel pins were successfully irradiated in EBR-II to high burnup (12%) at high power levels (100 kW/m) with peak cladding midwall temperatures of 550 0 C. The wire-wrapped pins were clad with 0.51-mm-thick, 20% cold-worked Type 316 stainless steel and contained hyperstoichiometric (Usub(0.8)Pusub(0.2))C fuel covering the smeared density range from 75-82% TD. Post-irradiation examinations revealed: extensive fuel-cladding mechanical interaction over the entire length of the fuel column, 35% fission gas release at 12% burnup, cladding carburization and fuel restructuring. (orig.)

Stability of Transition-metal Carbides in Liquid Phase Reactions Relevant for Biomass-Based Conversion

NARCIS (Netherlands)

Souza Macêdo, L.; Stellwagen, D.R.; Teixeira da Silva, V.; Bitter, J.H.

2015-01-01

Transition-metal carbides have been employed for biobased conversions aiming to replace the rare noble metals. However, when reactions are in liquid phase, many authors have observed catalyst deactivation. The main routes of deactivation in liquid phase biobased conversions are coke deposition,

A Study On The Metal Carbide Composite Diffusion Bonding For Mechanical Seal

Directory of Open Access Journals (Sweden)

Kim D.-K.

2015-06-01

Full Text Available Mechanical Seal use highly efficient alternative water having a great quantity of an aqueous solution and has an advantage no corrosion brine. Metal Carbide composites have been investigated as potential materials for high temperature structural applications and for application in the processing industry. The existing Mechanical seal material is a highly expensive carbide alloy, and it is difficult to take a price advantage. Therefore the study of replacing body area with inexpensive steel material excluding O-ring and contact area which demands high characteristics is needed.

Radionuclide Inventories for DOE SNF Waste Stream and Uranium/Thorium Carbide Fuels

International Nuclear Information System (INIS)

K.L. Goluoglu

2000-01-01

The objective of this calculation is to generate radionuclide inventories for the Department of Energy (DOE) spent nuclear fuel (SNF) waste stream destined for disposal at the potential repository at Yucca Mountain. The scope of this calculation is limited to the calculation of two radionuclide inventories; one for all uranium/thorium carbide fuels in the waste stream and one for the entire waste stream. These inventories will provide input in future screening calculations to be performed by Performance Assessment to determine important radionuclides

Nuclear reactor fuel element

International Nuclear Information System (INIS)

D'Eye, R.W.M.; Shennan, J.V.; Ford, L.H.

1977-01-01

Fuel element with particles from ceramic fissionable material (e.g. uranium carbide), each one being coated with pyrolitically deposited carbon and all of them being connected at their points of contact by means of an individual crossbar. The crossbar consists of silicon carbide produced by reaction of silicon metal powder with the carbon under the influence of heat. Previously the silicon metal powder together with the particles was kneaded in a solvent and a binder (e.g. epoxy resin in methyl ethyl ketone plus setting agent) to from a pulp. The reaction temperature lies at 1750 0 C. The reaction itself may take place in a nitrogen atmosphere. There will be produced a fuel element with a high overall thermal conductivity. (DG) [de

Gamma scanning of mixed carbide and oxide fuel pins irradiated in FBTR

International Nuclear Information System (INIS)

Jayaraj, V.V.; Padalakshmi, M.; Ulaganathan, T.; Venkiteswaran, C.N.; Divakar, R.; Joseph, Jojo; Bhaduri, A.K.

2016-01-01

Fission in nuclear fuels results in a number of fission products that are gamma emitters in the energy range of 100 keV to 3 MeV. The gamma emitting fission products are therefore amenable for detection by gamma detectors. Assessment of the fission product distribution and their migration behavior through gamma scanning is important for characterizing the in reactor behavior of the fuel. Gamma scanning is an important non destructive technique used to evaluate the behavior of irradiated fuels. As a part of Post Irradiation Examinations (PIE), axial gamma scanning has been carried out on selected fuel pins of the FBTR Mark I mixed carbide fuel sub-assemblies and PFBR MOX test fuel sub-assembly irradiated in FBTR. This paper covers the results of gamma scanning and correlation of gamma scanning results with other PIE techniques

Determining the minimum required uranium carbide content for HTGR UCO fuel kernels

International Nuclear Information System (INIS)

McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.; Reif, Tyler J.; Morris, Robert N.; Hunn, John D.

2017-01-01

Highlights: • The minimum required uranium carbide content for HTGR UCO fuel kernels is calculated. • More nuclear and chemical factors have been included for more useful predictions. • The effect of transmutation products, like Pu and Np, on the oxygen distribution is included for the first time. - Abstract: Three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from O release when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. In the HTGR kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium apart from UO 2 in the form of a carbide, UC x and this fuel form is designated UCO. Here general oxygen balance formulas were developed for calculating the minimum UC x content to ensure negligible CO formation for 15.5% enriched UCO taken to 16.1% actinide burnup. Required input data were obtained from CALPHAD (CALculation of PHAse Diagrams) chemical thermodynamic models and the Serpent 2 reactor physics and depletion analysis tool. The results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmuted Pu and Np oxides on the oxygen distribution as the fuel kernel composition evolves with burnup.

Hydrogen chemisorption and oxidation of transition metal carbides

International Nuclear Information System (INIS)

Bethin, J.R.

1979-01-01

A study was made of the catalytic activity of WC, focusing on the possible influence of point defects. The chemisorption of H on WC and titanium oxycarbides was studied with differential scanning calorimetry. The catalytic activity of these materials for oxidation of H was determined by potentiostatic steady-state and potentiodynamic measurements in acid electrolyte. Compositions of WC surfaces were determined by x-ray photoemission and related to the catalytic behavior. Titanium oxycarbide surfaces were analyzed by Auger electron spectroscopy. Of the carbides tested only one WC preparation was able to chemisorb H. Both WC powders investigated catalyzed H oxidation with similar specific activities. Spectroscopic studies showed that the active surface of WC was a mixture of WO 3 and a carbon-deficient WC phase. This result indicates that carbon vacancies are the active sites in tungsten carbide. Theoretical models of a carbon vacancy surrounded by metal atoms suggested by calculations by other workers support this assignment and identify the important role of the W6s level. The measured value of the heat of chemisorption is consistent with the proposed model

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

2011-08-15

Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

Metal-boride phase formation on tungsten carbide (WC-Co) during microwave plasma chemical vapor deposition

Energy Technology Data Exchange (ETDEWEB)

Johnston, Jamin M.; Catledge, Shane A., E-mail: catledge@uab.edu

2016-02-28

Graphical abstract: - Highlights: • A detailed phase analysis after PECVD boriding shows WCoB, CoB and/or W{sub 2}CoB{sub 2}. • EDS of PECVD borides shows boron diffusion into the carbide grain structure. • Nanoindentation hardness and modulus of borides is 23–27 GPa and 600–780 GPa. • Scratch testing shows hard coating with cracking at 40N and spallation at 70N. - Abstract: Strengthening of cemented tungsten carbide by boriding is used to improve the wear resistance and lifetime of carbide tools; however, many conventional boriding techniques render the bulk carbide too brittle for extreme conditions, such as hard rock drilling. This research explored the variation in metal-boride phase formation during the microwave plasma enhanced chemical vapor deposition process at surface temperatures from 700 to 1100 °C. We showed several well-adhered metal-boride surface layers consisting of WCoB, CoB and/or W{sub 2}CoB{sub 2} with average hardness from 23 to 27 GPa and average elastic modulus of 600–730 GPa. The metal-boride interlayer was shown to be an effective diffusion barrier against elemental cobalt; migration of elemental cobalt to the surface of the interlayer was significantly reduced. A combination of glancing angle X-ray diffraction, electron dispersive spectroscopy, nanoindentation and scratch testing was used to evaluate the surface composition and material properties. An evaluation of the material properties shows that plasma enhanced chemical vapor deposited borides formed at substrate temperatures of 800 °C, 850 °C, 900 °C and 1000 °C strengthen the material by increasing the hardness and elastic modulus of cemented tungsten carbide. Additionally, these boride surface layers may offer potential for adhesion of ultra-hard carbon coatings.

Metal-boride phase formation on tungsten carbide (WC-Co) during microwave plasma chemical vapor deposition

International Nuclear Information System (INIS)

Johnston, Jamin M.; Catledge, Shane A.

2016-01-01

Graphical abstract: - Highlights: • A detailed phase analysis after PECVD boriding shows WCoB, CoB and/or W_2CoB_2. • EDS of PECVD borides shows boron diffusion into the carbide grain structure. • Nanoindentation hardness and modulus of borides is 23–27 GPa and 600–780 GPa. • Scratch testing shows hard coating with cracking at 40N and spallation at 70N. - Abstract: Strengthening of cemented tungsten carbide by boriding is used to improve the wear resistance and lifetime of carbide tools; however, many conventional boriding techniques render the bulk carbide too brittle for extreme conditions, such as hard rock drilling. This research explored the variation in metal-boride phase formation during the microwave plasma enhanced chemical vapor deposition process at surface temperatures from 700 to 1100 °C. We showed several well-adhered metal-boride surface layers consisting of WCoB, CoB and/or W_2CoB_2 with average hardness from 23 to 27 GPa and average elastic modulus of 600–730 GPa. The metal-boride interlayer was shown to be an effective diffusion barrier against elemental cobalt; migration of elemental cobalt to the surface of the interlayer was significantly reduced. A combination of glancing angle X-ray diffraction, electron dispersive spectroscopy, nanoindentation and scratch testing was used to evaluate the surface composition and material properties. An evaluation of the material properties shows that plasma enhanced chemical vapor deposited borides formed at substrate temperatures of 800 °C, 850 °C, 900 °C and 1000 °C strengthen the material by increasing the hardness and elastic modulus of cemented tungsten carbide. Additionally, these boride surface layers may offer potential for adhesion of ultra-hard carbon coatings.

Lattice vibrational properties of transition metal carbides (TiC, ZrC

Indian Academy of Sciences (India)

Lattice vibrational properties of transition metal carbides (TiC, ZrC and HfC) have been presented by including the effects of free-carrier doping and three-body interactions in the rigid shell model. The short-range overlap repulsion is operative up to the second neighbour ions. An excellent agreement has been obtained ...

Performance of metallic fuels in liquid-metal fast reactors

International Nuclear Information System (INIS)

Seidel, B.R.; Walters, L.C.; Kittel, J.H.

1984-01-01

Interest in metallic fuels for liquid-metal fast reactors has come full circle. Metallic fuels are once again a viable alternative for fast reactors because reactor outlet temperature of interest to industry are well within the range where metallic fuels have demonstrated high burnup and reliable performance. In addition, metallic fuel is very tolerant of off-normal events of its high thermal conductivity and fuel behavior. Futhermore, metallic fuels lend themselves to compact and simplified reprocessing and refabrication technologies, a key feature in a new concept for deployment of fast reactors called the Integral Fast Reactor (IFR). The IFR concept is a metallic-fueled pool reactor(s) coupled to an integral-remote reprocessing and fabrication facility. The purpose of this paper is to review recent metallic fuel performance, much of which was tested and proven during the twenty years of EBR-II operation

Utility industry evaluation of the metal fuel facility and metal fuel performance for liquid metal reactors

International Nuclear Information System (INIS)

Burstein, S.; Gibbons, J.P.; High, M.D.; O'Boyle, D.R.; Pickens, T.A.; Pilmer, D.F.; Tomonto, J.R.; Weinberg, C.J.

1990-02-01

A team of utility industry representatives evaluated the liquid metal reactor metal fuel process and facility conceptual design being developed by Argonne National Laboratory (ANL) under Department of Energy sponsorship. The utility team concluded that a highly competent ANL team was making impressive progress in developing high performance advanced metal fuel and an economic processing and fabrication technology. The utility team concluded that the potential benefits of advanced metal fuel justified the development program, but that, at this early stage, there are considerable uncertainties in predicting the net overall economic benefit of metal fuel. Specific comments and recommendations are provided as a contribution towards enhancing the development program. 6 refs

A review of the breeding potentials of carbide, nitride and oxide fueled LMFBRs and GCFRs

International Nuclear Information System (INIS)

Handa, Muneo

1977-11-01

The effects of design parameters in large variation on compound system doubling time of large advanced-fueled LMFBR are described on the base of recent U.S. results. The fuel element design by Combustion Engineering Inc. in step-by-step substitution of the initial oxide fuel subassemblies with carbide ones is explained. Breeding characteristics of the oxide-fueled LMFBR and its potential design modifications are expounded. The gas cooled fast breeder program in West Germany and in the United States are briefed. Definitions of the breeding ratio and doubling time in overall fuel cycle are given. (auth.)

Transition metal carbide and boride abrasive particles

International Nuclear Information System (INIS)

Valdsaar, H.

1978-01-01

Abrasive particles and their preparation are discussed. The particles consist essentially of a matrix of titanium carbide and zirconium carbide, at least partially in solid solution form, and grains of crystalline titanium diboride dispersed throughout the carbide matrix. These abrasive particles are particularly useful as components of grinding wheels for abrading steel. 1 figure, 6 tables

Performance of a sphere-pac mixed carbide fuel pin irradiated in the Dounreay Fast Reactor (DFR 527/1 experiment)

International Nuclear Information System (INIS)

Bischoff, K.; Smith, L.; Stratton, R.W.

1980-10-01

The DFR 527/1 experiment was the first irradiation of EIR sphere-pac uranium-plutonium mixed carbide fuel in a fast flux. The experiment has been successfully irradiated to a burn-up of 7.3% FIMA at ratings between 45 and 62 kW m - 1 and clad temperatures between 300 and 600 0 C. Restructuring and elemental redistribution has been found to be similar to the pattern established for pellet type fuel and follows effects seen in earlier sphere-pac carbide tests. Gas release of 12-14% has been measured. A preliminary comparison of radial temperature distribution calculations using a first version of the fuel behaviour modelling code SPECKLE with the actual metallography has been attempted. (Auth.)

Performance of HVOF carbide coatings under erosion/corrosion

International Nuclear Information System (INIS)

Simard, S.; Arsenault, B.; Legoux, J.G.; Hawthorne, H.M.

1999-01-01

Cermet based materials are known to have an excellent performance under several wear conditions. High velocity oxy-fuel (HVOF) technology allows the deposition of such hard materials in the form of protective coatings onto different surfaces. Under slurry erosion, the performance of the coatings is influenced by the occurrence of corrosion reactions on the metallic matrix. Indeed, wet conditions promote the dissolution of metallic binder resulting in a potential synergic effect between the corrosion and wear mechanisms. The composition of the metallic matrix plays a key role on the stability of the coatings and their degradation rate. In this work, four coatings based on tungsten carbide embedded in different metallic binders were evaluated with regard to corrosion and wear. (author)

Metallic fuel development

International Nuclear Information System (INIS)

Walters, L.C.

1987-01-01

Metallic fuels are capable of achieving high burnup as a result of design modifications instituted in the late 1960's. The gap between the fuel slug and the cladding is fixed such that by the time the fuel swells to the cladding the fission gas bubbles interconnect and release the fission gas to an appropriately sized plenum volume. Interconnected porosity thus provides room for the fuel to deform from further swelling rather than stress the cladding. In addition, the interconnected porosity allows the fuel pin to be tolerant to transient events because as stresses are generated during a transient event the fuel flows rather than applying significant stress to the cladding. Until 1969 a number of metallic fuel alloys were under development in the US. At that time the metallic fuel development program in the US was discontinued in favor of ceramic fuels. However, development had proceeded to the point where it was clear that the zirconium addition to uranium-plutonium fuel would yield a ternary fuel with an adequately high solidus temperature and good compatibility with austenitic stainless steel cladding. Furthermore, several U-Pu-Zr fuel pins had achieved about 6 at.% bu by the late 1960's, without failure, and thus the prospect for high burnup was promising

One-step synthesis of 2D-layered carbon wrapped transition metal nitrides from transition metal carbides (MXenes) for supercapacitors with ultrahigh cycling stability.

Science.gov (United States)

Yuan, Wenyu; Cheng, Laifei; Wu, Heng; Zhang, Yani; Lv, Shilin; Guo, Xiaohui

2018-03-13

A novel one-step method to synthesize 2D carbon wrapped TiN (C@TiN) was proposed via using 2D metal carbides (MXenes) as precursors. This study provides a novel approach to synthesize carbon wrapped metal nitrides.

Carbide-forming groups IVB-VIB metals: a new territory in the periodic table for CVD growth of graphene.

Science.gov (United States)

Zou, Zhiyu; Fu, Lei; Song, Xiuju; Zhang, Yanfeng; Liu, Zhongfan

2014-07-09

Early transition metals, especially groups IVB-VIB metals, can form stable carbides, which are known to exhibit excellent "noble-metal-like" catalytic activities. We demonstrate herein the applications of groups IVB-VIB metals in graphene growth using atmospheric pressure chemical vapor deposition technique. Similar to the extensively studied Cu, Ni, and noble metals, these transition-metal foils facilitate the catalytic growth of single- to few-layer graphene. The most attractive advantage over the existing catalysts is their perfect control of layer thickness and uniformity with highly flexible experimental conditions by in situ converting the dissolved carbons into stable carbides to fully suppress the upward segregation/precipitation effect. The growth performance of graphene on these transition metals can be well explained by the periodic physicochemical properties of elements. Our work has disclosed a new territory of catalysts in the periodic table for graphene growth and is expected to trigger more interest in graphene research.

New catalysts for coal processing: Metal carbides and nitrides

Energy Technology Data Exchange (ETDEWEB)

S. Ted Oyama; David F. Cox

1999-12-03

The subject of this research project was to investigate the catalytic properties of a new class of materials, transition metal carbides and nitrides, for treatment of coal liquid and petroleum feedstocks. The main objectives were: (1) preparation of catalysts in unsupported and supported form; (2) characterization of the materials; (3) evaluation of their catalytic properties in HDS and HDN; (4) measurement of the surface properties; and (5) observation of adsorbed species. All of the objectives were substantially carried out and the results will be described in detail below. The catalysts were transition metal carbides and nitrides spanning Groups 4--6 in the Periodic Table. They were chosen for study because initial work had shown they were promising materials for hydrotreating. The basic strategy was first to prepare the materials in unsupported form to identify the most promising catalyst, and then to synthesize a supported form of the material. Already work had been carried out on the synthesis of the Group VI compounds Mo{sub 2}C, Mo{sub 2}N, and WC, and new methods were developed for the Group V compounds VC and NbC. All the catalysts were then evaluated in a hydrotreating test at realistic conditions. It was found that the most active catalyst was Mo{sub 2}C, and further investigations of the material were carried out in supported form. A new technique was employed for the study of the bulk and surface properties of the catalysts, near edge x-ray absorption spectroscopy (NEXAFS), that fingerprinted the electronic structure of the materials. Finally, two new research direction were explored. Bimetallic alloys formed between two transition metals were prepared, resulting in catalysts having even higher activity than Mo{sub 2}C. The performance of the catalysts in hydrodechloration was also investigated.

Tungsten carbide encapsulated in nitrogen-doped carbon with iron/cobalt carbides electrocatalyst for oxygen reduction reaction

Energy Technology Data Exchange (ETDEWEB)

Zhang, Jie; Chen, Jinwei, E-mail: jwchen@scu.edu.cn; Jiang, Yiwu; Zhou, Feilong; Wang, Gang; Wang, Ruilin, E-mail: rl.wang@scu.edu.cn

2016-12-15

Graphical abstract: A hybrid catalyst was prepared via a quite green and simple method to achieve an one-pot synthesis of the N-doping carbon, tungsten carbides, and iron/cobalt carbides. It exhibited comparable electrocatalytic activity, higher durability and ability to methanol tolerance compared with commercial Pt/C to ORR. - Highlights: • A novel type of hybrid Fe/Co/WC@NC catalysts have been successfully synthesized. • The hybrid catalyst also exhibited better durability and methanol tolerance. • Multiple effective active sites of Fe{sub 3}C, Co{sub 3}C, WC, and NC help to improve catalytic performance. - Abstract: This work presents a type of hybrid catalyst prepared through an environmental and simple method, combining a pyrolysis of transition metal precursors, a nitrogen-containing material, and a tungsten source to achieve a one-pot synthesis of N-doping carbon, tungsten carbides, and iron/cobalt carbides (Fe/Co/WC@NC). The obtained Fe/Co/WC@NC consists of uniform Fe{sub 3}C and Co{sub 3}C nanoparticles encapsulated in graphitized carbon with surface nitrogen doping, closely wrapped around a plate-like tungsten carbide (WC) that functions as an efficient oxygen reduction reaction (ORR) catalyst. The introduction of WC is found to promote the ORR activity of Fe/Co-based carbide electrocatalysts, which is attributed to the synergistic catalysts of WC, Fe{sub 3}C, and Co{sub 3}C. Results suggest that the composite exhibits comparable electrocatalytic activity, higher durability, and ability for methanol tolerance compared with commercial Pt/C for ORR in alkaline electrolyte. These advantages make Fe/Co/WC@NC a promising ORR electrocatalyst and a cost-effective alternative to Pt/C for practical application as fuel cell.

Status of LMR fuel development in the United States of America

International Nuclear Information System (INIS)

Leggett, R.D.; Walters, L.C.

1993-01-01

Three fuel systems oxide, metal, and carbide are shown to be reliable to high burnup and a fourth system, nitride, is shown to have promise for LMR applications. The excellent steady state performance of the oxide and metal driver fuels for FFTF and EBR-II, respectively, supported by the experience base on tens of thousands of test pins is provided. Achieving 300 MWd/kg in the oxide fuel system through the use of low swelling cladding and duct materials and the Integral Fast Reactor (IFR) concept that utilizes metallic fuel are described. Arguments for economic viability are presented. Responses to operational transients and severe over-power events are shown to have large safety margins and run-beyond-cladding-breach (RBCB), is shown to be non-threatening to LMR reactor system. Results from a joint U.S.-Swiss carbide test that operated successfully at high power and burnup in FFTF are also presented. (orig.)

Status of LMR fuel development in the United States of America

International Nuclear Information System (INIS)

Leggett, R.D.; Walters, L.C.

1992-01-01

Three fuel systems - oxide, metal and carbide - are shown to be reliable to high burnup and a fourth system, nitride, is shown to have promise for LMR applications. The excellent steady state performance of the oxide and metal driver fuels for FFTF and EBR-II, respectively, as well as that of tens of thousands of test pins is provided. Achieving 300 MWd/kg in the oxide fuel system through the use of low swelling cladding and duct materials is described and arguments for economic viability are presented. Responses to operational transients and severe overpower events are shown to have large safety margins and run beyond cladding breach, RBCB, likewise, is shown to be nonthreatening to LMR reactor systems. The Integral Fast Reactor (IFR) concept that utilizes metallic fuel and the commercial viability of this concept are discussed. Results from a joint US-Swiss carbide test that operated successfully at high power and burnup in FFTF are also presented

Production of metal and metal-ceramic coatings on D-Gun Ob

International Nuclear Information System (INIS)

Gavrilenko, T.P.; Nikolaev, Y.A.; Ulianitsky, V.Y.

1995-01-01

Optimization of the detonation spraying process has been made for the production of metal and metal-ceramics coatings with the D-Gun Ob. Owing to the ability of Ob to work with several fuels and an inert diluent simultaneously, variation of detonation regimes in a wide range is possible, and because of localized powder injection in the D-Gun barrel, high uniformity of parameters of powder particles is achieved. The best conditions for particle heating and acceleration were calculated with the help of mathematical simulation, and the corresponding regimes were realized on D-Gun Ob. High-quality aluminum, copper, nickel, and nickel-chromium-silicon-carbon-boron alloy coatings were produced by using only propane fuel. Chromium carbide with nickel and tungsten carbide with cobalt coatings were produced with addition of acetylene. Optimal efficiency and high bonding strength were achieved for all powders. Data on microhardness, bonding strength, and efficiency are presented

Development, Fabrication and Characterization of Fuels for Indian Fast Reactor Programme

International Nuclear Information System (INIS)

Kumar, Arun

2013-01-01

Development of Fast Reactor fuels in India started in early Seventies. The successful development of Mixed Carbide fuels for FBTR and MOX fuel for PFBR have given confidence in manufacture of fuels for Fast Reactors. Effort is being put to develop high Breeding Ratio Metallic fuel (binary/ternary). Few fuel pins have been fabricated and is under test irradiation. However, this is only a beginning and complete fuel cycle activities are under development. Metal fuelled Fast Reactors will provide high growth rate in Indian Fast Reactor programme

Metal fuel safety performance

International Nuclear Information System (INIS)

Miles, K.J. Jr.; Tentner, A.M.

1988-01-01

The current development of breeder reactor systems has lead to the renewed interest in metal fuels as the driver material. Modeling efforts were begun to provide a mechanistic description of the metal fuel during anticipated and hypothetical transients within the context of the SAS4A accident analysis code system. Through validation exercises using experimental results of metal fuel TREAT tests, confidence is being developed on the nature and accuracy of the modeling and implementation. Prefailure characterization, transient pin response, margins to failure, axial in-pin fuel relocation prior to cladding breach, and molten fuel relocation after cladding breach are considered. Transient time scales ranging from milliseconds to many hours can be studied with all the reactivity feedbacks evaluated

Corrosion behaviour of 2124 aluminium alloy-silicon carbide metal matrix composites in sodium chloride environment

International Nuclear Information System (INIS)

Singh, Nirbhay; Vadera, K.K.; Ramesh Kumar, A.V.; Singh, R.S.; Monga, S.S.; Mathur, G.N.

1999-01-01

Aluminium alloy based particle reinforced metal matrix composites (MMCs) are being considered for a range of applications. Their mechanical properties have been investigated in detail, but more information about their corrosion resistance is needed. In this investigation, the corrosion behaviour of silicon carbide particulates (SiC p )-2124 aluminium metal matrix composites was studied in 3 wt% sodium chloride solution by means of electrochemical technique and optical microscope. The effects of weight percentages and particle size of silicon carbide particulates on corrosion behaviour of the composite were studied in NaCl and it was observed that corrosion rate increases linearly with the increasing weight percentage of SiC p . The corrosion rate of the MMC increases by increasing the size of SiC particles. Anodization improved corrosion resistance of the composites. (author)

Fuel fabrication processes, design and experimental conditions for the joint US-Swiss mixed carbide test in FFTF (AC-3 test)

International Nuclear Information System (INIS)

Stratton, R.W.; Ledergerber, G.; Ingold, F.; Latimer, T.W.; Chidester, K.M.

1993-01-01

The preparation of mixed carbide fuel for a joint (US-Swiss) irradiation test in the US Fast Flux Test Facility (FFTF) is described, together with the experiment design and the irradiation conditions. Two fabrication routes were compared. The US produced 66 fuel pins containing pellet fuel via the powder-pellet (dry) route, and the Swiss group produced 25 sphere pac pins of mixed carbide using the internal gelation (wet) route. Both sets of fuel met all t the requirements of the specifications concerning soichiometry, chemical composition and structure. The pin designs were as similar as possible. The test operated successfully in the FFTF for 620 effective full power days until October 1988 and reached over 8% burn up with peak powers of around 80 kW/m. The conclusions were that the choice of sphere pac or pellet fuel for reactor application is dependent on preferred differences in fabrication (e.g. economics and environmental factors) and not on differences in irradiation behaviour. (orig.)

Atomic structure of non-stoichiometric transition metal carbides

International Nuclear Information System (INIS)

Moisy-Maurice, Virginie.

1981-10-01

Different kinds of experimental studies of the atomic arrangement in non-stoichiometric transition metal carbides are proposed: the ordering of carbon vacancies and the atomic static displacements are the main subjects studied. Powder neutron diffraction on TiCsub(1-x) allowed us to determine the order-disorder transition critical temperature -Tsub(c) approximately 770 0 C- in the TiCsub(0.52-0.67) range, and to analyze at 300 K the crystal structure of long-range ordered samples. A neutron diffuse scattering quantitative study at 300 K of short-range order in TiCsub(0.76), TiCsub(0.79) and NbCsub(0.73) single crystals is presented: as in Ti 2 Csub(1+x) and Nb 6 C 5 superstructures, vacancies avoid to be on each side of a metal atom. Besides, the mean-square carbon atom displacements from their sites are small, whereas metal atoms move radially about 0.03 A away from vacancies. These results are in qualitative agreement with EXAFS measurements at titanium-K edge of TiCsub(1-x). An interpretation of ordering in term of short-range interaction pair potentials between vacancies is proposed [fr

Metallic Reactor Fuel Fabrication for SFR

Energy Technology Data Exchange (ETDEWEB)

Song, Hoon; Kim, Jong-Hwan; Ko, Young-Mo; Woo, Yoon-Myung; Kim, Ki-Hwan; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

The metal fuel for an SFR has such advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant, and inherent passive safety 1. U-Zr metal fuel for SFR is now being developed by KAERI as a national R and D program of Korea. The fabrication technology of metal fuel for SFR has been under development in Korea as a national nuclear R and D program since 2007. The fabrication process for SFR fuel is composed of (1) fuel slug casting, (2) loading and fabrication of the fuel rods, and (3) fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycled streams in this fabrication process. Fabrication on the rod type metallic fuel was carried out for the purpose of establishing a practical fabrication method. Rod-type fuel slugs were fabricated by injection casting. Metallic fuel slugs fabricated showed a general appearance was smooth.

Silver diffusion through silicon carbide in microencapsulated nuclear fuels TRISO; Difusion de plata a traves de carburo de silicio en combustibles nucleares microencapsulados TRISO

Energy Technology Data Exchange (ETDEWEB)

Cancino T, F.; Lopez H, E., E-mail: Felix.cancino@cinvestav.edu.mx [IPN, Centro de Investigacion y de Estudios Avanzados, Unidad Saltillo, Av. Industria Metalurgica No. 1062, Col. Ramos Arizpe, 25900 Saltillo, Coahuila (Mexico)

2013-10-15

The silver diffusion through silicon carbide is a challenge that has persisted in the development of microencapsulated fuels TRISO (Tri structural Isotropic) for more than four decades. The silver is known as a strong emitter of gamma radiation, for what is able to diffuse through the ceramic coatings of pyrolytic coal and silicon carbide and to be deposited in the heat exchangers. In this work we carry out a recount about the art state in the topic of the diffusion of Ag through silicon carbide in microencapsulated fuels and we propose the role that the complexities in the grain limit can have this problem. (Author)

Binary and ternary carbides and nitrides of the transition metals and their phase relations

International Nuclear Information System (INIS)

Holleck, H.

1981-01-01

The occurrance and the structure of the binary and ternary transition metal carbides and nitrides are described. Phase diagrams are assessed for most of the binary and ternary systems. Many ternary phase diagrams are published in this report for the first time. (orig.) [de

Device for fracturing silicon-carbide coatings on nuclear-fuel particles

Science.gov (United States)

Turner, L.J.; Willey, M.G.; Tiegs, S.M.; Van Cleve, J.E. Jr.

This invention is a device for fracturing particles. It is designed especially for use in hot cells designed for the handling of radioactive materials. In a typical application, the device is used to fracture a hard silicon-carbide coating present on carbon-matrix microspheres containing nuclear-fuel materials, such as uranium or thorium compounds. To promote remote control and facilitate maintenance, the particle breaker is pneumatically operated and contains no moving parts. It includes means for serially entraining the entrained particles on an anvil housed in a leak-tight chamber. The flow rate of the gas is at a value effecting fracture of the particles; preferably, it is at a value fracturing them into product particulates of fluidizable size. The chamber is provided with an outlet passage whose cross-sectional area decreases in the direction away from the chamber. The outlet is connected tangentially to a vertically oriented vortex-flow separator for recovering the product particulates entrained in the gas outflow from the chamber. The invention can be used on a batch or continuous basis to fracture the silicon-carbide coatings on virtually all of the particles fed thereto.

Method for fracturing silicon-carbide coatings on nuclear-fuel particles

Science.gov (United States)

Turner, Lloyd J.; Willey, Melvin G.; Tiegs, Sue M.; Van Cleve, Jr., John E.

1982-01-01

This invention is a device for fracturing particles. It is designed especially for use in "hot cells" designed for the handling of radioactive materials. In a typical application, the device is used to fracture a hard silicon-carbide coating present on carbon-matrix microspheres containing nuclear-fuel material, such as uranium or thorium compounds. To promote remote control and facilitate maintenance, the particle breaker is pneumatically operated and contains no moving parts. It includes means for serially entraining the entrained particles on an anvil housed in a leak-tight chamber. The flow rate of the gas is at a value effecting fracture of the particles; preferably, it is at a value fracturing them into product particulates of fluidizable size. The chamber is provided with an outlet passage whose cross-sectional area decreases in the direction away from the chamber. The outlet is connected tangentially to a vertically oriented vortex-flow separator for recovering the product particulates entrained in the gas outflow from the chamber. The invention can be used on a batch or continuous basis to fracture the silicon-carbide coatings on virtually all of the particles fed thereto.

Factors controlling metal fuel lifetime

International Nuclear Information System (INIS)

Porter, D.L.; Hofman, G.L.; Seidel, B.R.; Walters, L.C.

1986-01-01

The reliability of metal fuel elements is determined by a fuel burnup at which a statistically predicted number of fuel breaches would occur, the number of breaches determined by the amount of free fission gas which a particular reactor design can tolerate. The reliability is therefore measured using experimentally determined breach statistics, or by modelling fuel element behavior and those factors which contribute to cladding breach. The factors are fuel/cladding mechanical and chemical interactions, fission gas pressure, fuel phase transformations involving volume changes, and fission product effects on cladding integrity. Experimental data for EBR-II fuel elements has shown that the primary, and perhaps the only significant factor affecting metal fuel reliability, is the pressure-induced stresses caused by fission gas release. Other metal fuel/cladding systems may perform similarly

Advanced Silicon Carbide from Molecular Engineering and Actinide Fuels

International Nuclear Information System (INIS)

Meyer, D.J.M.; Garcia, J.; Guillaneux, D.; Wong-Chi-Man, M.; Moreau, J.J.E.

2008-01-01

In the frame of nuclear fuels studies for generation IV, carbides or oxycarbides assemblies are one of the engaged material for high temperature reactors. The design of the fuels is not yet defined but some structures are actually considered with SiC as matrix for the actinide fuel. In this work we have studied the synthesis of a multi-scale structure controlled SiC matrix using molecular silicon organometallic precursors. The aim of this work was to develop a way to obtain multi-scale SiC matrix material which could be engineered to fit in any fuel structure defined for generation IV fuels. The control of this multi-scale structure was done using several simulation methods specific of the low temperature solution synthesis of the precursor. In a first step, we have focused our effort on the synthesis of the SiC material. A first level of template was successfully done by the use of solid silica 500 nm balls. A second level of template was studied by the use of meso-porous silica, structured at a 50 nm level. At least, supra-molecular simulation in non aqueous media was considered with the difficulty to build a molecular assembly (inverse micelles). In a second step, we have functionalized the primary silane phase with actinide complexing agent in order to blend directly the actinide inside this primary phase in a controlled way. During these studies, a new one pot synthesis route to obtain the functionalized primary silane phase was developed. (authors)

History of fast reactor fuel development

Energy Technology Data Exchange (ETDEWEB)

Kittel, J.H. (Argonne National Lab., IL (United States)); Frost, B.R.T. (Argonne National Lab., IL (United States)); Mustelier, J.P. (COGEMA, Velizy-Villacoublay (France)); Bagley, K.Q. (AEA Reactor Services, Risley (United Kingdom)); Crittenden, G.C. (AEA Reactor Services, Dounreay (United Kingdom)); Dievoet, J. van (Belgonucleaire, Brussels (Belgium))

1993-09-01

The first fast breeder eactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s. (orig.)

Fast breeder fuel cycle

International Nuclear Information System (INIS)

1978-09-01

Basic elements of the ex-reactor part of the fuel cycle (reprocessing, fabrication, waste handling and transportation) are described. Possible technical and proliferation measures are evaluated, including current methods of accountability, surveillance and protection. The reference oxide based cycle and advanced cycles based on carbide and metallic fuels are considered utilizing conventional processes; advanced nonaqueous reprocessing is also considered. This contribution provides a comprehensive data base for evaluation of proliferation risks

Effect of Carbide Dissolution on Chlorine Induced High Temperature Corrosion of HVOF and HVAF Sprayed Cr3C2-NiCrMoNb Coatings

Science.gov (United States)

Fantozzi, D.; Matikainen, V.; Uusitalo, M.; Koivuluoto, H.; Vuoristo, P.

2018-01-01

Highly corrosion- and wear-resistant thermally sprayed chromium carbide (Cr3C2)-based cermet coatings are nowadays a potential highly durable solution to allow traditional fluidized bed combustors (FBC) to be operated with ecological waste and biomass fuels. However, the heat input of thermal spray causes carbide dissolution in the metal binder. This results in the formation of carbon saturated metastable phases, which can affect the behavior of the materials during exposure. This study analyses the effect of carbide dissolution in the metal matrix of Cr3C2-50NiCrMoNb coatings and its effect on chlorine-induced high-temperature corrosion. Four coatings were thermally sprayed with HVAF and HVOF techniques in order to obtain microstructures with increasing amount of carbide dissolution in the metal matrix. The coatings were heat-treated in an inert argon atmosphere to induce secondary carbide precipitation. As-sprayed and heat-treated self-standing coatings were covered with KCl, and their corrosion resistance was investigated with thermogravimetric analysis (TGA) and ordinary high-temperature corrosion test at 550 °C for 4 and 72 h, respectively. High carbon dissolution in the metal matrix appeared to be detrimental against chlorine-induced high-temperature corrosion. The microstructural changes induced by the heat treatment hindered the corrosion onset in the coatings.

Irradiation performance of metallic fuels

International Nuclear Information System (INIS)

Pahl, R.G.; Lahm, C.E.; Porter, D.L.; Batte, G.L.; Hofman, G.L.

1989-01-01

Argonne National Laboratory has been working for the past five years to develop and demonstrate the Integral Fast Reactor (IFR) concept. The concept involves a closed system for fast-reactor power generation and on-site fuel reprocessing, both designed specifically around the use of metallic fuel. The Experimental Breeder Reactor-II (EBR-II) has used metallic fuel for all of its 25-year life. In 1985, tests were begun to examine the irradiation performance of advanced-design metallic fuel systems based on U-Zr or U-Pu-Zr fuels. These tests have demonstrated the viable performance of these fuel systems to high burnup. The initial testing program will be described in this paper. 2 figs

Synthesis of IV-VI Transition Metal Carbide and Nitride Nanoparticles Using a Reactive Mesoporous Template for Electrochemical Hydrogen Evolution Reaction

KAUST Repository

Alhajri, Nawal Saad

2016-01-01

Interstitial carbides and nitrides of early transition metals in Groups IV-VI exhibit platinum-like behavior which makes them a promising candidate to replace noble metals in a wide variety of reactions. Most synthetic methods used to prepare these materials lead to bulk or micron size powder which limits their use in reactions in particular in catalytic applications. Attempts toward the production of transition metal carbide and nitride nanoparticles in a sustainable, simple and cheap manner have been rapidly increasing. In this thesis, a new approach was presented to prepare nano-scale transition metal carbides and nitrides of group IV-VI with a size as small as 3 nm through the reaction of transition metal precursor with mesoporous graphitic carbon nitride (mpg-C3N4) that not only provides confined spaces for nanoparticles formation but also acts as a chemical source of nitrogen and carbon. The produced nanoparticles were characterized by powder X-ray diffraction (XRD), temperature-programmed reaction with mass spectroscopy (MS), CHN elemental analyses, thermogravimetric analyses (TGA), nitrogen sorption, X-ray photoelectron spectroscopy (XPS), and transmission electron microscopy (TEM). The effects of the reaction temperature, the ratio of the transition metal precursor to the reactive template (mpg-C3N4), and the selection of the carrier gas (Ar, N2, and NH3) on the resultant crystal phases and structures were investigated. The results indicated that different tantalum phases with cubic structure, TaN, Ta2CN, and TaC, can be formed under a flow of nitrogen by changing the reaction temperatures. Two forms of tantalum nitride, namely TaN and Ta3N5, were selectively formed under N2 and NH3 flow, respectively. Significantly, the formation of TaC, Ta2CN, and TaN can be controlled by altering the weight ratio of the C3N4 template relative to the Ta precursor at 1573 K under a flow of nitrogen where high C3N4/Ta precursor ratio generally resulted in high carbide

Irradiation behavior of metallic fast reactor fuels

International Nuclear Information System (INIS)

Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

1991-01-01

Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985

Design and Thermal Analysis for Irradiation of Pyrolytic Carbon/Silicon Carbide Diffusion Couples in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Gerczak, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

Tristructural-isotropic (TRISO)–coated particle fuel is a promising advanced fuel concept consisting of a spherical fuel kernel made of uranium oxide and uranium carbide, surrounded by a porous carbonaceous buffer layer and successive layers of dense inner pyrolytic carbon (IPyC), silicon carbide (SiC) deposited by chemical vapor , and dense outer pyrolytic carbon (OPyC). This fuel concept is being considered for advanced reactor applications such as high temperature gas-cooled reactors (HTGRs) and molten salt reactors (MSRs), as well as for accident-tolerant fuel for light water reactors (LWRs). Development and implementation of TRISO fuel for these reactor concepts support the US Department of Energy (DOE) Office of Nuclear Energy mission to promote safe, reliable nuclear energy that is sustainable and environmentally friendly. During operation, the SiC layer serves as the primary barrier to metallic fission products and actinides not retained in the kernel. It has been observed that certain fission products are released from TRISO fuel during operation, notably, Ag, Eu, and Sr [1]. Release of these radioisotopes causes safety and maintenance concerns.

A decade of advances in metallic fuel

International Nuclear Information System (INIS)

Lahm, C.E.; Pahl, R.G.; Porter, D.L.; Tsai, H.; Seidel, B.R.; Batte, G.L.; Dodds, N.E.; Hofman, G.L.; Walters, L.C.

1991-01-01

Significant advances in the understanding of behavior and performance of metallic fuels to high burnup have been achieved over the past four decades. Metallic fuels were the first fuels for liquid-metal-cooled fast reactors (LMR) but in the late 1960's worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved. Now metallic fuels are recognized as a preferred viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last decade and highlights the behavior and performance features which have demonstrated a much greater potential than previously expected

Studies relating to construction materials to be used in different options for head end treatment in reprocessing of mixed carbide fuel of plutonium and uranium

International Nuclear Information System (INIS)

Rajan, S.K.; Palamalai, A.; Ravi, T.N.; Sampath, M.; Raman, V.R.; Balasubramanian, G.R.

1993-01-01

Mixed carbide of uranium and plutonium has been chosen as the fuel for the first core of Fast Breeder Test Reactor, installed in the Indira Gandhi Centre for Atomic Research. Reprocessing of this fuel is one of the vital steps to prove the viability of the fuel cycle. The head end treatment process introduces constraints in the reprocessing of carbide fuel when compared to the commonly used mixed oxide fuel. Three head end processes, namely direct oxidation, pyrohydrolysis and direct dissolution in nitric acid with oxidation of organic acids were considered for study for exercising the choice. The paper briefly describes the three processes. In each process probable material of construction and related problems are discussed. (author). 3 refs, 5 figs, 7 tabs

PLUTONIUM METALLIC FUELS FOR FAST REACTORS

Energy Technology Data Exchange (ETDEWEB)

STAN, MARIUS [Los Alamos National Laboratory; HECKER, SIEGFRIED S. [Los Alamos National Laboratory

2007-02-07

Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

33 CFR 183.562 - Metallic fuel lines.

Science.gov (United States)

2010-07-01

...) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Manufacturer Requirements § 183.562 Metallic fuel lines. (a) Each metallic fuel line that is mounted to the boat structure must be connected to the engine by a flexible fuel line. (b) Each metallic fuel line must be attached to the boat's structure...

Investigation of iron adsorption on composite transition metal carbides in steel by first-principles calculation

Science.gov (United States)

Xiong, Hui-Hui; Gan, Lei; Tong, Zhi-Fang; Zhang, Heng-Hua; Zhou, Yang

2018-05-01

The nucleation potential of transition metal (TM) carbides formed in steel can be predicted by the behavior of iron adsorption on their surface. Therefore, Fe adsorption on the (001) surface of (A1-xmx)C (A = Nb, Ti, m = Mo, V) was investigated by the first-principles method to reveal the initialization of Fe nucleation. The Mulliken population and partial density of state (PDOS) were also calculated and analyzed in this work. The results show that Fe adsorption depends on the composition and configuration of the composite carbides. The adsorption energy (Wads) of Fe on most of (A1-xmx)C is larger than that of Fe on pure TiC or NbC. The maximum Wads is found for Fe on (Nb0.5Mo0.5)C complex carbide, indicating that this carbide has the high nucleation capacity at early stage. The Fe adsorption could be improved by the segregation of Cr and Mn atoms on the surfaces of (Nb0.5Mo0.5)C and (Ti0.5Mo0.5)C. The PDOS analysis of (Cr, Mn)-doped systems further explains the strong interactions between Fe and Cr or Mn atoms.

A decade of advances in metallic fuel

International Nuclear Information System (INIS)

Seidel, B.R.; Batte, G.L.; Dodds, N.E.; Hofman, G.L.; Lahm, C.E.; Pahl, R.G.; Porter, D.L.; Tsai, H.; Walters, L.C.

1990-01-01

Significant advances in the understanding of behavior and performance of metallic fuels to high burnup have been achieved over the past four decades. Metallic fuels were the first fuels for liquid-metal-cooled fast reactors (LMR) but in the late 1960s worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved. Now metallic fuels are recognized as a preferred viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last decade and highlights the behavior and performance features which have demonstrated a much greater potential than previously expected. 28 refs., 2 figs., 1 tab

A new metal electrocatalysts supported matrix: Palladium nanoparticles supported silicon carbide nanoparticles and its application for alcohol electrooxidation

International Nuclear Information System (INIS)

Dai Hong; Chen Yanling; Lin Yanyu; Xu Guifang; Yang Caiping; Tong Yuejin; Guo Longhua; Chen Guonan

2012-01-01

In this paper, we propose a facile approach for palladium nanoparticles load using silicon carbide nanoparticles as the new supported matrix and a familiar NaBH 4 as reducer. Detailed X-ray photoelectron spectrum (XPS) and transmission electron microscopy (TEM) analysis of the resultant products indicated that palladium nanoparticles are successfully immobilized onto the surface of the silicon carbide nanoparticles with uniform size distribution between 5 and 7 nm. The relative electrochemical characterization clearly demonstrated excellent electrocatalytic activity of this material toward alcohol in alkaline electrolytes. Investigation on the characteristics of the electrocatalytic activity of this material further indicated that the palladium nanoparticles supporting on SiC are very promising for direct alcohol fuel cells (DMFCs), biosensor and electronic devices. Moreover, it was proved that silicon carbide nanoparticles with outstanding properties as support for catalysis are of strong practical interest. And the silicon carbide could perform attractive role in adsorbents, electrodes, biomedical applications, etc.

Improvements in fabrication of metallic fuels

International Nuclear Information System (INIS)

Tracy, D.B.; Henslee, S.P.; Dodds, N.E.; Longua, K.J.

1989-12-01

Argonne National Laboratory is currently developing a new liquid- metal cooled breeder reactor known as the Integral Fast Reactor (IFR). IFR fuels represent the state-of-the-art in metal-fueled reactor technology. Improvements in the fabrication of metal fuel, to be discussed below, will support the fully remote fuel cycle facility that as an integral part of the IFR concept will be demonstrated at the EBR-II site. 3 refs

The analysis of fuel constituent redistribution for ternary metallic fuel slug

Energy Technology Data Exchange (ETDEWEB)

Lee, Byoung Oon; Lee, Dong Uk; Kim, Young Kyun; Chang, Jin Wook; Lee, Ki Bok; Kim, Young Il

2004-02-01

U-TRU-Zr metallic alloy is being considered as the fuel slug for the proliferation resistance core of KALIMER. The radial fuel constituent migration is a general phenomenon in the metallic alloys. This phenomenon may affect the in-reactor performance of metallic fuel rods, influencing such factors as melting temperature, thermal conductivity, power generation rate, phase boundaries and eutectic melting of the fuel slug. Thus, constituent redistribution modeling is essential when developing a metallic fuel performance code. The constituent migration model adopted in this report was based on the Ishida's model and Hofman's theory. A subroutine program has been made and installed into the MACSIS code to simulate constituent redistribution. The radial profile of Zr redistribution was calculated for the ternary metallic fuel, and compared with the measured data.

Synthesis of IV-VI Transition Metal Carbide and Nitride Nanoparticles Using a Reactive Mesoporous Template for Electrochemical Hydrogen Evolution Reaction

KAUST Repository

Alhajri, Nawal Saad

2016-01-01

Interstitial carbides and nitrides of early transition metals in Groups IV-VI exhibit platinum-like behavior which makes them a promising candidate to replace noble metals in a wide variety of reactions. Most synthetic methods used to prepare

Effect of metallic coating on the properties of copper-silicon carbide composites

Science.gov (United States)

Chmielewski, M.; Pietrzak, K.; Teodorczyk, M.; Nosewicz, S.; Jarząbek, D.; Zybała, R.; Bazarnik, P.; Lewandowska, M.; Strojny-Nędza, A.

2017-11-01

In the presented paper a coating of SiC particles with a metallic layer was used to prepare copper matrix composite materials. The role of the layer was to protect the silicon carbide from decomposition and dissolution of silicon in the copper matrix during the sintering process. The SiC particles were covered by chromium, tungsten and titanium using Plasma Vapour Deposition method. After powder mixing of components, the final densification process via Spark Plasma Sintering (SPS) method at temperature 950 °C was provided. The almost fully dense materials were obtained (>97.5%). The microstructure of obtained composites was studied using scanning electron microscopy as well as transmission electron microscopy. The microstructural analysis of composites confirmed that regardless of the type of deposited material, there is no evidence for decomposition process of silicon carbide in copper. In order to measure the strength of the interface between ceramic particles and the metal matrix, the micro tensile tests have been performed. Furthermore, thermal diffusivity was measured with the use of the laser pulse technique. In the context of performed studies, the tungsten coating seems to be the most promising solution for heat sink application. Compared to pure composites without metallic layer, Cu-SiC with W coating indicate the higher tensile strength and thermal diffusitivy, irrespective of an amount of SiC reinforcement. The improvement of the composite properties is related to advantageous condition of Cu-SiC interface characterized by well homogenity and low porosity, as well as individual properties of the tungsten coating material.

Status of SFR Metal Fuel Development

International Nuclear Information System (INIS)

Lee, Chan Bock; Lee, Byoung Oon; Kim, Ki Hwan; Kim, Sung Ho

2013-01-01

Conclusion: • Metal fuel recycling in SFR: - Enhanced utilization of uranium resource; - Efficient transmutation of minor actinides; - Inherent passive reactor safety; - Proliferation resistance with pyro-electrochemical fuel recycling. • Demonstration of technical feasibility of recycling TRU metal fuel by 2020: - Remote fuel fabrication; - Irradiation performance up to high burnup

Synthesis of Binary Transition Metal Nitrides, Carbides and Borides from the Elements in the Laser-Heated Diamond Anvil Cell and Their Structure-Property Relations

Directory of Open Access Journals (Sweden)

Lkhamsuren Bayarjargal

2011-09-01

Full Text Available Transition metal nitrides, carbides and borides have a high potential for industrial applications as they not only have a high melting point but are generally harder and less compressible than the pure metals. Here we summarize recent advances in the synthesis of binary transition metal nitrides, carbides and borides focusing on the reaction of the elements at extreme conditions generated within the laser-heated diamond anvil cell. The current knowledge of their structures and high-pressure properties like high-(p; T stability, compressibility and hardness is described as obtained from experiments.

Synthesis of Binary Transition Metal Nitrides, Carbides and Borides from the Elements in the Laser-Heated Diamond Anvil Cell and Their Structure-Property Relations

Science.gov (United States)

Friedrich, Alexandra; Winkler, Björn; Juarez-Arellano, Erick A.; Bayarjargal, Lkhamsuren

2011-01-01

Transition metal nitrides, carbides and borides have a high potential for industrial applications as they not only have a high melting point but are generally harder and less compressible than the pure metals. Here we summarize recent advances in the synthesis of binary transition metal nitrides, carbides and borides focusing on the reaction of the elements at extreme conditions generated within the laser-heated diamond anvil cell. The current knowledge of their structures and high-pressure properties like high-(p,T) stability, compressibility and hardness is described as obtained from experiments. PMID:28824101

Calculation of vapour pressures over mixed carbide fuels

International Nuclear Information System (INIS)

Joseph, M.; Mathews, C.K.

1988-01-01

Vapour pressure over the uranium-plutonium mixed carbide (Usub(l-p) Pusub(p C) was calculated in the temperature range of 1300-9000 for various compositions (p=0.1 to 0.7). Effects of variation of the sesquicarbide content were also studied. The principle of corresponding states was applied to UC and mixed carbides to obtain the equation of state. (author)

Apparatus for surface treatment of U-Pu carbide fuel samples

International Nuclear Information System (INIS)

Fukushima, Susumu; Arai, Yasuo; Handa, Muneo; Ohmichi, Toshihiko; Shiozawa, Ken-ichi.

1979-05-01

Apparatus has been constructed for treating the surface of U-Pu carbide fuel samples for EPMA. The treatment is to clean off oxide layer on the surface, then coat with an electric-conductive material. The apparatus, safe in handling plutonium, operates as follows. (1) To avoid oxidation of the analyzing surface by oxygen and water in the air, series of cleaning and coating, i.e. ion-etching and ion-coating or ion-etching and vacuum-evaporation is done at the same time in an inert gas atmosphere. (2) Ion-etching is possible on samples embedded in non-electric-conductive and low heat-conductive resin. (3) Since the temperature rise in (2) is negligible, there is no deterioration of the samples. (author)

Reference binding energies of transition metal carbides by core-level x-ray photoelectron spectroscopy free from Ar+ etching artefacts

Science.gov (United States)

Greczynski, G.; Primetzhofer, D.; Hultman, L.

2018-04-01

We report x-ray photoelectron spectroscopy (XPS) core level binding energies (BE's) for the widely-applicable groups IVb-VIb transition metal carbides (TMCs) TiC, VC, CrC, ZrC, NbC, MoC, HfC, TaC, and WC. Thin film samples are grown in the same deposition system, by dc magnetron co-sputtering from graphite and respective elemental metal targets in Ar atmosphere. To remove surface contaminations resulting from exposure to air during sample transfer from the growth chamber into the XPS system, layers are either (i) Ar+ ion-etched or (ii) UHV-annealed in situ prior to XPS analyses. High resolution XPS spectra reveal that even gentle etching affects the shape of core level signals, as well as BE values, which are systematically offset by 0.2-0.5 eV towards lower BE. These destructive effects of Ar+ ion etch become more pronounced with increasing the metal atom mass due to an increasing carbon-to-metal sputter yield ratio. Systematic analysis reveals that for each row in the periodic table (3d, 4d, and 5d) C 1s BE increases from left to right indicative of a decreased charge transfer from TM to C atoms, hence bond weakening. Moreover, C 1s BE decreases linearly with increasing carbide/metal melting point ratio. Spectra reported here, acquired from a consistent set of samples in the same instrument, should serve as a reference for true deconvolution of complex XPS cases, including multinary carbides, nitrides, and carbonitrides.

Process for the manufacture of a fuel catalyst made of tungsten carbide for electrochemical fuel cells. Verfahren zur Herstellung eines Brennstoffkatalysators aus Wolframcarbid fuer elektrochemische Brennstoffzellen

Energy Technology Data Exchange (ETDEWEB)

Baresel, D.; Gellert, W.; Scharner, P.

1982-05-19

The invention refers to a process for the manufacture of a fuel catalyst made of tungsten carbide for the direct generation of electrical energy by the oxidation of hydrogen, formaldehyde or formic acid in electrochemical fuel cells. Tungsten carbide is obtained by carburisation of tungsten or tungsten oxide by carbon monoxide. The steps of the process are as follows: dissolving the commercial-quality tungstic acid in ammonium hydroxide; precipitating the tungstic acid with concentrated hydrochloric acid; drying in a vacuum and then heating to 200/sup 0/C to remove the water of crystallisation forming tungsten trioxide; and mixing the tungsten trioxide with zinc powder and heating to 600/sup 0/C. The zinc oxide is dissolved with hydrochloric acid after cooling. The finely divided tungsten obtained in this way is converted with carbon monoxide in a quartz tube at 700/sup 0/C.

Metal-carbide multilayers for molten Pu containment

International Nuclear Information System (INIS)

Summers, T.S.E.; Curtis, P.G.; Juntz, R.S.; Krueger, R.L.

1991-12-01

Multilayers composed of nine or ten alternating layers of Ta or W and TaC were studied for the feasibility of their use in containing molten plutonium (Pu) at 1200 degrees C. Single layers of W and TaC were also investigated. A two-source electron beam evaporation process was developed to deposit these coatings onto the inside surface of hemispherical Ta cups about 38 mm in diameter. Pu testing was done by melting Pu in the coated hemispherical cups and holding them under vacuum at 1200 degrees C for two hours. Metallographic examination and microprobe analysis of cross sections showed that Pu had penetrated to the Ta substrate in all cases to some extent. Full penetration to the outer surface of the Ta substrate, however, occurred in only a few of the samples. The fact that full penetration occurred in any of the samples suggests that it would have occurred in uncoated Ta under these testing conditions which in turn suggests that the multilayer coatings do afford some protection against Pu attack. The TaC used for these specimens was wet by Pu under these testing conditions, and following testing, Pu was found uniformly distributed throughout the carbide layers which appeared to be rather porous. Pu was seen in the W and Ta layers only when exposed directly to molten Pu during testing or near defects suggesting that Pu penetrated the multilayers at defects in the coating and traveled parallel to the layers along the carbide layers. These results indicate that the use of alternating metal and ceramic layers for Pu containment should be possible through the use of nonporous ceramic that is not wet by molten Pu and defect-free films

Fuel performance of DOE fuels in water storage

International Nuclear Information System (INIS)

Hoskins, A.P.; Scott, J.G.; Shelton-Davis, C.V.; McDannel, G.E.

1993-01-01

Westinghouse Idaho Nuclear Company operates the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory. In April of 1992, the U.S. Department of Energy (DOE) decided to end the fuel reprocessing mission at ICPP. Fuel performance in storage received increased emphasis as the fuel now needs to be stored until final dispositioning is defined and implemented. Fuels are stored in four main areas: an original underwater storage facility, a modern underwater storage facility, and two dry fuel storage facilities. As a result of the reactor research mission of the DOE and predecessor agencies, the Energy Research and Development Administration and the Atomic Energy Commission, many types of nuclear fuel have been developed, used, and assigned to storage at the ICPP. Fuel clad with stainless steel, zirconium, aluminum, and graphite are represented. Fuel matrices include uranium oxide, hydride, carbide, metal, and alloy fuels, resulting in 55 different fuel types in storage. Also included in the fuel storage inventory is canned scrap material

Performance of refractory alloy-clad fuel pins

International Nuclear Information System (INIS)

Dutt, D.S.; Cox, C.M.; Millhollen, M.K.

1984-12-01

This paper discusses objectives and basic design of two fuel-cladding tests being conducted in support of SP-100 technology development. Two of the current space nuclear power concepts use conventional pin type designs, where a coolant removes the heat from the core and transports it to an out-of-core energy conversion system. An extensive irradiation testing program was conducted in the 1950's and 1960's to develop fuel pins for space nuclear reactors. The program emphasized refractory metal clad uranium nitride (UN), uranium carbide (UC), uranium oxide (UO 2 ), and metal matrix fuels (UCZr and BeO-UO 2 ). Based on this earlier work, studies presented here show that UN and UO 2 fuels in conjunction with several refractory metal cladding materials demonstrated high potential for meeting space reactor requirements and that UC could serve as an alternative but higher risk fuel

High temperature evaporation of titanium, zirconium and hafnium carbides

International Nuclear Information System (INIS)

Gusev, A.I.; Rempel', A.A.

1991-01-01

Evaporation of cubic nonstoichiometric carbides of titanium, zirconium and hafnium in a comparatively low-temperature interval (1800-2700) with detailed crystallochemical sample certification is studied. Titanium carbide is characterized by the maximum evaporation rate: at T>2300 K it loses 3% of sample mass during an hour and at T>2400 K titanium carbide evaporation becomes extremely rapid. Zirconium and hafnium carbide evaporation rates are several times lower than titanium carbide evaporation rates at similar temperatures. Partial pressures of metals and carbon over the carbides studied are calculated on the base of evaporation rates

From Electronic Structure to Catalytic Activity: A Single Descriptor for Adsorption and Reactivity on Transition-Metal Carbides

DEFF Research Database (Denmark)

Vojvodic, A.; Hellman, Anders; Ruberto, C.

2009-01-01

Adsorption and catalytic properties of the polar (111) surface of transition-metal carbides (TMC's) are investigated by density-functional theory. Atomic and molecular adsorption are rationalized with the concerted-coupling model, in which two types of TMC surface resonances (SR's) play key roles...

Development of Metallic Fuels for Actinide Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hayes, Steven Lowe [Idaho National Laboratory; Fielding, Randall Sidney [Idaho National Laboratory; Benson, Michael Timothy [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory; Carmack, William Jonathan [Idaho National Laboratory

2015-09-01

Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimized for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr between 10

Surface metallurgy of cemented carbide tools

International Nuclear Information System (INIS)

Chopra, K.L.; Kashyap, S.C.; Rao, T.V.; Rajagopalan, S.; Srivastava, P.K.

1983-01-01

Transition metal carbides, owing to their high melting point, hardness and wear resistance, are potential candidates for specific application in rockets, nuclear engineering equipment and cutting tools. Tungsten carbide sintered with a binder (either cobalt metal or a mixture of Co + TiC and/or TaC(NbC)) is used for cutting tools. The surface metallurgy of several commercially available cemented carbide tools was studied by Auger electron spectroscopy and X-ray photoelectron spectroscopy techniques. The tool surfaces were contaminated by adsorbed oxygen up to a depth of nearly 0.3 μm causing deterioration of the mechanical properties of the tools. Studies of fractured samples indicated that the tool surfaces were prone to oxygen adsorption. The fracture path passes through the cobalt-rich regions. The ineffectiveness of a worn cutting tool is attributed to the presence of excessive iron from the steel workpiece and carbon and oxygen in the surface layers of the tool. The use of appropriate hard coatings on cemented carbide tools is suggested. (Auth.)

Research and development of thorium fuel cycle

International Nuclear Information System (INIS)

Oishi, Jun.

1994-01-01

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

Joining of boron carbide using nickel interlayer

International Nuclear Information System (INIS)

Vosughi, A.; Hadian, A. M.

2008-01-01

Carbide ceramics such as boron carbide due to their unique properties such as low density, high refractoriness, and high strength to weight ratio have many applications in different industries. This study focuses on direct bonding of boron carbide for high temperature applications using nickel interlayer. The process variables such as bonding time, temperature, and pressure have been investigated. The microstructure of the joint area was studied using electron scanning microscope technique. At all the bonding temperatures ranging from 1150 to 1300 d eg C a reaction layer formed across the ceramic/metal interface. The thickness of the reaction layer increased by increasing temperature. The strength of the bonded samples was measured using shear testing method. The highest strength value obtained was about 100 MPa and belonged to the samples bonded at 1250 for 75 min bonding time. The strength of the joints decreased by increasing the bonding temperature above 1250 d eg C . The results of this study showed that direct bonding technique along with nickel interlayer can be successfully utilized for bonding boron carbide ceramic to itself. This method may be used for bonding boron carbide to metals as well.

Synthesis of carbide fuels from nano-structured precursors: impact on carbo-reduction and physico-chemical properties

International Nuclear Information System (INIS)

Saravia, Alvaro

2015-01-01

The classical way classically used for manufacturing carbide fuels consists of carbo-reducing at high temperature (1600 C) and under primary vacuum a mixture of AnO 2 and graphite powders. These conditions are disadvantageous for the synthesis of mixed (U,Pu)C carbides on account of plutonium volatilization. Therefore, one of the main aims of these studies is to decrease the carbo-reduction temperature. The experiments focused mainly on the lowering of the uranium oxide temperature. This result has been obtained with the use of uranium oxide and carbon nano-structured precursors. To achieve this goal colloidal suspensions of uranium oxide have been prepared and stabilized by cellulosic ethers. Cellulosic ethers are both stabiliser for uranium oxide nanoparticles and carbon source for carbo-reduction. It has been shown that these precursors are more efficient for carbo-reduction than the standard precursors: a reduction of 300 C of carbo-reduction temperature has been obtained. The impact of these precursors on carbo-reduction and on physico-chemical properties as well as the structural and microstructural characterizations of the obtained carbides have been carried out. (author) [fr

Development of advanced nuclear fuels in the Indian context: advantages and challenges

International Nuclear Information System (INIS)

Ganesan, V.

2012-01-01

The ever increasing demand on power requirement in the country has opened up need for exploring use of nuclear fuels that could meet such demands. This makes the mission of the department to shift from the first stage of nuclear programme employing natural uranium in PHWRs to the second stage of deploying a large number of fast reactors with plutonium based fuels capable of realising high breeding ratios in addition to energy production. The transition to fast reactors with advanced fuels, capable of higher breeding ratio, opens up a number of scientific and technological challenges in design and operation of such fast reactors. In the Indian context, after successful demonstration of natural uranium based PHWRs, the performance of U-Pu based carbide fuel, as a unique experience in the world, has been demonstrated in FBTR at Kalpakkam. This paper deals with the performance of carbide fuel in FBTR and the programme on development of metallic fuels with appreciably high breeding ratio that would result in considerable reduction in doubling time thereby addressing the increasing demands of power production as well as pave way for introduction of a large number of such fast reactors to provide energy security to the country. The advantages of introduction of metallic fuels as well as the scientific and technological challenges to be faced in doing so and the ongoing efforts towards metallic fuel development are also described in the paper. (author)

Medium temperature reaction between lanthanide and actinide carbides and hydrogen

International Nuclear Information System (INIS)

Dean, G.; Lorenzelli, R.; Pascard, R.

1964-01-01

Hydrogen is fixed reversibly by the lanthanide and actinide mono carbides in the range 25 - 400 C, as for pure corresponding metals. Hydrogen goes into the carbides lattice through carbon vacancies and the total fixed amount is approximately equal to two hydrogen atoms per initial vacancy. Final products c.n thus be considered as carbo-hydrides of general formula M(C 1-x , H 2x ). The primitive CFC, NaCl type, structure remains unchanged but expands strongly in the case of actinide carbides. With lanthanide carbides, hydrogenation induces a phase transformation with reappearance of the metal structure (HCP). Hydrogen decomposition pressures of all the studied carbo-hydrides are greater than those of the corresponding di-hydrides. (authors) [fr

Improvements in the fabrication of metallic fuels

International Nuclear Information System (INIS)

Tracy, D.B.; Henslee, S.P.; Dodds, N.E.; Longua, K.J.

1989-01-01

Argonne National Laboratory (ANL) is currently developing a new liquid-metal-cooled breeder reactor known as the Integral Fast Reactor (IFR). The IFR represents the state of the art in metal-fueled reactor technology. Improvements in the fabrication of metal fuel, discussed in this paper, will support ANL-West's (ANL-W) fully remote fuel cycle facility, which is an integral part of the IFR concept

Microsegregation in Nodular Cast Iron with Carbides

Directory of Open Access Journals (Sweden)

S. Pietrowski

2012-12-01

Full Text Available In this paper results of microsegregation in the newly developed nodular cast iron with carbides are presented. To investigate the pearlitic and bainitic cast iron with carbides obtained by Inmold method were chosen. The distribution of linear elements on the eutectic cell radius was examined. To investigate the microsegregation pearlitic and bainitic cast iron with carbides obtained by Inmold method were chosen.The linear distribution of elements on the eutectic cell radius was examined. Testing of the chemical composition of cast iron metal matrix components, including carbides were carried out. The change of graphitizing and anti-graphitizing element concentrations within eutectic cell was determined. It was found, that in cast iron containing Mo carbides crystallizing after austenite + graphite eutectic are Si enriched.

Microsegregation in Nodular Cast Iron with Carbides

Directory of Open Access Journals (Sweden)

Pietrowski S.

2012-12-01

Full Text Available In this paper results of microsegregation in the newly developed nodular cast iron with carbides are presented. To investigate the pearlitic and bainitic cast iron with carbides obtained by Inmold method were chosen. The distribution of linear elements on the eutectic cell radius was examined. To investigate the microsegregation pearlitic and bainitic cast iron with carbides obtained by Inmold method were chosen. The linear distribution of elements on the eutectic cell radius was examined. Testing of the chemical composition of cast iron metal matrix components, including carbides were carried out. The change of graphitizing and anti-graphitizing element concentrations within eutectic cell was determined. It was found, that in cast iron containing Mo carbides crystallizing after austenite + graphite eutectic are Si enriched.

Development of metallic fuel fabrication

Energy Technology Data Exchange (ETDEWEB)

Kang, Young Ho; Lee, Chong Yak; Lee, Myung Ho and others

1999-03-01

With the vacuum melting and casting of the U-10wt%Zr alloy which is metallic fuel for liquid metal fast breeder reactor, we studied the microstructure of the alloy and the parameters of the melting and casting for the fuel rods. Internal defects of the U-10wt%Zr fuel by gravity casting, were inspected by non-destructive test. U-10wt%Zr alloy has been prepared for the thermal stability test in order to estimate the decomposition of the lamellar structure with relation to swelling under irradiation condition. (author)

2D metal carbides and nitrides (MXenes) for energy storage

KAUST Repository

Anasori, Babak; Lukatskaya, Maria R.; Gogotsi, Yury

2017-01-01

The family of 2D transition metal carbides, carbonitrides and nitrides (collectively referred to as MXenes) has expanded rapidly since the discovery of Ti3C2 in 2011. The materials reported so far always have surface terminations, such as hydroxyl, oxygen or fluorine, which impart hydrophilicity to their surfaces. About 20 different MXenes have been synthesized, and the structures and properties of dozens more have been theoretically predicted. The availability of solid solutions, the control of surface terminations and a recent discovery of multi-transition-metal layered MXenes offer the potential for synthesis of many new structures. The versatile chemistry of MXenes allows the tuning of properties for applications including energy storage, electromagnetic interference shielding, reinforcement for composites, water purification, gas- and biosensors, lubrication, and photo-, electro- and chemical catalysis. Attractive electronic, optical, plasmonic and thermoelectric properties have also been shown. In this Review, we present the synthesis, structure and properties of MXenes, as well as their energy storage and related applications, and an outlook for future research.

2D metal carbides and nitrides (MXenes) for energy storage

KAUST Repository

Anasori, Babak

2017-01-17

The family of 2D transition metal carbides, carbonitrides and nitrides (collectively referred to as MXenes) has expanded rapidly since the discovery of Ti3C2 in 2011. The materials reported so far always have surface terminations, such as hydroxyl, oxygen or fluorine, which impart hydrophilicity to their surfaces. About 20 different MXenes have been synthesized, and the structures and properties of dozens more have been theoretically predicted. The availability of solid solutions, the control of surface terminations and a recent discovery of multi-transition-metal layered MXenes offer the potential for synthesis of many new structures. The versatile chemistry of MXenes allows the tuning of properties for applications including energy storage, electromagnetic interference shielding, reinforcement for composites, water purification, gas- and biosensors, lubrication, and photo-, electro- and chemical catalysis. Attractive electronic, optical, plasmonic and thermoelectric properties have also been shown. In this Review, we present the synthesis, structure and properties of MXenes, as well as their energy storage and related applications, and an outlook for future research.

Control of electronic properties of 2D carbides (MXenes) by manipulating their transition metal layers

KAUST Repository

Anasori, Babak

2016-02-24

In this study, a transition from metallic to semiconducting-like behavior has been demonstrated in two-dimensional (2D) transition metal carbides by replacing titanium with molybdenum in the outer transition metal (M) layers of M3C2 and M4C3 MXenes. The MXene structure consists of n + 1 layers of near-close packed M layers with C or N occupying the octahedral site between them in an [MX]nM arrangement. Recently, two new families of ordered 2D double transition metal carbides MXenes were discovered, M′2M′′C2 and M′2M′′2C3 – where M′ and M′′ are two different early transition metals, such as Mo, Cr, Ta, Nb, V, and Ti. The M′ atoms only occupy the outer layers and the M′′ atoms fill the middle layers. In other words, M′ atomic layers sandwich the middle M′′–C layers. Using X-ray atomic pair distribution function (PDF) analysis on Mo2TiC2 and Mo2Ti2C3 MXenes, we present the first quantitative analysis of structures of these novel materials and experimentally confirm that Mo atoms are in the outer layers of the [MC]nM structures. The electronic properties of these Mo-containing MXenes are compared with their Ti3C2 counterparts, and are found to be no longer metallic-like conductors; instead the resistance increases mildly with decreasing temperatures. Density functional theory (DFT) calculations suggest that OH terminated Mo–Ti MXenes are semiconductors with narrow band gaps. Measurements of the temperature dependencies of conductivities and magnetoresistances have confirmed that Mo2TiC2Tx exhibits semiconductor-like transport behavior, while Ti3C2Tx is a metal. This finding opens new avenues for the control of the electronic and optical applications of MXenes and for exploring new applications, in which semiconducting properties are required.

Alternative Fabrication of Recycling Fast Reactor Metal Fuel

International Nuclear Information System (INIS)

Kim, Ki-Hwan; Kim, Jong Hwan; Song, Hoon; Kim, Hyung-Tae; Lee, Chan-Bock

2015-01-01

Metal fuels such as U-Zr/U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. In order to develop innovative fabrication method of metal fuel for preventing the evaporation of volatile elements such as Am, modified casting under inert atmosphere has been applied for metal fuel slugs for SFR. Alternative fabrication method of fuel slugs has been introduced to develop an improved fabrication process of metal fuel for preventing the evaporation of volatile elements. In this study, metal fuel slugs for SFR have been fabricated by modified casting method, and characterized to evaluate the feasibility of the alternative fabrication method. In order to prevent evaporation of volatile elements such as Am and improve quality of fuel slugs, alternative fabrication methods of metal fuel slugs have been studied in KAERI. U-10Zr-5Mn fuel slug containing volatile surrogate element Mn was soundly cast by modified injection casting under modest pressure. Evaporation of Mn during alternative casting could not be detected by chemical analysis. Mn element was most recovered with prevention of evaporation by alternative casting. Modified injection casting has been selected as an alternative fabrication method in KAERI, considering evaporation prevention, and proven benefits of high productivity, high yield, and good remote control

Thermal-Hydraulic Aspects of Changing the Nuclear Fuel-Cladding Materials from Zircaloy to Silicon Carbides

International Nuclear Information System (INIS)

Niceno, Bojan; Pouchon, Manuel

2014-01-01

The accident in Fukushima has drastically shown the drawbacks of Zircaloy claddings despite their beneficial properties in normal use. The effect of the lack of cooling and the production of hydrogen would not have been so strong if the fuel cladding had not consisted of a zirconium (or metal) alloy. International activities have been started to search for an alternative to Zircaloy, however, still on a limited basis. A project sponsored by Swissnuclear has been conducted at Paul Scherrer Institute (PSI) with the aim to close the gap in knowledge on application of silicon carbides (SiC) as potential replacement for Zircaloys as material for nuclear fuel cladding. The work was interdisciplinary, result of collaboration between different laboratories at PSI, and has focused on SiC cladding material properties, implication of its usage on neutronics and on thermal-hydraulics. This paper summarizes thermal-hydraulic aspects of changing Zircaloy for SiC as the cladding material. The change of cladding material inevitably changes the surface properties thus making a significant impact on boiling curve, and critical heat flux (CHF). Low chemical reactivity of SiC means fewer particles in the flow (less crud), which leads to fewer failures, but also decreases the CHF. Due to differences in physical properties between SiC and Zircaloys, higher brittleness of SiC in particular, might have impact on fuel-rod assembly design, which has direct influence on flow patterns and heat transfer in the fuel assembly. Higher melting (i.e. decomposition) point for SiC means that severe accident management guidelines (SAMG) should have to be re-assessed. Not only would the core degrade later than in the case of conventional fuels, but the production of hydrogen would be quite different as well. All these issues are explored in this work in two steps; first the SiC properties which may have influence on thermal-hydraulics are outlined, then each thermal-hydraulic issues is explained from

New Icosahedral Boron Carbide Semiconductors

Science.gov (United States)

Echeverria Mora, Elena Maria

Novel semiconductor boron carbide films and boron carbide films doped with aromatic compounds have been investigated and characterized. Most of these semiconductors were formed by plasma enhanced chemical vapor deposition. The aromatic compound additives used, in this thesis, were pyridine (Py), aniline, and diaminobenzene (DAB). As one of the key parameters for semiconducting device functionality is the metal contact and, therefore, the chemical interactions or band bending that may occur at the metal/semiconductor interface, X-ray photoemission spectroscopy has been used to investigate the interaction of gold (Au) with these novel boron carbide-based semiconductors. Both n- and p-type films have been tested and pure boron carbide devices are compared to those containing aromatic compounds. The results show that boron carbide seems to behave differently from other semiconductors, opening a way for new analysis and approaches in device's functionality. By studying the electrical and optical properties of these films, it has been found that samples containing the aromatic compound exhibit an improvement in the electron-hole separation and charge extraction, as well as a decrease in the band gap. The hole carrier lifetimes for each sample were extracted from the capacitance-voltage, C(V), and current-voltage, I(V), curves. Additionally, devices, with boron carbide with the addition of pyridine, exhibited better collection of neutron capture generated pulses at ZERO applied bias, compared to the pure boron carbide samples. This is consistent with the longer carrier lifetimes estimated for these films. The I-V curves, as a function of external magnetic field, of the pure boron carbide films and films containing DAB demonstrate that significant room temperature negative magneto-resistance (> 100% for pure samples, and > 50% for samples containing DAB) is possible in the resulting dielectric thin films. Inclusion of DAB is not essential for significant negative magneto

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

International Nuclear Information System (INIS)

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

2012-01-01

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

Metallic fuel design development

International Nuclear Information System (INIS)

Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

1999-04-01

This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual design technology on metallic fuel

Laser deposition of carbide-reinforced coatings

International Nuclear Information System (INIS)

Cerri, W.; Martinella, R.; Mor, G.P.; Bianchi, P.; D'Angelo, D.

1991-01-01

CO 2 laser cladding with blown powder presents many advantages: fusion bonding with the substrate with low dilution, metallurgical continuity in the metallic matrix, high solidification rates, ease of automation, and reduced environmental contamination. In the present paper, laser cladding experimental results using families of carbides (tungsten and titanium) mixed with metallic alloys are reported. As substrates, low alloy construction steel (AISI 4140) (austenitic stainless steel) samples have been utilized, depending on the particular carbide reinforcement application. The coating layers obtained have been characterized by metallurgical examination. They show low dilution, absence of cracks, and high abrasion resistance. The WC samples, obtained with different carbide sizes and percentages, have been characterized with dry and rubber wheel abrasion tests and the specimen behaviour has been compared with the behaviour of materials used for similar applications. The abrasion resistance proved to be better than that of other widely used hardfacing materials and the powder morphology have a non-negligible influence on the tribological properties. (orig.)

Fracture and Residual Characterization of Tungsten Carbide Cobalt Coatings on High Strength Steel

National Research Council Canada - National Science Library

Parker, Donald S

2003-01-01

Tungsten carbide cobalt coatings applied via high velocity oxygen fuel thermal spray deposition are essentially anisotropic composite structures with aggregates of tungsten carbide particles bonded...

Fracture of coherent interfaces between an fcc metal matrix and the Cr23C6 carbide precipitate from first principles

Science.gov (United States)

Barbé, Elric; Fu, Chu-Chun; Sauzay, Maxime

2018-02-01

It is known that microcrack initiation in metallic alloys containing second-phase particles may be caused by either an interfacial or an intraprecipitate fracture. So far, the dependence of these features on properties of the precipitate and the interface is not clearly known. The present study aims to determine the key properties of carbide-metal interfaces controlling the energy and critical stress of fracture, based on density functional theory (DFT) calculations. We address coherent interfaces between a fcc iron or nickel matrix and a frequently observed carbide, the M23C6 , for which a simplified chemical composition Cr23C6 is assumed. The interfacial properties such as the formation and Griffith energies, and the effective Young's modulus are analyzed as functions of the magnetic state of the metal lattice, including the paramagnetic phase of iron. Interestingly, a simpler antiferromagnetic phase is found to exhibit similar interfacial mechanical behavior to the paramagnetic phase. A linear dependence is determined between the surface (and interface) energy and the variation of the number of chemical bonds weighted by the respective bond strength, which can be used to predict the relative formation energy for the surface and interface with various chemical terminations. Finally, the critical stresses of both intraprecipitate and interfacial fractures due to a tensile loading are estimated via the universal binding energy relation (UBER) model, parametrized on the DFT data. The validity of this model is verified in the case of intraprecipitate fracture, against results from DFT tensile test simulations. In agreement with experimental evidences, we predict a much stronger tendency for an interfacial fracture for this carbide. In addition, the calculated interfacial critical stresses are fully compatible with available experimental data in steels, where the interfacial carbide-matrix fracture is only observed at incoherent interfaces.

Studies and manufacture of plutonium fuel

International Nuclear Information System (INIS)

Bussy, P.; Mustelier, J.P.; Pascard, R.

1964-01-01

The studies carried out at the C.E.A. on the properties of fast neutron reactor fuels, the manufacture of fuel elements and their behaviour under irradiation are broadly outlined. The metal fuels studied are the ternary alloys U Pu Mo, U Pu Nb, U Pa Ti, U Pa Zr, the ceramic fuels being mixed uranium and plutonium oxides, carbides and nitrides obtained by sintering. Results are given on the manufacture of uranium fuel elements containing a small proportion of plutonium, used in a critical experiment, and on the first experiments in the manufacture of fuel elements for the reactor Rapsodie. Finally the results of irradiation tests carried out on the prototype fuel pins for Rapsodie are described. (authors) [fr

Recent metal fuel safety tests in TREAT

International Nuclear Information System (INIS)

Wright, A.E.; Bauer, T.H.; Lo, R.K.; Robinson, W.R.; Palm, R.G.

1986-01-01

In-reactor safety tests have been performed on metal-alloy reactor fuel to study its response to transient-overpower conditions, in particular, the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Uranium-fissium EBR-II driver fuel elements of several burnups were tested, some to cladding breach and others to incipient breach. Transient fuel motions were monitored, and time and location of breach were measured. The test results and computations of fuel extrusion and cladding failure in metal-alloy fuel are described

Structure-Property Relationship in Metal Carbides and Bimetallic Alloys

Energy Technology Data Exchange (ETDEWEB)

Chen, Jingguan [University of Delaware

2014-03-04

The primary objective of our DOE/BES sponsored research is to use carbide and bimetallic catalysts as model systems to demonstrate the feasibility of tuning the catalytic activity, selectivity and stability. Our efforts involve three parallel approaches, with the aim at studying single crystal model surfaces and bridging the “materials gap” and “pressure gap” between fundamental surface science studies and real world catalysis. The utilization of the three parallel approaches has led to the discovery of many intriguing catalytic properties of carbide and bimetallic surfaces and catalysts. During the past funding period we have utilized these combined research approaches to explore the possibility of predicting and verifying bimetallic and carbide combinations with enhanced catalytic activity, selectivity and stability.

Ultra-high-rate pseudocapacitive energy storage in two-dimensional transition metal carbides

Science.gov (United States)

Lukatskaya, Maria R.; Kota, Sankalp; Lin, Zifeng; Zhao, Meng-Qiang; Shpigel, Netanel; Levi, Mikhael D.; Halim, Joseph; Taberna, Pierre-Louis; Barsoum, Michel W.; Simon, Patrice; Gogotsi, Yury

2017-08-01

The use of fast surface redox storage (pseudocapacitive) mechanisms can enable devices that store much more energy than electrical double-layer capacitors (EDLCs) and, unlike batteries, can do so quite rapidly. Yet, few pseudocapacitive transition metal oxides can provide a high power capability due to their low intrinsic electronic and ionic conductivity. Here we demonstrate that two-dimensional transition metal carbides (MXenes) can operate at rates exceeding those of conventional EDLCs, but still provide higher volumetric and areal capacitance than carbon, electrically conducting polymers or transition metal oxides. We applied two distinct designs for MXene electrode architectures with improved ion accessibility to redox-active sites. A macroporous Ti3C2Tx MXene film delivered up to 210 F g-1 at scan rates of 10 V s-1, surpassing the best carbon supercapacitors known. In contrast, we show that MXene hydrogels are able to deliver volumetric capacitance of ˜1,500 F cm-3 reaching the previously unmatched volumetric performance of RuO2.

Development of Plasma-Sprayed Molybdenum Carbide-Based Anode Layers with Various Metal Oxides for SOFC

Science.gov (United States)

Faisal, N. H.; Ahmed, R.; Katikaneni, S. P.; Souentie, S.; Goosen, M. F. A.

2015-12-01

Air plasma-sprayed (APS) coatings provide an ability to deposit a range of novel fuel cell materials at competitive costs. This work develops three separate types of composite anodes (Mo-Mo2C/Al2O3, Mo-Mo2C/ZrO2, Mo-Mo2C/TiO2) using a combination of APS process parameters on Hastelloy®X for application in intermediate temperature proton-conducting solid oxide fuel cells. Commercially available carbide of molybdenum powder catalyst (Mo-Mo2C) and three metal oxides (Al2O3, ZrO2, TiO2) was used to prepare three separate composite feedstock powders to fabricate three different anodes. Each of the modified composition anode feedstock powders included a stoichiometric weight ratio of 0.8:0.2. The coatings were characterized by scanning electron microscopy, energy dispersive spectroscopy, x-ray diffraction, nanoindentation, and conductivity. We report herein that three optimized anode layers of thicknesses between 200 and 300 µm and porosity as high as 20% for Mo-Mo2C/Al2O3 (250-µm thick) and Mo-Mo2C/TiO2 (300 µm thick) and 17% for Mo-Mo2C/ZrO2 (220-µm thick), controllable by a selection of the APS process parameters with no addition of sacrificial pore-forming material. The nanohardness results indicate the upper layers of the coatings have higher values than the subsurface layers in coatings with some effect of the deposition on the substrate. Mo-Mo2C/ZrO2 shows high electrical conductivity.

Recent progress in the development of metallic fuel

International Nuclear Information System (INIS)

Seidel, B.R.; Batte, G.L.; Dodds, N.E.; Lahm, C.E.; Pahl, R.G.; Tsai, H.C.

1990-01-01

Tests to date demonstrate that metallic fuel for advanced liquid metal reactors performs well, is easily reprocessed and refabricated and provides inherent reactor safety within an economic design. The behavior and performance of metallic fuel is key to the demonstration of the Integral Fast Reactor (IFR) concept at Argonne National Laboratory. Since 1985, more than 40 assemblies of experimental fuel in addition to the standard metallic driver fuel for Experimental Breeder Reactor 2 (EBR-2)have been irradiated; several more continue to be designed and fabricated. Results have characterized the influence of a wide range of fabrication, design and material variables upon irradiation behavior throughout the fuel lifetime under normal and upset conditions including operation with breached cladding. Results of test, both in- and out-of-reactor, indicate that metallic fuel is readily and economically fabricated, capable of achieving high exposure and long reactor residence times, and possesses unique and promising safety features. 9 refs., 6 figs

Direct electrical heating of irradiated metal fuel

International Nuclear Information System (INIS)

Fenske, G.R.; Emerson, J.E.; Savoie, F.E.; Johanson, E.W.

1985-01-01

The Integral Fast Reactor (IFR) concept proposed by Argonne National Laboratory utilizes a metal fuel core. Reactor safety analysis requires information on the potential for fuel axial expansion during severe thermal transients. In addition to a comparatively large thermal expansion coefficient, metallic fuel has a unique potential for enhanced pre-failure expansion driven by retained fission gas and ingested bond sodium. In this paper, the authors present preliminary results from three direct electrical heating (DEH) experiments performed on irradiated metal fuel to investigate axial expansion behavior. The test samples were from Experimental Breeder Reactor II (EBR-II) driver fuel ML-11 irradiated to 8 at.% burnup. Preliminary analysis of the results suggest that enhanced expansion driven by trapped fission gas can occur

Gas chromatographic determination of Di-n-butyl phosphate in radioactive lean organic solvent of FBTR carbide fuel reprocessing

International Nuclear Information System (INIS)

Velavendan, P.; Ganesh, S.; Pandey, N.K.; Kamachi Mudali, U.; Natarajan, R.

2011-01-01

In the present work Di-n- butyl phosphate (DBP) a degraded product of Tri-n-butyl phosphate (TBP) formed by acid hydrolysis and radiolysis in the PUREX process was analyzed. Lean organic streams of different fuel burn-up FBTR carbide fuel reprocessing solution was determined by standard Gas Chromatographic technique. The method involves the conversion of non-volatile Di-n-butyl phosphate into volatile and stable derivatives by the action of diazomethane and then determined by Gas Chromatograph (GC). A calibration graph was made for DBP concentration range of 200-2000 ppm with correlation coefficient of 0.99587 and RSD 1.2 %. (author)

Fuel element for a nuclear reactor

International Nuclear Information System (INIS)

Linning, D.L.

1977-01-01

An improvement of the fuel element for a fast nuclear reactor described in patent 15 89 010 is proposed which should avoid possible damage due to swelling of the fuel. While the fuel element according to patent 15 89 010 is made in the form of a tube, here a further metal jacket is inserted in the centre of the fuel rod and the intermediate layer (ceramic uranium compound) is provided on both sides, so that the nuclear fuel is situated in the centre of the annular construction. Ceramic uranium or plutonium compounds (preferably carbide) form the fuel zone in the form of circular pellets, which are surrounded by annular gaps, so that gaseous fission products can escape. (UWI) [de

Advanced Characterization Techniques for Silicon Carbide and Pyrocarbon Coatings on Fuel Particles for High Temperature Reactors (HTR)

Energy Technology Data Exchange (ETDEWEB)

Basini, V.; Charollais, F. [CEA Cadarache, DEN/DEC/SPUA, BP 1, 13108 St Paul Lez Durance (France); Dugne, O. [CEA Marcoule, DEN/DTEC/SCGS BP 17171 30207 Bagnols sur Ceze (France); Garcia, C. [Laboratoire des Composites Thermostructuraux (LCTS), UMR CNRS 5801, 3 allee de La Boetie, 33600 Pessac (France); Perez, M. [CEA Grenoble DRT/DTH/LTH, 17 rue des Martyrs, 38054 Grenoble cedex 9 (France)

2008-07-01

Cea and AREVA NP have engaged an extensive research and development program on HTR (high temperature reactor) fuel. The improving of safety of (very) high temperature reactors (V/HTR) is based on the quality of the fuel particles. This requires a good knowledge of the properties of the four-layers TRISO particles designed to retain the uranium and fission products during irradiation or accident conditions. The aim of this work is to characterize exhaustively the structure and the thermomechanical properties of each unirradiated layer (silicon carbide and pyrocarbon coatings) by electron microscopy (SEM, TEM), selected area electronic diffraction (SEAD), thermo reflectance microscopy and nano-indentation. The long term objective of this study is to define pertinent parameters for fuel performance codes used to better understand the thermomechanical behaviour of the coated particles. (authors)

METHOD FOR PRODUCING CEMENTED CARBIDE ARTICLES

Science.gov (United States)

Onstott, E.I.; Cremer, G.D.

1959-07-14

A method is described for making molded materials of intricate shape where the materials consist of mixtures of one or more hard metal carbides or oxides and matrix metals or binder metals thereof. In one embodiment of the invention 90% of finely comminuted tungsten carbide powder together with finely comminuted cobalt bonding agent is incorporated at 60 deg C into a slurry with methyl alcohol containing 1.5% paraffin, 3% camphor, 3.5% naphthalene, and 1.8% toluene. The compact is formed by the steps of placing the slurry in a mold at least one surface of which is porous to the fluid organic system, compacting the slurry, removing a portion of the mold from contact with the formed object and heating the formed object to remove the remaining organic matter and to sinter the compact.

Metal Matrix Composite Material by Direct Metal Deposition

Science.gov (United States)

Novichenko, D.; Marants, A.; Thivillon, L.; Bertrand, P. H.; Smurov, I.

Direct Metal Deposition (DMD) is a laser cladding process for producing a protective coating on the surface of a metallic part or manufacturing layer-by-layer parts in a single-step process. The objective of this work is to demonstrate the possibility to create carbide-reinforced metal matrix composite objects. Powders of steel 16NCD13 with different volume contents of titanium carbide are tested. On the base of statistical analysis, a laser cladding processing map is constructed. Relationships between the different content of titanium carbide in a powder mixture and the material microstructure are found. Mechanism of formation of various precipitated titanium carbides is investigated.

Plasma spraying process of disperse carbides for spraying and facing

International Nuclear Information System (INIS)

Blinkov, I.V.; Vishnevetskaya, I.A.; Kostyukovich, T.G.; Ostapovich, A.O.

1989-01-01

A possibility to metallize carbides in plasma of impulsing capacitor discharge is considered. Powders granulation occurs during plasma spraying process, ceramic core being completely capped. X-ray phase and chemical analyses of coatings did not show considerable changes of carbon content in carbides before and after plasma processing. This distinguishes the process of carbides metallization in impulsing plasma from the similar processing in arc and high-frequency plasma generator. Use of powder composites produced in the impulsing capacitor discharge, for plasma spraying and laser facing permits 2-3 times increasing wear resistance of the surface layer as against the coatings produced from mechanical powders mixtures

Pyroprocessing of IFR Metal Fuel

International Nuclear Information System (INIS)

Laidler, J.J.

1993-01-01

The Integral Fast Reactor (IFR) fuel cycle features the use of an innovative reprocessing method, known as open-quotes pyroprocessingclose quotes featuring fused-salt electrofining of the spent fuel. Electrofining of IFR spent fuel involves uranium recovery by electro-transport to a solid steel cathode. The thermodynamics of the system preclude plutonium recovery in the same way, so a liquid cadmium cathode located in the electrolyte salt phase is utilized. The deposition of Pu, Am, Np, and Cm takes place at the liquid cadmium cathode in the form of cadmium intermetallic compounds (e.g, PuCd 6 ), and uranium deposits as the pure metal when cadmium saturation is reached. A small amount of rare earth fission products deposit together with the heavy metals at both the solid and liquid cadmium cathodes, providing a significant degree of self-protection. A full scope demonstration of the IFR fuel cycle will begin in 1993, using fuel irradiated in EBR-II

Evaluation of Metal-Fueled Surface Reactor Concepts

International Nuclear Information System (INIS)

Poston, David I.; Marcille, Thomas F.; Kapernick, Richard J.; Hiatt, Matthew T.; Amiri, Benjamin W.

2007-01-01

Surface fission power systems for use on the Moon and Mars may provide the first use of near-term reactor technology in space. Most near-term surface reactor concepts specify reactor temperatures <1000 K to allow the use of established material and power conversion technology and minimize the impact of the in-situ environment. Metal alloy fuels (e.g. U-10Zr and U-10Mo) have not traditionally been considered for space reactors because of high-temperature requirements, but they might be an attractive option for these lower temperature surface power missions. In addition to temperature limitations, metal fuels are also known to swell significantly at rather low fuel burnups (∼1 a/o), but near-term surface missions can mitigate this concern as well, because power and lifetime requirements generally keep fuel burnups <1 a/o. If temperature and swelling issues are not a concern, then a surface reactor concept may be able to benefit from the high uranium density and relative ease of manufacture of metal fuels. This paper investigates two reactor concepts that utilize metal fuels. It is found that these concepts compare very well to concepts that utilize other fuels (UN, UO2, UZrH) on a mass basis, while also providing the potential to simplify material safeguards issues

A state of the art on metallic fuel technology development

International Nuclear Information System (INIS)

Hwang, Woan; Kang, Hee Young; Nam, Cheol; Kim, Jong Oh

1997-01-01

Since worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved in the late 1960's, the development of metallic fuels continued throughout the 1970's at ANL's experimental breeder reactor II (EBR-II) because EBR-II continued to be fueled with the metallic uranium-fissium alloy, U-5Fs. During this decade the performance limitations of metallic fuel were satisfactorily resolved resolved at EBR-II. The concept of the IFR developed at ANL since 1984. The technical feasibility had been demonstrated and the technology database had been established to support its practicality. One key features of the IFR is that the fuel is metallic, which brings pronounced benefits over oxide in improved inherent safety and lower processing costs. At the outset of the 1980's, it appeared that metallic fuels are recognized as a professed viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last score and summarizes the state-of the art on metallic fuel technology development. (author). 29 refs., 1 tab

Metallic fuel design development

Energy Technology Data Exchange (ETDEWEB)

Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

1999-04-01

This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual

Device for reprocessing nuclear fuels

International Nuclear Information System (INIS)

Hatano, Mamoru.

1981-01-01

Purpose: To readily discharge a nuclear fuel by burning the nuclear fuel as it is without a pulverizing step and removing the graphite and other coated fuel particles. Constitution: An oxygen supply pipe is connected to the lower portion of a discharge chamber having an inlet for the fuel, and an exhaust pipe is connected to the upper portion of the chamber. The fuel mounted on a metallic gripping member made of metallic material is inserted from the inlet, the gripping member is connected through a conductor to a voltage supply unit, oxygen is then supplied through the oxygen supply tube to the discharge chamber, the voltage supply unit is subsequently operated, and discharge takes place among the fuels. Thus, high heat is generated by the discharge, the graphite carbon of the fuel is burnt, silicon carbide is destroyed and decomposed, the isolated nuclear fuel particles are discharged from the exhaust port, and the combustion gas and small embers are exhausted from the exhaust tube. Accordingly, radioactive dusts are not so much generated as when using a mechanical pulverizing means, and prescribed objective can be achieved. (Yoshino, Y.)

Synthesis, characterization and thermoelectric properties of metal borides, boron carbides and carbaborides; Synthese, Charakterisierung und thermoelektrische Eigenschaften ausgewaehlter Metallboride, Borcarbide und Carbaboride

Energy Technology Data Exchange (ETDEWEB)

Guersoy, Murat

2015-07-06

This work reports on the solid state synthesis and structural and thermoelectrical characterization of hexaborides (CaB{sub 6}, SrB{sub 6}, BaB{sub 6}, EuB{sub 6}), diboride dicarbides (CeB{sub 2}C{sub 2}, LaB{sub 2}C{sub 2}), a carbaboride (NaB{sub 5}C) and composites of boron carbide. The characterization was performed by X-ray diffraction methods and Rietveld refinements based on structure models from literature. Most of the compounds were densified by spark plasma sintering at 100 MPa. As high-temperature thermoelectric properties the Seebeck coefficients, electrical conductivities, thermal diffusivities and heat capacities were measured between room temperature and 1073 K. ZT values as high as 0.5 at 1273 K were obtained for n-type conducting EuB{sub 6}. High-temperature X-ray diffraction also confirmed its thermal stability. The solid solutions Ca{sub x}Sr{sub 1-x}B{sub 6}, Ca{sub x}Ba{sub 1-x}B{sub 6} and Sr{sub x}Ba{sub 1-x}B{sub 6} (x = 0, 0.25, 0.5, 0.75, 1) are also n-type but did not show better ZT values for the ternary compounds compared to the binaries, but for CaB{sub 6} the values of the figure of merit (ca. 0.3 at 1073 K) were significantly increased (ca. 50 %) compared to earlier investigations which is attributed to the densification process. Sodium carbaboride, NaB{sub 5}C, was found to be the first p-type thermoelectric material that crystallizes with the hexaboride-structure type. Seebeck coefficients of ca. 80 μV . K{sup -1} were obtained. Cerium diboride dicarbide, CeB{sub 2}C{sub 2}, and lanthanum diboride dicarbide, LaB{sub 2}C{sub 2}, are metallic. Both compounds were used as model compounds to develop compacting strategies for such layered borides. Densities obtained at 50 MPa were determined to be higher than 90 %. A new synthesis route using single source precursors that contain boron and carbon was developed to open the access to new metal-doped boron carbides. It was possible to obtain boron carbide, but metal-doping could not be

Reaction phases and diffusion paths in SiC/metal systems

Energy Technology Data Exchange (ETDEWEB)

Naka, M.; Fukai, T. [Osaka Univ., Osaka (Japan); Schuster, J.C. [Vienna Univ., Vienna (Austria)

2004-07-01

The interface structures between SiC and metal are reviewed at SiC/metal systems. Metal groups are divided to carbide forming metals and non-carbide forming metals. Carbide forming metals form metal carbide granular or zone at metal side, and metal silicide zone at SiC side. The further diffusion of Si and C from SiC causes the formation of T ternary phase depending metal. Non-carbide forming metals form silicide zone containing graphite or the layered structure of metal silicide and metal silicide containing graphite. The diffusion path between SiC and metal are formed along tie-lines connecting SiC and metal on the corresponding ternary Si-C-M system. The reactivity of metals is dominated by the forming ability of carbide or silicide. Te reactivity tendency of elements are discussed on the periodical table of elements, and Ti among elements shows the highest reactivity among carbide forming metals. For non-carbide forming metals the reactivity sequence of metals is Fe>Ni>Co. (orig.)

Irradiation performance of full-length metallic IFR fuels

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.

1992-07-01

An assembly irradiation of 169 full-length U-Pu-Zr metallic fuel pins was successfully completed in FFTF to a goal burnup of 10 at.%. All test fuel pins maintained their cladding integrity during the irradiation. Postirradiation examination showed minimal fuel/cladding mechanical interaction and excellent stability of the fuel column. Fission-gas release was normal and consistent with the existing data base from irradiation testing of shorter metallic fuel pins in EBR-II

A state of the art on metallic fuel technology development

Energy Technology Data Exchange (ETDEWEB)

Hwang, Woan; Kang, Hee Young; Nam, Cheol; Kim, Jong Oh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

Since worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved in the late 1960`s, the development of metallic fuels continued throughout the 1970`s at ANL`s experimental breeder reactor II (EBR-II) because EBR-II continued to be fueled with the metallic uranium-fissium alloy, U-5Fs. During this decade the performance limitations of metallic fuel were satisfactorily resolved resolved at EBR-II. The concept of the IFR developed at ANL since 1984. The technical feasibility had been demonstrated and the technology database had been established to support its practicality. One key features of the IFR is that the fuel is metallic, which brings pronounced benefits over oxide in improved inherent safety and lower processing costs. At the outset of the 1980`s, it appeared that metallic fuels are recognized as a professed viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last score and summarizes the state-of the art on metallic fuel technology development. (author). 29 refs., 1 tab.

Casting of metallic fuel containing minor actinide additions

International Nuclear Information System (INIS)

Trybus, C.L.; Henslee, S.P.; Sanecki, J.E.

1992-01-01

A significant attribute of the Integral Fast Reactor (IFR) concept is the transmutation of long-lived minor actinide fission products. These isotopes require isolation for thousands of years, and if they could be removed from the waste, disposal problems would be reduced. The IFR utilizes pyroprocessing of metallic fuel to separate auranium, plutonium, and the minor actinides from nonfissionable constituents. These materials are reintroduced into the fuel and reirradiated. Spent IFR fuel is expected to contain low levels of americium, neptunium, and curium because the hard neutron spectrum should transmute these isotopes as they are produced. This opens the possibility of using an IFR to trnasmute minor actinide waste from conventional light water reactors (LWRs). A standard IFR fuel is based on the alloy U-20% Pu-10% Zr (in weight percent). A metallic fuel system eases the requirements for reprocessing methods and enables the minor actinide metals to be incorporated into the fuel with simple modifications to the basic fuel casting process. In this paper, the authors report the initial casting experience with minor actinide element addition to an IFR U-Pu-Zr metallic fuel

Ultra-high-rate pseudocapacitive energy storage in two-dimensional transition metal carbides

Energy Technology Data Exchange (ETDEWEB)

Lukatskaya, Maria R. [Drexel Univ., Philadelphia, PA (United States); Dept. of Chemical Engineering, Stanford, CA (United States); Kota, Sankalp [Drexel Univ., Philadelphia, PA (United States); Lin, Zifeng [Univ. Paul Sabatier, Toulouse (France); Reseau sur le Stockage Electrochimique de l' Energie (RS2E) (France); Zhao, Meng -Qiang [Drexel Univ., Philadelphia, PA (United States); Shpigel, Netanel [Bar-Ilan Univ., Ramat-Gan (Israel); Levi, Mikhael D. [Bar-Ilan Univ., Ramat-Gan (Israel); Halim, Joseph [Drexel Univ., Philadelphia, PA (United States); Taberna, Pierre -Louis [Univ. Paul Sabatier, Toulouse (France); Reseau sur le Stockage Electrochimique de l' Energie (RS2E) (France); Barsoum, Michel W. [Drexel Univ., Philadelphia, PA (United States); Simon, Patrice [Univ. Paul Sabatier, Toulouse (France); Reseau sur le Stockage Electrochimique de l' Energie (RS2E) (France); Gogotsi, Yury G. [Drexel Univ., Philadelphia, PA (United States)

2017-07-10

In this study, the use of fast surface redox storage (pseudocapacitive) mechanisms can enable devices that store much more energy than electrical double-layer capacitors (EDLCs) and, unlike batteries, can do so quite rapidly. Yet, few pseudocapacitive transition metal oxides can provide a high power capability due to their low intrinsic electronic and ionic conductivity. Here we demonstrate that two-dimensional transition metal carbides (MXenes) can operate at rates exceeding those of conventional EDLCs, but still provide higher volumetric and areal capacitance than carbon, electrically conducting polymers or transition metal oxides. We applied two distinct designs for MXene electrode architectures with improved ion accessibility to redox-active sites. A macroporous Ti₃C₂T_x MXene film delivered up to 210 F g^–1 at scan rates of 10 V s^–1, surpassing the best carbon supercapacitors known. In contrast, we show that MXene hydrogels are able to deliver volumetric capacitance of ~1,500 F cm^–3 reaching the previously unmatched volumetric performance of RuO₂.

A Swiss contribution to a secure LMFBR fuel cycle

International Nuclear Information System (INIS)

Nicolet, M.; Bischoff, K.; Hausmann, W.; Stofer, B.

1978-12-01

Since 1967, EIR has been using the sphere-pac fuel concept, which takes advantage of the wet route fabrication of (U,Pu) carbide-microspheres using an internal gelation method, followed by carbothermic reduction of the precipitated metal-oxides. Some of the promises of the wet process are a shorter fabrication route than for pellet manufacture, no dust problems, reduced fire hazard for carbides, and last but not least the improvement of Pu safeguards. The method is particularly suitable for direct coupling to a reprocessing plant, where coprocessing of both U and Pu and spiked solutions will be possible. (Auth.)

Advances in Metallic Fuels for High Burnup and Actinide Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hayes, S. L.; Harp, J. M.; Chichester, H. J. M.; Fielding, R. S.; Mariani, R. D.; Carmack, W. J.

2016-10-01

Research and development activities on metallic fuels in the US are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is a desire to demonstrate a multifold increase in burnup potential. A number of metallic fuel design innovations are under investigation with a view toward significantly increasing the burnup potential of metallic fuels, since higher discharge burnups equate to lower potential actinide losses during recycle. Promising innovations under investigation include: 1) lowering the fuel smeared density in order to accommodate the additional swelling expected as burnups increase, 2) utilizing an annular fuel geometry for better geometrical stability at low smeared densities, as well as the potential to eliminate the need for a sodium bond, and 3) minor alloy additions to immobilize lanthanide fission products inside the metallic fuel matrix and prevent their transport to the cladding resulting in fuel-cladding chemical interaction. This paper presents results from these efforts to advance metallic fuel technology in support of high burnup and actinide transmutation objectives. Highlights include examples of fabrication of low smeared density annular metallic fuels, experiments to identify alloy additions effective in immobilizing lanthanide fission products, and early postirradiation examinations of annular metallic fuels having low smeared densities and palladium additions for fission product immobilization.

Fuel-cladding chemical interaction

International Nuclear Information System (INIS)

Gueneau, C.; Piron, J.P.; Dumas, J.C.; Bouineau, V.; Iglesias, F.C.; Lewis, B.J.

2015-01-01

The chemistry of the nuclear fuel is very complex. Its chemical composition changes with time due to the formation of fission products and depends on the temperature level history within the fuel pellet and the clad during operation. Firstly, in thermal reactors, zircaloy oxidation from reaction with UO 2 fuel under high-temperature conditions will be addressed. Then other fuel-cladding interaction phenomena occurring in fast reactors will be described. Large thermal gradients existing between the centre and the periphery of the pellet induce the radial redistribution of the fuel constituents. The fuel pellet can react with the clad by different corrosion processes which can involve actinide and/or fission product transport via gas, liquid or/and solid phases. All these phenomena are briefly described in the case of different kinds of fuels (oxide, carbide, nitride, metallic) to be used in fast reactors. The way these phenomena are taken into account in fuel performance codes is presented. (authors)

Fabrication and characterization of fully ceramic microencapsulated fuels

Energy Technology Data Exchange (ETDEWEB)

Terrani, K.A., E-mail: kurt.terrani@gmail.com [Fuel Cycle and Isotopes Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kiggans, J.O.; Katoh, Y. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Shimoda, K. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Montgomery, F.C.; Armstrong, B.L.; Parish, C.M. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Hinoki, T. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hunn, J.D. [Fuel Cycle and Isotopes Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Snead, L.L. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

2012-07-15

The current generation of fully ceramic microencapsulated fuels, consisting of Tristructural Isotropic fuel particles embedded in a silicon carbide matrix, is fabricated by hot pressing. Matrix powder feedstock is comprised of alumina-yttria additives thoroughly mixed with silicon carbide nanopowder using polyethyleneimine as a dispersing agent. Fuel compacts are fabricated by hot pressing the powder-fuel particle mixture at a temperature of 1800-1900 Degree-Sign C using compaction pressures of 10-20 MPa. Detailed microstructural characterization of the final fuel compacts shows that oxide additives are limited in extent and are distributed uniformly at silicon carbide grain boundaries, at triple joints between silicon carbide grains, and at the fuel particle-matrix interface.

Microencapsulated fuel technology for commercial light water and advanced reactor application

International Nuclear Information System (INIS)

Terrani, Kurt A.; Snead, Lance L.; Gehin, Jess C.

2012-01-01

The potential application of microencapsulated fuels to light water reactors (LWRs) has been explored. The specific fuel manifestation being put forward is for coated fuel particles embedded in silicon carbide or zirconium metal matrices. Detailed descriptions of these concepts are presented, along with a review of attributes, potential benefits, and issues with respect to their application in LWR environments, specifically from the standpoints of materials, neutronics, operations, and economics. Preliminary experiment and modeling results imply that with marginal redesign, significant gains in operational reliability and accident response margins could be potentially achieved by replacing conventional oxide-type LWR fuel with microencapsulated fuel forms.

Metal fuel manufacturing and irradiation performance

International Nuclear Information System (INIS)

Pedersen, D.R.; Walters, L.C.

1992-01-01

The advances in metal fuel by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, and improved passive safety. The goals and the safety philosophy of the Integral Fast Reactor Program are stressed

FFTF metal fuel pin sodium bond quality verification

International Nuclear Information System (INIS)

Pitner, A.L.; Dittmer, J.O.

1988-12-01

The Fast Flux Test Facility (FFTF) Series III driver fuel design consists of U-10Zr fuel slugs contained in a ferritic alloy cladding. A liquid metal, sodium bond between the fuel and cladding is required to prevent unacceptable temperatures during operation. Excessive voiding or porosity in the sodium thermal bond could result in localized fuel melting during irradiation. It is therefore imperative that bond quality be verified during fabrication of these metal fuel pins prior to irradiation. This document discusses this verification

The oxidative corrosion of carbide inclusions at the surface of uranium metal during exposure to water vapour.

Science.gov (United States)

Scott, T B; Petherbridge, J R; Harker, N J; Ball, R J; Heard, P J; Glascott, J; Allen, G C

2011-11-15

The reaction between uranium and water vapour has been well investigated, however discrepancies exist between the described kinetic laws, pressure dependence of the reaction rate constant and activation energies. Here this problem is looked at by examining the influence of impurities in the form of carbide inclusions on the reaction. Samples of uranium containing 600 ppm carbon were analysed during and after exposure to water vapour at 19 mbar pressure, in an environmental scanning electron microscope (ESEM) system. After water exposure, samples were analysed using secondary ion mass spectrometry (SIMS), focused ion beam (FIB) imaging and sectioning and transmission electron microscopy (TEM) with X-ray diffraction (micro-XRD). The results of the current study indicate that carbide particles on the surface of uranium readily react with water vapour to form voluminous UO(3) · xH(2)O growths at rates significantly faster than that of the metal. The observation may also have implications for previous experimental studies of uranium-water interactions, where the presence of differing levels of undetected carbide may partly account for the discrepancies observed between datasets. Crown Copyright © 2011. Published by Elsevier B.V. All rights reserved.

Overview of chemical characterization of FBTR fuel

International Nuclear Information System (INIS)

Venkatesan, V.; Nandi, C.; Patil, A.B.; Prakash, Amrit; Khan, K.B.; Arun Kumar

2015-01-01

Uranium Plutonium mixed carbide fuel is the driver fuel for Fast Breeder Test Reactor (FBTR) at IGCAR. The fuel is being fabricated at Radiometallurgy Division, BARC by conventional powder metallurgy route. During the fabrication of fuel, chemical quality control of process intermediates is very important to reach stringent specification of the final fuel product. Different steps are involved in the fabrication of uranium-plutonium carbide (MC) for FBTR. The main steps in the fabrication of MC fuel pellets are carbothermic reduction (CR) of mixture of uranium oxide, plutonium oxide and graphite powder to prepare MC clinkers, crushing and milling of MC clinkers and consolidation of MC powders into fuel pellets and sintering. As a part of process control, analysis of uranium (U), plutonium (Pu), carbon in oxide graphite mixture and U, Pu, carbon, oxygen, nitrogen, MC, M 2 C 3 contents in mixed carbide powder (MC clinkers) are carried out at our laboratory. Analysis of U, Pu, carbon, oxygen, nitrogen, MC and M 2 C 3 contents in mixed carbide sintered pellets are carried out as a part of quality control. This paper describes an overview of analytical instruments used during chemical quality control of mixed carbide fuel

Microprobe study of fission product behavior in high-burnup HTR fuels

International Nuclear Information System (INIS)

Kleykamp, H.

Electron microprobe analysis of irradiated coated particles with high burnup (greater than 50 percent fima) gives detailed information on the chemical state and the transport behavior of the fission products in UO 2 and UC 2 kernels and in the coatings. In oxide fuel kernels, metallic inclusions and ceramic precipitations are observed. The solubility behavior of the fission products in the fuel matrix has been investigated. Fission product inclusions could not be detected in carbide fuel kernels; post irradiation annealed UC 2 kernels, however, give information on the element combinations of some fission product phases. Corresponding to the chemical state in the kernel, Cs, Sr, Ba, Pd, Te and the rare earths are released easily and diffuse through the entire pyrocarbon coating. These fission products can be retained by a silicon carbide layer. The initial stage of a corrosive attack of the SiC coating by the fission products is evidenced

Neutronics performances study of silicon carbide as an inert matrix to achieve very high burn-up for light water reactor fuels

International Nuclear Information System (INIS)

Chabert, C.; Coulon-Picard, E.; Pelletier, M.

2007-01-01

In order to extend the actual limits of light water reactors, the Cea has put emphasis on the exploration of major fuel innovations that would allow us to increase the competitiveness, the safety and flexibility, while keeping the standard PWR environment. Different fuel concepts have been chosen and are actually studied to evaluate their advantages and drawbacks. The objectives of these new fuels are to increase the safety performances and to achieve a very high burn-up. One concept is a CERCER fuel with silicon carbide (SiC) as an inert matrix devoted to reduce the fuel temperature at nominal conditions. Besides the investigation of the neutronic performance, analyses on the thermomechanical performances, the fuel fabrication, the fuel reprocessing and economic aspects have been performed. This paper presents particularly neutronic results obtained for the CERCER fuel. The results show that a very high burn-up, a high safety performance and a better competitiveness cannot be achieved with this fuel concept. (authors)

A new electrothermal-chemical method for metals, carbides, and ceramics hard coating: Experiment and theory

Energy Technology Data Exchange (ETDEWEB)

Zoler, D.; Bruma, C.; Cuperman, S.

1999-07-01

A new method and an experimental device for powders of metals, carbides and ceramics coating of various substrates are presented. The powder-particles are accelerated and heated by a mixture of plasma and gases resulted from the burning of an energetic (propellant). The operating prototype already allows one to obtain coatings of metals, carbides and ceramics. Some of the coatings obtained, especially those by carbides powders, indicate even at the present stage of research, properties (as hardness, porosity) which are comparable to those provided by the presently industrial methods in use. The accelerating-heating agent in the device (the plasma-gas mixture) is characterized by very high densities (up to 120 kg/m{sup 3}), temperatures (up to 20,000 K) and velocities (more than 1,500 m/s). Due to these characteristics, the powder particles are accelerated to velocities significantly higher than those reached in other coating devices as, for example, the detonation (D) gun. Some preliminary experimental data show that the accelerated particle can reach velocities higher than 1,000 m/s. In parallel, in order to better understand the phenomena taking place inside the device and to determine the optimal process parameters leading to high quality coatings an appropriate theoretical model was developed. The model is able to describe the complex processes of plasma-gas-propellant interaction, gas flow and powder particles heating and acceleration. The model gives a detailed description of the gas, propellant and accelerated particle parameters, their spatial distribution and temporal evolution; predicts their dependence on the values of some input quantities such as: the plasma energy, propellant characteristics and accelerated particles type and geometry. The computational results the authors obtained show that, indeed, during the acceleration process the particles are heated, melted and eventually vaporized. One of the most interesting theoretical results is that the

Studies of Lanthanide Transport in Metallic Fuel

Energy Technology Data Exchange (ETDEWEB)

Zhang, Jinsuo; Taylor, Christopher

2018-04-02

Metallic nuclear fuels were tested in fast reactor programs and performed well. However, metallic fuels have shown the phenomenon of FCCI that are due to deleterious reactions between lanthanide fission products and cladding material. As the burnup is increased, lanthanide fission products that contact with the cladding could react with cladding constituents such as iron and chrome. These reactions produce higher-melting intermetallic compounds and low-melting alloys, and weaken the mechanical integrity. The lanthanide interaction with clad in metallic fuels is recognized as a long-term, high-burnup cause of the clad failures. Therefore, one of the key concerns of using metallic fuels is the redistribution of lanthanide fission products and migration to the fuel surface. It is believed that lanthanide migration is in part due to the thermal gradient between the center and the fuel-cladding interface, but also largely in part due to the low solubility of lanthanides within the uranium-based metal fuel. PIE of EBR-II fuels shows that lanthanides precipitate directly and do not dissolve to an appreciable extent in the fuel matrix. Based on the PIE data from EBR-II, a recent study recommended a so-called “liquid-like” transport mechanism for lanthanides and certain other species. The liquid-like transport model readily accounts for redistribution of Ln, noble metal fission products, and cladding components in the fuel matrix. According to the novel mechanism, fission products can transport as solutes in liquid metals, such as liquid cesium or liquid cesium–sodium, and on pore surfaces and fracture surfaces for metals near their melting temperatures. Transport in such solutions is expected to be much more rapid than solid-state diffusion. The mechanism could explain the Ln migration to the fuel slug peripheral surface and their deposition with a sludge-like form. Lanthanides have high solubility in liquid cesium but have low solubility in liquid sodium. As a

A review of carbide fuel corrosion for nuclear thermal propulsion applications

Energy Technology Data Exchange (ETDEWEB)

Pelaccio, D.G.; El-Genk, M.S. [Univ. of New Mexico, Albuquerque, NM (United States). Inst. for Space Nuclear Power Studies; Butt, D.P. [Los Alamos National Lab., NM (United States)

1993-12-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico`s Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

A Review of Carbide Fuel Corrosion for Nuclear Thermal Propulsion Applications

Science.gov (United States)

Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

1994-07-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

Design report for an annular fuel element for accommodation of a carbide test bundle on the ring position of the KNK II/2 test zone

International Nuclear Information System (INIS)

Haefner, H.E.

1982-03-01

This report describes an annular oxide element with Mark II rods for accommodation of a 19-pin carbide test bundle on position 201 in the test zone of the second core of KNK II as well as its behavior during the period of operation. The ring element comprises within a driver wrapper in three rows of pins 102 fuel pins of 7.6 mm diameter and six structural rods for fixing the spark eroded spacers. The report deals with the ring element with its individual components fuel rod, bundle, wrappers, head and foot and describes methods, criteria and results concerning the design. The carbide test bundle to be accommodated by the annular carrier element will be treated in a separate report. The loadability of the annular element with its components is demonstrated by generally valid standards for strength criteria

Weld Joint Design for SFR Metallic Fuel Element Closures

Energy Technology Data Exchange (ETDEWEB)

Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Kim, Ki Hwan; Yoon, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The sodium-cooled fast reactor (SFR) system is among the six systems selected for Gen-IV promising systems and expected to become available for commercial introduction around 2030. In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the joint designs for endplug welding were investigated. For the irradiation test of SFR metallic fuel element, the TIG welding technique was adopted and the welding joint design was developed based on the welding conditions and parameters established. In order to make SFR metallic fuel elements, the weld joint design was developed based on the TIG welding technique.

Endplug Welding Techniques developed for SFR Metallic Fuel Elements

Energy Technology Data Exchange (ETDEWEB)

Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established.

Endplug Welding Techniques developed for SFR Metallic Fuel Elements

International Nuclear Information System (INIS)

Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan

2013-01-01

In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established

Atom-vacancy ordering and magnetic susceptibility of nonstoichiometric hafnium carbide

International Nuclear Information System (INIS)

Gusev, A.I.; Zyryanova, A.N.

1999-01-01

Experimental results on magnetic susceptibility of nonstoichiometric hafnium carbide HfC y (0.6 0.71 , HfC 0.78 and HfC 0.83 in the range of 870-930 K the anomalies are revealed which are associated with superstructure short-range ordering in a non-metallics sublattice. It is shown that a short-range order in HfC 0.71 and HfC 0.78 carbides corresponds to Hf 3 C 2 ordered phase, and in HfC 0.83 carbide - to Hf 6 C 5 ordered phase. HfC 0.78 carbide is found to possesses zero magnetic susceptibility in temperature range 910-980 K [ru

Hardness of carbides, nitrides, and borides

International Nuclear Information System (INIS)

Schroeter, W.

1981-01-01

Intermetallic compounds of metals with non-metals such as C, N, and B show different hardness. Wagner's interaction parameter characterizes manner and extent of the interaction between the atoms of the substance dissolved and the additional elements in metallic mixed phases. An attempt has been made to correlate the hardness of carbides, nitrides, and borides (data taken from literature) with certain interaction parameters and associated thermodynamic quantities (ΔH, ΔG). For some metals of periods 4, 5, and 6 corresponding relations were found between microhardness, interaction parameters, heat of formation, and atomic number

Advanced methods for fabrication of PHWR and LMFBR fuels

International Nuclear Information System (INIS)

Ganguly, C.

1988-01-01

For self-reliance in nuclear power, the Department of Atomic Energy (DAE), India is pursuing two specific reactor systems, namely the pressurised heavy water reactors (PHWR) and the liquid metal cooled fast breeder reactors (LMFBR). The reference fuel for PHWR is zircaloy-4 clad high density (≤ 96 per cent T.D.) natural UO 2 pellet-pins. The advanced PHWR fuels are UO 2 -PuO 2 (≤ 2 per cent), ThO 2 -PuO 2 (≤ 4 per cent) and ThO 2 -U 233 O 2 (≤ 2 per cent). Similarly, low density (≤ 85 per cent T.D.) (UPu)O 2 pellets clad in SS 316 or D9 is the reference fuel for the first generation of prototype and commercial LMFBRs all over the world. However, (UPu)C and (UPu)N are considered as advanced fuels for LMFBRs mainly because of their shorter doubling time. The conventional method of fabrication of both high and low density oxide, carbide and nitride fuel pellets starting from UO 2 , PuO 2 and ThO 2 powders is 'powder metallurgy (P/M)'. The P/M route has, however, the disadvantage of generation and handling of fine powder particles of the fuel and the associated problem of 'radiotoxic dust hazard'. The present paper summarises the state-of-the-art of advanced methods of fabrication of oxide, carbide and nitride fuels and highlights the author's experience on sol-gel-microsphere-pelletisation (SGMP) route for preparation of these materials. The SGMP process uses sol gel derived, dust-free and free-flowing microspheres of oxides, carbide or nitride for direct pelletisation and sintering. Fuel pellets of both low and high density, excellent microhomogeneity and controlled 'open' or 'closed' porosity could be fabricated via the SGMP route. (author). 5 tables, 14 figs., 15 refs

Natural precursor based hydrothermal synthesis of sodium carbide for reactor applications

Science.gov (United States)

Swapna, M. S.; Saritha Devi, H. V.; Sebastian, Riya; Ambadas, G.; Sankararaman, S.

2017-12-01

Carbides are a class of materials with high mechanical strength and refractory nature which finds a wide range of applications in industries and nuclear reactors. The existing synthesis methods of all types of carbides have problems in terms of use of toxic chemical precursors, high-cost, etc. Sodium carbide (Na2C2) which is an alkali metal carbide is the least explored one and also that there is no report of low-cost and low-temperature synthesis of sodium carbide using the eco-friendly, easily available natural precursors. In the present work, we report a simple low-cost, non-toxic hydrothermal synthesis of refractory sodium carbide using the natural precursor—Pandanus. The formation of sodium carbide along with boron carbide is evidenced by the structural and morphological characterizations. The sample thus synthesized is subjected to field emission scanning electron microscopy (FESEM), x-ray powder diffraction (XRD), ultraviolet (UV)—visible spectroscopy, Fourier transform infrared spectroscopy (FTIR), Raman, and photoluminescent (PL) spectroscopic techniques.

Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

Energy Technology Data Exchange (ETDEWEB)

Yacout, A. M. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

2016-09-16

The US sodium cooled fast reactor (SFR) metallic fuel performance data that are of interest to advanced fast reactors applications, can be attributed mostly to the Integral Fast Reactor (IFR) program between 1984 and 1994. Metallic fuel data collected prior to the IFR program were associated with types of fuel that are not of interest to future advanced reactors deployment (e.g., previous U-Fissium alloy fuel). The IFR fuels data were collected from irradiation of U-Zr based fuel alloy, with and without Pu additions, and clad in different types of steels, including HT9, D9, and 316 stainless-steel. Different types of data were generated during the program, and were based on the requirements associated with the DOE Advanced Liquid Metal Cooled Reactor (ALMR) program.

Advanced breeder cycle uses metallic fuel

International Nuclear Information System (INIS)

Chang, Y.I.

1991-01-01

Scientists from Argonne National Laboratory have been developing a concept called the Integral fast Reactor (IFR). This fast breeder reactor could effectively increase Uranium resources a hundred fold making nuclear power essentially an inexhaustible energy source. The IFR is outlined. In the IFR, the inherent properties of liquid metal cooling are combined with a new metallic fuel which is allowed to swell and gives an improved burnup level and a radically different refining process to allow breakthroughs in passive safety, fuel cycle economics and waste management. (author)

Metal-deactivating additives for liquid fuels

Energy Technology Data Exchange (ETDEWEB)

Boneva, M.I. [Institute of Organic Chemistry, Sofia (Bulgaria); Ivanov, S.K.; Kalitchin, Z.D. [SciBulCom, Ltd., Sofia (Bulgaria); Tanielyan, S.K. [Seton Hall Univ., South Orange, NJ (United States); Terebenina, A.; Todorova, O.I. [Institute of Inorganic Chemistry, Sofia (Bulgaria)

1995-05-01

The metal-deactivating and the antioxidant properties of 1-phenyl-3-methylpyrazolone-5 derivatives have been investigated both in the model reaction of low temperature oxidation of ethylbenzene and in gasoline oxidation. The study of the ability of these derivatives to reduce the catalytic effect of copper naphthenate demonstrates that they are promising as metal deactivating additives for light fuels. Some of the pyrazolone compounds appear to be of special interest for the long-term storage of liquid fuels due to their action as multifunctional inhibitors.

Iron Carbides in Fischer–Tropsch Synthesis: Theoretical and Experimental Understanding in Epsilon-Iron Carbide Phase Assignment

International Nuclear Information System (INIS)

Liu, Xing-Wu; Cao, Zhi; Zhao, Shu; Gao, Rui

2017-01-01

As active phases in low-temperature Fischer–Tropsch synthesis for liquid fuel production, epsilon iron carbides are critically important industrial materials. However, the precise atomic structure of epsilon iron carbides remains unclear, leading to a half-century of debate on the phase assignment of the ε-Fe 2 C and ε’-Fe 2.2 C. Here, we resolve this decades-long question by a combining theoretical and experimental investigation to assign the phases unambiguously. First, we have investigated the equilibrium structures and thermal stabilities of ε-Fe x C, (x = 1, 2, 2.2, 3, 4, 6, 8) by first-principles calculations. We have also acquired X-ray diffraction patterns and Mössbauer spectra for these epsilon iron carbides, and compared them with the simulated results. These analyses indicate that the unit cell of ε-Fe 2 C contains only one type of chemical environment for Fe atoms, while ε’-Fe 2.2 C has six sets of chemically distinct Fe atoms.

Fabrication of metallic fuel for fast breeder reactor

International Nuclear Information System (INIS)

Saify, M.T.; Jha, S.K.; Abdulla, K.K.; Kumar, Arbind; Mittal, R.K.; Prasad, R.S.; Mahule, N.; Kumar, Arun; Prasad, G.J.

2012-01-01

Natural uranium oxide fuelled PHWRs comprises of first stage of Indian nuclear power programme. Liquid metal fast breeder reactors fuelled by Pu (from PHWR's) form the second stage. A shorter reactor doubling time is essential in order to accelerate the nuclear power growth in India. Metallic fuels are known to provide shorter doubling times, necessitating to be used as driver fuel for fast breeder reactors. One of the fabrication routes for metallic fuels having random grain orientation, is injection casting technique. The technique finds its basis in an elementary physical concept - the possibility of supporting a liquid column within a tube, by the application of a pressure difference across the liquid interface inside and outside the tube. At AFD, BARC a facility has been set-up for injection casting of uranium rods in quartz tube moulds, demoulding of cast rods, end-shearing of rods and an automated inspection system for inspection of fuel rods with respect to mass, length, diameter and diameter variation along the length and internal and external porosities/voids. All the above facilities have been set-up in glove boxes and have successfully been used for fabrication of uranium bearing fuel rods. The facility has been designed for fabrication and inspection of Pu-bearing metallic fuels also, if required

Tantalum and niobium carbides obtention by carbothermic reduction of columbotantalite ores

International Nuclear Information System (INIS)

Gordo, E.; Garcia-Carcedo, F.; Torralba, J.M.

1998-01-01

Tantalum and niobium carbides are characterized by its high hardness and chemical corrosion resistance. Both carbides, but mainly TaC, are used in hard metals (sintered carbides), together with their carbides, to manufacture cutting tools and dies in special machining applications involving mechanical shock at high temperature. Its use as reinforcement of wear resistant materials through powder metallurgy techniques are being investigated. However, the use of TaC is usually limited because of its high cost. Therefore tantalum carbide with niobium content, which is cheaper, is used. In this work the obtention of complex tantalum and niobium carbides from a Spanish columbotantalite ore is studied through relatively cheap and simple process as it is carbothermic reduction. Concentration of the ore, its reduction and the characterization of products are described. (Author) 11 refs

Formation mechanism of spheroidal carbide in ultra-low carbon ductile cast iron

Directory of Open Access Journals (Sweden)

Bin-guo Fu

2016-09-01

Full Text Available The formation mechanism of the spheroidal carbide in the ultra-low carbon ductile cast iron fabricated by the metal mold casting technique was systematically investigated. The results demonstrated that the spheroidal carbide belonged to eutectic carbide and crystallized in the isolated eutectic liquid phase area. The formation process of the spheroidal carbide was related to the contact and the intersection between the primary dendrite and the secondary dendrite of austenite. The oxides of magnesium, rare earths and other elements can act as heterogeneous nucleation sites for the spheroidal carbide. It was also found that the amount of the spheroidal carbide would increase with an increase in carbon content. The cooling rate has an important influence on the spheroidal carbide under the same chemical composition condition.

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

Hardened over-coating fuel particle and manufacture of nuclear fuel using its fuel particle

International Nuclear Information System (INIS)

Yoshimuda, Hideharu.

1990-01-01

Coated-fuel particles comprise a coating layer formed by coating ceramics such as silicon carbide or zirconium carbide and carbons, etc. to a fuel core made of nuclear fuel materials. The fuel core generally includes oxide particles such as uranium, thorium and plutonium, having 400 to 600 μm of average grain size. The average grain size of the coated-fuel particle is usually from 800 to 900 μm. The thickness of the coating layer is usually from 150 to 250 μm. Matrix material comprising a powdery graphite and a thermosetting resin such as phenol resin, etc. is overcoated to the surface of the coated-fuel particle and hardened under heating to form a hardened overcoating layer to the coated-fuel particle. If such coated-fuel particles are used, cracks, etc. are less caused to the coating layer of the coated-fuel particles upon production, thereby enabling to prevent the damages to the coating layer. (T.M.)

A general salt-templating method to fabricate vertically aligned graphitic carbon nanosheets and their metal carbide hybrids for superior lithium ion batteries and water splitting.

Science.gov (United States)

Zhu, Jixin; Sakaushi, Ken; Clavel, Guylhaine; Shalom, Menny; Antonietti, Markus; Fellinger, Tim-Patrick

2015-04-29

The synthesis of vertically aligned functional graphitic carbon nanosheets (CNS) is challenging. Herein, we demonstrate a general approach for the fabrication of vertically aligned CNS and metal carbide@CNS composites via a facile salt templating induced self-assembly. The resulting vertically aligned CNS and metal carbide@CNS structures possess ultrathin walls, good electrical conductivity, strong adhesion, excellent structural robustness, and small particle size. In electrochemical energy conversion and storage such unique features are favorable for providing efficient mass transport as well as a large and accessible electroactive surface. The materials were tested as electrodes in a lithium ion battery and in electrochemical water splitting. The vertically aligned nanosheets exhibit remarkable lithium ion storage properties and, concurrently, excellent properties as electrocatalysts for hydrogen evolution.

The use of metal hydrides in fuel cell applications

Directory of Open Access Journals (Sweden)

Mykhaylo V. Lototskyy

2017-02-01

Full Text Available This paper reviews state-of-the-art developments in hydrogen energy systems which integrate fuel cells with metal hydride-based hydrogen storage. The 187 reference papers included in this review provide an overview of all major publications in the field, as well as recent work by several of the authors of the review. The review contains four parts. The first part gives an overview of the existing types of fuel cells and outlines the potential of using metal hydride stores as a source of hydrogen fuel. The second part of the review considers the suitability and optimisation of different metal hydrides based on their energy efficient thermal integration with fuel cells. The performances of metal hydrides are considered from the viewpoint of the reversible heat driven interaction of the metal hydrides with gaseous H2. Efficiencies of hydrogen and heat exchange in hydrogen stores to control H2 charge/discharge flow rates are the focus of the third section of the review and are considered together with metal hydride – fuel cell system integration issues and the corresponding engineering solutions. Finally, the last section of the review describes specific hydrogen-fuelled systems presented in the available reference data.

Postirradiation results and evaluation of helium-bonded uranium--plutonium carbide fuel elements irradiated in EBR-II. Interim report

International Nuclear Information System (INIS)

Latimer, T.W.; Barner, J.O.; Kerrisk, J.F.; Green, J.L.

1976-02-01

An evaluation was made of the performance of 74 helium-bonded uranium-plutonium carbide fuel elements that were irradiated in EBR-II at 38-96 kW/m to 2-12 at. percent burnup. Only 38 of these elements have completed postirradiation examination. The higher failure rate found in fuel elements which contained high-density (greater than 95 percent theoretical density) fuel than those which contained low-density (77-91 percent theoretical density) fuel was attributed to the limited ability of the high-density fuel to swell into the void space provided in the fuel element. Increasing cladding thickness and original fuel-cladding gap size were both found to influence the failure rates for elements containing low-density fuel. Lower cladding strain and higher fission-gas release were found in high-burnup fuel elements having smear densities of less than 81 percent. Fission-gas release was usually less than 5 percent for high-density fuel, but increased with burnup to a maximum of 37 percent in low-density fuel. Maximum carburization in elements attaining 5-10 at. percent burnup and clad in Types 304 or 316 stainless steel and Incoloy 800 ranged from 36-80 μm and 38-52 μm, respectively. Strontium and barium were the fission products most frequently found in contact with the cladding but no penetration of the cladding by uranium, plutonium, or fission products was observed

Plastic deformation of particles of zirconium and titanium carbide subjected to vibration grinding

Energy Technology Data Exchange (ETDEWEB)

Kravchik, A.E.; Neshpor, V.S.; Savel' ev, G.A.; Ordan' yan, S.S.

1976-12-01

A study is made of the influence of stoichiometry on the characteristics of microplastic deformation in powders of zirconium and titanium carbide subjected to vibration grinding. The carbide powders were produced by direct synthesis from the pure materials: metallic titanium and zirconium and acetylene black. As to the nature of their elastic deformation, zirconium and titanium carbides can be considered elastic-isotropic materials. During vibration grinding, the primary fracture planes are the (110) planes. Carbides of nonstoichiometric composition are more brittle.

Analysis of metallic fuel pin behaviors under transient conditions of liquid metal reactors

International Nuclear Information System (INIS)

Nam, Cheol; Kwon, Hyoung Mun; Hwang, Woan

1999-02-01

Transient behavior of metallic fuel pins in liquid metal reactor is quite different to that in steady state conditions. Even in transient conditions, the fuel may behave differently depending on its accident situation and/or accident sequence. This report describes and identifies the possible and hypothetical transient events at the aspects of fuel pin behavior. Furthermore, the transient experiments on HT9 clad metallic fuel have been analyzed, and then failure assessments are performed based on accident classes. As a result, the failure mechanism of coolant-related accidents, such as LOF, is mainly due to plenum pressure and cladding thinning caused by eutectic penetration. In the reactivity-related accidents, such as TOP, the reason to cladding failure is believed to be the fuel swelling as well as plenum pressure. The probabilistic Weibull analysis is performed to evaluate the failure behavior of HT9 clad-metallic fuel pin on coolant related accidents.The Weibull failure function is derived as a function of cladding CDF. Using the function, a sample calculation for the ULOF accident of EBR-II fuel is performed, and the results indicate that failure probability is less the 0.3%. Further discussion on failure criteria of accident condition is provided. Finally, it is introduced the state-of-arts for developing computer codes of reactivity-related fuel pin behavior. The development efforts for a simple model to predict transient fuel swelling is described, and the preliminary calculation results compared to hot pressing test results in literature.This model is currently under development, and it is recommended in the future that the transient swelling model will be combined with the cladding model and the additional development for post-failure behavior of fuel pin is required. (Author). 36 refs., 9 tabs., 18 figs

The valve effect of the carbide interlayer of an electric resistance plug

International Nuclear Information System (INIS)

Lakomskii, V.

1998-01-01

The welded electric resistance plug (ERP) usually contains a carbide interlayer at the plug-carbon material interface. The interlayer forms during welding the contact metallic alloy with the carbon material when the oxide films of the alloy are reduced on the interface surface by carbon to the formation of carbides and the surface layer of the plug material dissolves carbon to saturation. Subsequently, during solidification of the plug material it forms carbides with the alloy components. The structural composition of the carbide interlayer is determined by the chemical composition of the contact alloy. In alloys developed by the author and his colleagues the carbide forming elements are represented in most cases by silicon and titanium and, less frequently, by chromium and manganese. Therefore, the carbide interlayers in the ERP consisted mainly of silicon and titanium carbides

High performance nuclear fuel element

International Nuclear Information System (INIS)

Mordarski, W.J.; Zegler, S.T.

1980-01-01

A fuel-pellet composition is disclosed for use in fast breeder reactors. Uranium carbide particles are mixed with a powder of uraniumplutonium carbides having a stable microstructure. The resulting mixture is formed into fuel pellets. The pellets thus produced exhibit a relatively low propensity to swell while maintaining a high density

Fabrication of particulate metal fuel for fast burner reactors

International Nuclear Information System (INIS)

Ryu, Ho Jin; Lee, Sun Yong; Kim, Jong Hwan; Woo, Yoon Myung; Ko, Young Mo; Kim, Ki Hwan; Park, Jong Man; Lee, Chan Bok

2012-01-01

U Zr metallic fuel for sodium cooled fast reactors is now being developed by KAERI as a national R and D program of Korea. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. Therefore, innovative fuel concepts should be developed to address the fabrication challenges pertaining to TRU while maintaining good performances of metallic fuel. Particulate fuel concepts have already been proposed and tested at several experimental fast reactor systems and vipac ceramic fuel of RIAR, Russia is one of the examples. However, much less work has been reported for particulate metallic fuel development. Spherical uranium alloy particles with various diameters can be easily produced by the centrifugal atomization technique developed by KAERI. Using the atomized uranium and uranium zirconium alloy particles, we fabricated various kinds of powder pack, powder compacts and sintered pellets. The microstructures and properties of the powder pack and pellets are presented

Consequences of pipe ruptures in metal fueled, liquid metal cooled reactors

International Nuclear Information System (INIS)

Dunn, F.E.

1990-01-01

The capability to simulate pipe ruptures has recently been added to the SASSYS-1 LMR systems analysis code. Using this capability, the consequences of severe pipe ruptures in both loop-type and pool-type reactors using metal fuel were investigated. With metal fuel, if the control rods scram then either type of reactor can easily survive a complete double-ended break of a single pipe; although, as might be expected, the consequences are less severe for a pool-type reactor. A pool-type reactor can even survive a protected simultaneous breaking of all of its inlet pipes without boiling of the coolant or melting of the fuel or cladding. 2 refs., 16 figs., 1 tab

High surface area synthesis, electrochemical activity, and stability of tungsten carbide supported Pt during oxygen reduction in proton exchange membrane fuel cells

Science.gov (United States)

Chhina, H.; Campbell, S.; Kesler, O.

The oxidation of carbon catalyst supports to carbon dioxide gas leads to degradation in catalyst performance over time in proton exchange membrane fuel cells (PEMFCs). The electrochemical stability of Pt supported on tungsten carbide has been evaluated on a carbon-based gas diffusion layer (GDL) at 80 °C and compared to that of HiSpec 4000™ Pt/Vulcan XC-72R in 0.5 M H 2SO 4. Due to other electrochemical processes occurring on the GDL, detailed studies were also performed on a gold mesh substrate. The oxygen reduction reaction (ORR) activity was measured both before and after accelerated oxidation cycles between +0.6 V and +1.8 V vs. RHE. Tafel plots show that the ORR activity remained high even after accelerated oxidation tests for Pt/tungsten carbide, while the ORR activity was extremely poor after accelerated oxidation tests for HiSpec 4000™. In order to make high surface area tungsten carbide, three synthesis routes were investigated. Magnetron sputtering of tungsten on carbon was found to be the most promising route, but needs further optimization.

High surface area synthesis, electrochemical activity, and stability of tungsten carbide supported Pt during oxygen reduction in proton exchange membrane fuel cells

Energy Technology Data Exchange (ETDEWEB)

Chhina, H. [Automotive fuel cell corporation, 9000 Glenlyon Parkway, Burnaby, BC (Canada); Department of Mechanical and Industrial Engineering, 5 King' s College Road, University of Toronto, Toronto, Ontario (Canada); Campbell, S. [Automotive fuel cell corporation, 9000 Glenlyon Parkway, Burnaby, BC (Canada); Kesler, O. [Department of Mechanical and Industrial Engineering, 5 King' s College Road, University of Toronto, Toronto, Ontario (Canada)

2008-04-15

The oxidation of carbon catalyst supports to carbon dioxide gas leads to degradation in catalyst performance over time in proton exchange membrane fuel cells (PEMFCs). The electrochemical stability of Pt supported on tungsten carbide has been evaluated on a carbon-based gas diffusion layer (GDL) at 80 C and compared to that of HiSpec 4000 trademark Pt/Vulcan XC-72R in 0.5 M H{sub 2}SO{sub 4}. Due to other electrochemical processes occurring on the GDL, detailed studies were also performed on a gold mesh substrate. The oxygen reduction reaction (ORR) activity was measured both before and after accelerated oxidation cycles between +0.6 V and +1.8 V vs. RHE. Tafel plots show that the ORR activity remained high even after accelerated oxidation tests for Pt/tungsten carbide, while the ORR activity was extremely poor after accelerated oxidation tests for HiSpec 4000 trademark. In order to make high surface area tungsten carbide, three synthesis routes were investigated. Magnetron sputtering of tungsten on carbon was found to be the most promising route, but needs further optimization. (author)

Development of metallic fuel fabrication - A study on the interdiffusion behavior between ternary metallic fuel and cladding materials

Energy Technology Data Exchange (ETDEWEB)

Lee, Byung Soo; Seol, Kyung Won; Shon, In Jin [Chonbuk National University, Chonju (Korea)

1999-04-01

To study a new ternary metallic fuel for liquid metal reactor, various U-Zr-X alloys have been made by induction melting. The specimens were prepared for thermal stability tests at 630 deg. C upto 5000 hours in order to estimate the decomposition of the lamellar structure. Interdiffusion studies were carried out at 700 deg. C for 200 hours for the diffusion couples assembled with U-Zr-X ternary fuel versus austenitic stainless steel D9 and martensitic stainless steel HT9, respectively, to investigate the fuel-cladding compatibility. The ternary alloy, especially U-Zr-Mo and U-Zr-Nb alloys showed relatively good thermal stability as long as 5000hrs at 630 deg. C. From the composition profiles of the interdiffusion study, Fe penetrated deeper to the fuel side than other cladding elements such as Ni and Cr, whereas U did to the cladding side of fuel elements in the fuel/D9 couples. On the contrary, the reaction layers of Fuel/HT9 couple were thinner than that of Fuel/D9 couples and were less affected by cladding element, which was believed to be due to Zr rich layer between the fuel-cladding interface. HT9 is considered to be superior to D9 and a favorable choice as a cladding material in terms of fuel-cladding compatibility. 21 refs., 24 figs., 7 tabs. (Author)

Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

Energy Technology Data Exchange (ETDEWEB)

William J. Carmack

2012-05-01

Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in

Pilot production of 325 kg of uranium carbide

International Nuclear Information System (INIS)

Clozet, C.; Dessus, J.; Devillard, J.; Guibert, M.; Morlot, G.

1969-01-01

This report describes the pilot fabrication of uranium carbide rods to be mounted in bundles and assayed in two channels of the EL 4 reactor. The fabrication process includes: - elaboration of uranium carbide granules by carbothermic reduction of uranium dioxide; - electron bombardment melting and continuous casting of the granules; - machining of the raw ingots into rods of the required dimensions; finally, the rods will be piled-up to make the fuel elements. Both qualitative and quantitative results of this pilot production chain are presented and discussed. (authors) [fr

Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel

International Nuclear Information System (INIS)

Wang, Lumin; Wierschke, Jonathan Brett

2015-01-01

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 (H 3 BO 3 ). 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.

Low Loss Advanced Metallic Fuel Casting Evaluation

International Nuclear Information System (INIS)

Kim, Kihwan; Ko, Youngmo; Kim, Jonghwan; Song, Hoon; Lee Chanbock

2014-01-01

The fabrication process for SFR fuel is composed of fuel slug casting, loading and fabrication of the fuel rods, and the fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycles streams in the fabrication process. Recycle streams include fuel slug reworks, returned scraps, and fuel casting heels, which are a special concern in the counter gravity injection casting process because of the large masses involved. Large recycle and waste streams result in lowering the productivity and the economic efficiency of fuel production. To increase efficiency the fuel losses in the furnace chamber, crucible, and the mold, after casting a considerable amount of fuel alloy in the casting furnace, will be quantitatively evaluated. After evaluation the losses will be identified and minimized. It is expected that this study will contribute to the minimization of fuel losses and the wastes streams in the fabrication process of the fuel slugs. Also through this study the technical readiness level of the metallic fuel fabrication process will be further enhanced. In this study, U-Zr alloy system fuel slugs were fabricated by a gravity casting method. Metallic fuel slugs were successfully fabricated with 19 slugs/batch with diameter of 5mm and length of 300mm. Fuel losses was quantitatively evaluated in casting process for the fuel slugs. Fuel losses of the fuel slugs were so low, 0.1∼1.0%. Injection casting experiments have been performed to reduce the fuel loss and improve the casting method. U-Zr fuel slug having φ5.4-L250mm was soundly fabricated with 0.1% in fuel loss. The fuel losses could be minimized to 0.1%, which showed that casting technology of fuel slugs can be a feasible approach to reach the goal of the fuel losses of 0.1% or less in commercial scale

Carbon potential measurement on some actinide carbides

International Nuclear Information System (INIS)

Anthonysamy, S.; Ananthasivan, K.; Kaliappan, I.; Chandramouli, V.; Vasudeva Rao, P.R.; Mathews, C.K.; Jacob, K.T.

1994-01-01

Uranium-Plutonium mixed carbides with a Pu/(U+Pu) ratio of 0.55 are to be used as the fuel in the Fast Breeder Test Reactor (FBTR) at Kalpakkam, India. Carburization of the stainless steel clad by this fuel is determined by its carbon potential. Because the carbon potential of this fuel composition is not available in the literature, it was measured by the methane-hydrogen gas equilibration technique. The sample was equilibrated with purified hydrogen and the equilibrium methane-to-hydrogen ratio in the gas phase was measured with a flame ionization detector. The carbon potential of the ThC-ThC 2 as well as Mo-Mo 2 C system, which is an important binary in the actinide-fission product-carbon systems, were also measured by this technique in the temperature range 973 to 1,173 K. The data for the Mo-Mo 2 C system are in agreement with values reported in the literature. The results for the ThC-ThC 2 system are different from estimated values with large uncertainty limits given in the literature. The data on (U, Pu) mixed carbides indicates the possibility of stainless steel clad attack under isothermal equilibrium conditions

Low temperature CVD deposition of silicon carbide

International Nuclear Information System (INIS)

Dariel, M.; Yeheskel, J.; Agam, S.; Edelstein, D.; Lebovits, O.; Ron, Y.

1991-04-01

The coating of graphite on silicon carbide from the gaseous phase in a hot-well, open flow reactor at 1150degC is described. This study constitutes the first part of an investigation of the process for the coating of nuclear fuel by chemical vapor deposition (CVD)

The German carbide program: Performance, experimental findings, and evaluation of irradiation results

International Nuclear Information System (INIS)

Steiner, H.; Freund, D.; Geithoff, D.

1982-09-01

In this report a synopsis of the German carbide program is presented. The program comprises the irradiation of about 100 carbide pins equipped with pelletted fuel. Most of these fuel pins were He-bonded, the sodium bonding concept taken as a back-up solution. The main design parameters such as smear and pellet density, gap size, pin diameter and wall thickness as well as the irradiation conditions were varied mostly within wide ranges. Based on a compilation of relevant pin parameters, irradiation conditions, and the results of various irradiation experiments conclusions on the optimum ranges of the main design parameters are drawn. Furthermore, some important aspects of fuel pin behaviour are discussed based on quantitative results from post irradiation examinations. (orig.) [de

Metal-fuel modeling for inherently safe reactor designs

International Nuclear Information System (INIS)

Miles, K.J. Jr.

1987-01-01

Current development of breeder reactor systems has led to the renewed interest in metal fuels. These fuels have properties that enhance the inherent safety of the system, such as high thermal conductivity, compatibility with liquid sodium, and low fuel/cladding mechanical interaction. While metal-fuel irradiation behavior is well understood, there are some areas where more information is needed to fully understand the various safety-related phenomena, such as fuel/cladding chemical interaction, eutectic melting and penetration, and axial relocation of molten fuel prior to cladding breach. Because many of these phenomena can cause changes in the reactivity state of the system, their effects on whole-core normal, anticipated, and hypothetical accident scenarios need to be studied. The metal-fuel behavior model DEFORM-5 is being developed to provide the necessary phenomenological basis for these studies. The first stage in the DEFORM-5 development has been completed. Presently, DEFORM-5 calculates the cladding strain, life fraction, and eutectic penetration thinning for Types D9, HT9, or 316 steels. This first stage of DEFORM-5 has been used to analyze the TREAT M2, M3, and M4 transients with irradiated Experimental Breeder Reactor-II driver fuel. The paper shows the DEFORM-5 and experimental results for failure times for the test pins. The results provide confidence and validation of the DEFORM-5 modeling of the cladding behavior

Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

Science.gov (United States)

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

2011-03-01

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Extended fuel swelling models and ultra high burn-up fuel behavior of U–Pu–Zr metallic fuel using FEAST-METAL

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydın, E-mail: karahan@alum.mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-215, Cambridge, MA 02139 (United States); Andrews, Nathan C., E-mail: nandrews@mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-215, Cambridge, MA 02139 (United States)

2013-05-15

Highlights: ► Improved fuel swelling models in phase structure dependent form. ► A probabilistic verification exercise for the open porosity formation threshold. ► Satisfactory validation effort for available EBR-II database. ► Ultra high burn-up behavior of U–6Zr fuel with 60% smear density fuel. -- Abstract: Computational models in FEAST-METAL U–Pu–Zr metallic fuel behavior code have been upgraded to improve fission gas, solid fission product swelling, and pore sintering behavior in a microstructure dependent form. First, fission gas bubble growth is modeled by selecting small and large bubble groups according to a fixed number of gas atoms per bubble group. Small bubbles nucleated at phase boundaries grow via gas migration and turn into large bubbles. Furthermore, bubble morphology for each phase structure is captured by selecting the number of atoms per bubble and the shape of the bubbles in a phase dependent form. The gas diffusion coefficients for the single gamma phase and effective dual (α + δ) and (β + γ) phase structures are modeled separately, using the activation energy of the corresponding phase structure. In this study, it is found that pressure sintering of the interconnected porosity in dual phases should be less effective than the reference model in order to match clad strain and fission gas release behavior. In addition to these improvements, a probabilistic approach is taken to verify the fission gas-swelling threshold at which interconnected porosity begins. This fracture problem is treated as a function of critical crack length formed via bubble coalescence. It was found that a 10% gas-swelling threshold is appropriate for a wide range of gas bubble sizes. The new version of FEAST-METAL predicts the burn-up, smear density, and axial variation of the clad hoop strain and fission gas release behavior satisfactorily for selected test pins under EBR-II conditions. The code is used to predict ultra-high burn-up U–Pu–6Zr vented

Extended fuel swelling models and ultra high burn-up fuel behavior of U–Pu–Zr metallic fuel using FEAST-METAL

International Nuclear Information System (INIS)

Karahan, Aydın; Andrews, Nathan C.

2013-01-01

Highlights: ► Improved fuel swelling models in phase structure dependent form. ► A probabilistic verification exercise for the open porosity formation threshold. ► Satisfactory validation effort for available EBR-II database. ► Ultra high burn-up behavior of U–6Zr fuel with 60% smear density fuel. -- Abstract: Computational models in FEAST-METAL U–Pu–Zr metallic fuel behavior code have been upgraded to improve fission gas, solid fission product swelling, and pore sintering behavior in a microstructure dependent form. First, fission gas bubble growth is modeled by selecting small and large bubble groups according to a fixed number of gas atoms per bubble group. Small bubbles nucleated at phase boundaries grow via gas migration and turn into large bubbles. Furthermore, bubble morphology for each phase structure is captured by selecting the number of atoms per bubble and the shape of the bubbles in a phase dependent form. The gas diffusion coefficients for the single gamma phase and effective dual (α + δ) and (β + γ) phase structures are modeled separately, using the activation energy of the corresponding phase structure. In this study, it is found that pressure sintering of the interconnected porosity in dual phases should be less effective than the reference model in order to match clad strain and fission gas release behavior. In addition to these improvements, a probabilistic approach is taken to verify the fission gas-swelling threshold at which interconnected porosity begins. This fracture problem is treated as a function of critical crack length formed via bubble coalescence. It was found that a 10% gas-swelling threshold is appropriate for a wide range of gas bubble sizes. The new version of FEAST-METAL predicts the burn-up, smear density, and axial variation of the clad hoop strain and fission gas release behavior satisfactorily for selected test pins under EBR-II conditions. The code is used to predict ultra-high burn-up U–Pu–6Zr vented

Behavior of metallic fuel in treat transient overpower tests

International Nuclear Information System (INIS)

Bauer, T.H.; Wright, A.E.; Robinson, W.R.; Klickman, A.E.

1988-01-01

Results and analyses are reported for TREAT in-pile transient overpower tests of margin to cladding failure and pre-failure axial expansion of metallic fuel. In all cases the power rise was exponential on an 8 s period until either incipient or actual cladding failure was achieved. Test fuel included EBR-II driver fuel and ternary alloy, the reference fuel of the Intergral Fast Reactor concept. Test pin burnup spanned the widest range available. The nature of the observed cladding failure and resultant fuel dispersals is described. Simple models are presented which describe observed cladding failures and pre-failure axial expansions yet are general enough to apply to all metal fuel types

Carbothermic reduction of refractory metals

International Nuclear Information System (INIS)

Anderson, R.N.; Parlee, N.A.D.

1976-01-01

The reduction of stable refractory metal oxides by carbon is generally unacceptable since the product is usually contaminated with carbides. The carbide formation may be avoided by selecting a solvent metal to dissolve the reactive metal as it is produced and reduce its chemical activity below that required for carbide formation. This approach has been successfully applied to the oxides of Si, Zr, Ti, Al, Mg, and U. In the case where a volatile suboxide, a carbonyl reaction, or a volatile metal occur, the use of the solvent metal appears satisfactory to limit the loss of material at low pressures. In several solute--solvent systems, vacuum evaporation is used to strip the solvent metal from the alloy to give the pure metal

Synthesis and characterization of group V metal carbide and nitride catalysts

Science.gov (United States)

Kwon, Heock-Hoi

1998-11-01

Group V transition metal carbides and nitrides were prepared via the temperature programmed reaction (TPR) of corresponding oxides with NHsb3 or a CHsb4/Hsb2 mixture. Except for the tantalum compounds, phase-pure carbides and nitrides were prepared. The vanadium carbides and nitrides were the most active and selective catalysts. Therefore the principal focus of the research was the preparation, characterization, and evaluation of high surface area vanadium nitride catalysts. A series of vanadium nitrides with surface areas up to 60 msp2/g was prepared. Thermal gravimetric analysis coupled with x-ray diffraction and scanning electron microscopy indicated that the solid-state reaction proceeded by the sequential reduction of Vsb2Osb5 to VOsb{0.9} and concluded with the topotactic substitution of nitrogen for oxygen in VOsb{0.9}. The transformation of Vsb2Osb5 to VN was pseudomorphic. An experimental design was executed to determine effects of the heating rates and space velocities on the VN microstructures. The heating rates had minor effects on the surface areas and pore size distributions; however, increasing the space velocity significantly increased the surface area. The materials were mostly mesoporous. Oxygen chemisorption on the vanadium nitrides scaled linearly with the surface area. The corresponding O/Vsbsurface ratio was ≈0.6. The vanadium nitrides were active for butane activation and pyridine hydrodenitrogenation. During butane activation, their selectivities towards dehydrogenation products were as high as 98%. The major product in pyridine hydrodenitrogenation was pentane. The reaction rates increased almost linearly with the surface area suggesting that these reactions were structure insensitive. The vanadium nitrides were not active for crotonaldehyde hydrogenation; however, they catalyzed an interesting ring formation reaction that produced methylbenzaldehyde and xylene from crotonaldehyde. A new method was demonstrated for the production of very

A basic research on the transient behavior for a metallic fuel FBR

International Nuclear Information System (INIS)

Baba, Mamoru; Hirano, Go; Kawada, Ken-ichi; Niwa, Hajime

1999-03-01

A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku university and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled 'A basic research on the transient behavior for a metallic fuel FBR'. The results are the following. (1) Target and Results of analysis: The accident initiator considered is a LOF accident without scram. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transition phase. It is shown that the moderator is quite elective to mitigate the accident at the initiation phase. However, it is necessary to analyze the transition phase to know if the re-criticality is totally avoided after the initiation phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large scale material movement in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics, and examined

Thermionic emission of cermets made of refractory carbides

International Nuclear Information System (INIS)

Samsonow, G.W.; Bogomol, I.W.; Ochremtschuk, L.N.; Podtschernjajewa, I.A.; Fomenko, W.S.

1975-01-01

In order to improve the resistance to thermal variations of refractory carbides having good behavior for thermionic emission, they have been combined with transition metals d. Thermionic emission was studied with cermets in compact samples. Following systems were examined: TiC-Nb, TiC-Mo, TiC-W, ZrC-Nb, ZrC-Mo, ZrC-W, WC-Mo with compositions of: 75% M 1 C-25% M 2 , 50%M 1 C-50%M 2 , 25%M 1 C-75%M 2 . When following the variation of electron emission energy phi versus the composition, it appears that in the range of mixed crystals (M 1 M 2 )C, phi decreases and the resistance to thermal variations of these phases is higher than that of individual carbides. The study of obtained cermets shows that their resistance to thermal variations is largely superior to the one of starting carbides; TiC and ZrC carbides, combined with molybdenum and tungsten support the highest number of thermic cycles

Design characteristics of metallic fuel rod on its in-LMR performance

International Nuclear Information System (INIS)

Hwang, Woan; Kang Hee Young; Nam, Cheol; Kim, Jong Oh

1997-01-01

Fuel design is a key feature to assure LMR safety goals. To date, a large effort had been devoted to develop metallic fuels at ANL's experimental breeder reactor (EBR-II). The major design and performance parameters investigated include; thermal conductivity and temperature profile; smear density; axial plenum; FCMI and cladding deformation including creep, and fission gas release. In order to evaluate the sensitivity of each parameter, in-LMR performances of metallic fuels are not only reviewed by the experiment results in literatures, but also key design characteristics according to the variation of metallic fuel rod design parameters are analyzed by using the MACSIS code which simulates in-reactor behaviors of metal fuel rod. In this study, key design characteristics and the criteria which must be considered to design fuel rod in LMR, are proposed and discussed. (author). 14 refs., 4 figs

Spheroidization of transition metal carbides in low temperature plasma

International Nuclear Information System (INIS)

Klinskaya, N.A.; Koroleva, E.B.; Petrunichev, V.A.; Rybalko, O.F.; Solov'ev, P.V.; Ugol'nikova, T.A.

1986-01-01

Plasma process of preparation of titanium, tungsten and chromium carbide spherical powders with the main particle size 40-80 μm is considered. Spheroidization degree, granulometric and phase composition of the product are investigated

Sensitivity Analysis of FEAST-Metal Fuel Performance Code: Initial Results

International Nuclear Information System (INIS)

Edelmann, Paul Guy; Williams, Brian J.; Unal, Cetin; Yacout, Abdellatif

2012-01-01

This memo documents the completion of the LANL milestone, M3FT-12LA0202041, describing methodologies and initial results using FEAST-Metal. The FEAST-Metal code calculations for this work are being conducted at LANL in support of on-going activities related to sensitivity analysis of fuel performance codes. The objective is to identify important macroscopic parameters of interest to modeling and simulation of metallic fuel performance. This report summarizes our preliminary results for the sensitivity analysis using 6 calibration datasets for metallic fuel developed at ANL for EBR-II experiments. Sensitivity ranking methodology was deployed to narrow down the selected parameters for the current study. There are approximately 84 calibration parameters in the FEAST-Metal code, of which 32 were ultimately used in Phase II of this study. Preliminary results of this sensitivity analysis led to the following ranking of FEAST models for future calibration and improvements: fuel conductivity, fission gas transport/release, fuel creep, and precipitation kinetics. More validation data is needed to validate calibrated parameter distributions for future uncertainty quantification studies with FEAST-Metal. Results of this study also served to point out some code deficiencies and possible errors, and these are being investigated in order to determine root causes and to improve upon the existing code models.

Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

Energy Technology Data Exchange (ETDEWEB)

Ammirabile, L.; Tsige-Tamirat, H. [European Commission, JRC, Inst. for Energy, Petten (Netherlands)

2011-07-01

In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

International Nuclear Information System (INIS)

Ammirabile, L.; Tsige-Tamirat, H.

2011-01-01

In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

The neutronic and fuel cycle performance of interchangeable 3500 MWth metal and oxide fueled LMRs

International Nuclear Information System (INIS)

Fujita, E.K.; Wade, D.C.

1990-01-01

This study summarizes the neutronic and fuel cycle analysis performed at Argonne National Laboratory for an oxide and a metal fueled 3500 MWth LMR. These reactor designs formed the basis for a joint US/European study of LMR ATWS events. The oxide and metal core designs were developed to meet reactor performance specifications that are constrained by requirements for core loading interchangeability and for a small burnup reactivity swing. Differences in the computed performance parameters of the oxide and metal cores, arising from basic differences in their neutronic characteristics, are identified and discussed. It is shown that metal and oxide cores designed to the same ground rules exhibit many similar performance characteristics; however, they differ substantially in reactivity coefficients, control strategies, and fuel cycle options. 12 refs., 2 figs., 12 tabs

Two-Dimensional Titanium Carbide (MXene) as Surface-Enhanced Raman Scattering Substrate

Energy Technology Data Exchange (ETDEWEB)

Sarycheva, Asia [Drexel Univ., Philadelphia, PA (United States); Makaryan, Taron [Drexel Univ., Philadelphia, PA (United States); Maleski, Kathleen [Drexel Univ., Philadelphia, PA (United States); Satheeshkumar, Elumalai [National Cheng Kung Univ., Tainan (Taiwan); National Institute of Technology-Trichy, Tamil Nadu (India); Melikyan, Armen [Russian-Armenian (Slavonic) State Univ., Yerevan (Armenia); Minassian, Hayk [A. Alikhanian National Science Lab., Yerevan (Armenia); Yoshimura, Masahiro [National Cheng Kung Univ., Tainan (Taiwan); Gogotsi, Yury G. [Drexel Univ., Philadelphia, PA (United States)

2017-08-22

Here, noble metal (gold or silver) nanoparticles or patterned films are typically used as substrates for surface-enhanced Raman spectroscopy (SERS). Two-dimensional (2D) carbides and nitrides (MXenes) exhibit unique electronic and optical properties, including metallic conductivity and plasmon resonance in the visible or near-infrared range, making them promising candidates for a wide variety of applications. Herein, we show that 2D titanium carbide, Ti₃C₂T_x, enhances Raman signal from organic dyes on a substrate and in solution. As a proof of concept, MXene SERS substrates were manufactured by spray-coating and used to detect several common dyes, with calculated enhancement factors reaching ~10⁶. Titanium carbide MXene demonstrates SERS effect in aqueous colloidal solutions, suggesting the potential for biomedical or environmental applications, where MXene can selectively enhance positively charged molecules.

Performance of HT9 clad metallic fuel at high temperature

International Nuclear Information System (INIS)

Pahl, R.G.; Lahm, C.E.; Hayes, S.L.

1992-01-01

Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching ∼660 degree C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area

The basic research on the CDA initiation phase for a metallic fuel FBR

International Nuclear Information System (INIS)

Hirano, Go; Hirakawa, Naohiro; Kawada, Ken-ichi; Niwa, Hazime

1998-03-01

A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku University and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled. (1) Target and Results of analysis: The accident initiator considered is a LOF accident with ATWS. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transient phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large material movement to in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics. The results of a sample case, that is a metallic fueled core at the beginning of cycle, show this improvement is appropriate. (3) Conclusion: The behavior at CDA of a metallic fueled core of a fast reactor was analyzed using the CDA initiation phase analysis code and the knowledge of the important characteristics at the CDA initiation phase was obtained

Metal-like self-organization of periodic nanostructures on silicon and silicon carbide under femtosecond laser pulses

International Nuclear Information System (INIS)

Gemini, Laura; Hashida, Masaki; Shimizu, Masahiro; Miyasaka, Yasuhiro; Inoue, Shunsuke; Tokita, Shigeki; Sakabe, Shuji; Limpouch, Jiri; Mocek, Tomas

2013-01-01

Periodic structures were generated on Si and SiC surfaces by irradiation with femtosecond laser pulses. Self-organized structures with spatial periodicity of approximately 600 nm appear on silicon and silicon carbide in the laser fluence range just above the ablation threshold and upon irradiation with a large number of pulses. As in the case of metals, the dependence of the spatial periodicity on laser fluence can be explained by the parametric decay of laser light into surface plasma waves. The results show that the proposed model might be universally applicable to any solid state material

A design study of high breeding ratio sodium cooled metal fuel core without blanket fuels

International Nuclear Information System (INIS)

Kobayashi, Noboru; Ogawa, Takashi; Ohki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari

2009-01-01

The metal fuel core is superior to the mixed oxide fuel core because of its high breeding ratio and compact core size resulting from hard neutron spectrum and high heavy metal densities. Utilizing these characteristics, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8 $, a core height of less than 150 cm, the maximum cladding temperature of 650degC, and the maximum fuel pin bundle pressure drop of 0.4 MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40. (author)

Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

International Nuclear Information System (INIS)

Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Pal, S.; Mishra, G.P.

2012-01-01

CERMET fuel with either PuO 2 or enriched UO 2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO 2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO 2 , U–25 wt%UO 2 and U–30 wt%UO 2 . It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO 2 compositions.

Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

Science.gov (United States)

Sinha, V. P.; Hegde, P. V.; Prasad, G. J.; Pal, S.; Mishra, G. P.

2012-08-01

CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

Nonmetal effect on ordering structures in titanium carbide

International Nuclear Information System (INIS)

Tashmetov, M.Yu.; Ehm, V.T.; Savenko, B.M.

1997-01-01

The effect of oxygen and nitrogen atoms on formation of intermediate, cubic and trigonal ordering structures in the titanium carbide is studied through the roentgenography and neutron radiography methods. Metal atoms in the TiC 0.545 O 0.08 , TiC 0.545 N 0.09 samples under study are shifted from ideal positions in the direction from vacancies to metalloid atoms. In the intermediate cubic phase the values of the titanium atoms free parameter in both samples are identical, but they differ from analogous values in the titanium carbide

Preliminary study on metallic inclusion in nuclear fuel

International Nuclear Information System (INIS)

Fukuzawa, Takashi; Tanaka, Masahiro; Tanabe, Tetsuo; Imoto, Shosuke

1984-01-01

In recent postirradiation tests, metallic fission products such as Mo, Ru, Rh and Pd are known to precipitate as metallic inclusions in the fuel. These inclusions remain as insoluble residues and provide various problems in different fields of the reprocessing. In this report are presented preliminary results of the study on the ternary phase diagram of Mo-Ru-Pd system and on their properties in nitric acid or various oxidative environments. It is concluded that (1) most of metallic inclusions which are insoluble in nitric acid show epsilon phase, Ru base hcp alloy, in which a large amount of Mo and Pd are soluble, (2) Pd, however, seems to deposit separately in the fuel pin because of its high vapor pressure, (3) Mo fraction in the inclusion would be highly dependent on oxygen potential in the fuel pin. (author)

The coloring problem in the solid-state metal boride carbide ScB{sub 2}C{sub 2}. A theoretical analysis

Energy Technology Data Exchange (ETDEWEB)

Lassoued, Souheila [Universite de Rennes, Ecole Nationale Superieure de Chimie, UMR 6226 CNRS (France). Inst. des Sciences Chimiques; Universite Kasdi Merbah-Ouargla (Algeria). Faculte des Mathematiques et des Sciences de la Matiere; Boucher, Benoit [Universite de Rennes, Ecole Nationale Superieure de Chimie, UMR 6226 CNRS (France). Inst. des Sciences Chimiques; Max-Planck-Institut fuer Chemische Physik Fester Stoffe, Dresden (Germany); Boutarfaia, Ahmed [Universite Kasdi Merbah-Ouargla (Algeria). Faculte des Mathematiques et des Sciences de la Matiere; Gautier, Regis; Halet, Jean-Francois [Universite de Rennes, Ecole Nationale Superieure de Chimie, UMR 6226 CNRS (France). Inst. des Sciences Chimiques

2016-08-01

The electronic properties of the layered ternary metal boride carbide ScB{sub 2}C{sub 2}, the structure of which consists of B/C layers made of fused five- and seven-membered rings alternating with scandium sheets, are analyzed. In particular, the respective positions of the B and C atoms (the so-called coloring problem) are tackled using density functional theory, quantum theory of atoms in molecules, and electron localizability indicator calculations. Results reveal that (i) the most stable coloring minimizes the number of B-B and C-C contacts and maximizes the number of boron atoms in the heptagons, (ii) the compound is metallic in character, and (iii) rather important covalent bonding occurs between the metallic sheets and the boron-carbon network.

ENTIRELY AQUEOUS SOLUTION-GEL ROUTE FOR THE PREPARATION OF ZIRCONIUM CARBIDE, HAFNIUM CARBIDE AND THEIR TERNARY CARBIDE POWDERS

Directory of Open Access Journals (Sweden)

Zhang Changrui

2016-07-01

Full Text Available An entirely aqueous solution-gel route has been developed for the synthesis of zirconium carbide, hafnium carbide and their ternary carbide powders. Zirconium oxychloride (ZrOCl₂.8H₂O, malic acid (MA and ethylene glycol (EG were dissolved in water to form the aqueous zirconium carbide precursor. Afterwards, this aqueous precursor was gelled and transformed into zirconium carbide at a relatively low temperature (1200 °C for achieving an intimate mixing of the intermediate products. Hafnium and the ternary carbide powders were also synthesized via the same aqueous route. All the zirconium, hafnium and ternary carbide powders exhibited a particle size of ∼100 nm.

Simulation of the injection casting of metallic fuels

International Nuclear Information System (INIS)

Nakagawa, Tomokazu; Ogata, Takanari; Tokiwai, Moriyasu.

1989-01-01

For the fabrication of metallic fuel pins, injection casting is a preferable process because the simplicity of the process is suitable for remote operation. In this process, the molten metal in the crucible is injected into evacuated molds (suspended above the crucible) by pressurizing the casting furnace. Argonne National Laboratory has already adopted this process in the Integral Fast Reactor program. To obtain fuel pins with good quality, the casting parameters, such as the molten metal temperature, the magnitude of the pressure applied, the pressurizing rate, the cooling time, etc., must be optimized. Otherwise, bad-quality castings (short castings, rough surfaces, shrinkage cavities, mold fracture) may result. Therefore, it is very important in designing the casting equipment and optimizing the operation conditions to be able to predict the fluid and thermal behavior of the castings. This paper describes methods to simulate the heat and mass transfer in the molds and molten metallic fuel during injection casting. The results obtained by simulation are compared with experimental ones. Also, appropriate casting conditions for the uranium-plutonium-zirconium alloy are discussed based on the simulated results

Irradiation and examination results of the AC-3 mixed-carbide test

International Nuclear Information System (INIS)

Mason, R.E.; Hoth, C.W.; Stratton, R.W.; Botta, F.

1992-01-01

The AC-3 test was a cooperative Swiss/US irradiation test of mixed-carbide, (U,Pr)C, fuel pins in the Fast Flux Test Facility. The test included 25 Swiss-fabricated sphere-pac-type fuel pins and 66 U.S. fabricated pellet-type fuel pins. The test was designed to operate at prototypical fast reactor conditions to provide a direct comparison of the irradiation performance of the two fuel types. The test design and fuel fabrication processes used for the AC-3 test are presented

Non-noble metal fuel cell catalysts

CERN Document Server

Chen, Zhongwei; Zhang, Jiujun

2014-01-01

Written and edited by a group of top scientists and engineers in the field of fuel cell catalysts from both industry and academia, this book provides a complete overview of this hot topic. It covers the synthesis, characterization, activity validation and modeling of different non-noble metal and metalfree electrocatalysts for the reduction of oxygen, as well as their integration into acid or alkaline polymer exchange membrane (PEM) fuel cells and their performance validation, while also discussing those factors that will drive fuel cell commercialization. With its well-structured app

Pyroelectrochemical process for reprocessing irradiated nuclear fuels

International Nuclear Information System (INIS)

Brambilla, G.; Sartorelli, A.

1982-01-01

A pyroelectrochemical process for reprocessing irradiated fast reactor mixed oxide or carbide fuels is described. The fuel is dissolved in a bath of molten alkali metal sulfates. The Pu(SO 4 ) 2 formed in the bath is thermally decomposed, leaving crystalline PuO 2 on the bottom of the reaction vessel. Electrodes are then introduced into the bath, and UO 2 is deposited on the cathode. Alternatively, both UO 2 and PuO 2 may be electrodeposited. The molten salts, after decontamination by precipitating the fission products dissolved in the bath by introducing basic agents such as oxides, carbonates, or hydroxides, may be recycled. Since it is not possible to remove cesium from the molten salt bath, periodic disposal and partial renewal with fresh salts is necessary. The melted salts that contain the fission products are conditioned for disposal by embedding them in a metallic matrix

Metal waste forms from treatment of EBR-II spent fuel

International Nuclear Information System (INIS)

Abraham, D. P.

1998-01-01

Demonstration of Argonne National Laboratory's electrometallurgical treatment of spent nuclear fuel is currently being conducted on irradiated, metallic driver fuel and blanket fuel elements from the Experimental Breeder Reactor-II (EBR-II) in Idaho. The residual metallic material from the electrometallurgical treatment process is consolidated into an ingot, the metal waste form (MWF), by employing an induction furnace in a hot cell. Scanning electron microscopy (SEM) and chemical analyses have been performed on irradiated cladding hulls from the driver fuel, and on samples from the alloy ingots. This paper presents the microstructures of the radioactive ingots and compares them with observations on simulated waste forms prepared using non-irradiated material. These simulated waste forms have the baseline composition of stainless steel - 15 wt % zirconium (SS-15Zr). Additions of noble metal elements, which serve as surrogates for fission products, and actinides are made to that baseline composition. The partitioning of noble metal and actinide elements into alloy phases and the role of zirconium for incorporating these elements is discussed in this paper

Technology of the production of breeder fuel elements

International Nuclear Information System (INIS)

Funke, P.

1976-01-01

A survey is presented of the fabrication of oxide and carbide fuels and of the fuel rod for fast breeders (KNK, SNR-300). The advantages of the chosen methods are explained. The main points of development concerning the oxide fuel rod are gone into. The process sequence for plutonium oxide and plutonium carbide processing is presented in a flow chart. (HR) [de

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.

Chemical compatibility between cladding alloys and advanced fuels

International Nuclear Information System (INIS)

Fee, D.C.; Johnson, C.E.

1975-05-01

The National Advanced Fuels Program requires chemical, mechanical, and thermophysical properties data for cladding alloys. The compatibility behavior of cladding alloys with advanced fuels is critically reviewed. in carbide fuel pins, the principal compatibility problem is cladding carburization, diffusion of carbon into the cladding matrix accompanied by carbide precipitation. Carburization changes the mechanical properties of the cladding alloy. The extent of carburization increases in sodium (versus gas) bonded fuels. The depth of carburization increases with increasing sesquicarbide (M 2 C 3 ) content of the fuel. In nitride fuel pins, the principal compatibility problem is cladding nitriding, diffusion of nitrogen into the cladding matrix accompanied by nitride precipitation. Nitriding changes the mechanical properties of the cladding alloy. In both carbide and nitride fuel pins, fission products do not migrate appreciably to the cladding and do not appear to contribute to cladding attack. 77 references. (U.S.)

State-of-the-art Report on Innovative Fuels for Advanced Nuclear Systems

International Nuclear Information System (INIS)

Chauvin, N.; Minato, K.; Ogata, T.; Lee, C.B.; Pouchon, M.A.; Pasamehmetoglu, K.O.; Choi, Y.J.; Kennedy, J.R.; Massara, S.; Cornet, S.; ); Sommers, J.; ); McClellan, K.

2014-01-01

Development of innovative fuels such as homogeneous and heterogeneous fuels, ADS fuels, and oxide, metal, nitride and carbide fuels is an important stage in the implementation process of advanced nuclear systems. Several national and international R and D programmes are investigating minor actinide-bearing fuels due to their ability to help reduce the radiotoxicity of spent fuel and therefore decrease the burden on geological repositories. Minor actinides can be converted into a suitable fuel form for irradiation in reactor systems where they are transmuted into fission products with a significantly shorter half-life. This report compares recent studies of fuels containing minor actinides for use in advanced nuclear systems. The studies review different fuels for several types of advanced reactors by examining various technical issues associated with fabrication, characterisation, irradiation performance, design and safety criteria, as well as technical maturity. (authors)

Advancing liquid metal reactor technology with nitride fuels

International Nuclear Information System (INIS)

Lyon, W.F.; Baker, R.B.; Leggett, R.D.; Matthews, R.B.

1991-08-01

A review of the use of nitride fuels in liquid metal fast reactors is presented. Past studies indicate that both uranium nitride and uranium/plutonium nitride possess characteristics that may offer enhanced performance, particularly in the area of passive safety. To further quantify these effects, the analysis of a mixed-nitride fuel system utilizing the geometry and power level of the US Advanced Liquid Metal Reactor as a reference is described. 18 refs., 2 figs., 2 tabs

Benchmark physics tests in the metallic-fueled assembly ZPPR-15

International Nuclear Information System (INIS)

McFarlane, H.F.; Brumbach, S.B.; Carpenter, S.G.; Collins, P.J.

1989-01-01

Results of the first benchmark physics tests of a metallic-fueled, demonstration-size liquid-metal reactor (LMR) are reported. A simple, two-zone, cylindrical conventional assembly was built with three distinctly different compositions to represent the stages of the Integral Fast Reactor fuel cycle. Experiments included criticality, control, power distribution, reaction rate ratios, reactivity coefficients, shielding, kinetics, and spectrum. Analysis was done with three-dimensional nodal diffusion calculations and ENDF/B-V.2 cross sections. Predictions of the ZPPR-15 reactor physics parameters agreed sufficiently well with the measured values to justify confidence in design analyses for metallic-fueled LMRs

Hydrotreatment activities of supported molybdenum nitrides and carbides

Energy Technology Data Exchange (ETDEWEB)

Dolce, G.M.; Savage, P.E.; Thompson, L.T. [University of Michigan, Ann Arbor, MI (United States). Dept. of Chemical Engineering

1997-05-01

The growing need for alternative sources of transportation fuels encourages the development of new hydrotreatment catalysts. These catalysts must be active and more hydrogen efficient than the current commercial hydrotreatment catalysts. Molybdenum nitrides and carbides are attractive candidate materials possessing properties that are comparable or superior to those of commercial sulfide catalysts. This research investigated the catalytic properties of {gamma}-Al{sub 2}O{sub 3}-supported molybdenum nitrides and carbides. These catalysts were synthesized via temperature-programmed reaction of supported molybdenum oxides with ammonia or methane/hydrogen mixtures. Phase constituents and compositions were determined by X-ray diffraction, elemental analysis, and neutral activation analysis. Oxygen chemisorption was used to probe the surface properties of the catalysts. Specific activities of the molybdenum nitrides and carbides were competitive with those of a commercial sulfide catalyst for hydrodenitrogenation (HDN), hydrodesulfurization (HDS), and hydrodeoxygenation (HDO). For HDN and HDS, the catalytic activity on a molybdenum basis was a strong inverse function of the molybdenum loading. Product distributions of the HDN, HDO and HDS of a variety of heteroatom compounds indicated that several of the nitrides and carbides were more hydrogen efficient than the sulfide catalyst. 35 refs., 8 figs., 7 tabs.

High pressure sintering (HP-HT) of diamond powders with titanium and titanium carbide

International Nuclear Information System (INIS)

Jaworska, L.

1999-01-01

Polycrystalline diamond compacts for cutting tools are mostly manufactured using high pressure sintering (HP-HT). The standard diamond compacts are prepared by diamond powders sintering with metallic binding phase. The first group of metallic binder are metals able to solve carbon - Co, Ni. The second group of metal binders are carbide forming elements - Ti, Cr, W and others. The paper describes high pressure sintering of diamond powder with titanium and nonstoichiometry titanium carbide for cutting tool application. A type of binding phase has the significant influence on microstructure and mechanical properties of diamond compacts. Very homogeneous structure was achieved in case of compacts obtained from metalized diamond where diamond-TiC-diamond connection were predominant. In the case of compacts prepared by mechanical mixing of diamond with titanium powders the obtained structure was nonhomogeneous with titanium carbide clusters. They had more diamond to diamond connections. These compacts compared to the compact made of metallized diamond have greater wear resistance. In the case of the diamond and TiC 0.92 sintering the strong bonding of TiC diamond grains was obtained. The microstructure observations for diamond with 5% wt. Ti and diamond with 5% wt. TiC 0.92 (the initial composition) compacts were performed in transmission microscope. For two type of compacts the strong bonding phase TiC without defects is creating. (author)

The effect of heat treatment on structure and properties of hard metals on a tungsten carbide basis with iron-nickel-binders

International Nuclear Information System (INIS)

Chaporova, I.N.; Kudryavtseva, V.I.; Sapronova, Z.N.; Sychkova, L.V.

1980-06-01

In the present paper the effect of storage and quenching on structure and properties of WC(Fe,Ni)-hardmetals was investigated. Starting materials were powders of tungsten carbide, iron and nickel, commonly used for the hard metal production. WC(Fe,Ni)-specimens (Fe: Ni = 80:20, 85:15) with 8, 11, 80, 85, 89 and 92 percent by weight were produced for the investigation. (orig.) [de

Synthesis of carbon fibre-reinforced, silicon carbide composites by ...

Indian Academy of Sciences (India)

carbon fibre (Cf) reinforced, silicon carbide matrix composites which are ... eral applications, such as automotive brakes, high-efficiency engine systems, ... The PIP method is based on the use of organo metallic pre-ceramic precursors.

Exploration on Wire Discharge Machining Added Powder for Metal-Based Diamond Grinding Wheel on Wire EDM Dressing and Truing of Grinding Tungsten Carbide Material

Science.gov (United States)

Chow, H. M.; Yang, L. D.; Lin, Y. C.; Lin, C. L.

2017-12-01

In this paper, the effects of material removal rate and abrasive grain protrusion on the metal-based diamond grinding wheel were studied to find the optimal parameters for adding powder and wire discharge. In addition, this kind of electric discharge method to add powder on the metal-based diamond grinding wheel on line after dressing and truing will be applied on tungsten carbide to study the grinding material removal rate, grinding wheel wear, surface roughness, and surface micro-hardness.

Structural stability, electronic structure and mechanical properties of actinide carbides AnC (An = U, Np)

International Nuclear Information System (INIS)

Manikandan, M.; Santhosh, M.; Rajeswarapalanichamy, R.

2016-01-01

Ab initio calculations are performed to investigate the structural stability, electronic structure and mechanical properties of actinide carbides AnC (An=U, Np) for three different crystal structures, namely NaCl, CsCl and ZnS. Among the considered structures, NaCl structure is found to be the most stable structure for these carbides at normal pressure. A pressure induced structural phase transition from NaCl to ZnS is observed. The electronic structure reveals that these carbides are metals. The calculated elastic constants indicate that these carbides are mechanically stable at normal pressure.

Irradiation performance of U-Pu-Zr metal fuels for liquid-metal-cooled reactors

International Nuclear Information System (INIS)

Tsai, H.; Cohen, A.B.; Billone, M.C.; Neimark, L.A.

1994-10-01

This report discusses a fuel system utilizing metallic U-Pu-Zr alloys which has been developed for advanced liquid metal-cooled reactors (LMRs). Result's from extensive irradiation testing conducted in EBR-II show a design having the following key features can achieve both high reliability and high burnup capability: a cast nominally U-20wt %Pu-10wt %Zr slug with the diameter sized to yield a fuel smear density of ∼75% theoretical density, low-swelling tempered martensitic stainless steel cladding, sodium bond filling the initial fuel/cladding gap, and an as-built plenum/fuel volume ratio of ∼1.5. The robust performance capability of this design stems primarily from the negligible loading on the cladding from either fuel/cladding mechanical interaction or fission-gas pressure during the irradiation. The effects of these individual design parameters, e.g., fuel smear density, zirconium content in fuel, plenum volume, and cladding types, on fuel element performance were investigated in a systematic irradiation experiment in EBR-II. The results show that, at the discharge burnup of ∼11 at. %, variations on zirconium content or plenum volume in the ranges tested have no substantial effects on performance. Fuel smear density, on the other hand, has pronounced but countervailing effects: increased density results in greater cladding strain, but lesser cladding wastage from fuel/cladding chemical interaction

Contribution of metallic fission product inclusions to axial fuel motion potential

International Nuclear Information System (INIS)

Sasa, P.; Cronenberg, A.; Stevenson, M.

1979-01-01

In the analysis of postulated nuclear reactor accidents, axial fuel motion within the fuel pin prior to cladding failure can have an important mitigating effect. The question of primary importance is whether or not metallic inclusions have the potential to vaporize during an overheating event and thus contribute to fuel motion. To assess this potential, two limiting calculations were made: 1) The inclusion constituent assumed insoluble in one another and 2) The constituents assumed totally miscible in one another. Thermodynamic considerations indicate that the metallic fission products found within inclusions of fuel rods irradiated in a fast neutron spectrum, would form homogeneous solutions. Therefore, it is concluded that the metallic fission products would not enhance fuel swelling during an overheating event. 16 refs

APS- and XPS-investigations of vanadium, vanadium carbide and graphite

Energy Technology Data Exchange (ETDEWEB)

Bradshaw, A M; Krause, U [Technische Univ. Muenchen (F.R. Germany). Inst. fuer Physikalische Chemie und Theoretische Chemie

1975-11-01

Soft X-ray appearance potential spectroscopy (APS) and X-ray photoelectron spectroscopy (XPS) have been used to study vanadium, vanadium carbide, and graphite. The chemical shifts for vanadium carbide with respect to metallic vanadium and graphite are compared for the two methods. The Csub(K) structure in APS and the valence band in XPS for vanadium carbide show good agreement with the band structure calculations of Neckel and co-workers. Using the band structure calculations of Painter et al. it is also shown how the multi-peak structure in the APS spectrum of graphite is possibly due to density of states effects. It would therefore appear that plasmon coupling plays only a minor role.

Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

Science.gov (United States)

Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

2013-09-03

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Tungsten carbide nanoparticles as efficient cocatalysts for photocatalytic overall water splitting

KAUST Repository

Garcia Esparza, Angel T.

2012-12-17

Tungsten carbide exhibits platinum-like behavior, which makes it an interesting potential substitute for noble metals in catalytic applications. Tungsten carbide nanocrystals (≈5 nm) are directly synthesized through the reaction of tungsten precursors with mesoporous graphitic C3N 4 (mpg-C3N4) as the reactive template in a flow of inert gas at high temperatures. Systematic experiments that vary the precursor compositions and temperatures used in the synthesis selectively generate different compositions and structures for the final nanocarbide (W 2C or WC) products. Electrochemical measurements demonstrate that the WC phase with a high surface area exhibits both high activity and stability in hydrogen evolution over a wide pH range. The WC sample also shows excellent hydrogen oxidation activity, whereas its activity in oxygen reduction is poor. These tungsten carbides are successful cocatalysts for overall water splitting and give H2 and O2 in a stoichiometric ratio from H 2O decomposition when supported on a Na-doped SrTiO3 photocatalyst. Herein, we present tungsten carbide (on a small scale) as a promising and durable catalyst substitute for platinum and other scarce noble-metal catalysts in catalytic reaction systems used for renewable energy generation. Platinum replacement: The phase-controlled synthesis of tungsten carbide nanoparticles from the nanoconfinement of a mesoporous graphite C 3N4 (mpg-C3N4) reactive template is shown. The nanomaterials catalyze hydrogen evolution/oxidation reactions, but are inactive in the oxygen reduction reaction. Tungsten carbide is an effective cocatalyst for photocatalytic overall water splitting (see picture). Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Tungsten carbide nanoparticles as efficient cocatalysts for photocatalytic overall water splitting

KAUST Repository

Garcia Esparza, Angel T.; Cha, Dong Kyu; Ou, Yiwei; Kubota, Jun; Domen, Kazunari; Takanabe, Kazuhiro

2012-01-01

Tungsten carbide exhibits platinum-like behavior, which makes it an interesting potential substitute for noble metals in catalytic applications. Tungsten carbide nanocrystals (≈5 nm) are directly synthesized through the reaction of tungsten precursors with mesoporous graphitic C3N 4 (mpg-C3N4) as the reactive template in a flow of inert gas at high temperatures. Systematic experiments that vary the precursor compositions and temperatures used in the synthesis selectively generate different compositions and structures for the final nanocarbide (W 2C or WC) products. Electrochemical measurements demonstrate that the WC phase with a high surface area exhibits both high activity and stability in hydrogen evolution over a wide pH range. The WC sample also shows excellent hydrogen oxidation activity, whereas its activity in oxygen reduction is poor. These tungsten carbides are successful cocatalysts for overall water splitting and give H2 and O2 in a stoichiometric ratio from H 2O decomposition when supported on a Na-doped SrTiO3 photocatalyst. Herein, we present tungsten carbide (on a small scale) as a promising and durable catalyst substitute for platinum and other scarce noble-metal catalysts in catalytic reaction systems used for renewable energy generation. Platinum replacement: The phase-controlled synthesis of tungsten carbide nanoparticles from the nanoconfinement of a mesoporous graphite C 3N4 (mpg-C3N4) reactive template is shown. The nanomaterials catalyze hydrogen evolution/oxidation reactions, but are inactive in the oxygen reduction reaction. Tungsten carbide is an effective cocatalyst for photocatalytic overall water splitting (see picture). Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

The characterization and monitoring of metallic fuel breaches in EBR-2

International Nuclear Information System (INIS)

Pahl, R.G.; Batte, G.L.; Mikaili, R.; Lambert, J.D.B.; Hofman, G.L.

1991-01-01

This paper discusses the characterization and monitoring of metallic fuel breaches which is now a significant part of the Integral Fast Reactor fuel testing program at Argonne National Laboratory. Irradiation experience with failed metallic fuel now includes natural breaches in the plenum and fuel column regions in lead ''endurance'' tests as well as fuel column breaches in artificially-defected fuel which have operated for months in the run-beyond-cladding breach (RBCB) mode. Analyses of the fission gas (FG) release-to-birth (R/B) ratios of selected historical breaches have been completed and have proven to be very useful in differentiating between plenum and fuel column breaches

Analytical and experimental evaluation of joining silicon nitride to metal and silicon carbide to metal for advanced heat engine applications. Final report

Energy Technology Data Exchange (ETDEWEB)

Kang, S.; Selverian, J.H.; O`Neil, D.; Kim, H. [GTE Labs., Inc., Waltham, MA (US); Kim, K. [Brown Univ., Providence, RI (US). Div. of Engineering

1993-05-01

This report summarizes the results of Phase 2 of Analytical and Experimental Evaluation of Joining Silicon Nitride to Metal and Silicon Carbide to Metal for Advanced Heat Engine Applications. A general methodology was developed to optimize the joint geometry and material systems for 650{degrees}C applications. Failure criteria were derived to predict the fracture of the braze and ceramic. Extensive finite element analyses (FEA) were performed to examine various joint geometries and to evaluate the affect of different interlayers on the residual stress state. Also, material systems composed of coating materials, interlayers, and braze alloys were developed for the program based on the chemical stability and strength of the joints during processing, and service. The FEA results were compared with experiments using two methods: (1) an idealized strength relationship of the ceramic, and (2) a probabilistic analysis of the ceramic strength (NASA CARES). The results showed that the measured strength of the joint reached 30--80% of the strength predicted by FEA. Also, potential high-temperature braze alloys were developed and evaluated for the high-temperature application of ceramic-metal joints. 38 tabs, 29 figs, 20 refs.

Development of 4S and related technologies (2). Long life metallic fuel

International Nuclear Information System (INIS)

Yacout, A.M.; Tsuboi, Y.; Ueda, N.

2009-01-01

This paper provides an overview of the long life metallic fuel to be used in the 4S reactor. The 4S fuel design is presented and implications of its characteristics on fuel performance are discussed. Main design characteristics include the long fuel life time of 30 years and the wider and longer fuel pins compared to EBR-II and FFTF fuel pins. The LIFE-METAL fuel performance code was used to evaluate the performance of the 4S fuel design. The code has been validated using post irradiation examination data of metallic fuel irradiated in EBR-II. The performance evaluation shows the benign nature of the design. The design enables the fuel to perform adequately during reactor operations without violating any of a conservative set of steady state design criteria. A survey evaluation of the fuel performance is also presented. This performance bounding evaluation took into account possible fuel swelling behavior and cladding temperature range that represents worst case scenarios. The evaluation showed that the fuel maintains its integrity even under those worst case conditions. (author)

Performance of metal and oxide fuels during accidents in a large liquid metal cooled reactor

International Nuclear Information System (INIS)

Cahalan, J.; Wigeland, R.; Friedel, G.; Kussmaul, G.; Royl, P.; Moreau, J.; Perks, M.

1990-01-01

In a cooperative effort among European and US analysts, an assessment of the comparative safety performance of metal and oxide fuels during accidents in a large (3500 MWt), pool-type, liquid-metal-cooled reactor (LMR) was performed. The study focused on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected loss-of-heat-sink (ULOHS). Emphasis was placed on identification of design features that provide passive, self-limiting responses to upset conditions, and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than oxide-fueled reactors of the same design. 3 refs., 4 figs

Fuel damage during off-normal transients in metal-fueled fast reactors

International Nuclear Information System (INIS)

Kramer, J.M.; Bauer, T.H.

1990-01-01

Fuel damage during off-normal transients is a key issue in the safety of fast reactors because the fuel pin cladding provides the primary barrier to the release of radioactive materials. Part of the Safety Task of the Integral Fast Reactor Program is to provide assessments of the damage and margins to failure for metallic fuels over the wide range of transients that must be considered in safety analyses. This paper reviews the current status of the analytical and experimental programs that are providing the bases for these assessments. 13 refs., 2 figs

Irradiation experience with HT9-clad metallic fuel

International Nuclear Information System (INIS)

Pahl, R.G.; Lahm, C.E.; Tsai, H.; Billone, M.C.

1991-01-01

The safe and reliable performance of metallic fuel is currently under study and demonstration in the Integral Fast Reactor program. In-reactor tests of HT9-clad metallic fuel have now reached maturity and have all shown good performance characteristics to burnups exceeding 17.5 at. % in the lead assembly. Because this low-swelling tempered martensitic alloy is the cladding of choice for high fluence applications, the experimental observations and performance modelling efforts reported in this paper play an important role in demonstrating reliability

Measuring the noble metal and iodine composition of extracted noble metal phase from spent nuclear fuel using instrumental neutron activation analysis

International Nuclear Information System (INIS)

Palomares, R.I.; Dayman, K.J.; Landsberger, S.; Biegalski, S.R.; Soderquist, C.Z.; Casella, A.J.; Brady Raap, M.C.; Schwantes, J.M.

2015-01-01

Masses of noble metal and iodine nuclides in the metallic noble metal phase extracted from spent fuel are measured using instrumental neutron activation analysis. Nuclide presence is predicted using fission yield analysis, and radionuclides are identified and the masses quantified using neutron activation analysis. The nuclide compositions of noble metal phase derived from two dissolution methods, UO 2 fuel dissolved in nitric acid and UO 2 fuel dissolved in ammonium-carbonate and hydrogen-peroxide solution, are compared. - Highlights: • The noble metal phase was chemically extracted from spent nuclear fuel and analyzed non-destructively. • Noble metal phase nuclides and long-lived iodine were identified and quantified using neutron activation analysis. • Activation to shorter-lived radionuclides allowed rapid analysis of long-lived fission products in spent fuel using gamma spectrometry

The evaluation of the use of metal alloy fuels in pressurized water reactors

International Nuclear Information System (INIS)

Lancaster, D.

1992-01-01

The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ''advanced reactors,'' it became clear that reactor design optimization has been under emphasized. Current ''advanced reactors'' are severely constrained. The AP-600 required the use of a fuel design from the 1970's. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing

Growth and structure of carbide nanorods

International Nuclear Information System (INIS)

Lieber, C.M.; Wong, E.W.; Dai, H.; Maynor, B.W.; Burns, L.D.

1996-01-01

Recent research on the growth and structure of carbide nanorods is reviewed. Carbide nanorods have been prepared by reacting carbon nanotubes with volatile transition metal and main group oxides and halides. Using this approach it has been possible to obtain solid carbide nanorods of TiC, SiC, NbC, Fe 3 C, and BC x having diameters between 2 and 30 nm and lengths up to 20 microm. Structural studies of single crystal TiC nanorods obtained through reactions of TiO with carbon nanotubes show that the nanorods grow along both [110] and [111] directions, and that the rods can exhibit either smooth or saw-tooth morphologies. Crystalline SiC nanorods have been produced from reactions of carbon nanotubes with SiO and Si-iodine reactants. The preferred growth direction of these nanorods is [111], although at low reaction temperatures rods with [100] growth axes are also observed. The growth mechanisms leading to these novel nanomaterials have also been addressed. Temperature dependent growth studies of TiC nanorods produced using a Ti-iodine reactant have provided definitive proof for a template or topotactic growth mechanism, and furthermore, have yielded new TiC nanotube materials. Investigations of the growth of SiC nanorods show that in some cases a catalytic mechanism may also be operable. Future research directions and applications of these new carbide nanorod materials are discussed

Innovative coating of nanostructured vanadium carbide on the F/M cladding tube inner surface for mitigating the fuel cladding chemical interactions

Energy Technology Data Exchange (ETDEWEB)

Yang, Yong [Univ. of Florida, Gainesville, FL (United States); Phillpot, Simon [Univ. of Florida, Gainesville, FL (United States)

2017-11-29

Fuel cladding chemical interactions (FCCI) have been acknowledged as a critical issue in a metallic fuel/steel cladding system due to the formation of low melting intermetallic eutectic compounds between the fuel and cladding steel, resulting in reduction in cladding wall thickness as well as a formation of eutectic compounds that can initiate melting in the fuel at lower temperature. In order to mitigate FCCI, diffusion barrier coatings on the cladding inner surface have been considered. In order to generate the required coating techniques, pack cementation, electroplating, and electrophoretic deposition have been investigated. However, these methods require a high processing temperature of above 700 oC, resulting in decarburization and decomposition of the martensites in a ferritic/martensitic (F/M) cladding steel. Alternatively, organometallic chemical vapor deposition (OMCVD) can be a promising process due to its low processing temperature of below 600 oC. The aim of the project is to conduct applied and fundamental research towards the development of diffusion barrier coatings on the inner surface of F/M fuel cladding tubes. Advanced cladding steels such as T91, HT9 and NF616 have been developed and extensively studied as advanced cladding materials due to their excellent irradiation and corrosion resistance. However, the FCCI accelerated by the elevated temperature and high neutron exposure anticipated in fast reactors, can have severe detrimental effects on the cladding steels through the diffusion of Fe into fuel and lanthanides towards into the claddings. To test the functionality of developed coating layer, the diffusion couple experiments were focused on using T91 as cladding and Ce as a surrogate lanthanum fission product. By using the customized OMCVD coating equipment, thin and compact layers with a few micron between 1.5 µm and 8 µm thick and average grain size of 200 nm and 5 µm were successfully obtained at the specimen coated between 300oC and

1982 Annual Status Report Plutonium Fuels and Actinide Programme

International Nuclear Information System (INIS)

Lindner, R.

1983-01-01

The programme of the Transuranium Institute has long included work on advanced fuels for fast breeder reactors. Study of the swelling of carbide and nitride fuels is now nearing completion, the retention of fission gases in bubbles of different sizes in the fuel having been quantified as function of burn-up and temperature. An important step forward has been achieved in the studies of the Equation of State of Nuclear Fuels up to 5000 K. Formation of some of the less abundant isotopes in PWR fuel has been determined experimentally. Aerosol formation during the fabrication of plutonium containing fuels, part of the activity Safe Handling of Plutonium Fuel has been studied. Head-End Processing of carbide fuels has continued experiments with high burn up mixed carbides. In the field of actinide research the preparation and characterisation of pure specimens is carried out. Effect of actinides on the properties of waste glasses is investigated

Fabrication of carbide and nitride pellets and the nitride irradiations Niloc 1 and Niloc 2

International Nuclear Information System (INIS)

Blank, H.

1991-01-01

Besides the relatively well-known advanced LMFBR mixed carbide fuel an advanced mixed nitride is also an attractive candidate for the optimised fuel cycle of the European Fast Reactor, but the present knowledge about the nitride is still insufficient and should be raised to the level of the carbide. For such an optimised fuel cycle the following general conditions have been set up for the fuel: (i) the burnup of the optimised MN and MC should be at least 15 a/o or even beyond, at moderate linear ratings of less than 75 kW/m (ii) the fuel will be used in a He-bonding pin concept and (iii) as far as available an advanced economic pellet fabrication method should be employed. (iv) The fuel structure must contain 15 - 20% porosity in order to accomodate the fission product swelling at high burnup. This report gives a comprehensive description of fuel and pellet fabrication and characterization, irradiation, and post-irradiation examination. From the results important conclusions can be drawn about future work on nitrides

Metal membrane-type 25-kW methanol fuel processor for fuel-cell hybrid vehicle

Science.gov (United States)

Han, Jaesung; Lee, Seok-Min; Chang, Hyuksang

A 25-kW on-board methanol fuel processor has been developed. It consists of a methanol steam reformer, which converts methanol to hydrogen-rich gas mixture, and two metal membrane modules, which clean-up the gas mixture to high-purity hydrogen. It produces hydrogen at rates up to 25 N m 3/h and the purity of the product hydrogen is over 99.9995% with a CO content of less than 1 ppm. In this fuel processor, the operating condition of the reformer and the metal membrane modules is nearly the same, so that operation is simple and the overall system construction is compact by eliminating the extensive temperature control of the intermediate gas streams. The recovery of hydrogen in the metal membrane units is maintained at 70-75% by the control of the pressure in the system, and the remaining 25-30% hydrogen is recycled to a catalytic combustion zone to supply heat for the methanol steam-reforming reaction. The thermal efficiency of the fuel processor is about 75% and the inlet air pressure is as low as 4 psi. The fuel processor is currently being integrated with 25-kW polymer electrolyte membrane fuel-cell (PEMFC) stack developed by the Hyundai Motor Company. The stack exhibits the same performance as those with pure hydrogen, which proves that the maximum power output as well as the minimum stack degradation is possible with this fuel processor. This fuel-cell 'engine' is to be installed in a hybrid passenger vehicle for road testing.

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

Institute of Scientific and Technical Information of China (English)

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

2017-01-01

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

Design of metallic bipolar plates for PEM fuel cells.

Science.gov (United States)

2012-01-01

This project focused on the design and production of metallic bipolar plates for use in PEM fuel cells. Different metals were explored : and stainless steel was found out to be best suited to our purpose. Following the selection of metal, it was calc...

Design study on metal fuel FBR cores

International Nuclear Information System (INIS)

Yokoo, T.; Tanaka, Y.; Ogata, T.

1991-01-01

A design approach for metal fuel FBR core to maintain fuel integrity during transient events by limiting eutectic/liquid phase formation is proposed based on the current status of metallic fuel development. Its impact as the limitation on the core outlet temperature is assessed through its application to two of CRIEPI's core concepts, high linear power 1000 MWe homogeneous design and medium linear power 300 MWe radially heterogeneous design. SESAME/SALT code is used in this study to analyze steady state and transient fuel behavior. SE2-FA code is developed based on SUPERENERGY-2 and used to analyze core thermal-hydraulics with uncertainties. As the result, the core outlet temperatures of both designs are found to be limited to ≤500degC if it is required to prevent eutectic/liquid phase formation during operational transients in order to guarantee the fuel integrity. Additional assessment is made assuming an advanced limiting condition that allows small liquid phase formation based on the liquid phase penetration rate derived from existing experimental results. The result indicates possibility of raising core outlet temperature to ∼ 530degC. Also, it is found that core design technology improvements such as hot spot factors reduction can contribute to the core outlet temperature extension by 10 ∼ 20degC. (author)

Fabrication of Metallic Fuel Slugs for Irradiation Experiments in Fast Breeder Test Reactor

International Nuclear Information System (INIS)

Saify, M.T.; Jha, S.K.; Abdulla, K.K.; Kumar, Arun; Prasad, G.J.

2013-01-01

Advantages of Metallic fuels for future FBR: → High heavy metal atom density; → Higher thermal conductivity at room temperature that increases with temperature; → Metal fuels can be relatively easily fabricated with close dimensional tolerances; → They have excellent compatibility with liquid metal coolants

Dimensional, microstructural and compositional stability of metal fuels

International Nuclear Information System (INIS)

Solomon, A.A.; Dayananda, M.A.

1993-01-01

The projects undertaken were to address two areas of concern for metal-fueled fast reactors: metallurgical compatibility of fuel and its fission products with the stainless steel cladding, and effects of porosity development in the fuel on fuel/cladding interactions and on sodium penetration in fuel. The following studies are reported on extensively in appendices: hot isostatic pressing of U-10Zr by coupled boundary diffusion/power law creep cavitation, liquid Na intrusion into porous U-10Zr fuel alloy by differential capillarity, interdiffusion between U-Zr fuel and selected Fe-Ni-Cr alloys, interdiffusion between U-Zr fuel vs selected cladding steels, and interdiffusion of Ce in Fe-base alloys with Ni or Cr

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

Energy Technology Data Exchange (ETDEWEB)

Nagle, Denis [Johns Hopkins Univ., Baltimore, MD (United States); Zhang, Dajie [Johns Hopkins Univ., Baltimore, MD (United States)

2015-10-22

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

Experimental specifications for eutectic reaction between metallic fuel and HT-9

International Nuclear Information System (INIS)

Hwang, Woan; Nam, Cheol; Lee, Byoung Oon; Ryu, Woo Seog

1998-10-01

The chemical interaction between metallic fuel and cladding is important in designing the fuel pin of the KALIMER. When metal fuel and cladding are contacted, the elements in fuel and cladding are inter-diffuse each other, forming the reaction layers at interface. The reaction layers may cause two important factors in aspects of fuel pin integrity. Firstly, it degrades cladding strength by reducing effective cladding thickness. Secondly, these layers accelerate eutectic reaction at transient conditions. To evaluate these phenomena, the diffusion couple experiment is planned by using metal fuels with various zirconium contents and HT-9 steel. The U-Zr fuel alloys will be used for the experiment with the different zirconium contents, these are 8, 10 and 12 weight %. This experiment aims to evaluate the effects of zirconium content on the chemical reaction. Furthermore, the reaction rate and threshold temperature of the eutectic melting will be determined as a function of the zirconium content. This document describes the detail experimental specifications for the eutectic reaction such as test setup, test requirements and test procedure. (author). 10 refs

Cumulative damage fraction design approach for LMFBR metallic fuel elements

International Nuclear Information System (INIS)

Johnson, D.L.; Einziger, R.E.; Huchman, G.D.

1979-01-01

The cumulative damage fraction (CDF) analytical technique is currently being used to analyze the performance of metallic fuel elements for proliferation-resistant LMFBRs. In this technique, the fraction of the total time to rupture of the cladding is calculated as a function of the thermal, stress, and neutronic history. Cladding breach or rupture is implied by CDF = 1. Cladding wastage, caused by interactions with both the fuel and sodium coolant, is assumed to uniformly thin the cladding wall. The irradiation experience of the EBR-II Mark-II driver fuel with solution-annealed Type 316 stainless steel cladding provides an excellent data base for testing the applicability of the CDF technique to metallic fuel. The advanced metal fuels being considered for use in LMFBRs are U-15-Pu-10Zr, Th-20Pu and Th-2OU (compositions are given in weight percent). The two cladding alloys being considered are Type 316 stainless steel and a titanium-stabilized Type 316 stainless steel. Both are in the cold-worked condition. The CDF technique was applied to these fuels and claddings under the assumed steady-state operating conditions

Development of advanced spent fuel management process. The fabrication and oxidation behavior of simulated metallized spent fuel

Energy Technology Data Exchange (ETDEWEB)

Ro, Seung Gy; Shin, Y.J.; You, G.S.; Joo, J.S.; Min, D.K.; Chun, Y.B.; Lee, E.P.; Seo, H.S.; Ahn, S.B

1999-03-01

The simulated metallized spent fuel ingots were fabricated and evaluated the oxidation rates and the activation energies under several temperature conditions to develop an advanced spent fuel management process. It was also checked the alloying characteristics of the some elements with metal uranium. (Author). 3 refs., 1 tab., 36 figs.

Metal Matrix Microencapsulated Fuel Technology for LWR Applications

International Nuclear Information System (INIS)

Terrani, Kurt A.; Bell, Gary L.; Kiggans, Jim; Snead, Lance Lewis

2012-01-01

An overview of the metal matrix microencapsulated (M3) fuel concept for the specific LWR application has been provided. Basic fuel properties and characteristics that aim to improve operational reliability, enlarge performance envelope, and enhance safety margins under design-basis accident scenarios are summarized. Fabrication of M3 rodlets with various coated fuel particles over a temperature range of 800-1300 C is discussed. Results from preliminary irradiation testing of LWR M3 rodlets with surrogate coated fuel particles are also reported.

The All Boron Carbide Diode Neutron Detector: Experiment and Modeling Approach

International Nuclear Information System (INIS)

Sabirianov, Ildar F.; Brand, Jennifer I.; Fairchild, Robert W.

2008-01-01

Boron carbide diode detectors, fabricated from two different polytypes of semiconducting boron carbide, will detect neutrons in reasonable agreement with theoretical expectations. The performance of the all boron carbide neutron detector differs, as expected, from devices where a boron rich neutron capture layer is distinct from the diode charge collection region (i.e. a conversion layer solid state detector). Diodes were fabricated from natural abundance boron (20% 10 B and 80% 11 B.) directly on the metal substrates and metal contacts applied to the films as grown. The total boron depth was on the order of 2 microns. This is clearly not a conversion-layer configuration. The diodes were exposed to thermal neutrons generated from a paraffin moderated plutonium-beryllium source in moderated and un-moderated, as well as shielded and unshielded experimental configurations, where the expected energy peaks at at 2.31 MeV and 2.8 MeV were clearly observed, albeit with some incomplete charge collection typical of thinner diode structures. The results are compared with other boron based thin film detectors and literature models. (authors)

Surface modification of the hard metal tungsten carbide-cobalt by boron ion implantation

International Nuclear Information System (INIS)

Mrotchek, I.

2007-01-01

In the present thesis ion beam implantation of boron is studied as method for the increasement of the hardness and for the improvement of the operational characteristics of cutting tools on the tungsten carbide-cobalt base. For the boron implantation with 40 keV energy and ∼5.10 17 ions/cm 2 fluence following topics were shown: The incoerporation of boron leads to a deformation and remaining strain of the WC lattice, which possesses different stregth in the different directions of the elementary cell. The maximum of the deformation is reached at an implantation temperature of 450 C. The segregation of the new phases CoWB and Co 3 W was detected at 900 C implantation temperature. At lower temperatures now new phases were found. The tribological characteristics of WC-Co are improved. Hereby the maxiaml effect was measured for implantation temperatures from 450 C to 700 C: Improvement of the microhardness by the factor 2..2.5, improvement of the wear resistance by the factor 4. The tribological effects extend to larger depths than the penetration depth of the boron implantation profile. The detected property improvements of the hard metal H3 show the possibility of a practical application of boron ion implantation in industry. The effects essential for a wer decreasement are a hardening of the carbide phase by deformation of the lattice, a hardening of the cobalt binding material and the phase boundaries because of the formation of a solid solution of the implanted boron atoms in Co and by this a blocking of the dislocation movement and the rupture spreading under load

Analytical chemistry methods for boron carbide absorber material. [Standard

Energy Technology Data Exchange (ETDEWEB)

DELVIN WL

1977-07-01

This standard provides analytical chemistry methods for the analysis of boron carbide powder and pellets for the following: total C and B, B isotopic composition, soluble C and B, fluoride, chloride, metallic impurities, gas content, water, nitrogen, and oxygen. (DLC)

Molecular-Dynamic Simulation In Substation Of Advanced Fuel With Improved Properties

Energy Technology Data Exchange (ETDEWEB)

Kolokol, Alexander S.; Shimkevich, Alexander L. [Russian Research Center ' Kurchatov Institute' , 1 Kurchatov Sq. Moscow 123182 (Russian Federation)

2008-07-01

A disadvantage of the uranium dioxide fuel is very low thermal conductivity than the one of nitride, carbide, metal fuel, and cermets as composites, UO{sub 2}+Me, due to the portion in thermal conductivity of their electronic conductivity and high phonon mobility. An investigation of the microstructure and atomic dynamics of solid solutions as well as the physical and chemical processes in them will make it possible to adjust the properties of the solutions in steps according to prescribed indicators by using alloying additives. The concept for designing an oxide fuel may be promising for the development of a new generation of nuclear reactors. In developing the methods for designing reactor materials as to the nuclear fuel, microscopic structure improving its thermal and physical properties is formulated here. (authors)

Molecular-Dynamic Simulation In Substation Of Advanced Fuel With Improved Properties

International Nuclear Information System (INIS)

Kolokol, Alexander S.; Shimkevich, Alexander L.

2008-01-01

A disadvantage of the uranium dioxide fuel is very low thermal conductivity than the one of nitride, carbide, metal fuel, and cermets as composites, UO 2 +Me, due to the portion in thermal conductivity of their electronic conductivity and high phonon mobility. An investigation of the microstructure and atomic dynamics of solid solutions as well as the physical and chemical processes in them will make it possible to adjust the properties of the solutions in steps according to prescribed indicators by using alloying additives. The concept for designing an oxide fuel may be promising for the development of a new generation of nuclear reactors. In developing the methods for designing reactor materials as to the nuclear fuel, microscopic structure improving its thermal and physical properties is formulated here. (authors)

Stoichiometric effects on performance of high-temperature gas-cooled reactor fuels from the U--C--O system

International Nuclear Information System (INIS)

Homan, F.J.; Lindemer, T.B.; Long, E.L. Jr.; Tiegs, T.N.; Beatty, R.L.

1977-01-01

Two fuel failure mechanisms were identified for coated particle fuels that are directly related to fuel kernel stoichiometry. These mechanisms are thermal migration of the kernel through the coating layers and chemical interaction between rare-earth fission products and the silicon carbide (SiC) layer leading to failure of the SiC layer. Thermal migration appears to be most severe for oxide fuels, while chemical interaction is most severe with carbide systems. Thermodynamic calculations indicated that oxide-carbide fuel kernels may permit a stoichiometry that reduces both problems to manageable levels for currently planned high-temperature gas-cooled reactors. Such stoichiometry adjustment is possible over the complete spectrum from UO 2 to UC 2 for the present recycle fuel, a weak acid resin (WAR)-derived fissile kernel. Thermodynamic calculations indicate that WAR kernels containing less than 15 percent UC 2 (greater than 85 percent UO 2 ) will develop excessive CO overpressures within the particle during irradiation. In 100 percent UO 2 particles, thermal migration and oxidation of the SiC layer were observed after irradiation. The calculations also indicate that WAR kernels containing greater than 70 percent UC 2 (less than 30 percent UC 2 ) contain insufficient oxygen to oxidize the rare-earth fission products formed in fuel operated to the maximum burnup levels of 75 percent fissions per initial metal atom (75 percent FIMA). Instead, the rare earths are present in part or completely as dicarbides. As such, they were observed to segregate from the kernel and collect at the SiC interface on the cold side of the particle, react with the SiC, and eventually fail this coating

Thermal conversion of biomass to valuable fuels, chemical feedstocks and chemicals

Science.gov (United States)

Peters, William A [Lexington, MA; Howard, Jack B [Winchester, MA; Modestino, Anthony J [Hanson, MA; Vogel, Fredreric [Villigen PSI, CH; Steffin, Carsten R [Herne, DE

2009-02-24

A continuous process for the conversion of biomass to form a chemical feedstock is described. The biomass and an exogenous metal oxide, preferably calcium oxide, or metal oxide precursor are continuously fed into a reaction chamber that is operated at a temperature of at least 1400.degree. C. to form reaction products including metal carbide. The metal oxide or metal oxide precursor is capable of forming a hydrolizable metal carbide. The reaction products are quenched to a temperature of 800.degree. C. or less. The resulting metal carbide is separated from the reaction products or, alternatively, when quenched with water, hydolyzed to provide a recoverable hydrocarbon gas feedstock.

Fission gas retention in irradiated metallic fuel

International Nuclear Information System (INIS)

Fenske, G.R.; Gruber, E.E.; Kramer, J.M.

1987-01-01

Theoretical calculations and experimental measurements of the quantity of retained fission gas in irradiated metallic fuel (U-5Fs) are presented. The calculations utilize the Booth method to model the steady-state release of gases from fuel grains and a simplified grain-boundary gas model to predict the gas release from intergranular regions. The quantity of gas retained in as-irradiated fuel was determined by collecting the gases released from short segments of EBR-II driver fuel that were melted in a gas-tight furnace. Comparison of the calculations to the measurements shows quantitative agreement with both the magnitude and the axial variation of the retained gas content

In-pile measurement of the thermal conductivity of irradiated metallic fuel

International Nuclear Information System (INIS)

Bauer, T.H.; Holland, J.W.

1995-01-01

Transient test data and posttest measurements from recent in-pile overpower transient experiments are used for an in situ determination of metallic fuel thermal conductivity. For test pins that undergo melting but remain intact, a technique is described that relates fuel thermal conductivity to peak pin power during the transient and a posttest measured melt radius. Conductivity estimates and their uncertainty are made for a database of four irradiated Integral Fast Reactor-type metal fuel pins of relatively low burnup (<3 at.%). In the assessment of results, averages and trends of measured fuel thermal conductivity are correlated to local burnup. Emphasis is placed on the changes of conductivity that take place with burnup-induced swelling and sodium logging. Measurements are used to validate simple empirically based analytical models that describe thermal conductivity of porous media and that are recommended for general thermal analyses of irradiated metallic fuel

Complex plasmochemical processing of solid fuel

Directory of Open Access Journals (Sweden)

Vladimir Messerle

2012-12-01

Full Text Available Technology of complex plasmaochemical processing of solid fuel by Ecibastuz bituminous and Turgay brown coals is presented. Thermodynamic and experimental study of the technology was fulfilled. Use of this technology allows producing of synthesis gas from organic mass of coal and valuable components (technical silicon, ferrosilicon, aluminum and silicon carbide and microelements of rare metals: uranium, molybdenum, vanadium etc. from mineral mass of coal. Produced a high-calorific synthesis gas can be used for methanol synthesis, as high-grade reducing gas instead of coke, as well as energy gas in thermal power plants.

Tungsten carbide and tungsten-molybdenum carbides as automobile exhaust catalysts

International Nuclear Information System (INIS)

Leclercq, L.; Daubrege, F.; Gengembre, L.; Leclercq, G.; Prigent, M.

1987-01-01

Several catalyst samples of tungsten carbide and W, Mo mixed carbides with different Mo/W atom ratios, have been prepared to test their ability to remove carbon monoxide, nitric oxide and propane from a synthetic exhaust gas simulating automobile emissions. Surface characterization of the catalysts has been performed by X-ray photoelectron spectroscopy (XPS) and selective chemisorption of hydrogen and carbon monoxide. Tungsten carbide exhibits good activity for CO and NO conversion, compared to a standard three-way catalyst based on Pt and Rh. However, this W carbide is ineffective in the oxidation of propane. The Mo,W mixed carbides are markedly different having only a very low activity. 9 refs.; 10 figs.; 5 tabs

Development status of metallic, dispersion and non-oxide advanced and alternative fuels for power and research reactors

International Nuclear Information System (INIS)

2003-09-01

The current thermal power reactors use less than 1% of the energy contained in uranium. Long term perspectives aiming at a better economical extraction of the potential supplied by uranium motivated the development of new reactor types and, of course, new fuel concepts. Most of them dated from the sixties including liquid metal cooled fast (FR) and high temperature gas cooled (HTGR) reactors. Unfortunately, these impulses slowed down during the last twenty years; nuclear energy had to face political and consensus problems, in particular in the United States of America and in Europe, resulting from the consequences of the TMI and Chernobyl accidents. Good economical results obtained by the thermal power reactors also contributed to this process. During the last twenty years mainly France, India, Japan and the Russian Federation have maintained a relatively high level of technological development with appropriate financial items, in particular, in fuel research for the above mentioned reactor types. China and South Africa are now progressing in development of FR/HTGR and HTGR technologies, respectively. The purpose of this report is not only to summarise knowledge accumulated in the fuel research since the beginning of the sixties. This subject has been well covered in literature up to the end of the eighties. This report rather concentrates on the 'advanced fuels 'for the current different types of reactors including metallic, carbide and nitride fuels for fast reactors, so-called 'cold' fuels and fuels to burn excessive ex-weapons plutonium in thermal power reactors, alternative fuels for small size and research reactors. Emphasis has been put on the aspects of fabrication and irradiation behaviour of these fuels; available basic data concerning essential properties that help to understand the phenomena have been mentioned as well. This report brings complementary information to the earlier published monographs and concerns developments carried out after the early

Characterization and performances of cobalt-tungsten and molybdenum-tungsten carbides as anode catalyst for PEFC

International Nuclear Information System (INIS)

Izhar, Shamsul; Yoshida, Michiko; Nagai, Masatoshi

2009-01-01

The preparation of carbon-supported cobalt-tungsten and molybdenum-tungsten carbides and their activity as an anode catalyst for a polymer electrolyte fuel cell were investigated. The electrocatalytic activity for the hydrogen oxidation reaction over the catalysts was evaluated using a single-stack fuel cell and a rotating disk electrode. The characterization of the catalysts was performed by XRD, temperature-programmed carburization, temperature-programmed reduction and X-ray photoelectron spectroscopy. The maximum power densities of the 30 wt% 873 K-carburized cobalt-tungsten and molybdenum-tungsten mixed with Ketjen carbon (cobalt-tungsten carbide (CoWC)/Ketjen black (KB) and molybdenum-tungsten carbide (MoWC)/KB) were 15.7 and 12.0 mW cm -2 , respectively, which were 14 and 11%, compared to the in-house membrane electrode assembly (MEA) prepared from a 20 wt% Pt/C catalyst. The CoWC/KB catalyst exhibited the highest maximum power density compared to the MoWC/KB and WC/KB catalysts. The 873 K-carburized CoW/KB catalyst formed the oxycarbided and/or carbided CoW that are responsible for the excellent hydrogen oxygen reaction

Laser cladding in-situ carbide particle reinforced Fe-based composite coatings with rare earth oxide addition

Institute of Scientific and Technical Information of China (English)

吴朝锋; 马明星; 刘文今; 钟敏霖; 张红军; 张伟明

2009-01-01

Particulate reinforced metal matrix composite(PR-MMC) has excellent properties such as good wear resistance,corrosion resistance and high temperature properties.Laser cladding is usually used to form PR-MMC on metal surface with various volume fractions of ceramic particles.Recent literatures showed that laser melting of powder mixture containing carbon and carbide-forming elements,was favorable for the formation of in-situ synthesized carbide particles.In this paper,rare earth oxide(RE2O3) was added into t...

Microstructural evaluation of the NbC-20Ni cemented carbides during sintering

International Nuclear Information System (INIS)

Rodrigues, D.; Cannizza, E.

2016-01-01

Full text: Fine carbides in a metallic matrix (binder) form the microstructure of the cemented carbides. Grain size and binder content are the main variables to adjust hardness and toughness. These products are produced by Powder Metallurgy, and traditional route involves mixing carbides with binder by high energy milling, pressing and sintering. During sintering, a liquid phase promotes densification, and a final relative density higher than 99% is expected. Sintering is carried out at high temperatures, and dissolution of the carbides changes the chemical composition of the binder. To control grain growth of the main carbide, which reduces hardness, small quantities of secondary carbides are used. These additives limit dissolution and precipitation of the main carbides reducing the final grain size. This paper focused the structural and chemical evolution during sintering using NbC-20Ni cermets. Mixtures of very fine NbC carbides and carbonyl Ni powders were produce by intense milling. These mixtures were pressed using uniaxial pressures from 50 to 200MPa. Shrinkage was evaluated using dilatometric measurements under an atmosphere of dynamic argon. Samples were also sintered under vacuum in high temperature industrial furnace. The sintered samples were characterized in terms of density hardness, toughness and microstructure. DRX was the main tool used to evaluate the structural evolution of the binder. In situ chemical analysis helped to understand the dissolution mechanisms. (author)

Economic Viability of Metallic Sodium-Cooled Fast Reactor Fuel in Korea

Directory of Open Access Journals (Sweden)

S. K. Kim

2013-01-01

Full Text Available This paper evaluates whether SFR metallic nuclear fuel can be economical. To make this determination, the cost of SFCF (SFR fuel cycle facilities was estimated, and the break-even point of the manufacturing cost of SFR metallic nuclear fuel for direct disposal option was then calculated. As a result of the cost estimation, the levelized unit cost (LUC for SFCF was calculated to be 5,311 $/kgHM, and the break-even point was calculated to be $5,267/kgHM. Therefore, the cost difference between LUC and the break-even point is not only small but is also within the relevant range of the uncertainty level of Class 3 in accordance with a generic cost estimate classification matrix of AACE (the Association for the Advancement of Cost Engineering. This means it is very difficult to judge the economical feasibility of SFR metallic nuclear fuel because as of today there are no commercial facilities in Korea or the world. The economic feasibility of SFR metallic nuclear fuel, however, will be enhanced if the mass production of SFCF becomes possible in the future.

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

International Nuclear Information System (INIS)

Grabaskas, David; Bucknor, Matthew; Jerden, James

2016-01-01

The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

Energy Technology Data Exchange (ETDEWEB)

Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States); Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-02-01

The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

The significance of strength of silicon carbide for the mechanical integrity of coated fuel particles for HTRs

International Nuclear Information System (INIS)

Bongartz, K.; Scheer, A.; Schuster, H.; Taeuber, K.

1975-01-01

Silicon carbide (SiC) and pyrocarbon are used as coating material for the HTR fuel particles. The PyC shell having a certain strength acts as a pressure vessel for the fission gases whereas the SiC shell has to retain the solid fission products in the fuel kernel. For measuring the strength of coating material the so-called Brittle Ring Test was developed. Strength and Young's modulus can be measured simultaneously with this method on SiC or PyC rings prepared out of the coating material of real fuel particles. The strength measured on the ring under a certain stress distribution which is characteristic for this method is transformed with the aid of the Weibull formalism for brittle fracture into the equivalent strength of the spherical coating shell on the fuel particle under uniform stress caused by the fission gas pressure. The values measured for the strength of the SiC were high (400-700MN/m 2 ), it could therefore be assumed that a SiC layer might contribute significantly also to the mechanical strength of the fuel coating. This assumption was confirmed by an irradiation test on coated particles with PyC-SiC-PyC coatings. There were several particles with all PyC layers broken during the irradiation, whereas the SiC layers remained intact having to withstand the fission gas pressure alone. This fact can only be explained assuming that the strength of the SiC is within the range of the values measured with the brittle ring test. The result indicates that, in optimising the coating of a fuel particle, the PyC layers of a multilayer coating should be considered alone as prospective layers for the SiC. The SiC shell, besides acting as a fission product barrier, is then also responsible for the mechanical integrity of the particle

Development and testing of metallic fuels with high minor actinide content

International Nuclear Information System (INIS)

Meyer, M.K.; Hayes, S.L.; Kennedy, J.R.; Keiser, D.D.; Hilton, B.A.; Frank, S.M.; Kim, Y.-S.; Chang, G.; Ambrosek, R.G.

2003-01-01

Metallic alloys are promising candidates for use as fuels for transmutation and in advanced closed nuclear cycles. Metallic alloys have high heavy metal atom density, relatively high thermal conductivity, favorable gas release behavior, and lend themselves to remote recycle processes. Both non-fertile and uranium-bearing metal fuels containing minor actinide are under consideration for use as transmutation fuels by the U.S. Advanced Fuel Cycle (AFC) program, however, little irradiation performance data exists for fuel forms containing significant fractions of minor actinides. The first irradiation tests of non-fertile high-actinide-content fuels are scheduled to begin in early 2003 in the Advanced Test Reactor (ATR). The irradiation test matrix was designed to provide basic information on the irradiation behavior of binary Pu-Zr alloy fuel and the effect of the minor actinides americium and neptunium on alloy fuel behavior, together and separately. Five variants of transuranic containing zirconium-based alloy fuels are included in the AFC-1 irradiation test matrix. These are (in wt.%) Pu-40Zr, Pu-60Zr, Pu-12Am-40Zr, Pu-10Np-40Zr and Pu-10Np-10Am-40Zr. PuN-ZrN based fuels containing Am and Np are also included. All five of the fuel alloys have been fabricated in the form of cylindrical fuel slugs by arc-casting. Short melt times, on the order or 5-20 seconds, prevent the volatilization of significant quantities of americium metal, despite the high melt temperatures characteristic of the arc-melting process. Alloy microstructure have been characterized by x-ray diffraction and scanning electron microscopy. Thermal analysis has also been performed. The AFC-1 irradiation experiment configuration consists of twenty-four sodium bonded fuel specimens sealed in helium filled secondary capsules. The first capsule has a design burnup to 7 at.% 239 Pu; goal peak burnup of the second capsule is ∼18 at%. Capsule assemblies are placed within an aluminum flow-through basket

Evaluation of the use of metal alloy fuels in pressurized water reactors

International Nuclear Information System (INIS)

1990-01-01

The project concentrated on model development. Reactor physics modeling involved establishing accurate models with PC versions of COMBINE and VENTURE. Fuel performance analysis will start with METAL- LIFE. In order to justify the change of fuel to metal alloy, large benefits will have to be found; the cost benefit reported is not sufficient. The fuel pin will be annular and contact the clad; the clad thickness will force the fuel to grow toward the central hole. This report reports: design improvements, neutronic model development, COBRA modifications, reactor kinetics model development, RELAP code, and fuel performance

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

Science.gov (United States)

Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2018-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are discussed. The authors demonstrated success in reaching desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and define a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

Response of Salix alba L. to heavy metals and diesel fuel ...

African Journals Online (AJOL)

Response of Salix alba L. to heavy metals and diesel fuel contamination. ... Ni and Pb led to reduced accumulation of Pb. Presence of 5 g/kg of diesel fuel in soil significantly increased toxic influence of applied heavy metals by further reducing ... Key words: Cd, Ni, Pb, phytoremediation, phytoextraction, willow, Salix alba.

Contribution to the study of zirconium self-diffusion in zirconium carbide

International Nuclear Information System (INIS)

An, Chul

1972-01-01

The objective of this research thesis is to determine experimental conditions allowing the measurement of the self-diffusion coefficient of zirconium in zirconium carbide. The author reports the development of a method of preparation of zirconium carbide samples. He reports the use of ion implantation as technique to obtain a radio-tracer coating. The obtained results give evidence of the impossibility to use sintered samples with small grains because of the demonstrated importance of intergranular diffusion. The self-diffusion coefficient is obtained in the case of zirconium carbide with grains having a diameter of few millimetres. The presence of 95 Nb from the disintegration of 95 Zr indicates that these both metallic elements have very close diffusion coefficients at 2.600 C [fr

Radiation stability of proton irradiated zirconium carbide

International Nuclear Information System (INIS)

Yang, Yong; Dickerson, Clayton A.; Allen, Todd R.

2009-01-01

The use of zirconium carbide (ZrC) is being considered for the deep burn (DB)-TRISO fuel as a replacement for the silicon carbide coating. The radiation stability of ZrC was studied using 2.6 MeV protons, across the irradiation temperature range from 600 to 900degC and to doses up to 1.75 dpa. The microstructural characterization shows that the irradiated microstructure is comprised of a high density of nanometer-sized dislocation loops, while no irradiation induced amorphization or voids are observed. The lattice expansion induced by point defects is found to increase as the dose increases for the samples irradiated at 600 and 800degC, while for the 900degC irradiation, a slight lattice contraction is observed. The radiation hardening is also quantified using a micro indentation technique for the temperature and doses studies. (author)

Fuel assembly for FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki.

1995-01-01

Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

Degradation of solid oxide fuel cell metallic interconnects in fuels containing sulfur

Energy Technology Data Exchange (ETDEWEB)

Ziomek-Moroz, M.; Hawk, Jeffrey A.

2005-01-01

Hydrogen is the main fuel for all types of fuel cells except direct methanol fuel cells. Hydrogen can be generated from all manner of fossil fuels, including coal, natural gas, diesel, gasoline, other hydrocarbons, and oxygenates (e.g., methanol, ethanol, butanol, etc.). Impurities in the fuel can cause significant performance problems and sulfur, in particular, can decrease the cell performance of fuel cells, including solid oxide fuel cells (SOFC). In the SOFC, the high (800-1000°C) operating temperature yields advantages (e.g., internal fuel reforming) and disadvantages (e.g., material selection and degradation problems). Significant progress in reducing the operating temperature of the SOFC from ~1000 ºC to ~750 ºC may allow less expensive metallic materials to be used for interconnects and as balance of plant (BOP) materials. This paper provides insight on the material performance of nickel, ferritic steels, and nickel-based alloys in fuels containing sulfur, primarily in the form of H2S, and seeks to quantify the extent of possible degradation due to sulfur in the gas stream.

Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

International Nuclear Information System (INIS)

Karahan, Aydın; Kazimi, Mujid S.

2013-01-01

The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors

Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydın, E-mail: karahan@alum.mit.edu; Kazimi, Mujid S.

2013-10-15

The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

Thermal bonding of light water reactor fuel using nonalkaline liquid-metal alloy

International Nuclear Information System (INIS)

Wright, R.F.; Tulenko, J.S.; Schoessow, G.J.; Connell, R.G. Jr.; Dubecky, M.A.; Adams, T.

1996-01-01

Light water reactor (LWR) fuel performance is limited by thermal and mechanical constraints associated with the design, fabrication, and operation of fuel in a nuclear reactor. A technique is explored that extends fuel performance by thermally bonding LWR fuel with a nonalkaline liquid-metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Because of the low thermal conductivity of the gas, the gas space thermally insulates the fuel pellets from the reactor coolant outside the fuel rod, elevating the fuel temperatures. Filling the gap between the fuel and clad with a high-conductivity liquid metal thermally bonds the fuel to the cladding and eliminates the large temperature change across the gap while preserving the expansion and pellet-loading capabilities. The application of liquid-bonding techniques to LWR fuel is explored to increase LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) is developed to analyze the in-reactor performance of the liquid-metal-bonded fuel. An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of liquid-bonded LWR fuel. The results show that liquid-bonded boiling water reactor peak fuel temperatures are 400 F lower at beginning of life and 200 F lower at end of life compared with conventional fuel

Characterization of IFR metal fuel fragmentation

International Nuclear Information System (INIS)

Gabor, J.D.; Purviance, R.T.; Aeschlimann, R.W.; Spencer, B.W.

1987-01-01

The integral fast reactor (IFR) employs a reactor design that has inherent safety features. An important safety advantage is derived from its pool configuration, which facilitates passive decay heat removal and isolates the core from accidents that might occur elsewhere in the plant. The metal-alloy fuel has superior heat transfer properties compared to oxide fuels. While the IFR design has these inherent safety features, a complete analysis of reactor safety requires assessment of the consequences of the melting of the uranium alloy fuel in the core and the contact of molten core materials with sodium. A series of eight tests was conducted in which the fragmentation and interaction behavior of kilogram quantities of uranium-zirconium alloy in sodium was studied

Geochemical Interactions in failed Co-Disposal Waste Packages for N Reactor and Ft. St. Vrain Spent Fuel and the Melt and Dilute Waste Form

International Nuclear Information System (INIS)

Arthur, S.E.; McNeish, J.

2002-01-01

The objective of this scientific analysis is to calculate the long-term geochemical behavior in a failed co-disposal waste package (WP) containing U. S. Department of Energy (DOE) spent nuclear fuel (SNF) and high level waste (HLW) glass. This analysis was prepared according to a Technical Work Plan (BSC 2002). Specifically the scope of these calculations is to determine: (1) The geochemical characteristics of the fluids inside the WP after breach, including the corrosion/dissolution of the initial WP configuration; (2) The transport of radionuclides of concern to performance assessment out of the degraded WP by infiltrating water; and (3) The range of parameter variation for additional laboratory and numerical evaluations. This analysis is limited to three SNF groups, uranium (U)/thorium (Th) carbide SNF (Group 5), U metal SNF (Group 7), and aluminum(Al)-based fuels (Group 9). Group 5 is represented by Ft. St. Vrain (FSV) U/Th carbide SNF, Group 7 is represented by N-Reactor U metal SNF, and Group 9 is represented by the Melt and Dilute (MandD) waste form developed from Al-based SNF. The DOE (2001a, Appendix A) describes all of these fuels. Table 1 shows the groups of DOE SNF, the representative SNF for each group, and the metric tons of heavy metal (MTHM) of SNF in each group

Hafnium carbide formation in oxygen deficient hafnium oxide thin films

Energy Technology Data Exchange (ETDEWEB)

Rodenbücher, C. [Forschungszentrum Jülich GmbH, Peter Grünberg Institute (PGI-7), JARA-FIT, 52425 Jülich (Germany); Hildebrandt, E.; Sharath, S. U.; Kurian, J.; Komissinskiy, P.; Alff, L. [Technische Universität Darmstadt, Institute of Materials Science, 64287 Darmstadt (Germany); Szot, K. [Forschungszentrum Jülich GmbH, Peter Grünberg Institute (PGI-7), JARA-FIT, 52425 Jülich (Germany); University of Silesia, A. Chełkowski Institute of Physics, 40-007 Katowice (Poland); Breuer, U. [Forschungszentrum Jülich GmbH, Central Institute for Engineering, Electronics and Analytics (ZEA-3), 52425 Jülich (Germany); Waser, R. [Forschungszentrum Jülich GmbH, Peter Grünberg Institute (PGI-7), JARA-FIT, 52425 Jülich (Germany); RWTH Aachen, Institute of Electronic Materials (IWE 2), 52056 Aachen (Germany)

2016-06-20

On highly oxygen deficient thin films of hafnium oxide (hafnia, HfO{sub 2−x}) contaminated with adsorbates of carbon oxides, the formation of hafnium carbide (HfC{sub x}) at the surface during vacuum annealing at temperatures as low as 600 °C is reported. Using X-ray photoelectron spectroscopy the evolution of the HfC{sub x} surface layer related to a transformation from insulating into metallic state is monitored in situ. In contrast, for fully stoichiometric HfO{sub 2} thin films prepared and measured under identical conditions, the formation of HfC{sub x} was not detectable suggesting that the enhanced adsorption of carbon oxides on oxygen deficient films provides a carbon source for the carbide formation. This shows that a high concentration of oxygen vacancies in carbon contaminated hafnia lowers considerably the formation energy of hafnium carbide. Thus, the presence of a sufficient amount of residual carbon in resistive random access memory devices might lead to a similar carbide formation within the conducting filaments due to Joule heating.

Characterisation of nuclear dispersion fuels. The non-destructive examination of silicon carbide by selenium immersion

Energy Technology Data Exchange (ETDEWEB)

Ambler, J.F.R.; Ferguson, I.F.

1974-07-15

The non-destructive microscopic examination of silicon-carbide-coated spheres containing uranium carbide, which involves immersing the coated spheres in selenium, is particularly suited for the examination of flaws in the coats but it is not possible to measure coating thicknesses by this method. Some coats are found to be opaque and this is related to their porosity. (auth)

Combined Photoemission Spectroscopy and Electrochemical Study of a Mixture of (Oxy)carbides as Potential Innovative Supports and Electrocatalysts.

Science.gov (United States)

Calvillo, Laura; Valero-Vidal, Carlos; Agnoli, Stefano; Sezen, Hikmet; Rüdiger, Celine; Kunze-Liebhäuser, Julia; Granozzi, Gaetano

2016-08-03

Active and stable non-noble metal materials, able to substitute Pt as catalyst or to reduce the Pt amount, are vitally important for the extended commercialization of energy conversion technologies, such as fuel cells and electrolyzers. Here, we report a fundamental study of nonstoichiometric tungsten carbide (WxC) and its interaction with titanium oxycarbide (TiOxCy) under electrochemical working conditions. In particular, the electrochemical activity and stability of the WxC/TiOxCy system toward the ethanol electrooxidation reaction (EOR) and hydrogen evolution reaction (HER) are investigated. The chemical changes caused by the applied potential are established by combining photoemission spectroscopy and electrochemistry. WxC is not active toward the ethanol electrooxidation reaction at room temperature but it is highly stable under these conditions thanks to the formation of a passive thin film on the surface, consisting mainly of WO2 and W2O5, which prevents the full oxidation of WxC. In addition, WxC is able to adsorb ethanol, forming ethoxy groups on the surface, which constitutes the first step for the ethanol oxidation. The interaction between WxC and TiOxCy plays an important role in the electrochemical stability of WxC since specific orientations of the substrate are able to stabilize WxC and prevent its corrosion. The beneficial interaction with the substrate and the specific surface chemistry makes tungsten carbide a good electrocatalyst support or cocatalyst for direct ethanol fuel cells. However, WxC is active toward the HER and chemically stable under hydrogen reduction conditions, since no changes in the chemical composition or dissolution of the film are observed. This makes tungsten carbide a good candidate as electrocatalyst support or cocatalyst for the electrochemical production of hydrogen.

Carbides precipitated from the melt in a Zr-2.5 Nb alloy

International Nuclear Information System (INIS)

Piotrkowski, R.; Garcia, E.A.; Vigna, G.L.; Bermudez, S.E.

1993-01-01

An experimental method is presented which leads to the formation of carbides similar in size (3 to 8 microns) and composition to those observed in some pressure tubes of CANDU type reactors. The method is based on melting the Zr-2.5 Nb alloy in a graphite crucible, where isothermal C diffusion in the Zr-Nb melt took place. It can be inferred that the carbides observed in pressure tubes could be originated in high temperature stages of the manufacture process. Otherwise, they could have been incorporated in the Zr sponge. As a result of the diffusion couple Liquid Zr-2.5 Nb/Solid Graphite, a carbide layer, up to 100μm thick, grew attached to the crucible wall, together with carbide particles whose size was in the some microns range. The smallest particles were arranged in rows determined by the prior β phase grains. The main carbide phase detected was the cubic MC 1-x ; the hexagonal M 2 C was also detected; M for metal. (Author)

Natural uranium metallic fuel elements: fabrication and operating experience

International Nuclear Information System (INIS)

Hammad, F.H.; Abou-Zahra, A.A.; Sharkawy, S.W.

1980-01-01

The main reactor types based on natural uranium metallic fuel element, particularly the early types, are reviewed in this report. The reactor types are: graphite moderated air cooled, graphite moderated gas cooled and heavy water moderated reactors. The design features, fabrication technology of these reactor fuel elements and the operating experience gained during reactor operation are described and discussed. The interrelation between operating experience, fuel design and fabrication was also discussed with emphasis on improving fuel performance. (author)

Magnetic susceptibility as a method of investigation of short-range order in strongly nonstoichiometric carbides

International Nuclear Information System (INIS)

Nazarova, S.Z.; Gusev, A.I.

2001-01-01

Magnetic susceptibility in disordered and ordered carbides of transition metals (M = Ti, Zr, Hf, Nb, Ta) was studied, the results are generalized. It was ascertained that the change in carbide susceptibility induced by deviation from stoichiometry stems from specific features of electronic spectra of the compounds. The use of magnetic susceptibility for determining structural disorder-order transitions is discussed. It is shown that change in the contribution made by orbital paramagnetism, resulting from short-range order formation, is the reason of decrease in susceptibility of nonstoichiometric carbides during the ordering. Experimentally obtained data on susceptibility permitted evaluating short- and far-range order parameters in NbC y , TaC y , TiC y and HfC y carbides [ru

Silicon-Carbide Power MOSFET Performance in High Efficiency Boost Power Processing Unit for Extreme Environments

Science.gov (United States)

Ikpe, Stanley A.; Lauenstein, Jean-Marie; Carr, Gregory A.; Hunter, Don; Ludwig, Lawrence L.; Wood, William; Del Castillo, Linda Y.; Fitzpatrick, Fred; Chen, Yuan

2016-01-01

Silicon-Carbide device technology has generated much interest in recent years. With superior thermal performance, power ratings and potential switching frequencies over its Silicon counterpart, Silicon-Carbide offers a greater possibility for high powered switching applications in extreme environment. In particular, Silicon-Carbide Metal-Oxide- Semiconductor Field-Effect Transistors' (MOSFETs) maturing process technology has produced a plethora of commercially available power dense, low on-state resistance devices capable of switching at high frequencies. A novel hard-switched power processing unit (PPU) is implemented utilizing Silicon-Carbide power devices. Accelerated life data is captured and assessed in conjunction with a damage accumulation model of gate oxide and drain-source junction lifetime to evaluate potential system performance at high temperature environments.

Electrometallurgical treatment of metallic spent nuclear fuel stored at the Hanford Site

International Nuclear Information System (INIS)

Laidler, J.J.; Gay, E.C.

1996-01-01

The major component of the DOE spent nuclear fuel inventory is the metallic fuel stored at the Hanford site in the southeastern part of the state of Washington. Most of this fuel was discharged from the N-Reactor; a small part of the inventory is fuel from the early Hanford production reactors. The U.S. Department of Energy (DOE) plans to remove these fuels from the spent fuel storage pools in which they are presently stored, dry them, and place them in interim storage at a location at the Hanford site that is far removed from the Columbia River. It is not yet certain that these fuels will be acceptable for disposal in a mined geologic repository without further treatment, due to their potential pyrophoric character. A practical method for treatment of the Hanford metallic spent fuel, based on an electrorefining process, has been developed and has been demonstrated with unirradiated N-Reactor fuel and with simulated single-pass reactor (SPR) spent fuel. The process can be operated with any desired throughput rates; being a batch process, it is simply a matter of setting the size of the electrorefiner modules and the number of such modules. A single module, prototypic of a production-scale module, has been fabricated and testing is in progress at a throughput rate of 150 kg (heavy metal) per day. The envisioned production version would incorporate additional anode baskets and cathode tubes and provide a throughput rate of 333 kgHM/day. A system with four of these modules would permit treatment of Hanford metallic fuels at a rate of at least 250 metric tons per year

Evaluation of the mechanical performance of silicon carbide in TRISO fuel at high temperatures

Energy Technology Data Exchange (ETDEWEB)

Rohbeck, Nadia, E-mail: nadia.rohbeck@manchester.ac.uk; Xiao, Ping, E-mail: p.xiao@manchester.ac.uk

2016-09-15

The HTR design envisions fuel operating temperatures of up to 1000 °C and in case of an accident even 1600 °C are conceivable. To ensure safety in all conditions a thorough understanding of the impact of an extreme temperature environment is necessary. This work assesses the high temperature mechanical performance of the silicon carbide (SiC) layer within the tristructural-isotropic (TRISO) fuel particle as it poses the main barrier against fission product release into the primary circuit. Therefore, simulated fuel was fabricated by fluidized bed chemical vapour deposition; varying the deposition conditions resulted in strongly differing SiC microstructures for the various samples. Subsequently the TRISO particles were annealed in inert atmosphere at temperatures ranging from 1600 °C up to 2200 °C. Scanning electron microscopy and Raman spectroscopy showed that strong disintegration of the SiC layer occurred from 2100 °C onwards, but initial signs of porosity formation were visible already at 1800 °C. Still, the elastic modulus and hardness as measured by nanoindentation were hardly impaired. After annealing stoichiometric SiC coatings showed a reduction in fracture strength as determined by a modified crush test, however the actual annealing temperature from 1600 °C to 2000 °C had no measureable effect. Furthermore, a technique was developed to measure the elastic modulus and hardness in situ up to 500 °C using a high temperature nanoindentation facility. This approach allows conducting tests while the specimen and indenter tip are heated to a specific measurement temperature, thus obtaining reliable values for the temperature dependent mechanical properties of the material. For the SiC layer in TRISO particles it was found that the elastic modulus decreased slightly from room temperature up to 500 °C, whereas the hardness was reduced more severely to approximately half of its ambient temperature value.

Evaluation of the Mechanical Performance of Silicon Carbide in TRISO Fuel at High Temperatures

International Nuclear Information System (INIS)

Rohbeck, N.; Xiao, P.

2014-01-01

The HTR design envisions fuel operating temperatures of up to 1000°C and in case of an accident even 1600°C are conceivable. To ensure safety in all conditions a thorough understanding of the impact of an extreme temperature environment is necessary. This work assesses the high temperature mechanical performance of the silicon carbide (SiC) layer within the tristructural-isotropic (TRISO) fuel particle as it poses the main barrier against fission product release into the primary circuit. Therefore simulated fuel was fabricated by fluidized bed chemical vapour deposition; varying the deposition conditions resulted in strongly differing SiC microstructures for the various samples. Subsequently the TRISO particles were annealed in inert atmosphere at temperatures ranging from 1600°C up to 2200°C. Scanning electron microscopy and Raman spectroscopy showed that strong disintegration of the SiC layer occurred from 2100°C onwards, but initial signs of porosity formation were visible already at 1800°C. Still, the elastic modulus and hardness as measured by nanoindentation were hardly impaired. After annealing stoichiometric SiC coatings showed a reduction in fracture strength as determined by a modified crush test, however the actual annealing temperature from 1600°C to 2000°C had no measureable effect. Furthermore, a technique was developed to measure the elastic modulus and hardness in-situ up to 500°C using a high temperature nanoindentation facility. This approach allows conducting numerous tests on small sample volumes and thus promises to improve our knowledge of irradiation effects on the mechanical properties. For the SiC layer in TRISO particles it was found that the elastic modulus decreased slightly from room temperature up to 500°C, whereas the hardness was reduced more severely to approximately half of its ambient temperature value. (author)

Evaluation of the mechanical performance of silicon carbide in TRISO fuel at high temperatures

International Nuclear Information System (INIS)

Rohbeck, Nadia; Xiao, Ping

2016-01-01

The HTR design envisions fuel operating temperatures of up to 1000 °C and in case of an accident even 1600 °C are conceivable. To ensure safety in all conditions a thorough understanding of the impact of an extreme temperature environment is necessary. This work assesses the high temperature mechanical performance of the silicon carbide (SiC) layer within the tristructural-isotropic (TRISO) fuel particle as it poses the main barrier against fission product release into the primary circuit. Therefore, simulated fuel was fabricated by fluidized bed chemical vapour deposition; varying the deposition conditions resulted in strongly differing SiC microstructures for the various samples. Subsequently the TRISO particles were annealed in inert atmosphere at temperatures ranging from 1600 °C up to 2200 °C. Scanning electron microscopy and Raman spectroscopy showed that strong disintegration of the SiC layer occurred from 2100 °C onwards, but initial signs of porosity formation were visible already at 1800 °C. Still, the elastic modulus and hardness as measured by nanoindentation were hardly impaired. After annealing stoichiometric SiC coatings showed a reduction in fracture strength as determined by a modified crush test, however the actual annealing temperature from 1600 °C to 2000 °C had no measureable effect. Furthermore, a technique was developed to measure the elastic modulus and hardness in situ up to 500 °C using a high temperature nanoindentation facility. This approach allows conducting tests while the specimen and indenter tip are heated to a specific measurement temperature, thus obtaining reliable values for the temperature dependent mechanical properties of the material. For the SiC layer in TRISO particles it was found that the elastic modulus decreased slightly from room temperature up to 500 °C, whereas the hardness was reduced more severely to approximately half of its ambient temperature value.

Fission gas retention in irradiated metallic fuel

International Nuclear Information System (INIS)

Fenske, G.R.; Gruber, E.; Kramer, J.M.

1987-01-01

Theoretical calculations and experimental measurements of the quantity of retained fission gas in irradiated metallic fuel (U-5 wt. % Fs) are presented. (The symbol 'Fs' designates fissium, a 'pseudo-element' which, in reality, is an alloy whose composition is representative of fission products that remain in reprocessed fuel). The calculations utilize the Booth method to model the steady-state release of gases from fuel grains and a simplified grain-boundary gas model to predict the gas release from intergranular regions. The quantity of gas retained in as-irradiated fuel was determined by collecting the gases released from short segments of EBR-II driver fuel that were melted in a gas-tight furnace. Comparison of the calculations with the measurements shows quantitative agreement in both the magnitude and the axial variation of the retained gas content. (orig.)

Method for increasing the activity of fuel cell electrodes containing tungsten carbide. Verfahren zur Steigerung der Aktivitaet von Brennstoffelektroden, die Wolframcarbid enthalten

Energy Technology Data Exchange (ETDEWEB)

Binder, H.; Koehling, A.; Kuhn, W.; Lindner, W.; Sandstede, G.

1977-10-13

An increase in the activity of electrodes containing tungsten carbide for a low-temperature fuel cell with sulfuric acid as electrolyte can be achieved, if one operates the electrodes for a few hours (5-20 h) in the presence of hydrogen and a means of reduction (formaldehyde, hydrazene) in a voltage range of between +500 and +800 mV (relative to the H/sub 2/ electrode). A corrosion resistant layer is formed, which is assumed to have the composition WC/sub X/O/sub y/H/sub z/.

Synthesis and phase transformation mechanism of Nb{sub 2}C carbide phases

Energy Technology Data Exchange (ETDEWEB)

Vishwanadh, B., E-mail: visubathula@gmail.com [Materials Science Division, Bhabha Atomic Research Centre, Mumbai 400 094 (India); Murthy, T.S.R.Ch. [Materials Processing Division, Bhabha Atomic Research Centre, Mumbai 400 094 (India); Arya, A.; Tewari, R.; Dey, G.K. [Materials Science Division, Bhabha Atomic Research Centre, Mumbai 400 094 (India)

2016-06-25

In the present work, Niobium carbide samples were prepared through powder metallurgy route using spark plasma sintering technique. Some of these samples were heat treated at 900 °C up to 7 days. In order to investigate the phase transformation in Nb{sub 2}C carbide, the as-prepared and heat treated samples were characterized by X-ray diffraction, scanning electron microscopy and electron back scattered diffraction (EBSD) and transmission electron microscopy techniques. EBSD could index the same area of the sample in terms of any of the three allotropes of Nb{sub 2}C carbide phases (γ-Nb{sub 2}C, β-Nb{sub 2}C and α-Nb{sub 2}C) with good confidence index. From the EBSD patterns orientation relationships (OR) among γ, β and α-Nb{sub 2}C have been determined. Based on this OR when crystals of the three allotropes were superimposed, it has revealed that the basic Nb metal atom lattice (hcp lattice) in all the Nb{sub 2}C phases is same. The only difference exists in the carbides is the ordering of carbon atoms and vacancies in the octahedral positions of the hcp Nb metal atom lattice. Crystallographic analysis showed that for the transformation of γ-Nb{sub 2}C → β-Nb{sub 2}C → α-Nb{sub 2}C, large movement of Nb atoms is not required; but only by ordering of carbon atoms ensues the phase transformation. Literature shows that in the Nb–C system formation of the α-Nb{sub 2}C is not well established. Therefore, first principle calculations were carried out on these carbides. It revealed that the formation energy for α-Nb{sub 2}C is lower than the β and γ-Nb{sub 2}C carbides which indicate that the formation of α-Nb{sub 2}C is thermodynamically feasible. - Highlights: • Nb{sub 2}C carbide was produced by Spark Plasma Sintering in a single process. • Phase transformation mechanism of different Nb{sub 2}C carbide phases is studied. • In all the three Nb{sub 2}C carbides (γ, β, α), the base Nb lattice remains same. • Among γ, β and α-Nb{sub 2}C

Effects of superfine refractory carbide additives on microstructure and mechanical properties of TiB2–TiC+Al2O3 composite ceramic cutting tool materials

International Nuclear Information System (INIS)

Zou, Bin; Ji, Wenbin; Huang, Chuanzhen; Wang, Jun; Li, Shasha; Xu, Kaitao

2014-01-01

Highlights: • The superfine carbides determined the mechanical properties of composites. • Superfine HfC or TaC caused some oxide impurities in composites. • Superfine VC or NbC refined and homogenized the microstructure. • Failure of composites containing HfC or TaC was produced by larger grains. • Composite containing VC exhibited more bridging and transcrystalline failure. -- Abstract: A study to increase the mechanical properties of TiB 2 –TiC+Al 2 O 3 composite ceramic cutting tool material by using superfine refractory carbide additives is presented. Four superfine refractory carbides are considered to investigate their effects on the phase composition, element distribution, grain size, fracture surface, crack propagation of the metal ceramic. The physicochemical properties of superfine carbides, such as chemical activities and atom radius, were found to have the significant effects on the microstructure and mechanical properties of the metal ceramic. Hafnium carbide (HfC) and Tantalum carbide (TaC) reduced the mechanical properties of the metal ceramic because of their poor solubility with the Ni binder phase and the formation of oxides. The mechanical properties of the metal ceramic were increased by the addition of superfine niobium carbide (NbC) and vanadium carbide (VC), and their optimum values were a flexural strength of 1100 ± 62 MPa, fracture toughness of 8.5 ± 0.8 MPa.m1/2 and hardness of 21.53 ± 0.36 GPa, respectively, when 3.2 wt% superfine VC was used

Properties of carbides and hard metals. Pt. 1

International Nuclear Information System (INIS)

Nazare, S.; Ondracek, G.

1977-01-01

Measurements of the electrical resistivity and of the Young modulus are reported for the carbides TiC, HfC, NbC, and TaC. The investigations included the solid solutions TiC-HfC, HfC-NbC, TaC-NbC, and TiC-NbC. For the solid solutions, measurements of the thermal expansion in the range 293-1173 K are also reported. All the samples were prepared by hot pressing of the commercial powders in graphite dies. The chemical composites are also presented. The measurement results are shown in figures. (GSC) [de

Electronic specific heat of transition metal carbides; Chaleur specifique electronique de carbures de metaux de transition

Energy Technology Data Exchange (ETDEWEB)

Conte, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1964-07-15

The experimental results that make it possible to define the band structure of transition metal carbides having an NaCI structure are still very few. We have measured the electronic specific heat of some of these carbides of varying electronic concentration (TiC, either stoichiometric or non-stoichiometric, TaC and mixed (Ti, Ta) - C). We give the main characteristics (metallography, resistivity, X-rays) of our samples and we describe the low temperature specific heat apparatus which has been built. In one of these we use helium as the exchange gas. The other is set up with a mechanical contact. The two use a germanium probe for thermometer. The measurement of the temperature using this probe is described, as well as the various measurement devices. The results are presented in the form of a rigid band model and show that the density of the states at the Fermi level has a minimum in the neighbourhood of the group IV carbides. (author) [French] Les donnees experimentales permettant de preciser la structure de bandes des carbures de metaux de transition de structure NaCI sont encore peu.nombreuses. Nous avons mesure la chaleur specifique electronique de certains de ces carbures, de differentes concentrations electroniques (TiC stoechiometrique ou non, TaC et mixtes (Ti, Ta) - C). Nous donnons les principales caracteristiques (metallographie, resistivite, rayon X), de nos echantillons, et nous decrivons l'appareillage de chaleur specifique a basse temperature realise. Dans l'un nous utilisons l'helium comme gaz d'echange. L'autre est monte avec un contact mecanique. Les deux utilisent une sonde au germanium comme thermometre. La mesure de la resistance de cette sonde est decrite, ainsi que les differents montages de mesure. Les resultats, presentes dans un modele de bande rigide, font apparaitre que la densite des etats au niveau de Fermi presente un minimum au voisinage des carbures du groupe IV. (auteur)

Analysis of metal fuel transient overpower experiments with the SAS4A accident analysis code

International Nuclear Information System (INIS)

Tentner, A.M.; Kalimullah; Miles, K.J.

1990-01-01

The results of the SAS4A analysis of the M7 TREAT Metal fuel experiment are presented. New models incorporated in the metal fuel version of SAS4A are described. The computational results are compared with the experimental observations and this comparison is used in the interpretation of physical phenomena. This analysis was performed using the integrated metal fuel SAS4A version and covers a wide range of events, providing an increased degree of confidence in the SAS4A metal fuel accident analysis capabilities

Analysis of carbides and inclusions in high speed tool steels

DEFF Research Database (Denmark)

Therkildsen, K.T.; Dahl, K.V.

2002-01-01

The fracture surfaces of fatigued specimens were investigated using scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDS). The aim was to quantify the distribution of cracked carbides and non-metallic inclusions on the fracturesurfaces as well as on polished cross...

The application of the German reg. guides ('elements of calculation') for radioactive discharges via exhaust air and waste water on fuel element fabrication

International Nuclear Information System (INIS)

Hille, R.; Rudolph, W.

1978-01-01

The fuel element fabricating plants at Hanau are handlung uranium, plutonium and thorium. The process essentially of converting these heavy metals into oxide, carbide or metal compounds. Thereby occur radioactive discharges into the exhaust air and the waste water. The most important pathway for exposure from these substances is inhalation, the released radionuclides mostly being α-emitters. Compared to this the external irradiation from immersion in γ, β, and neutron radiation is of less importance. (orig./HP) [de

The evaluation of the use of metal alloy fuels in pressurized water reactors. Final report

Energy Technology Data Exchange (ETDEWEB)

Lancaster, D.

1992-10-26

The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ``advanced reactors,`` it became clear that reactor design optimization has been under emphasized. Current ``advanced reactors`` are severely constrained. The AP-600 required the use of a fuel design from the 1970`s. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing.

Metallic uranium as fuel for fast reactors

International Nuclear Information System (INIS)

Moura Neto, C. de

1988-01-01

This paper presents a first overview of the use of metallic uranium and its alloys as an option for fuel for rapid reactors. Aspects are discussed concerning uranium alloys which present high solubility in the gamma phase. (author)

Phase equilibrium study on system uranium-plutonium-tungsten-carbon

International Nuclear Information System (INIS)

Ugajin, Mitsuhiro

1976-11-01

Metallurgical properties of the U-Pu-W-C system have been studied with emphasis on phases and reactions. Free energy of compound formation, carbon activity and U/Pu segregation in the W-doped carbide fuel are estimated using phase diagram data. The results indicate that tungsten metal is useful as a thermochemical stabilizer of the carbide fuel. Tungsten has high temperature stability in contact with uranium carbide and mixed uranium-plutonium carbide. (auth.)

Behavior of metallic uranium-fissium fuel in TREAT transient overpower tests

International Nuclear Information System (INIS)

Bauer, T.H.; Klickman, A.E.; Lo, R.K.; Rhodes, E.A.; Robinson, W.R.; Stanford, G.S.; Wright, A.E.

1986-01-01

TREAT tests M2, M3, and M4 were performed to obtain information on two key behavior characteristics of fuel under transient overpower accident conditions in metal-fueled fast reactors: the prefailure axial self-extrusion (elongation beyond thermal expansion) of fuel within intact cladding and the margin to cladding breach. Uranium-5 wt% fissium Experimental Breeder Reactor-II driver fuel pins were used for the tests since they were available as suitable stand-ins for the uranium-plutonium-zirconium ternary fuel, which is the reference fuel of the integral fast reactor (IFR) concept. The ternary fuel will be used in subsequent TREAT tests. Preliminary results from tests M2 and M3 were presented earlier. The present report includes significant advances in analysis as well as additional data from test M4. Test results and analysis have led to the development and validation of pin cladding failure and fuel extrusion models for metallic fuel, within reasonable uncertainties for the uranium-fissium alloy. Concepts involved are straightforward and readily extendable to ternary alloys and behavior in full-size reactors

Advanced fuel for fast breeder reactors: Fabrication and properties and their optimization

International Nuclear Information System (INIS)

1988-06-01

The present design for FBR fuel rods includes usually MOX fuel pellets cladded into stainless steel tubes, together with UO 2 axial blanket and stainless steel hexagonal wrappers. Mixed carbide, nitride and metallic fuels have been tested as alternative fuels in test reactors. Among others, the objectives to develop these alternative fuels are to gain a high breeding ratio, short doubling time and high linear ratings. Fuel rod and assembly designers are now concentrating on finding the combination of optimized fuel, cladding and wrapper materials which could result in improvement of fuel operational reliability under high burnups and load-follow mode of operation. The purpose of the meeting was to review the experience of advanced FBR fuel fabrication technology, its properties before, under and after irradiation, peculiarities of the back-end of the nuclear fuel cycle, and to outline future trends. As a result of the panel discussion, the recommendations on future Agency activities in the area of advanced FBR fuels were developed. A separate abstract was prepared for each of the 10 presentations of this meeting. Refs, figs and tabs

Interim design report: fuel particle crushing

International Nuclear Information System (INIS)

Baer, J.W.; Strand, J.B.; Cook, E.J.; Miller, C.M.

1977-11-01

The double-roll fuel particle crusher was developed to fracture the silicon carbide coatings of Fort St. Vrain (FSV) fertile and fissile and large high-temperature gas-cooled reactor (LHTGR) fissile fuel particles. The report details the design task for the fuel particle crusher, including historical test information on double-roll crushers for carbide-coated fuels and the design approach selected for the cold pilot plant crusher, and shows how the design addresses the equipment goals and operational objectives. Design calculations and considerations are included to support the selection of crusher drive and gearing, the materials chosen for crushing rolls and housing, and the bearing selection. The results of the initial testing for compliance with design objectives and operational capabilities are also presented. 8 figures, 4 tables

Carbide Precipitation in 2.25 Cr-1 Mo Bainitic Steel: Effect of Heating and Isothermal Tempering Conditions

Science.gov (United States)

Dépinoy, Sylvain; Toffolon-Masclet, Caroline; Urvoy, Stéphane; Roubaud, Justine; Marini, Bernard; Roch, François; Kozeschnik, Ernst; Gourgues-Lorenzon, Anne-Françoise

2017-05-01

The effect of the tempering heat treatment, including heating prior to the isothermal step, on carbide precipitation has been determined in a 2.25 Cr-1 Mo bainitic steel for thick-walled applications. The carbides were identified using their amount of metallic elements, morphology, nucleation sites, and diffraction patterns. The evolution of carbide phase fraction, morphology, and composition was investigated using transmission electron microscopy, X-ray diffraction, as well as thermodynamic calculations. Upon heating, retained austenite into the as-quenched material decomposes into ferrite and cementite. M7C3 carbides then nucleate at the interface between the cementite and the matrix, triggering the dissolution of cementite. M2C carbides precipitate separately within the bainitic laths during slow heating. M23C6 carbides precipitate at the interfaces (lath boundaries or prior austenite grain boundaries) and grow by attracting nearby chromium atoms, which results in the dissolution of M7C3 and, depending on the temperature, coarsening, or dissolution of M2C carbides, respectively.

HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

Energy Technology Data Exchange (ETDEWEB)

Sell, D. A.; Baily, C. E.; Malewitz, T. J.; Medvedev, P. G.; Porter, D. L.; Hilton, B. A.

2016-09-01

A system has been developed to perform measurements on irradiated, sodium bonded-metallic fuel elements to determine the amount of fission gas retained in the fuel material after release of the gas to the element plenum. During irradiation of metallic fuel elements, most of the fission gas developed is released from the fuel and captured in the gas plenums of the fuel elements. A significant amount of fission gas, however, remains captured in closed porosities which develop in the fuel during irradiation. Additionally, some gas is trapped in open porosity but sealed off from the plenum by frozen bond sodium after the element has cooled in the hot cell. The Retained fission Gas (RFG) system has been designed, tested and implemented to capture and measure the quantity of retained fission gas in characterized cut pieces of sodium bonded metallic fuel. Fuel pieces are loaded into the apparatus along with a prescribed amount of iron powder, which is used to create a relatively low melting, eutectic composition as the iron diffuses into the fuel. The apparatus is sealed, evacuated, and then heated to temperatures in excess of the eutectic melting point. Retained fission gas release is monitored by pressure transducers during the heating phase, thus monitoring for release of fission gas as first the bond sodium melts and then the fuel. A separate hot cell system is used to sample the gas in the apparatus and also characterize the volume of the apparatus thus permitting the calculation of the total fission gas release from the fuel element samples along with analysis of the gas composition.