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Sample records for minor actinide-containing cermet

  1. 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

  2. Fabrication of U-Pu-Zr metallic fuel containing minor actinides

    International Nuclear Information System (INIS)

    Kurata, Masaki; Sasahara, Akihiro; Inoue, Tadashi; Betti, M.; Babelot, J.F.; Spirlet, J.C.; Koch, L.

    1997-01-01

    Rods of UPuZr alloy containing 5% minor actinides, 2% minor actinides and 2% rare-earth elements, and 5% minor actinides and 5% rare-earth elements have been fabricated by casting in yttria molds. Parts of the ingots were cut off for quantitative analysis and the rods characterized to the required extent, which included measurement of length, weight, diameter, and bending. For selected samples, metallographic study was carried out to examine the dispersion of the various phases contained in the alloy. Finally, the rods were encapsulated in stainless steel pin with the UPuZr reference after sodium bonding for the irradiation study. (author)

  3. Electrochemical reduction of CerMet fuels for transmutation using surrogate CeO2-Mo pellets

    Science.gov (United States)

    Claux, B.; Souček, P.; Malmbeck, R.; Rodrigues, A.; Glatz, J.-P.

    2017-08-01

    One of the concepts chosen for the transmutation of minor actinides in Accelerator Driven Systems or fast reactors proposes the use of fuels and targets containing minor actinides oxides embedded in an inert matrix either composed of molybdenum metal (CerMet fuel) or of ceramic magnesium oxide (CerCer fuel). Since the sufficient transmutation cannot be achieved in a single step, it requires multi-recycling of the fuel including recovery of the not transmuted minor actinides. In the present work, a pyrochemical process for treatment of Mo metal inert matrix based CerMet fuels is studied, particularly the electroreduction in molten chloride salt as a head-end step required prior the main separation process. At the initial stage, different inactive pellets simulating the fuel containing CeO2 as minor actinide surrogates were examined. The main studied parameters of the process efficiency were the porosity and composition of the pellets and the process parameters as current density and passed charge. The results indicated the feasibility of the process, gave insight into its limiting parameters and defined the parameters for the future experiment on minor actinide containing material.

  4. Minor actinide transmutation using minor actinide burner reactors

    International Nuclear Information System (INIS)

    Mukaiyama, T.; Yoshida, H.; Gunji, Y.

    1991-01-01

    The concept of minor actinide burner reactor is proposed as an efficient way to transmute long-lived minor actinides in order to ease the burden of high-level radioactive waste disposal problem. Conceptual design study of minor actinide burner reactors was performed to obtain a reactor model with very hard neutron spectrum and very high neutron flux in which minor actinides can be fissioned efficiently. Two models of burner reactors were obtained, one with metal fuel core and the other with particle fuel core. Minor actinide transmutation by the actinide burner reactors is compared with that by power reactors from both the reactor physics and fuel cycle facilities view point. (author)

  5. Minor Actinide Laboratory at JRC-ITU: Fuel fabrication facility

    International Nuclear Information System (INIS)

    Fernandez, A.; McGinley, J.; Somers, J.

    2008-01-01

    The Minor Actinide Laboratory (MA-lab) of the Institute for Transuranium Elements is a unique facility for the fabrication of fuels and targets containing minor actinides (MA). It is of key importance for research on Partitioning and Transmutation in Europe, as it is one of the only dedicated facilities for the fabrication of MA containing materials, either for property measurements or for the production of test pins for irradiation experiments. In this paper a detailed description of the MA-Lab facility and the fabrication processes developed to fabricate fuels and samples containing high content of minor actinides is given. In addition, experience gained and improvements are also outlined. (authors)

  6. Development of CERMET fuels for minor actinides transmutation

    International Nuclear Information System (INIS)

    Haas, D.; Fernandez, A.; Naestren, C.; Staicu, D.; Somers, J.; Maschek, W.; Chen, X.

    2006-01-01

    The sub-critical Accelerator Driven System (ADS) is now being considered as a potential means to burn long-lived transuranium nuclides. The preferred fuel for such a fast neutron reactor is uranium-free, highly enriched with plutonium and minor actinides. Requirements for ADS transmutation fuels are linked with the core design and safety parameters, the fuel properties and the ease of reprocessing. This study concerns the properties of metals as matrices, with the particular case of Mo. To improve the neutronic characteristics, enriched molybdenum (Mo-92) is required. To overcome the high enrichment cost, it is proposed to recover the matrix by pellet dissolution, and to recycle it for further use. Irradiation programmes are also planned to examine the in-reactor properties of the material. Based on the current status of the research, the results are promising, but irradiation results are still missing. (authors)

  7. Development of fast reactor metal fuels containing minor actinides

    International Nuclear Information System (INIS)

    Ohta, Hirokazu; Ogata, Takanari; Kurata, Masaki; Koyama, Tadafumi; Papaioannou, Dimitrios; Glatz, Jean-Paul; Rondinella, Vincenzo V.

    2011-01-01

    Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are being developed by the Central Research Institute of Electric Power Industry (CRIEPI) in collaboration with the Institute for Transuranium Elements (ITU) in the METAPHIX project. The basic properties of U-Pu-Zr alloys containing MA (and RE) were characterized by performing ex-reactor experiments. On the basis of the results, test fuel pins including U-Pu-Zr-MA(-RE) alloy ingots in parts of the fuel stack were fabricated and irradiated up to a maximum burnup of ∼10 at% in the Phenix fast reactor (France). Nondestructive postirradiation tests confirmed that no significant damage to the fuel pins occurred. At present, detailed destructive postirradiation examinations are being carried out at ITU. (author)

  8. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

    2015-01-01

    Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

  9. Partitioning and Transmutation of minor actinides

    International Nuclear Information System (INIS)

    Koch, L.; Wellum, R.

    1991-01-01

    The partitioning of minor actinides from spent fuels and their transmutation into short-lived fission products has been the topic of two dedicated meetings organized jointly by the European Commission and the OECD. The conclusion of the last meeting in 1980, in short, was that partitioning and transmutation of minor actinides, especially in fast reactors, seemed possible. However, the incentive, which would be a reduction of the radiological hazard to the public, was too small if long-lived fission products were not included. Furthermore this meeting showed that minor actinide targets or possible nuclear fuels containing minor actinides for transmutation had not yet been developed. The European Institute for Transuranium Elements took up this task and has carried it out as a small activity for several years. Interests expressed recently by an expert meeting of the OECD/NEA (Paris, 25 April 1989), which was initiated by the proposed Japanese project Omega, led us to the conclusion that the present state of knowledge should be looked at in a workshop environment. Since the Japanese proposal within the project Omega is based on a broader approach we needed this evaluation to assess the relevance of our present activity and wanted to identifiy additional studies which might be needed to cover possible future demands from the public. This workshop was therefore organized, and participants active in the field from EC countries, the USA and Japan were invited

  10. Optimisation of composite metallic fuel for minor actinide transmutation in an accelerator-driven system

    Science.gov (United States)

    Uyttenhove, W.; Sobolev, V.; Maschek, W.

    2011-09-01

    A potential option for neutralization of minor actinides (MA) accumulated in spent nuclear fuel of light water reactors (LWRs) is their transmutation in dedicated accelerator-driven systems (ADS). A promising fuel candidate dedicated to MA transmutation is a CERMET composite with Mo metal matrix and (Pu, Np, Am, Cm)O 2-x fuel particles. Results of optimisation studies of the CERMET fuel targeting to increasing the MA transmutation efficiency of the EFIT (European Facility for Industrial Transmutation) core are presented. In the adopted strategy of MA burning the plutonium (Pu) balance of the core is minimized, allowing a reduction in the reactivity swing and the peak power form-factor deviation and an extension of the cycle duration. The MA/Pu ratio is used as a variable for the fuel optimisation studies. The efficiency of MA transmutation is close to the foreseen theoretical value of 42 kg TW -1 h -1 when level of Pu in the actinide mixture is about 40 wt.%. The obtained results are compared with the reference case of the EFIT core loaded with the composite CERCER fuel, where fuel particles are incorporated in a ceramic magnesia matrix. The results of this study offer additional information for the EFIT fuel selection.

  11. Optimisation of composite metallic fuel for minor actinide transmutation in an accelerator-driven system

    International Nuclear Information System (INIS)

    Uyttenhove, W.; Sobolev, V.; Maschek, W.

    2011-01-01

    A potential option for neutralization of minor actinides (MA) accumulated in spent nuclear fuel of light water reactors (LWRs) is their transmutation in dedicated accelerator-driven systems (ADS). A promising fuel candidate dedicated to MA transmutation is a CERMET composite with Mo metal matrix and (Pu, Np, Am, Cm)O 2-x fuel particles. Results of optimisation studies of the CERMET fuel targeting to increasing the MA transmutation efficiency of the EFIT (European Facility for Industrial Transmutation) core are presented. In the adopted strategy of MA burning the plutonium (Pu) balance of the core is minimized, allowing a reduction in the reactivity swing and the peak power form-factor deviation and an extension of the cycle duration. The MA/Pu ratio is used as a variable for the fuel optimisation studies. The efficiency of MA transmutation is close to the foreseen theoretical value of 42 kg TW -1 h -1 when level of Pu in the actinide mixture is about 40 wt.%. The obtained results are compared with the reference case of the EFIT core loaded with the composite CERCER fuel, where fuel particles are incorporated in a ceramic magnesia matrix. The results of this study offer additional information for the EFIT fuel selection.

  12. Minor actinide transmutation - a waste management option

    International Nuclear Information System (INIS)

    Koch, L.

    1986-01-01

    The incentive to recycle minor actinides results from the reduction of the long-term α-radiological risk rather than from a better utilization of the uranium resources. Nevertheless, the gain in generated electricity by minor actinide transmutation in a fast breeder reactor can compensate for the costs of their recovery and make-up into fuel elements. Different recycling options of minor actinides are discussed: transmutation in liquid metal fast breeder reactors (LMFBRs) is possible as long as plutonium is not recycled in light water reactors (LWRs). In this case a minor actinide burner with fuel of different composition has to be introduced. The development of appropriate minor actinide fuels and their properties are described. The irradiation experiments underway or planned are summarized. A review of minor actinide partitioning from the PUREX waste stream is given. From the present constraints of LMFBR technology a reduction of the long-term α-radiological risk by a factor of 200 is deduced relative to that from the direct storage of spent LWR fuel. Though the present accumulation of minor actinides is low, nuclear transmutation may be needed when nuclear energy production has grown. (orig.)

  13. Impact of minor actinide recycling on sustainable fuel cycle options

    Energy Technology Data Exchange (ETDEWEB)

    Heidet, F.; Kim, T. K.; Taiwo, T. A.

    2017-11-01

    the repository performance. On the other hand, recycling minor actinides also results in an increase of the recycled fuel characteristics and therefore of the charged fuel. The radioactivity is slightly increased while the decay heat and radiotoxicities are very significantly increased. Despite these differences, the characteristics of the fuel at time of discharge remain similar whether minor actinides are recycled or not, with the exception of the inhalation radiotoxicity which is significantly larger with minor actinide recycling. After some cooling the characteristics of the discharged fuel become larger when minor actinides are recycled, potentially affecting the reprocessing plant requirements. Recycling minor actinides has a negative impact on the characteristics of the fresh fuel and will make it more challenging to fabricate fuel containing minor actinides.

  14. Transmutation of minor actinide using thorium fueled BWR core

    International Nuclear Information System (INIS)

    Susilo, Jati

    2002-01-01

    One of the methods to conduct transmutation of minor actinide is the use of BWR with thorium fuel. Thorium fuel has a specific behaviour of producing a little secondary minor actinides. Transmutation of minor actinide is done by loading it in the BWR with thorium fuel through two methods, namely close recycle and accumulation recycle. The calculation of minor actinide composition produced, weigh of minor actinide transmuted, and percentage of reminder transmutation was carried SRAC. The calculations were done to equivalent cell modeling from one fuel rod of BWR. The results show that minor actinide transmutation is more effective using thorium fuel than uranium fuel, through both close recycle and accumulation recycle. Minor actinide transmutation weight show that the same value for those recycle for 5th recycle. And most of all minor actinide produced from 5 unit BWR uranium fuel can transmuted in the 6 t h of close recycle. And, the minimal value of excess reactivity of the core is 12,15 % Δk/k, that is possible value for core operation

  15. Cermets for high level waste containment

    International Nuclear Information System (INIS)

    Aaron, W.S.; Quinby, T.C.; Kobisk, E.H.

    1978-01-01

    Cermet materials are currently under investigation as an alternate for the primary containment of high level wastes. The cermet in this study is an iron--nickel base metal matrix containing uniformly dispersed, micron-size fission product oxides, aluminosilicates, and titanates. Cermets possess high thermal conductivity, and typical waste loading of 70 wt % with volume reduction factors of 2 to 200 and low processing volatility losses have been realized. Preliminary leach studies indicate a leach resistance comparable to other candidate waste forms; however, more quantitative data are required. Actual waste studies have begun on NFS Acid Thorex, SRP dried sludge and fresh, unneutralized SRP process wastes

  16. Reduction of minor actinides for recycling in a light water reactor

    International Nuclear Information System (INIS)

    Martinez C, E.; Ramirez S, J. R.; Alonso V, G.

    2015-09-01

    The aim of actinide transmutation from spent nuclear fuel is the reduction in mass of high-level waste which must be stored in geological repositories and the lifetime of high-level waste; these two achievements will reduce the number of repositories needed, as well as the duration of storage. The present work is directed towards the evaluation of an advanced nuclear fuel cycle in which the minor actinides (Np, Am and Cm) could be recycled to remove most of the radioactive material; a reference of actinides production in standard nuclear fuel of uranium at the end of its burning in a BWR is first established, after a design of fuel rod containing 6% of minor actinides in a matrix of uranium from the enrichment lines is proposed, then 4 fuel rods of standard uranium are replaced by 4 actinides bars to evaluate the production and transmutation of them and finally the minor actinides reduction in the fuel is evaluated. In the development of this work the calculation tool are the codes: Intrepin-3, Casmo-4 and Simulate-3. (Author)

  17. Burning minor actinides in a HTR energy spectrum

    International Nuclear Information System (INIS)

    Pohl, Christoph; Rütten, H. Jochem

    2012-01-01

    Highlights: ► Burn-up analysis for varying plutonium/minor actinide fuel compositions. ► The influence of varying heavy metal fuel element loads is investigated. ► Significant burn-up via radiative capture and subsequently fission is observed. ► Difference observed between fuel element burn-up and total actinide burning rate. - Abstract: The generation of nuclear energy by means of the existing nuclear reactor systems is based mainly on the fission of U-235. But this comes along with the capture of neutrons by the U-238 faction and results in a build-up of plutonium isotopes and minor actinides as neptunium, americium and curium. These actinides are dominant for the long time assessment of the radiological risk of a final disposal therefore a minimization of the long living isotopes is aspired. Burning the actinides in a high temperature helium cooled graphite moderated reactor (HTR) is one of these options. The use of plutonium isotopes to sustain the criticality of the system is intended to avoid on the one hand highly enriched uranium because of international regulations and on the other hand low enriched uranium because of the build up of new actinides from neutron capture in the U-238 fraction. Because initial minor actinide isotopes are typically not fissionable by thermal neutrons the idea is to fission instead the intermediate isotopes generated by the first neutron capture. This paper comprises calculations for plutonium/minor actinides/thorium fuel compositions and their correlated final burn-up for a generic pebble bed HTR based on the reference design of the 400 MW PBMR. In particular the cross sections and the neutron balance of the different minor actinide isotopes in the higher thermal energy spectrum of a HTR will be discussed. For a fuel mixture of plutonium and minor actinides a significant burn-up of these actinides up to 20% can be achieved but at the expense of a higher residual fraction of plutonium in the burned fuel. Combining

  18. Phoenix type concepts for transmutation of LWR waste minor actinides

    International Nuclear Information System (INIS)

    Segev, M.

    1994-01-01

    A number of variations on the original Phoenix theme were studied. The basic rationale of the Phoenix incinerator is making oxide fuel of the LWR waste minor actinides, loading it in an FFTF-like subcritical core, then bombarding the core with the high current beam accelerated protons to generate considerable energy through spallation and fission reactions. As originally assessed, if the machine is fed with 1600 MeV protons in a 102 mA current, then 8 core modules are driven to transmute the yearly minor actinides waste of 75 1000 MW LWRs into Pu 238 and fission products; in a 2 years cycle the energy extracted is 100000 MW d/T. This performance cannot be substantiated in a rigorous analysis. A calculational consistent methodology, based on a combined execution of the Hermes, NCNP, and Korigen codes, shows, nonetheless that changes in the original Phoenix parameters can upgrade its performance.The original Phoenix contains 26 tons minor actinides in 8 core modules; 1.15 m 3 module is shaped for 40% neutron leakage; with a beam of 102 mA the 8 modules are driven to 100000 MW/T in 10.5 years, burning out the yearly minor actinide waste of 15 LWRs; the operation must be assisted by grid electricity. If the 1.15 m 3 module is shaped to allow only 28% leakage, then a beam of 102 mA will drive the 8 modules to 100000 MW/T in 3.5 years, burning out the yearly minor actinides waste of 45 LWRs. Some net grid electricity will be generated. If 25 tons minor actinides are loaded into 5 modules, each 1.72 m 3 in volume and of 24% leakage, then a 97 mA beam will drive the module to 100000 MW/T in 2.5 years, burning out the yearly minor actinides waste of 70 LWRs. A considerable amount of net grid electricity will be generated. If the lattice is made of metal fuel, and 26 tons minor actinides are loaded into 32 small modules, 0.17 m 3 each, then a 102 mA beam will drive the modules to 100000 MW/T in 2 years, burning out the yearly minor actinides waste of 72 LWRs. A considerable

  19. Properties of minor actinide nitrides

    International Nuclear Information System (INIS)

    Takano, Masahide; Itoh, Akinori; Akabori, Mitsuo; Arai, Yasuo; Minato, Kazuo

    2004-01-01

    The present status of the research on properties of minor actinide nitrides for the development of an advanced nuclear fuel cycle based on nitride fuel and pyrochemical reprocessing is described. Some thermal stabilities of Am-based nitrides such as AmN and (Am, Zr)N were mainly investigated. Stabilization effect of ZrN was cleary confirmed for the vaporization and hydrolytic behaviors. New experimental equipments for measuring thermal properties of minor actinide nitrides were also introduced. (author)

  20. Development of nitride fuel and pyrochemical process for transmutation of minor actinides

    International Nuclear Information System (INIS)

    Arai, Yasuo; Akabori, Mitsuo; Minato, Kazuo; Uno, Masayoshi

    2010-01-01

    Nitride fuel cycle for transmutation of minor actinides has been investigated under the double-strata fuel cycle concept. Mononitride solid solutions containing minor actinides have been prepared and characterised. Thermo-physical properties, such as thermal expansion, heat capacity and thermal diffusivity, have been measured by use of minor actinide nitride and burn-up simulated nitride samples. Irradiation behaviour of nitride fuel has been examined by irradiation tests. Pyrochemical process for treatment of spent nitride fuel has been investigated mainly by electrochemical measurements and nitride formation behaviour in pyrochemical process has been studied for recycled fuel fabrication. Recent results of experimental study on nitride fuel and pyrochemical process are summarised in the paper. (authors)

  1. Evaluating the efficacy of a minor actinide burner

    International Nuclear Information System (INIS)

    Dobbin, K.D.; Kessler, S.F.; Nelson, J.V.; Omberg, R.P.; Wootan, D.W.

    1993-06-01

    The efficacy of a minor actinide burner can be evaluated by comparing safety and economic parameters to the support ratio. Minor actinide mass produced per unit time in this number of Light Water Reactors (LWRs) can be burned during the same time period in one burner system. The larger the support ratio for a given set of safety and economic parameters, the better. To illustrate this concept, the support ratio for selected Liquid Metal Reactor (LMR) burner core designs was compared with corresponding coolant void worths, a fundamental safety concern following the Chernobyl accident. Results can be used to evaluate the cost in reduced burning of minor actinides caused by LMR sodium void reduction efforts or to compare with other minor actinide burner systems

  2. Production and measurement of minor actinides in the commercial fuel cycle

    International Nuclear Information System (INIS)

    Stanbro, W.D.

    1997-03-01

    The minor actinide elements, particularly neptunium and americium, are produced as a normal byproduct of the operation of thermal power reactors. Because of the existence of long-lived isotopes of these elements, they constitute the major sources of the residual radiation in spent fuel or in wastes resulting from reprocessing. This has led to examinations by some countries of the possibility of separating the minor actinides from waste products. The papers found in this report address the production of minor actinides in common thermal power reactors as well as approaches to measure these materials in various media. The first paper in this volume, open-quotes Production of Minor Actinides in the Commercial Fuel Cycle,close quotes uses calculations with the ORIGEN2 reactor and decay code to estimate the amounts of minor actinides in spent fuel and separated plutonium as a function of reactor irradiation and the time after discharge. The second paper, open-quotes Destructive Assay of Minor Actinides,close quotes describes a number of promising approaches for the chemical analysis of minor actinides in the various forms in which they are found at reprocessing plants. The next paper, open-quotes Hybrid KED/XRF Measurement of Minor Actinides in Reprocessing Plants,close quotes uses the results of a simulation model to examine the possible applications of the hybrid KED/XRF instrument to the determination of minor actinides in some of the solutions found in reprocessing plants. In open-quotes Calorimetric Assay of Minor Actinides,close quotes the authors show some possible extensions of this powerful technique beyond the normal plutonium assays to include the minor actinides. Finally, the last paper in this volume, open-quotes Environment Measurements of Transuranic Nuclides,close quotes discusses what is known about the levels of the minor actinides in the environment and ways to analyze for these materials in environmental matrices

  3. Study of burned optimization for minor actinides in European Sodium Fast Reactor (ESFR) by use of moderator materials

    International Nuclear Information System (INIS)

    Ramos, R L; Villanueva, A J; Buiront, L

    2012-01-01

    The minor actinides (MA) burn up optimization in the European Sodium Fast Reactor (ESFR) core was studied by adding different moderating materials in the Minor Actinides Bearing Blanket subassemblies (MABB SA) using the ERANOS neutron code package. These SA are of hexagonal shape and are composed of pellets inside of pins. These pellets contain a mixture of uranium dioxide (UO 2 ) and americium dioxide (AmO 2 ). If some of these pins are replaced by other identical ones containing moderating material instead of minor actinides, a shift in the spectrum towards lower energies is expected, which might enhance the burn up performance. The results of this work demonstrated that the use of compounds of hydrogen and magnesium as moderators produces a shift in the neutron spectrum, improving the porcentual minor actinides consumption. ZrH 2 moderator material was found to exhibit the best performances for this propose, followed by MgO and MgAl 2 O 4 , in that order. The use of SiC, BeO, TiC, LiO 2 and ZrC material produced no effect on the shift of the neutron spectrum. For safety reasons, it seems hardly realistic to use hydrogenous compounds in sodium fast reactors. So, compounds with magnesium are selected to be placed into the pins to improve the porcentual minor actinides consumption. The ESFR core is composed by 817 SA, 453 of them are fuel SA, 247 are reflectors SA, 84 are MABB (Minor Actinides Bearing Blankets) SA and 33 are control and shutdown rods. When about half of the total pins in each MABB were substituted by moderator pins with MgO pellets (135 of 271 pins), the porcentual consumption of minor actinides was of 30.85 %, i.e., 227.22 kg of minor actinides were consumed out of 736.65 kg in the initial configuration. In the case where all the pins of the MABB contained pellets of minor actinides, the porcentual consumption of minor actinides was of 21.26 %, i.e., 312.13 kg of minor actinides were consumed of 1467.87 kg in the initial configuration (author)

  4. Thermochemical and thermophysical properties of minor actinide compounds

    International Nuclear Information System (INIS)

    Minato, Kazuo; Takano, Masahide; Otobe, Haruyoshi; Nishi, Tsuyoshi; Akabori, Mitsuo; Arai, Yasuo

    2009-01-01

    Burning or transmutation of minor actinides (MA: Np, Am, Cm) that are classified as the high-level radioactive waste in the current nuclear fuel cycle is an option for the advanced nuclear fuel cycle. Although the thermochemical and thermophysical properties of minor actinide compounds are essential for the design of MA-bearing fuels and analysis of their behavior, the experimental data on minor actinide compounds are limited. To support the research and development of the MA-bearing fuels, the property measurements were carried out on minor actinide nitrides and oxides. The lattice parameters and their thermal expansions were measured by high-temperature X-ray diffractometry. The specific heat capacities were measured by drop calorimetry and the thermal diffusivities by laser-flash method. The thermal conductivities were determined by the specific heat capacities, thermal diffusivities and densities. The oxygen potentials were measured by electromotive force method.

  5. Review of Integral Experiments for Minor Actinide Management

    International Nuclear Information System (INIS)

    Gil, C.S.; Glinatsis, G.; Hesketh, K.; Iwamoto, O.; Okajima, S.; Tsujimoto, K.; Jacqmin, R.; Khomyakov, Y.; Kochetkov, A.; Kormilitsyn, M.; Palmiotti, G.; Salvatores, M.; Perret, G.; Rineiski, A.; Romanello, V.; Sweet, D.

    2015-01-01

    Spent nuclear fuel contains minor actinides (MAs) such as neptunium, americium and curium, which require careful management. This becomes even more important when mixed oxide (MOX) fuel is being used on a large scale since more MAs will accumulate in the spent fuel. One way to manage these MAs is to transmute them in nuclear reactors, including in light water reactors, fast reactors or accelerator-driven subcritical systems. The transmutation of MAs, however, is not straightforward, as the loading of MAs generally affects physics parameters, such as coolant void, Doppler and burn-up reactivity. This report focuses on nuclear data requirements for minor actinide management, the review of existing integral data and the determination of required experimental work, the identification of bottlenecks and possible solutions, and the recommendation of an action programme for international co-operation. (authors)

  6. Plutonium and minor actinide transmutation by long irradiation in LWR

    International Nuclear Information System (INIS)

    Facchini, A.; Sanjust, V.

    1993-01-01

    An investigation was made on the conceptual possibility of burning in a thermal reactor MOX fuel together with special pins containing plutonium, minor actinides and long lived fission products, recovered from the reprocessing of previously irradiated MOX fuel and mixed with an inter matrix. Preliminary calculations showed that the long term radiotoxicity of the above special pins is reduced to reasonable levels when they are irradiated up to 20 divided-by 30 years, and cooled for some centuries. In particular, during the whole life such a reactor should be able to burn a considerable fraction of plutonium, minor actinides and long lived fission products recovered from the MOX fuel irradiated along the same period of time

  7. Synthesis of Uranium-based Microspheres for Transmutation of Minor Actinides

    International Nuclear Information System (INIS)

    Daniels, Henrik; Neumeier, Stefan; Modolo, Giuseppe

    2010-01-01

    Utilisation of the internal gelation process is a promising perspective for the fabrication of advanced nuclear fuels containing minor actinides (MA). The formulation of appropriate precursor solutions for this process is an important step towards a working process as the chemistry of uranium-MA systems is quite complex. In this work, actinide surrogates were utilised for basic research on their influence on the system. The ceramics obtained through thermal treatment of the gels were characterised to optimise the calcination and sintering process. (authors)

  8. Neodymium partitioning in zirconolite-based glass-ceramics designed for minor actinides immobilization

    International Nuclear Information System (INIS)

    Loiseau, P.; Caurant, D.; Baffier, N.; Fillet, C.

    2000-01-01

    This study deals with glass-ceramic matrices designed for the conditioning of minor actinides, in which zirconolite crystals (CaZrTi 2 O 7 ) are homogeneously dispersed in a residual glassy matrix. Good immobilization performances require a high enrichment of actinides in the crystalline phase (double containment principle). Glass-ceramics are obtained by controlled devitrification of an aluminosilicate parent glass containing large amounts of TiO 2 and ZrO 2 . Neodymium was selected to simulate the trivalent minor actinides. Crystallization was performed at 1200 deg. C for various Nd 2 O 3 contents (0 - 10 wt. %). In all cases, zirconolite crystallization is obtained in the bulk of glass-ceramics. The evolution of Nd 3+ location between the crystals and the residual glass was followed by electron spin resonance and optical absorption. Both techniques demonstrate that neodymium is partly incorporated in zirconolite crystals. Moreover, total Nd 2 O 3 content in parent glass has a strong effect on Nd 3+ ions distribution. (authors)

  9. Sensitivity analysis of minor actinides transmutation to physical and technological parameters

    International Nuclear Information System (INIS)

    Kooyman, T.; Buiron, L.

    2015-01-01

    Minor actinides transmutation is one of the 3 main axis defined by the 2006 French law for management of nuclear waste, along with long-term storage and use of a deep geological repository. Transmutation options for critical systems can be divided in two different approaches: (a) homogeneous transmutation, in which minor actinides are mixed with the fuel. This exhibits the drawback of 'polluting' the entire fuel cycle with minor actinides and also has an important impact on core reactivity coefficients such as Doppler Effect or sodium void worth for fast reactors when the minor actinides fraction increases above 3 to 5% depending on the core; (b) heterogeneous transmutation, in which minor actinides are inserted into transmutation targets which can be located in the center or in the periphery of the core. This presents the advantage of decoupling the management of the minor actinides from the conventional fuel and not impacting the core reactivity coefficients. In both cases, the design and analyses of potential transmutation systems have been carried out in the frame of Gen IV fast reactor using a 'perturbation' approach in which nominal power reactor parameters are modified to accommodate the loading of minor actinides. However, when designing such a transmutation strategy, parameters from all steps of the fuel cycle must be taken into account, such as spent fuel heat load, gamma or neutron sources or fabrication feasibility. Considering a multi-recycling strategy of minor actinides, an analysis of relevant estimators necessary to fully analyze a transmutation strategy has been performed in this work and a sensitivity analysis of these estimators to a broad choice of reactors and fuel cycle parameters has been carried out. No threshold or percolation effects were observed. Saturation of transmutation rate with regards to several parameters has been observed, namely the minor actinides volume fraction and the irradiation time. Estimators of interest that have been

  10. Strategies for minority actinides transmutation in fast reactors

    International Nuclear Information System (INIS)

    Perez-Martin, S.; Martin-Fuertes, F.; Alvarez-Velarde, F.

    2010-01-01

    Presentation of the strategies that can be followed in fast reactors designed for the fourth generation to reduce the inventory of minority actinides generated in current light water reactors, as the actinides generation in fast reactor.

  11. An optimization methodology for heterogeneous minor actinides transmutation

    Science.gov (United States)

    Kooyman, Timothée; Buiron, Laurent; Rimpault, Gérald

    2018-04-01

    In the case of a closed fuel cycle, minor actinides transmutation can lead to a strong reduction in spent fuel radiotoxicity and decay heat. In the heterogeneous approach, minor actinides are loaded in dedicated targets located at the core periphery so that long-lived minor actinides undergo fission and are turned in shorter-lived fission products. However, such targets require a specific design process due to high helium production in the fuel, high flux gradient at the core periphery and low power production. Additionally, the targets are generally manufactured with a high content in minor actinides in order to compensate for the low flux level at the core periphery. This leads to negative impacts on the fuel cycle in terms of neutron source and decay heat of the irradiated targets, which penalize their handling and reprocessing. In this paper, a simplified methodology for the design of targets is coupled with a method for the optimization of transmutation which takes into account both transmutation performances and fuel cycle impacts. The uncertainties and performances of this methodology are evaluated and shown to be sufficient to carry out scoping studies. An illustration is then made by considering the use of moderating material in the targets, which has a positive impact on the minor actinides consumption but a negative impact both on fuel cycle constraints (higher decay heat and neutron) and on assembly design (higher helium production and lower fuel volume fraction). It is shown that the use of moderating material is an optimal solution of the transmutation problem with regards to consumption and fuel cycle impacts, even when taking geometrical design considerations into account.

  12. The prediction of minor actinides amounts accumulated in the spent fuel in China

    International Nuclear Information System (INIS)

    Zhou Peide

    2000-01-01

    The amounts of the Minor Actinides accumulated in the spent fuel are predicted according to the Nuclear Power Plant development plan envisaged in China. The Minor Actinides generated in the spent fuel unloaded from a typical PWR per year are calculated. The decay characteristics of the Minor Actinides during storage and cooling period are also calculated. At last, the Minor Actinides amounts accumulated in all spent fuel which were unloaded before sometime are given

  13. Cermet fuel behaviour and properties in ADS reactors

    International Nuclear Information System (INIS)

    Haas, D.; Fernandez, A.; Staicu, D.; Somers, J.; Maschek, W.; Chen, X.

    2007-01-01

    Within the EUROTRANS Integrated Project co- financed within the 6th Framework Programme of the European commission, the sub-critical Accelerator Driven System (ADS) is now being considered as a potential means to burn long-lived transuranium nuclides. Within the EUROTRANS Programme, the domain AFTRA is responsible to develop and provide the data basis for the fuels to be used in the European Facility for Industrial Transmutation (EFIT). The preferred fuel for such a fast neutron reactor is uranium-free, highly enriched with plutonium and minor actinides. Requirements for ADS transmuter fuels are strongly linked with the core design and safety parameters, the fuel properties and the ease of fabrication and reprocessing. This study concerns the behaviour and properties of fuels with molybdenum as inert matrix. The status of the development work was presented at the last ICENES conference [1]. Since then, the design of the European Facility for Industrial Transmutation (EFIT) was developed and the transmutation capability, the burn-up behaviour, the reactivity swing and power peaking factors, and the safety performance were determined for different cores with inert matrix fuels like MgO and Mo. For the EFIT, the CERMET with the Mo matrix is recommended as the reference fuel and CERCER with the MgO matrix as a back-up solution. The thermal diffusivity and specific heat of the CERMET fuels (loaded with Pu and Am) were measured, and the thermal conductivity was deduced. The thermal conductivity of the CERMET fuels was also predicted using a model proposed in [1], with a microstructure corresponding to a random distribution of spheres, with overlapping. This model microstructure takes into account the negative effects arising from the possible formation of small agglomerates of inclusions in the CERMET fuels. The agreement between the theoretical and calculated values is relatively good (the error is between 0 and 15% of the value of the thermal conductivity

  14. Homogeneous Minor Actinide Transmutation in SFR: Neutronic Uncertainties Propagation with Depletion

    International Nuclear Information System (INIS)

    Buiron, L.; Plisson-Rieunier, D.

    2015-01-01

    In the frame of next generation fast reactor design, the minimisation of nuclear waste production is one of the key objectives for current R and D. Among the possibilities studied at CEA, minor actinides multi-recycling is the most promising industrial way achievable in the near-term. Two main management options are considered: - Multi-recycling in a homogeneous way (minor actinides diluted in the driver fuel). If this solution can help achieving high transmutation rates, the negative impact of minor actinides on safety coefficients allows only a small fraction of the total heavy mass to be loaded in the core (∼ few %). - Multi-recycling in heterogeneous way by means of Minor Actinide Bearing Blanket (MABB) located at the core periphery. This solution offers more flexibility than the previous one, allowing a total minor actinides decoupled management from the core fuel. As the impact on feedback coefficient is small larger initial minor actinide mass can be loaded in this configuration. Starting from a breakeven Sodium Fast Reactor designed jointly by CEA, Areva and EdF teams, the so called SFR V2B, transmutation performances have been studied in frame on the French fleet for both options and various specific isotopic management (all minor actinides, americium only, etc.). Using these results, a sensitivity study has been performed to assess neutronic uncertainties (i.e coming from cross section) on mass balance on the most attractive configurations. This work in based on a new implementation of sensitivity on concentration with depletion in the ERANOS code package. Uncertainties on isotopes masses at the end of irradiation using various variance-covariance is discussed. (authors)

  15. Neutronics design study on a minor actinide burner for transmuting spent fuel

    International Nuclear Information System (INIS)

    Choi, Hang Bok

    1998-08-01

    A liquid metal reactor was designed for the primary purpose of burning the minor actinide waste from commercial light water reactors. The design was constrained to maintain acceptable safety performance as measured by the burnup reactivity swing, the doppler coefficient, and the sodium void worth. Sensitivity studies were performed for homogeneous and decoupled core designs, and a minor actinide burner design was determined to maximize actinide consumption and satisfy safety constraints. One of the principal innovations was the use of two core regions, with a fissile plutonium outer core and an inner core consisting only of minor actinides. The physics studies performed here indicate that a 1200 MWth core is able to transmute the annual minor actinide inventory of about 16 LWRs and still exhibit reasonable safety characteristics. (author). 34 refs., 22 tabs., 14 figs

  16. Comparison calculations for an accelerator-driven minor actinide burner

    International Nuclear Information System (INIS)

    2002-01-01

    International interest in accelerator-driven systems (ADS) has recently been increasing in view of the important role that these systems may play as efficient minor actinide and long-lived fission-product (LLFP) burners and/or energy producers with an enhanced safety potential. However, the current methods of analysis and nuclear data for minor actinide and LLFP burners are not as well established as those for conventionally fuelled reactor systems. Hence, in 1999, the OECD/NEA Nuclear Science Committee organised a benchmark exercise for an accelerator-driven minor actinide burner to check the performances of reactor codes and nuclear data for ADS with unconventional fuel and coolant. The benchmark model was a lead-bismuth-cooled subcritical system driven by a beam of 1 GeV protons. This report provides an analysis of the results supplied by seven participants from eight countries. The analysis reveals significant differences in important neutronic parameters, indicating a need for further investigation of the nuclear data, especially minor actinide data, as well as the calculation methods. This report will be of particular interest to reactor physicists and nuclear data evaluators developing nuclear systems for nuclear waste management. (authors)

  17. Heterogeneous fuels for minor actinides transmutation: Fuel performance codes predictions in the EFIT case study

    Energy Technology Data Exchange (ETDEWEB)

    Calabrese, R., E-mail: rolando.calabrese@enea.i [ENEA, Innovative Nuclear Reactors and Fuel Cycle Closure Division, via Martiri di Monte Sole 4, 40129 Bologna (Italy); Vettraino, F.; Artioli, C. [ENEA, Innovative Nuclear Reactors and Fuel Cycle Closure Division, via Martiri di Monte Sole 4, 40129 Bologna (Italy); Sobolev, V. [SCK.CEN, Belgian Nuclear Research Centre, Boeretang 200, B-2400 Mol (Belgium); Thetford, R. [Serco Technical and Assurance Services, 150 Harwell Business Centre, Didcot OX11 0QB (United Kingdom)

    2010-06-15

    Plutonium recycling in new-generation fast reactors coupled with minor actinides (MA) transmutation in dedicated nuclear systems could achieve a decrease of nuclear waste long-term radiotoxicity by two orders of magnitude in comparison with current once-through strategy. In a double-strata scenario, purpose-built accelerator-driven systems (ADS) could transmute minor actinides. The innovative nuclear fuel conceived for such systems demands significant R and D efforts in order to meet the safety and technical performance of current fuel systems. The Integrated Project EUROTRANS (EUROpean research programme for the TRANSmutation of high level nuclear waste in ADS), part of the EURATOM Framework Programme 6 (FP6), undertook some of this research. EUROTRANS developed from the FP5 research programmes on ADS (PDS-XADS) and on fuels dedicated to MA transmutation (FUTURE, CONFIRM). One of its main objectives is the conceptual design of a small sub-critical nuclear system loaded with uranium-free fuel to provide high MA transmutation efficiency. These principles guided the design of EFIT (European Facility for Industrial Transmutation) in the domain DESIGN of IP EUROTRANS. The domain AFTRA (Advanced Fuels for TRAnsmutation system) identified two composite fuel systems: a ceramic-ceramic (CERCER) where fuel particles are dispersed in a magnesia matrix, and a ceramic-metallic (CERMET) with a molybdenum matrix in the place of MgO matrix to host a ceramic fissile phase. The EFIT fuel is composed of plutonium and MA oxides in solid solution with isotopic vectors typical of LWR spent fuel with 45 MWd/kg{sub HM} discharge burnup and 30 years interim storage before reprocessing. This paper is focused on the thermomechanical state of the hottest fuel pins of two EFIT cores of 400 MW{sub (th)} loaded with either CERCER or CERMET fuels. For calculations three fuel performance codes were used: FEMALE, TRAFIC and TRANSURANUS. The analysis was performed at the beginning of fuel life

  18. The analysis and handling concept of minor actinides of NPP’s waste by using Ads technology

    International Nuclear Information System (INIS)

    Silakhuddin

    2008-01-01

    The contents of minor actinide elements (americium, neptunium and curium) on the spent fuel inventory from PWR operation of NPP have been calculated using Vista program. The calculation used parameters: enrichment 3.968%, power 1000 M We and burn-up is 60 M Wd/kg. The result of calculation showed that the arising of minor actinide elements on the spent fuel is 16.205 kg/year and 43.471 kg/year for PWR-UOX and PWR-MOX respectively. It is also discussed a concept of the use of ADS technology for transmuting the minor actinide elements contained in spent fuels. The result of the discussion showed that an ADS of 400 M Wth will serve 7 PWRs-UOX, and on the PWR system using UOX and MOX fuels an ADS will serve 3 PWRs. (author)

  19. What fits best minor actinides as a die material?

    International Nuclear Information System (INIS)

    Hinfray, J.

    2003-01-01

    Zirconolite might be the die material used to confine actinides definitively. Cea's teams have been investigating the ability of zirconolite to trap actinide atoms in its own crystal structure. These studies have been performed with 239 Pu that presents the same ability to set chemical links with the constituents of the die as 3 minor actinides do. Crystal materials like zirconolite are more sensitive to self irradiation than glass. The next step of the characterization of zirconolite is to evaluate its capacity to sustain self alpha irradiation. In order to do so, 238 Pu is used since its relatively short period (T = 87 years) allows an acceleration of the process : damages cumulated in the die material in 2 years will be equivalent to those produced by minor actinides for millions years. The results will be known in 2004. (A.C.)

  20. Oxygen potential of a prototypic Mo-cermet fuel containing plutonium oxide

    Energy Technology Data Exchange (ETDEWEB)

    Miwa, Shuhei, E-mail: miwa.shuhei@jaea.go.jp [Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki, 311-1393 (Japan); Osaka, Masahiko [Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki, 311-1393 (Japan); Nozaki, Takahiro; Arima, Tatsumi; Idemitsu, Kazuya [Kyushu University, 744 Motooka Nishi-ku, Fukuoka, 819-0395 (Japan)

    2015-10-15

    Oxygen potential of a prototypic Mo-cermet fuel containing 50 vol.% PuO{sub 2−x} were investigated by the thermogravimetric analysis in the temperature range from 1273 K to 1473 K. It was shown that the oxygen potential and oxidation rate of the Mo-cermet were the same as those of pure PuO{sub 2−x} below the oxygen potential of Mo/MoO{sub 2} oxidation reaction. The same features of the Mo-cermet sample containing 50 vol.% PuO{sub 2−x} with those of pure PuO{sub 2−x} were discussed in terms of the microstructure. - Highlights: • Oxygen potential of Mo-cermet fuel was investigated by thermogravimetric analysis. • It was the same as that of pure PuO{sub 2−x} below the oxygen potential for Mo/MoO{sub 2}. • Gradual oxidation of Mo matrix occurred only above the oxygen potential for Mo/MoO{sub 2}. • Mo matrix and PuO{sub 2−x} in Mo-cermet fuel can thus be thermochemically individual.

  1. Oxygen potential of a prototypic Mo-cermet fuel containing plutonium oxide

    International Nuclear Information System (INIS)

    Miwa, Shuhei; Osaka, Masahiko; Nozaki, Takahiro; Arima, Tatsumi; Idemitsu, Kazuya

    2015-01-01

    Oxygen potential of a prototypic Mo-cermet fuel containing 50 vol.% PuO_2_−_x were investigated by the thermogravimetric analysis in the temperature range from 1273 K to 1473 K. It was shown that the oxygen potential and oxidation rate of the Mo-cermet were the same as those of pure PuO_2_−_x below the oxygen potential of Mo/MoO_2 oxidation reaction. The same features of the Mo-cermet sample containing 50 vol.% PuO_2_−_x with those of pure PuO_2_−_x were discussed in terms of the microstructure. - Highlights: • Oxygen potential of Mo-cermet fuel was investigated by thermogravimetric analysis. • It was the same as that of pure PuO_2_−_x below the oxygen potential for Mo/MoO_2. • Gradual oxidation of Mo matrix occurred only above the oxygen potential for Mo/MoO_2. • Mo matrix and PuO_2_−_x in Mo-cermet fuel can thus be thermochemically individual.

  2. BWR Assembly Optimization for Minor Actinide Recycling

    International Nuclear Information System (INIS)

    Maldonado, G. Ivan; Christenson, John M.; Renier, J.P.; Marcille, T.F.; Casal, J.

    2010-01-01

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs). A top-level objective of the Advanced Fuel Cycle Systems Analysis program element of the DOE NERI program is to investigate spent fuel treatment and recycling options for current light water reactors (LWRs). Accordingly, this project targets to expand the traditional scope of nuclear fuel management optimization into the following two complementary specific objectives: (1) To develop a direct coupling between the pin-by-pin within-bundle loading control variables and core-wide (bundle-by-bundle) optimization objectives, (2) to extend the methodology developed to explicitly encompass control variables, objectives, and constraints designed to maximize minor actinide incineration in BWR bundles and cycles. The first specific objective is projected to 'uncover' dormant thermal margin made available by employing additional degrees of freedom within the optimization process, while the addition of minor actinides is expected to 'consume' some of the uncovered thermal margin. Therefore, a key underlying goal of this project is to effectively invest some of the uncovered thermal margin into achieving the primary objective.

  3. Minor Actinides Recycling in PWRs

    International Nuclear Information System (INIS)

    Delpech, M.; Golfier, H.; Vasile, A.; Varaine, F.; Boucher, L.; Greneche, D.

    2006-01-01

    Recycling of minor actinides in current and near future PWR is considered as one of the options of the general waste management strategy. This paper presents the analysis of this option both from the core physics and fuel cycle point of view. A first indicator of the efficiency of different neutron spectra for transmutation purposes is the capture to fission cross sections ratio which is less favourable by a factor between 5 to 10 in PWRs compared to fast reactors. Another indicator presented is the production of high ranking isotopes like Curium, Berkelium or Californium in the thermal or epithermal spectrum conditions of PWR cores by successive neutron captures. The impact of the accumulation of this elements on the fabrication process of such PWR fuels strongly penalizes this option. The main constraint on minor actinides loadings in PWR (or fast reactors) fuels are related to their direct impact (or the impact of their transmutation products) on the reactivity coefficients, the reactivity control means and the core kinetics parameters. The main fuel cycle physical parameters like the neutron source, the alpha decay power, the gamma and neutrons dose rate and the criticality aspects are also affected. Recent neutronic calculations based on a reference core of the Evolutionary Pressurized Reactor (EPR), indicates typical maximum values of 1 % loadings. Different fuel design options for minor actinides transmutation purposes in PWRs are presented: UOX and MOX, homogeneous and heterogeneous assemblies. In this later case, Americium loading is concentrated in specific pins of a standard UOX assembly. Recycling of Neptunium in UOX and MOX fuels was also studied to improve the proliferation resistance of the fuel. The impact on the core physics and penalties on Uranium enrichment were underlined in this case. (authors)

  4. 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

  5. Utilization of Minor Actinides as a Fuel Component for Ultra-Long Life VHTR Configurations: Designs, Advantages and Limitations

    International Nuclear Information System (INIS)

    Tsvetkov, Pavel V.

    2009-01-01

    This project assessed the advantages and limitations of using minor actinides as a fuel component to achieve ultra-long life Very High Temperature Reactor (VHTR) configurations. Researchers considered and compared the capabilities of pebble-bed and prismatic core designs with advanced actinide fuels to achieve ultra-long operation without refueling. Since both core designs permit flexibility in component configuration, fuel utilization, and fuel management, it is possible to improve fissile properties of minor actinides by neutron spectrum shifting through configuration adjustments. The project studied advanced actinide fuels, which could reduce the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository and enable recovery of the energy contained in spent fuel. The ultra-long core life autonomous approach may reduce the technical need for additional repositories and is capable to improve marketability of the Generation IV VHTR by allowing worldwide deployment, including remote regions and regions with limited industrial resources. Utilization of minor actinides in nuclear reactors facilitates developments of new fuel cycles towards sustainable nuclear energy scenarios.

  6. Utilization of Minor Actinides as a Fuel Component for Ultra-Long Life Bhr Configurations: Designs, Advantages and Limitations

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Pavel V. Tsvetkov

    2009-05-20

    This project assessed the advantages and limitations of using minor actinides as a fuel component to achieve ultra-long life Very High Temperature Reactor (VHTR) configurations. Researchers considered and compared the capabilities of pebble-bed and prismatic core designs with advanced actinide fuels to achieve ultra-long operation without refueling. Since both core designs permit flexibility in component configuration, fuel utilization, and fuel management, it is possible to improve fissile properties of minor actinides by neutron spectrum shifting through configuration adjustments. The project studied advanced actinide fuels, which could reduce the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository and enable recovery of the energy contained in spent fuel. The ultra-long core life autonomous approach may reduce the technical need for additional repositories and is capable to improve marketability of the Generation IV VHTR by allowing worldwide deployment, including remote regions and regions with limited industrial resources. Utilization of minor actinides in nuclear reactors facilitates developments of new fuel cycles towards sustainable nuclear energy scenarios.

  7. Preliminary Study for Inventories of Minor Actinides in Thorium Molten Salt Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Choong Wie; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    It has different characteristic with the conventional reactors which use a solid fuel. It can continually supply the fuel by online refueling and reprocessing of minor actinides so that those can be separated and eliminated from the reactor. The MSR maintains steady state except initial stage and the reactor becomes stable. In this research, considering online refueling, bubbling and reprocessing, the basic concept for evaluation of the inventory of minor actinide in the molten salt reactor is driven using the Bateman equation. The simulation results, where REM and MCNP code from CNRS (Centre National de la Recherche Scientifique) applied to the concept equation are analyzed. The analysis of the basic concept was carried out for evaluation of the inventory of the minor actinides in MSR. It was thought that the inventories of the minor actinides should be evaluated by solving the modified Bateman equation due to the MSR characteristic of online refueling, chemical reprocessing and bubbling.

  8. Preliminary Study for Inventories of Minor Actinides in Thorium Molten Salt Reactor

    International Nuclear Information System (INIS)

    Lee, Choong Wie; Kim, Hee Reyoung

    2015-01-01

    It has different characteristic with the conventional reactors which use a solid fuel. It can continually supply the fuel by online refueling and reprocessing of minor actinides so that those can be separated and eliminated from the reactor. The MSR maintains steady state except initial stage and the reactor becomes stable. In this research, considering online refueling, bubbling and reprocessing, the basic concept for evaluation of the inventory of minor actinide in the molten salt reactor is driven using the Bateman equation. The simulation results, where REM and MCNP code from CNRS (Centre National de la Recherche Scientifique) applied to the concept equation are analyzed. The analysis of the basic concept was carried out for evaluation of the inventory of the minor actinides in MSR. It was thought that the inventories of the minor actinides should be evaluated by solving the modified Bateman equation due to the MSR characteristic of online refueling, chemical reprocessing and bubbling

  9. Dense cermets containing fine grained ceramics and their manufacture

    International Nuclear Information System (INIS)

    King, H.L.

    1986-01-01

    This patent describes a method of producing a ceramic-metal composite (cermet) containing boride-oxide ceramic having components of a first metal boride and a second metal oxide, which ceramic is in mixture in the cermet with elemental metal of the second metal, wherein the cermet is produced by sintering a reaction mixture of the first metal oxide, boron oxide and the elemental second metal. The improvement consists of: combining for the reaction mixture; A. (a) first metal oxide; (b) boron oxide; (c) ceramic component in very finely divided form; and (d) elemental second metal in very finely divided form and in an amount of at least a 100 percent molar excess beyond that amount stoichiometrically required to produce the second metal oxide during sintering; and B. sintering the reaction mixture in inert gas atmosphere

  10. Thermal neutrons core concepts for minor actinides inventory reduction

    International Nuclear Information System (INIS)

    Huang, Shio-Ling

    1996-01-01

    The goal of this thesis is to propose a solution to the problem of reducing the inventory of Minor Actinides, discharged from PWR spent fuel, in the framework of a Separation/ Transmutation strategy. The solution envisaged is based on the utilisation of Pressurised Water Reactors (PWR), of the same type as those used to produce energy. The suggested solution is original and based on a special Assembly ANDIAMO dedicated to transmutation, where Actinide incineration is performed with the help of a fissile support in a once-through strategy. During this study, we have also tackled the impact of some parameters which so far have been less carefully studied (like the unavoidable presence of Lanthanides in fuel containing Am and Cm and the consequences on the cycle parameters with Actinide recycle). Moreover, we have carried out a sensitivity study in order to analysis the impact of nuclear data uncertainties on some important parameters of the reactor (reactivity coefficients) and on the isotopic concentration. This original study allows us to assess the accuracy of the results, of the presented tendencies and of the propositions made in the present thesis. (author) [fr

  11. BWR Assembly Optimization for Minor Actinide Recycling

    Energy Technology Data Exchange (ETDEWEB)

    G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

    2010-03-22

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

  12. The EBR-II X501 Minor Actinide Burning Experiment

    Energy Technology Data Exchange (ETDEWEB)

    W. J. Carmack; M. K. Meyer; S. L. Hayes; H. Tsai

    2008-01-01

    The X501 experiment was conducted in EBR II as part of the Integral Fast Reactor program to demonstrate minor actinide burning through the use of a homogeneous recycle scheme. The X501 subassembly contained two metallic fuel elements loaded with relatively small quantities of americium and neptunium. Interest in the behavior of minor actinides (MA) during fuel irradiation has prompted further examination of existing X501 data and generation of new data where needed in support of the U.S. waste transmutation effort. The X501 experiment is one of the few MA bearing fuel irradiation tests conducted worldwide, and knowledge can be gained by understanding the changes in fuel behavior due to addition of MAs. Of primary interest are the effect of the MAs on fuel cladding chemical interaction and the redistribution behavior of americium. The quantity of helium gas release from the fuel and any effects of helium on fuel performance are also of interest. It must be stressed that information presented at this time is based on the limited PIE conducted in 1995–1996 and, currently, represents a set of observations rather than a complete understanding of fuel behavior. This report provides a summary of the X501 fabrication, characterization, irradiation, and post irradiation examination.

  13. Minor Actinides Burnup Enhancement in the European Sodium Fast Reactor through Moderator Material Addition

    International Nuclear Information System (INIS)

    Ramos, R.L.; Buiron, L.

    2013-01-01

    Conclusions: • ZrH 2 was the best moderator material, followed by MgO and MgAl 2 O 4 ; • When the number of moderator pins is increased: – the percentage of minor actinides consumed increases; – the total mass consumed of minor actinides decreases; – the decay heat generated decreases; – the neutron flux in the reactor varies very little. Perspectives: • For future studies it would be possible to evaluate the use of other materials with resonances in the scattering cross section in the fast range that would improve the results obtained with Mg. • It would be necessary to consider how to add moderator material without changing the initial mass of minor actinides. E.g., adding the moderator at the periphery of the minor actinide elements

  14. The neutronics design and analysis of a liquid metal reactor for burning minor actinides

    International Nuclear Information System (INIS)

    Choi, H.B.; Downar, T.J.

    1992-01-01

    A liquid metal reactor was designed for the primary purpose of burning the minor actinide waste from commercial light water reactors (LWR). The design was constrained to maintain acceptable safety performance as measured by the burnup reactivity swing, the Doppler coefficient, and the sodium void worth. One of the principal innovations was the use of two core regions, with a fissile plutonium outer core and an inner core consisting only of minor actinides. The physics studies performed here indicate that a 1200 MWth core is able to transmute the annual minor actinide inventory of about 26 LWRs and still exhibit reasonable safety characteristics. Sensitivity analysis of the final core design indicates deficiencies in the minor actinide nuclear data can introduce large uncertainties in the prediction of the core safety performance parameters

  15. Minor actinide transmutation in accelerator driven systems

    Energy Technology Data Exchange (ETDEWEB)

    Friess, Friederike [IANUS, TU Darmstadt (Germany)

    2015-07-01

    Transmutation of radioactive waste, the legacy of nuclear energy use, gains rising interest. This includes the development of facilities able to transmute minor actinides (MA) into stable or short-lived isotopes before final disposal. The most common proposal is to use a double-strata approach with accelerator-driven-systems (ADS) for the efficient transmutation of MA and power reactors to dispose plutonium. An ADS consists of a sub-critical core that reaches criticality with neutrons supplied by a spallation target. An MCNP model of the ADS system Multi Purpose Research Reactor for Hightech Applications will be presented. Depletion calculations have been performed for both standard MOX fuel and transmutation fuel with an increased content of minor actinides. The resulting transmutation rates for MAs are compared to published values. Special attention is given to selected fission products such as Tc-99 and I-129, which impact the radiation from the spent fuel significantly.

  16. Study of minor actinides transmutation in heavy water cooled tight-pitch lattice

    International Nuclear Information System (INIS)

    Xu Xiaoqin; Shiroya, S.

    2002-01-01

    Minor actinides inhere long half-life and high toxicity. It is an alternative technical pathway and helpful for reducing environmental impact to incinerate minor actinides in spent fuel of nuclear power plants. Because of its high neutron, γ and β emitting rates and heat generation rate, it is necessary to imply more severe control and shielding techniques in the chemical treatment and fabrication. From economic view-point, it is suitable to transmute minor actinides in concentrated way. A technique for MA transmutation by heavy water cooled tight-pitch lattice system is proposed, and calculated with SRAC95 code system. It is shown that tight-pitch heavy water lattice can transmute MA effectively. The accelerator-driven subcritical system is practical for MA transmutation because of its low fraction of effective delay neutrons

  17. Preparation of minor actinides targets or blankets by means of ionic exchange resin

    Energy Technology Data Exchange (ETDEWEB)

    Picart, S.; Mokhtari, H.; Jobelin, I. [CEA Marcoule, Nucl Energy Div, RadioChem and Proc Dept, Actinides Chem and Convers, F-30207 Bagnols Sur Ceze (France)

    2010-07-01

    The conversion of minor actinides to fuel starting materials for transmutation in a closed nuclear cycle is a big challenge for the next decades and the development of Gen(IV) nuclear systems. Conversion routes are numerous, but one needs to prove that they can be adapted to handle minor actinides. One of them is called the resin process and is particularly attractive because it stands for a 'dustless' process as it produces microspheres of oxide or carbide after thermal treatment of the loaded resin. The study presented herein focuses on the experiments and tests which enable us to optimize the fixation of minor actinides onto ionic exchange resin and their carbonization into oxide type materials. (authors)

  18. Preparation of minor actinides targets or blankets by means of ionic exchange resin

    International Nuclear Information System (INIS)

    Picart, S; Mokhtari, H; Jobelin, I; Ramiere, I

    2010-01-01

    The conversion of minor actinides to fuel starting materials for transmutation in a closed nuclear cycle is a big challenge for the next decades and the development of Gen(IV) nuclear systems. Conversion routes are numerous, but one needs to prove that they can be adapted to handle minor actinides. One of them is called the resin process and is particularly attractive because it stands for a 'dustless' process as it produces microspheres of oxide or carbide after thermal treatment of the loaded resin. The study presented herein focuses on the experiments and tests which enable us to optimize the fixation of minor actinides onto ionic exchange resin and their carbonization into oxide type materials.

  19. Synthesis of selective extractor for minor actinide elements

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Seung [Konyang University, Nonsan (Korea); Cho, Moon Hwan [Kangwon National University, Chunchon (Korea)

    1998-04-01

    To selectively co-separate the lanthanide and actinide elements (MA) such as Am or Cm ion from radioactive waste, synthesis of diamide derivatives has been accomplished. In addition, picoline amide derivatives were also synthesized for selectively separate the minor actinide elements from lanthanide elements. The content of research has don are as follows: (1) synthesis of diamide as co-extractant (2) introduction of n-tetradecyl to increase the lipophilicity (3) Picolyl chloride, intermediate of the final product, was synthesized by improved method rather than reported method. (4) The length of alkyl side chain was adjusted to increase the lipophilicity of free ligand and its derivatives able to selectively separate the actinide metal from lanthanide metal ions was successfully synthesized and determined their purity by analytical instruments. (author). 12 refs., 28 figs.

  20. Reduction of minor actinides for recycling in a light water reactor; Reduccion de actinidos menores por reciclado en un reactor de agua ligera

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.; Ramirez S, J. R.; Alonso V, G., E-mail: eduardo.martinez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    The aim of actinide transmutation from spent nuclear fuel is the reduction in mass of high-level waste which must be stored in geological repositories and the lifetime of high-level waste; these two achievements will reduce the number of repositories needed, as well as the duration of storage. The present work is directed towards the evaluation of an advanced nuclear fuel cycle in which the minor actinides (Np, Am and Cm) could be recycled to remove most of the radioactive material; a reference of actinides production in standard nuclear fuel of uranium at the end of its burning in a BWR is first established, after a design of fuel rod containing 6% of minor actinides in a matrix of uranium from the enrichment lines is proposed, then 4 fuel rods of standard uranium are replaced by 4 actinides bars to evaluate the production and transmutation of them and finally the minor actinides reduction in the fuel is evaluated. In the development of this work the calculation tool are the codes: Intrepin-3, Casmo-4 and Simulate-3. (Author)

  1. Irradiation test of fuel containing minor actinides in the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    Soga, Tomonori; Sekine, Takashi; Wootan, David; Tanaka, Kosuke; Kitamura, Ryoichi; Aoyama, Takafumi

    2007-01-01

    The mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast reactor Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted in the Joyo MK-III 3rd operational cycle. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX), and reference MOX fuel. The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes in order to confirm whether or not fuel melting occurred. After 10 minutes irradiation in May 2006, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins including neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. Linear heat rates for each pin were calculated using MCNP, accounting for both prompt and delayed heating components, and then adjusted using E/C for 10 B (n, α) reaction rates measured in the MK-III core neutron field characterization test. Post irradiation examination of these pins to confirm the fuel melting and the local concentration under irradiation of NpO 2-x or AmO 2-x in the (U, Pu)O 2-x fuel are underway. The test results are expected to reduce uncertainties on the design margin in the thermal design for MA-MOX fuel. (author)

  2. Prognosis and comparison of performances of composite CERCER and CERMET fuels dedicated to transmutation of TRU in an EFIT ADS

    Science.gov (United States)

    Sobolev, V.; Uyttenhove, W.; Thetford, R.; Maschek, W.

    2011-07-01

    The neutronic and thermomechanical performances of two composite fuel systems: CERCER with (Pu,Np,Am,Cm)O 2-x fuel particles in ceramic MgO matrix and CERMET with metallic Mo matrix, selected for transmutation of minor actinides in the European Facility for Industrial Transmutation (EFIT), were analysed aiming at their optimisation. The ALEPH burnup code system, based on MNCPX and ORIGEN codes and JEFF3.1 nuclear data library, and the modern version of the fuel rod performance code TRAFIC were used for this analysis. Because experimental data on the properties of the mixed minor-actinide oxides are scarce, and the in-reactor behaviour of the T91 steel chosen as cladding, as well as of the corrosion protective layer, is still not well-known, a set of "best estimates" provided the properties used in the code. The obtained results indicate that both fuel candidates, CERCER and CERMET, can satisfy the fuel design and safety criteria of EFIT. The residence time for both types of fuel elements can reach about 5 years with the reactivity swing within ±1000 pcm, and about 22% of the loaded MA is transmuted during this period. However, the fuel centreline temperature in the hottest CERCER fuel rod is close to the temperature above which MgO matrix becomes chemically instable. Moreover, a weak PCMI can appear in about 3 years of operation. The CERMET fuel can provide larger safety margins: the fuel temperature is more than 1000 K below the permitted level of 2380 K and the pellet-cladding gap remains open until the end of operation.

  3. Accuracy Improvement of Neutron Nuclear Data on Minor Actinides

    Science.gov (United States)

    Harada, Hideo; Iwamoto, Osamu; Iwamoto, Nobuyuki; Kimura, Atsushi; Terada, Kazushi; Nakao, Taro; Nakamura, Shoji; Mizuyama, Kazuhito; Igashira, Masayuki; Katabuchi, Tatsuya; Sano, Tadafumi; Takahashi, Yoshiyuki; Takamiya, Koichi; Pyeon, Cheol Ho; Fukutani, Satoshi; Fujii, Toshiyuki; Hori, Jun-ichi; Yagi, Takahiro; Yashima, Hiroshi

    2015-05-01

    Improvement of accuracy of neutron nuclear data for minor actinides (MAs) and long-lived fission products (LLFPs) is required for developing innovative nuclear system transmuting these nuclei. In order to meet the requirement, the project entitled as "Research and development for Accuracy Improvement of neutron nuclear data on Minor ACtinides (AIMAC)" has been started as one of the "Innovative Nuclear Research and Development Program" in Japan at October 2013. The AIMAC project team is composed of researchers in four different fields: differential nuclear data measurement, integral nuclear data measurement, nuclear chemistry, and nuclear data evaluation. By integrating all of the forefront knowledge and techniques in these fields, the team aims at improving the accuracy of the data. The background and research plan of the AIMAC project are presented.

  4. Utilization of fast reactor excess neutrons for burning minor actinides and long lived FPs

    International Nuclear Information System (INIS)

    Kawashima, K.; Kobayashi, K.; Kaneto, K.

    1995-01-01

    An evaluation is made on a large MOX fuel fast reactor's capability of burning minor actinides and long lived fission products (FPs) without imposing penalties on core nuclear and safety characteristics. The excess neutrons generated in the fast reactor core are fully utilized not only to generate the fissile material but also to transmute the minor actinides and long lived FPs. The FP target assemblies which consist of Tc-99 and I-129 are loaded into the selected blanket positions whereas the minor actinides are loaded to the rest of the blanket. A long term FP accumulation scenario is also considered in the mix of FP burner fast reactor and non-burner LWRs. (author)

  5. Summary report of consultants meeting on Minor Actinide Nuclear Reaction Data (MANREAD)

    International Nuclear Information System (INIS)

    Plompen, A.; Mengoni, A.

    2007-10-01

    A Consultants Meeting on reaction cross section data for minor actinides was held at the IAEA Headquarters, in Vienna on 23 and 24 November, 2007. The main objective of the initiative was to define the detailed plan for the Co-ordinated Research Project on Minor Actinide Neutron Reaction Data (MANREAD) CRP. The details of the discussions which took place at the reported meeting include a review of the current activities in the field, a list of recommendations and a proposed timescale for the CRP. (author)

  6. PIE analysis for minor actinide

    International Nuclear Information System (INIS)

    Suyama, Kenya

    2005-01-01

    Minor actinide (MA) is generated in nuclear fuel during the operation of power reactor. For fuel design, reactivity decrease due to it should be considered. Out of reactors, MA plays key role to define the property of spent fuel (SF) such as α-radioactivity, neutron emission rate, and criticality of SF. In order to evaluate the calculation codes and libraries for predicting the amount of MA, comparison between calculation results and experimentally obtained data has been conducted. In this report, we will present the status of PIE data of MA taken by post irradiation examinations (PIE) and several calculation results. (author)

  7. A liquid-metal reactor for burning minor actinides of spent light water reactor fuel. 1: Neutronics design study

    International Nuclear Information System (INIS)

    Choi, H.; Downar, T.J.

    1999-01-01

    A liquid-metal reactor was designed for the primary purpose of burning the minor actinide waste from commercial light water reactors (LWRs). The design was constrained to maintain acceptable safety performance as measured by the burnup reactivity swing, the Doppler constant, and the sodium void worth. Sensitivity studies were performed for homogeneous and decoupled core designs, and a minor actinide burner design was determined to maximize actinide consumption and satisfy safety constraints. One of the principal innovations was the use of two core regions, with a fissile plutonium outer core and an inner core consisting only of minor actinides. The physics studies performed here indicate that a 1200-MW(thermal) core is able to consume the annual minor actinide inventory of about 16 LWRs and still exhibit reasonable safety characteristics

  8. Preparation of minor actinides targets or blankets by the means of Ionic Exchange Resin

    Energy Technology Data Exchange (ETDEWEB)

    Picart, S.; Mokhtari, H.; Ramiere, I.; Jobelin, I. [CEA, Nuclear Energy Division, RadioChemistry and Process Department, Actinides Chemistry Laboratory, BP17171, Bagnols-sur-Ceze, 30207 (France)

    2009-06-15

    The objective of our R and D work is the elaboration by the use of ionic exchange resin of minor actinide precursors for target or blanket dedicated to their transmutation in sodium fast reactor. From the beginning, the resin process called WAR (acronym of Weak Acid Resin) was developed in the 70's at the ORNL for the making of uranium carbide kernels for the high temperature gas reactor [1] [2]. By now, our aim is to extend this concept to the manufacturing of minor actinides oxide mixed with uranium oxides [3]. More precisely, this process can be divided in two major steps: the loading of the resin and the thermal treatment of the fully loaded resin driving either to oxide or carbide phases depending on the gas atmosphere. The difficulty stems from the preparation of the loading solutions which must fulfill precise conditions of pH in presence of actinides cations prone to hydrolysis. Furthermore, the proportions of uranium and minor actinides in solutions must be adjusted to fit the right ratio in the solid. The study presented here will then focus on the experiments and tests which enable us to optimize the fixing of minor actinides on ionic exchange resin and their carbonization in oxide. [1] G. W. Weber, R. L. Beatty et V. J. Tennery, Nuclear Technology, 35, 217-226, (1977), 'Processing and composition control of weak-acid-resin derived fuel microspheres'. [2] K. J. Notz, P. A. Haas, J. H. Shaffer, Radiochimica Acta, 25, 153-160, (1978). 'The preparation of HTGR Fissile Fuel Kernels by Uranium Loading of Ion Exchange Resin'. [3] S. Picart, H. Mokhtari, I. Ramiere, 'Plutonium Futures, The Science 2008', 7-11 july 2008, Dijon, France. 'Modelling of the ionic Exchange between a weak acid resin in its ammonium form and a minor actinide'. (authors)

  9. Study of nuclear energy systems and double strata scenarios for minor actinides transmutation in ADS

    International Nuclear Information System (INIS)

    Clavel, J.B.

    2012-01-01

    The French law of 28 June 2006 regarding advanced nuclear waste management requires a scientific assessment to define future industrial strategies. The present PhD thesis was carried in this framework and concerns specifically the research axis of minor actinides transmutation. A high power Accelerator Driven System (ADS) concept is developed at SUBATECH for this purpose. A 1 GeV proton beam feeds three liquid lead-bismuth spallation targets. The Multiple Spallation Target (MUST) ADS reaches the thermal powers up to 1 GW with a high specific power. A nuclear reactor dimensioning method has been developed and applied to different double strata scenarios. In these scenarios, SFR (Sodium Fast Reactors) or PWR (Pressurized Water Reactors) power reactors produce minor actinides that will be transmuted into ADS. In each core (SFR and ADS), the plutonium multi-reprocessing strategy is performed while ADS subcritical core also multi-reprocesses minor actinides. To limit the core reactivity and improve the fuel thermal conductivity, the minor actinides fuel is mixed with MgO inert matrix. Nuclear branches with lead and sodium coolants for the ADS, have been studied for different irradiation times and two transmutation strategies have been assessed: whether whole minor actinides, whether americium only is transmuted. The thesis presents precisely the MUST ADS design methodology and the calculations to get a fuel composition at equilibrium. Then a one cycle evolution is performed and analysed for the fuel and the multiplication factor. Radiotoxicity and thermal power of the waste produced are then compared. Finally, the study of double strata scenarios is performed to analyse the plutonium and minor actinides inventories in cycle and also the waste produced according to the transmutation strategies applied and the first stratum evolution. (author)

  10. Status of measurements of fission neutron spectra of Minor Actinides

    Energy Technology Data Exchange (ETDEWEB)

    Drapchinsky, L.; Shiryaev, B. [V.G. Khlopin Radium Inst., Saint Petersburg (Russian Federation)

    1997-03-01

    The report considers experimental and theoretical works on studying the energy spectra of prompt neutrons emitted in spontaneous fission and neutron induced fission of Minor Actinides. It is noted that neutron spectra investigations were done for only a small number of such nuclei, most measurements, except those of Cf-252, having been carried out long ago by obsolete methods and imperfectapparatus. The works have no detailed description of experiments, analysis of errors, detailed numerical information about results of experiments. A conclusion is made that the available data do not come up to modern requirements. It is necessary to make new measurements of fission prompt neutron spectra of transuranium nuclides important for the objectives of working out a conception of minor actinides transmutation by means of special reactors. (author)

  11. The cross section sensitivity of the minor actinides on a lead-bismuth cooled accelerator-driven burner system

    International Nuclear Information System (INIS)

    Gil, Choong-Sup; Kim, Jung-Do; Chang, Jonghwa

    2002-01-01

    In order to validate the detailed sensitivity of each minor actinide datum in ENDF/B-VI Release 6, JEF-2.2 and JENDL-3.2 on an accelerator-driven minor actinide burner benchmark system, a lead-bismuth cooled sub-critical system was analyzed. The impacts on the system by the ten minor actinides were compared. The k eff values and reaction rates were calculated by exchanging the data sets of each minor actinide from ENDF/B-VI.6 to JEF-2.2 or JENDL-3.2. At the equilibrium core, the k eff differences from ENDF/B-VI.6 by the ten minor actinides can cause more than 5,500 pcm for JEF-2.2 and 3,500 pcm for JENDL-3.2. The fission reaction rates of 242m Am and 243 Cm with ENDF/B-VI.6 show differences of more than 15% from those with JEF-2.2 and JENDL-3.2. 241 Am, 243 Am and 245 Cm in JEF-2.2 and americium isotope data and 245 Cm in JENDL-3.2 are sensitive to the fission spectrum. (author)

  12. Sintering of Mo2FeB2 based cermet and its layered composites containing Sic fibers

    International Nuclear Information System (INIS)

    Rao, D.; Upadhyaya, G.S.

    2001-01-01

    In the present investigation Mo 2 FeB 2 based cermet (KH-C50) and its composites containing SiC fibers were sintered in two different atmospheres namely hydrogen and vacuum. It was observed that vacuum sintered samples have remarkably lower porosities than the hydrogen sintered ones. Two different sintering cycles were employed for each of the atmosphere and properties of the material were studied. Introduction of fibers in the composite imparts shrinkage anisotropy during sintering. Fiber containing cermets have rather poor densification and transverse rupture strength (TRS). TRS, macro and microhardness, and boride grain size measurements were also carried out for the cermets sintered in different atmospheres. (author)

  13. Development of a fast reactor for minor actinides transmutation - (1) Overview and method development - 5092

    International Nuclear Information System (INIS)

    Takeda, T.; Usami, S.; Fujimura, K.; Takakuwa, M.

    2015-01-01

    The Ministry of Education, Culture, Sports, Science and Technology in Japan has launched a national project entitled 'technology development for the environmental burden reduction' in 2013. The present study is one of the studies adopted as the national project. The objective of the study is the efficient and safe transmutation and volume reduction of minor actinides (MA) with long-lived radioactivity and high decay heat contained in high level radioactive wastes by using sodium cooled fast reactors. We are developing MA transmutation core concepts which harmonize efficient MA transmutation with core safety. To accurately design the core concepts we have improved calculation methods for estimating the transmutation rate of individual MA nuclides, and estimating and reducing uncertainty of MA transmutation. The overview of the present project is first described. Then the method improvement is presented with numerical results for a minor-actinide transmutation fast reactor. The analysis is based on Monju reactor data. (authors)

  14. Limitations of actinide recycle and waste disposal consequences

    International Nuclear Information System (INIS)

    Baetsle, L.H.; Raedt, C. de

    1994-01-01

    The paper emphasizes the impact of Light Water Reactor - Mixed Oxides introduction on the subsequent actinide management and fate of reprocessed and depleted uranium. The spent fuel from LWR-MOX contains in principle 75% of the initially produced plutonium. This new source term has to be considered together with the minor actinides from the conventional reprocessing. Subsequent LWR-MOX reprocessing in the first step in a very long term Pu + minor actinides management. Recycling of Pu + minor actinides in fast reactors to significantly reduce the Pu and minor actinides inventory (e.g. a factor of 10) is a very slow process which requires the development and operation of a large park of actinide burner reactors during an extended period of time. The overall feasibility of the P and T option will greatly depend on the massive introduction during the next century of fast neutron reactors as a replacement to the present LWR generation of nuclear power plants. (authors). 11 refs., 6 tabs., 2 figs

  15. Detailed studies of Minor Actinide transmutation-incineration in high-intensity neutron fluxes

    International Nuclear Information System (INIS)

    Bringer, O.; Al Mahamid, I.; Blandin, C.; Chabod, S.; Chartier, F.; Dupont, E.; Fioni, G.; Isnard, H.; Letourneau, A.; Marie, F.; Mutti, P.; Oriol, L.; Panebianco, S.; Veyssiere, C.

    2006-01-01

    The Mini-INCA project is dedicated to the measurement of incineration-transmutation chains and potentials of minor actinides in high-intensity thermal neutron fluxes. In this context, new types of detectors and methods of analysis have been developed. The 241 Am and 232 Th transmutation-incineration chains have been studied and several capture and fission cross sections measured very precisely, showing some discrepancies with existing data or evaluated data. An impact study was made on different based-like GEN-IV reactors. It underlines the necessity to proceed to precise measurements for a large number of minor-actinides that contribute to these future incineration scenarios. (authors)

  16. Detailed studies of Minor Actinide transmutation-incineration in high-intensity neutron fluxes

    Energy Technology Data Exchange (ETDEWEB)

    Bringer, O. [CEA/Saclay/DSM/DAPNIA, Gif-sur-Yvette (France); Al Mahamid, I. [Lawrence Berkeley National Laboratory, E.H. and S. Div., CA (United States); Blandin, C. [CEA/Cadarache/DEN/DER/SPEX, Saint-Paul-lez-Durances (France); Chabod, S. [CEA/Saclay/DSM/DAPNIA, Gif-sur-Yvette (France); Chartier, F. [CEA/Cadarache/DEN/DPC/SECR, Gif-sur-Yvette (France); Dupont, E.; Fioni, G. [CEA/Saclay/DSM/DAPNIA, Gif-sur-Yvette (France); Isnard, H. [CEA/Cadarache/DEN/DPC/SECR, Gif-sur-Yvette (France); Letourneau, A.; Marie, F. [CEA/Saclay/DSM/DAPNIA, Gif-sur-Yvette (France); Mutti, P. [Institut Laue-Langevin, Grenoble (France); Oriol, L. [CEA/Cadarache/DEN/DER/SPEX, Saint-Paul-lez-Durances (France); Panebianco, S.; Veyssiere, C. [CEA/Saclay/DSM/DAPNIA, Gif-sur-Yvette (France)

    2006-07-01

    The Mini-INCA project is dedicated to the measurement of incineration-transmutation chains and potentials of minor actinides in high-intensity thermal neutron fluxes. In this context, new types of detectors and methods of analysis have been developed. The {sup 241}Am and {sup 232}Th transmutation-incineration chains have been studied and several capture and fission cross sections measured very precisely, showing some discrepancies with existing data or evaluated data. An impact study was made on different based-like GEN-IV reactors. It underlines the necessity to proceed to precise measurements for a large number of minor-actinides that contribute to these future incineration scenarios. (authors)

  17. Fission cross section measurements for minor actinides

    Energy Technology Data Exchange (ETDEWEB)

    Fursov, B. [IPPE, Obninsk (Russian Federation)

    1997-03-01

    The main task of this work is the measurement of fast neutron induced fission cross section for minor actinides of {sup 238}Pu, {sup 242m}Am, {sup 243,244,245,246,247,248}Cm. The task of the work is to increase the accuracy of data in MeV energy region. Basic experimental method, fissile samples, fission detectors and electronics, track detectors, alpha counting, neutron generation, fission rate measurement, corrections to the data and error analysis are presented in this paper. (author)

  18. Characterization of high level waste for minor actinides by chemical separation and alpha spectrometry

    International Nuclear Information System (INIS)

    Murali, M.S.; Bhattacharayya, A.; Kar, A.S.; Tomar, B.S.; Manchanda, V.K.

    2010-01-01

    Quantification of minor actinides present in of High Level Waste (HLW) solutions originating from the power reactors is important in view of management of radioactive wastes and actinide partitioning. Several methods such as ICP-MS, X-ray fluorescence methods, ICP-AES, alpha spectrometry are used in characterizing such types of wastes. As alpha spectrometry is simple and reliable, this technique has been used for the estimation of minor actinides after devising steps of separation for estimating Np and Pu present in HLW solutions of PHWR origin. Using a wealth of knowledge appropriate to the solution chemistry of actinides, the task of separation, though appears easy, it is challenging job for a radiochemist handling high-dose HLW samples, for obtaining clean alpha peaks for Np and Pu. This paper reports on the successful attempt made to quantify 241 Am, 244 Cm, Pu (239 mainly) and 237 Np present in HLW-PHWR obtained from PREFRE, Tarapur

  19. Plutonium Management, Minor Actinides Partitioning and Transmutation R and D in France

    International Nuclear Information System (INIS)

    Cavedon, Jean-Marc; Courtois, Charles

    2003-01-01

    Jean-Marc Cavedon (CEA, France) then presented the developments concerning Plutonium management and minor actinides P and T research and development in France. By the 1991 law on high-level long-lived radioactive waste a research programme was launched in the areas: (i) geological disposal, (ii) conditioning and long-term storage, and (iii) radiotoxicity reduction by P and T. The results of the work in these areas will be presented to the French Government and Parliament in 2006. The control of Plutonium stocks generated by the French PWRs is proposed to increase Plutonium consumption in reactors and minimise radioactive waste production, and requires the recycling of actinides, especially Plutonium. In the long term, CEA intends to develop a new technology based on gas cooled reactors and their associated fuel cycle, including multiple recycling of Plutonium. The advantages of this development consist in the optimisation of the use of natural resources and the concentration of Plutonium in limited quantities of fuel rods. If needed, the minor actinides could also be recycled. The planned CEA developments depend on new fuel types and will lead to novel waste types (light glasses) with a reduction of long-term radiotoxicity. Radiotoxicity reductions by a factor of 3 to 5 are expected for Plutonium recycling scenarios, and by up to a factor of a few hundreds for Plutonium and minor actinides recycling scenarios. This gain is nearly independent on the reactor type used, but needs about 100 years of application to become effective in terms of making a difference in the total waste inventory to be disposed of

  20. Ability to burn plutonium and minor actinides. Interest of accelerator driven system compared to critical reactor

    International Nuclear Information System (INIS)

    Vergnes, J.; Mouney, H.

    1998-01-01

    In the frame of the French Act of December 1991, EDF is presently assessing the interest of Acceleration Driven System (ADS) for the Transmutation of the Plutonium and Minor Actinides (MA) produced by its park of nuclear reactors. The studies presented here assess the efficiency of ADS and critical reactors to incinerate Pu and MA (Minor Actinides) and the potential interest of ADS for that purpose. (author)

  1. Analysis and optimization of minor actinides transmutation blankets with regards to neutron and gamma sources

    International Nuclear Information System (INIS)

    Kooyman, T.; Buiron, L.; Rimpault, G.

    2017-01-01

    Heterogeneous loading of minor actinides in radial blankets is a potential solution to implement minor actinides transmutation in fast reactors. However, to compensate for the lower flux level experienced by the blankets, the fraction of minor actinides to be loaded in the blankets must be increased to maintain acceptable performances. This severely increases the decay heat and neutron source of the blanket assemblies, both before and after irradiation, by more than an order of magnitude in the case of neutron source for instance. We propose here to implement an optimization methodology of the blankets design with regards to various parameters such as the local spectrum or the mass to be loaded, with the objective of minimizing the final neutron source of the spent assembly while maximizing the transmutation performances of the blankets. In a first stage, an analysis of the various contributors to long- and short-term neutron and gamma source is carried out whereas in a second stage, relevant estimators are designed for use in the effective optimization process, which is done in the last step. A comparison with core calculations is finally done for completeness and validation purposes. It is found that the use of a moderated spectrum in the blankets can be beneficial in terms of final neutron and gamma source without impacting minor actinides transmutation performances compared to more energetic spectrum that could be achieved using metallic fuel for instance. It is also confirmed that, if possible, the use of hydrides as moderating material in the blankets is a promising option to limit the total minor actinides inventory in the fuel cycle. If not, it appears that focus should be put upon an increased residence time for the blankets rather than an increase in the acceptable neutron source for handling and reprocessing. (authors)

  2. Analysis and optimization of minor actinides transmutation blankets with regards to neutron and gamma sources

    Directory of Open Access Journals (Sweden)

    Kooyman Timothée

    2017-01-01

    Full Text Available Heterogeneous loading of minor actinides in radial blankets is a potential solution to implement minor actinides transmutation in fast reactors. However, to compensate for the lower flux level experienced by the blankets, the fraction of minor actinides to be loaded in the blankets must be increased to maintain acceptable performances. This severely increases the decay heat and neutron source of the blanket assemblies, both before and after irradiation, by more than an order of magnitude in the case of neutron source for instance. We propose here to implement an optimization methodology of the blankets design with regards to various parameters such as the local spectrum or the mass to be loaded, with the objective of minimizing the final neutron source of the spent assembly while maximizing the transmutation performances of the blankets. In a first stage, an analysis of the various contributors to long- and short-term neutron and gamma source is carried out whereas in a second stage, relevant estimators are designed for use in the effective optimization process, which is done in the last step. A comparison with core calculations is finally done for completeness and validation purposes. It is found that the use of a moderated spectrum in the blankets can be beneficial in terms of final neutron and gamma source without impacting minor actinides transmutation performances compared to more energetic spectrum that could be achieved using metallic fuel for instance. It is also confirmed that, if possible, the use of hydrides as moderating material in the blankets is a promising option to limit the total minor actinides inventory in the fuel cycle. If not, it appears that focus should be put upon an increased residence time for the blankets rather than an increase in the acceptable neutron source for handling and reprocessing.

  3. Analysis and optimization of minor actinides transmutation blankets with regards to neutron and gamma sources

    Science.gov (United States)

    Kooymana, Timothée; Buiron, Laurent; Rimpault, Gérald

    2017-09-01

    Heterogeneous loading of minor actinides in radial blankets is a potential solution to implement minor actinides transmutation in fast reactors. However, to compensate for the lower flux level experienced by the blankets, the fraction of minor actinides to be loaded in the blankets must be increased to maintain acceptable performances. This severely increases the decay heat and neutron source of the blanket assemblies, both before and after irradiation, by more than an order of magnitude in the case of neutron source for instance. We propose here to implement an optimization methodology of the blankets design with regards to various parameters such as the local spectrum or the mass to be loaded, with the objective of minimizing the final neutron source of the spent assembly while maximizing the transmutation performances of the blankets. In a first stage, an analysis of the various contributors to long and short term neutron and gamma source is carried out while in a second stage, relevant estimators are designed for use in the effective optimization process, which is done in the last step. A comparison with core calculations is finally done for completeness and validation purposes. It is found that the use of a moderated spectrum in the blankets can be beneficial in terms of final neutron and gamma source without impacting minor actinides transmutation performances compared to more energetic spectrum that could be achieved using metallic fuel for instance. It is also confirmed that, if possible, the use of hydrides as moderating material in the blankets is a promising option to limit the total minor actinides inventory in the fuel cycle. If not, it appears that focus should be put upon an increased residence time for the blankets rather than an increase in the acceptable neutron source for handling and reprocessing.

  4. Minor actinides transmutation scenario studies with PWRs, FRs and moderated targets

    International Nuclear Information System (INIS)

    Grouiller, J.P.; Pillon, S.; Saint Jean, C. de; Varaine, F.; Leyval, L.; Vambenepe, G.; Carlier, B.

    2003-01-01

    Using current technologies, we have demonstrated in this study that it is theoretically possible to obtain different minor actinide transmutation scenarios with a significant gain on the waste radiotoxicity inventory. The handling of objects with Am+Cm entails the significant increase of penetrating radiation sources (neutron and γ) whatever mixed scenario is envisioned; the PWR and FR scenario involving the recycling of Am + Cm in the form of targets results in the lowest flow. In the light of these outcomes, the detailed studies has allowed to design a target sub assembly with a high fission rate (90%) and define a drawing up of reprocessing diagram with the plant head, the minor actinide separation processes (PUREX, DIAMEX and SANEX). Some technological difficulties appear in manipulating curium, principally in manufacturing where the wet process ('sol-gel') is not acquired for (Am+Cm). (author)

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  6. Neutronic analysis of the PBMR-400 full core using thorium fuel mixed with plutonium or minor actinides

    International Nuclear Information System (INIS)

    Acır, Adem; Coşkun, Hasan

    2012-01-01

    Highlights: ► Neutronic calculations for PBMR 400 were conducted with the computer codes MCNP and MONTEBURNS 2.0. ► The criticality and burnup were investigated for reactor grade plutonium and minor actinides. ► We found that the use of these new fuels in PBMRs would reduce the nuclear waste repository significantly. -- Abstract: Time evolution of criticality and burnup grades of the PBMR were investigated for reactor grade plutonium and minor actinides in the spent fuel of light water reactors (LWRs) mixed with thoria. The calculations were performed by employing the computer codes MCNP and MONTEBURNS 2.0 and using the ENDF/B-V nuclear data library. Firstly, the plutonium–thorium and minor actinides–thorium ratio was determined by using the initial k eff value of the original uranium fuel design. After the selection of the plutonium/minor actinides–thorium mixture ratio, the time-dependent neutronic behavior of the reactor grade plutonium and minor actinides and original fuels in a PBMR-400 reactor was calculated by using the MCNP code. Finally, k eff , burnup and operation time values of the fuels were compared. The core effective multiplication factor (k eff ) for the original fuel which has 9.6 wt.% enriched uranium was computed as 1.2395. Corresponding to this k eff value the reactor grade plutonium/thorium and minor actinide/thorium oxide mixtures were found to be 30%/70% and 50%/50%, respectively. The core lives for the original, the reactor grade plutonium/thorium and the minor actinide/thorium fuels were calculated as ∼3.2, ∼6.5 and ∼5.5 years, whereas, the corresponding burnups came out to be 99,000, ∼190,000 and ∼166,000 MWD/T, respectively, for an end of life k eff set equal to 1.02.

  7. Determination of minor actinides fission cross sections by means of transfer reactions

    Energy Technology Data Exchange (ETDEWEB)

    Jurado, B.; Aiche, M.; Barreau, G.; Boyer, S.; Czajkowski, S.; Dassie, D.; Grosjean, C.; Guiral, A.; Haas, B.; Osmanov, B.; Petit, M. [CENBG - UMR 5795 CNRS/IN2P3-Univ. Bordeaux 1- Le Haut Vigneau, 33175 Gradignan (France); Berthoumieux, E.; Gunsing, F.; Perrot, L.; Theisen, Ch. [CEN Saclay, DSM/DAPNIA/SPhN, 91191 Gif-sur-Yvette cedex (France); Bauge, E. [CEA, SPhN, BP12 91680 Bruyeres-le-Chatel (France); Michel-Sendis, F. [IPN, 15 rue G. Clemenceau, 91406 Orsay cedex (France); Billebaud, A. [LPSC, 53 Avenue des Martyrs, 38026 Grenoble cedex (France); Wilson, J. N. [IPN, 15 rue G. Clemenceau, 91406 Orsay cedex (France); LPSC, 53 Avenue des Martyrs, 38026 Grenoble cedex (France); Ahmad, I.; Greene, J.P.; Janssens, R. V. F. [ANL, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2005-07-01

    We present an original method that allows to determine neutron-induced cross sections of very short-lived minor actinides. This indirect method, based on the use of transfer reactions, has already been applied with success for the determination of the neutron-induced fission and capture cross section of {sup 233}Pa, a key nucleus in the {sup 232}Th - {sup 233}U fuel cycle. A recent experiment using this technique has been performed to determine the neutron-induced fission cross sections of {sup 242,243,244}Cm and {sup 241}Am which are present in the nuclear waste of the current U-Pu fuel cycle. These cross sections are highly relevant for the design of reactors capable to incinerate minor actinides. The first results will be illustrated. (authors)

  8. Design and safety studies on the European Facility for Industrial Transmutation (EFIT) with CERMET fuel

    International Nuclear Information System (INIS)

    Chen, X.N.; Rineiski, A.; Liu, P.; Matzerath Boccaccini, C.; Flad, M.; Gabrielli, F.; Maschek, W.; Morita, K.

    2008-01-01

    European R and D for ADS design and fuel development is driven in the 6 th FP of the EU by the EUROTRANS Programme [1]. In EUROTRANS two ADS design routes are followed, the XT-ADS and the EFIT. The XT-ADS is designed to provide the experimental demonstration of transmutation in an Accelerator Driven System. The EFIT development, the European Facility for Industrial Transmutation, aims at a generic conceptual design of a full transmuter. A key issue of the R and D work is the choice of an adequate fuel to be used in an Accelerator Driven Transmuter (ADT) like EFIT. Various fuel forms have been assessed. CERCER and CERMET fuels, specifically with the matrices MgO and Mo, have finally been selected and are now under closer investigation. Within EUROTRANS, a special domain named 'AFTRA', is responsible to more deeply assess the behavior of these dedicated fuels and to provide the fuel data base for the core design of the EFIT. The EFIT concept has to be optimized towards: a good transmutation efficiency, high burnup, low reactivity swing, low power peaking, adequate subcriticality, reasonable beam requirements and a high safety level. The final recommendation on fuels by AFTRA gave a ranking of these fuels based on the mentioned criteria. The composite CERMET fuel (Pu 0.5 ,Am 0.5 )O 2-x - Mo (with the isotope 92 Mo comprising 93% of the molybdenum) has been recommended as the primary candidate for the EFIT. This CERMET fuel fulfils adopted criteria for fabrication and reprocessing, and provides excellent safety margins. Disadvantages include the cost for enrichment of 92 Mo and a lower specific transmutation rate of minor actinides, because of the higher neutron absorption cross-section of the matrix. The composite CERCER fuel (Pu 0.4 ,Am 0.6 )O 2-x - MgO has therefore been recommended as a backup solution as it might offer a higher consumption rate of minor actinides, and can be manufactured for a lower unit cost. This paper is in fact a sequel to our last paper [2

  9. Separation of minor actinides from a genuine MA/LN fraction

    International Nuclear Information System (INIS)

    Satmark, B.; Courson, O.; Malmbeck, R.; Pagliosa, G.; Romer, K.; Glatz, J.P.

    2001-01-01

    Separation of the trivalent Minor Actinides (MA), Am and Cm, has been performed from a genuine MA(III) + Ln(III) solution using Bis-Triazine-Pyridine (BTP) as organic extractant. The representative MA/Ln fraction was obtained from a dissolved commercial LWR fuel (45.2 GWd/tM) submitted subsequently too a PUREX process followed by a DIAMEX process. A centrifugal extractor set-up (16-stages), working in a continuous counter-current mode, was used for the liquid-liquid separation. In the nPr-BTP process, feed decontamination factors for Am and Cm above 96 and 65, respectively were achieved. The back-extraction was more efficient for Am (99.1% recovery) than for Cm (97.5%). This experiment, using the Bis-Triazine-Pyridine molecule is the first successful demonstration of the separation of MA from lanthanides in a genuine MA/Ln fraction with a nitric acid concentration of ca. 1 M. It represents an important break through in the difficult field of minor actinide partitioning of high level liquid waste. (author)

  10. Hydrometallurgical minor actinide separation in hollow fiber modules

    International Nuclear Information System (INIS)

    Geist, A.; Weigl, M.; Gompper, K.

    2004-01-01

    Hollow fiber modules (HFM) were used as phase contacting devices for hydrometallurgical minor actinide separation in the Partitioning and Transmutation context. Two single-HFM setups, one using commercially available HFM, the other one using miniature HFM, have been developed and manufactured. Several very successful DIAMEX and SANEX once-through tests were performed. The major advantage of the new miniature HFM is their size drastically reducing chemicals consumption: only several 10 mL of feed phases are required for a test. (authors)

  11. The uncertainty analysis of a liquid metal reactor for burning minor actinides from light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    The neutronics analysis of a liquid metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWth actinide burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities were generated using depletion perturbation methods for the equilibrium cycle of the reactor and covariance data was taken ENDF-B/V and other published sources. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180%, 97%, and 46%, respectively. 5 refs., 1 fig., 3 tabs. (Author)

  12. The uncertainty analysis of a liquid metal reactor for burning minor actinides from light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The neutronics analysis of a liquid metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWth actinide burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities were generated using depletion perturbation methods for the equilibrium cycle of the reactor and covariance data was taken ENDF-B/V and other published sources. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180%, 97%, and 46%, respectively. 5 refs., 1 fig., 3 tabs. (Author)

  13. Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

    International Nuclear Information System (INIS)

    Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

    2007-01-01

    Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at ∼2.4, ∼7 and ∼11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of ∼7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10 15 n/cm 2 /s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between ∼410 deg. C and ∼645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

  14. Minor Actinide Burning in Thermal Reactors. A Report by the Working Party on Scientific Issues of Reactor Systems

    International Nuclear Information System (INIS)

    Hesketh, K.; Porsch, D.; Rimpault, G.; Taiwo, T.; Worrall, A.

    2013-01-01

    The actinides (or actinoids) are those elements in the periodic table from actinium upwards. Uranium (U) and plutonium (Pu) are two of the principal elements in nuclear fuel that could be classed as major actinides. The minor actinides are normally taken to be the triad of neptunium (Np), americium (Am) and curium (Cm). The combined masses of the remaining actinides (i.e. actinium, thorium, protactinium, berkelium, californium, einsteinium and fermium) are small enough to be regarded as very minor trace contaminants in nuclear fuel. Those elements above uranium in the periodic table are known collectively as the transuranics (TRUs). The operation of a nuclear reactor produces large quantities of irradiated fuel (sometimes referred to as spent fuel), which is either stored prior to eventual deep geological disposal or reprocessed to enable actinide recycling. A modern light water reactor (LWR) of 1 GWe capacity will typically discharge about 20-25 tonnes of irradiated fuel per year of operation. About 93-94% of the mass of uranium oxide irradiated fuel is comprised of uranium (mostly 238 U), with about 4-5% fission products and ∼1% plutonium. About 0.1-0.2% of the mass is comprised of neptunium, americium and curium. These latter elements accumulate in nuclear fuel because of neutron captures, and they contribute significantly to decay heat loading and neutron output, as well as to the overall radio-toxic hazard of spent fuel. Although the total minor actinide mass is relatively small - approximately 20-25 kg per year from a 1 GWe LWR - it has a disproportionate impact on spent fuel disposal, and thus the longstanding interest in transmuting these actinides either by fission (to fission products) or neutron capture in order to reduce their impact on the back end of the fuel cycle. The combined masses of the trace actinides actinium, thorium, protactinium, berkelium and californium in irradiated LWR fuel are only about 2 parts per billion, which is far too low for

  15. The technical and economic impact of minor actinide transmutation in a sodium fast reactor

    International Nuclear Information System (INIS)

    Gautier, G. M.; Morin, F.; Dechelette, F.; Sanseigne, E.; Chabert, C.

    2012-01-01

    Within the frame work of the French National Act of June 28, 2006 pertaining to the management of high activity, long-lived radioactive waste, one of the proposed processes consists in transmuting the Minor Actinides (MA) in the radial blankets of a Sodium Fast Reactor (SFR). With this option, we may assess the additional cost of the reactor by comparing two SFR designs, one with no Minor Actinides, and the other involving their transmutation. To perform this exercise, we define a reference design called SFRref, of 1500 MWe that is considered to be representative of the Reactor System. The SFRref mainly features a pool architecture with three pumps, six loops with one steam generator per loop. The reference core is the V2B core that was defined by the CEA a few years ago for the Reactor System. This architecture is designed to meet current safety requirements. In the case of transmutation, for this exercise we consider that the fertile blanket is replaced by two rows of assemblies having either 20% of Minor Actinides or 20% of Americium. The assessment work is performed in two phases. - The first consists in identifying and quantifying the technical differences between the two designs: the reference design without Minor Actinides and the design with Minor Actinides. The main differences are located in the reactor vessel, in the fuel handling system and in the intermediate storage area for spent fuel. An assessment of the availability is also performed so that the impact of the transmutation can be known. - The second consists in making an economic appraisal of the two designs. This work is performed using the CEA's SEMER code. The economic results are shown in relative values. For a transmutation of 20% of MA in the assemblies (S/As) and a hypothesis of 4 kW allowable for the washing device, there is a large external storage demanding a very long cooling time of the S/As. In this case, the economic impact may reach 5% on the capital part of the Levelized Unit

  16. Target fuels for plutonium and minor actinide transmutation in pressurized water reactors

    International Nuclear Information System (INIS)

    Washington, J.; King, J.; Shayer, Z.

    2017-01-01

    Highlights: • We evaluate transmutation fuels for plutonium and minor actinide destruction in LWRs. • We model a modified AP1000 fuel assembly in SCALE6.1. • We evaluate spectral shift absorber coatings to improve transmutation performance. - Abstract: The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a nearer-term solution. This study considers a method for plutonium and minor actinide transmutation in existing light water reactors and evaluates a variety of transmutation fuels to provide a common basis for comparison and to determine if any single target fuel provides superior transmutation properties. A model developed using the NEWT module in the SCALE 6.1 code package provided performance data for the burnup of the target fuel rods in the present study. The target fuels (MOX, PuO_2, Pu_3Si_2, PuN, PuUZrH, PuZrH, PuZrHTh, and PuZrO_2) are evaluated over a 1400 Effective Full Power Days (EFPD) interval to ensure each assembly remained critical over the entire burnup period. The MOX (5 wt% PuO_2), Pu_0_._3_1ZrH_1_._6Th_1_._0_8, and PuZrO_2MgO (8 wt% Pu) fuels result in the highest rate of plutonium transmutation with the lowest rate of curium-244 production. This study selected eleven different burnable absorbers (B_4C, CdO, Dy_2O_3, Er_2O_3, Eu_2O_3, Gd_2O_3, HfO_2, In_2O_3, Lu_2O_3, Sm_2O_3, and TaC) for evaluation as spectral shift absorber coatings on the outside of the fuel pellets to determine if an absorber coating can improve the transmutation properties of the target fuels. The PuZrO_2MgO (8 wt% Pu) target fuel with a coating of Lu_2O_3 resulted in the highest rate of plutonium transmutation with the greatest reduction in curium

  17. Target fuels for plutonium and minor actinide transmutation in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Washington, J., E-mail: jwashing@gmail.com [Nuclear Science and Engineering Program, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); King, J., E-mail: kingjc@mines.edu [Nuclear Science and Engineering Program, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); Shayer, Z., E-mail: zshayer@mines.edu [Department of Physics, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States)

    2017-03-15

    Highlights: • We evaluate transmutation fuels for plutonium and minor actinide destruction in LWRs. • We model a modified AP1000 fuel assembly in SCALE6.1. • We evaluate spectral shift absorber coatings to improve transmutation performance. - Abstract: The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a nearer-term solution. This study considers a method for plutonium and minor actinide transmutation in existing light water reactors and evaluates a variety of transmutation fuels to provide a common basis for comparison and to determine if any single target fuel provides superior transmutation properties. A model developed using the NEWT module in the SCALE 6.1 code package provided performance data for the burnup of the target fuel rods in the present study. The target fuels (MOX, PuO{sub 2}, Pu{sub 3}Si{sub 2}, PuN, PuUZrH, PuZrH, PuZrHTh, and PuZrO{sub 2}) are evaluated over a 1400 Effective Full Power Days (EFPD) interval to ensure each assembly remained critical over the entire burnup period. The MOX (5 wt% PuO{sub 2}), Pu{sub 0.31}ZrH{sub 1.6}Th{sub 1.08}, and PuZrO{sub 2}MgO (8 wt% Pu) fuels result in the highest rate of plutonium transmutation with the lowest rate of curium-244 production. This study selected eleven different burnable absorbers (B{sub 4}C, CdO, Dy{sub 2}O{sub 3}, Er{sub 2}O{sub 3}, Eu{sub 2}O{sub 3}, Gd{sub 2}O{sub 3}, HfO{sub 2}, In{sub 2}O{sub 3}, Lu{sub 2}O{sub 3}, Sm{sub 2}O{sub 3}, and TaC) for evaluation as spectral shift absorber coatings on the outside of the fuel pellets to determine if an absorber coating can improve the transmutation properties of the target fuels. The PuZrO{sub 2}MgO (8 wt% Pu) target

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  19. Optimization of SFR Reactor design with recycling or minor actinides

    International Nuclear Information System (INIS)

    Martin-Fuertes, F.; Vazquez, M.; Alvarez, F.

    2012-01-01

    In this paper we show results of the design features and ESFR optimized in three configurations: the reference, load the minority actinides homogeneous throughout the reactor and the high content of AM on a radial mantle. Was calculated reactivity evolution in five cycles burned (2050 days) to recharge One approach. To do this, we have employed EVOLCODE2 a development tool of CIEMAT own coupling MCNPX and ORIGEN.

  20. Minor actinides transmutation potential: state of art for GEN IV sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Buiron, Laurent

    2015-01-01

    In the frame of the R and D program relative to the 1991 French act on nuclear waste management, fast neutron systems have shown relevant characteristics that meet both requirements on sustainable resources management and waste minimization. They also offer flexibility by mean of burner or breeder configurations allowing mastering plutonium inventory without significant impact on core safety. From the technological point of view, sodium cooled fast reactor are considered in order to achieve mean term industrial deployment. The present document summaries the main results of R and D program on minor actinides transmutation in sodium fast reactor since 2006 following recommendation of the first part of the 1991 French act. Both homogeneous and heterogeneous management achievable performances are presented for 'evolutionary' SFR V2B core as well as low void worth CFV core for industrial scale configurations (1500 MWe). Minor actinides transmutation could be demonstrated in the ASTRID reactor with the following configurations: - a 2%vol Americium content for the homogeneous mode, - a 10%vol Americium content for the heterogeneous mode, without any substantial modification of the main core safety parameters and only limited impacts on the associated fuel cycle (manufacturing issues are not considered here). In order to achieve such goal, a wide range of experimental irradiations driven by transmutation scenarios have to be performed for both homogeneous and heterogeneous minor actinides management. (author) [fr

  1. Benchmark Evaluation of Dounreay Prototype Fast Reactor Minor Actinide Depletion Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Hess, J. D.; Gauld, I. C.; Gulliford, J.; Hill, I.; Okajima, S.

    2017-01-01

    Historic measurements of actinide samples in the Dounreay Prototype Fast Reactor (PFR) are of interest for modern nuclear data and simulation validation. Samples of various higher-actinide isotopes were irradiated for 492 effective full-power days and radiochemically assayed at Oak Ridge National Laboratory (ORNL) and Japan Atomic Energy Research Institute (JAERI). Limited data were available regarding the PFR irradiation; a six-group neutron spectra was available with some power history data to support a burnup depletion analysis validation study. Under the guidance of the Organisation for Economic Co-Operation and Development Nuclear Energy Agency (OECD NEA), the International Reactor Physics Experiment Evaluation Project (IRPhEP) and Spent Fuel Isotopic Composition (SFCOMPO) Project are collaborating to recover all measurement data pertaining to these measurements, including collaboration with the United Kingdom to obtain pertinent reactor physics design and operational history data. These activities will produce internationally peer-reviewed benchmark data to support validation of minor actinide cross section data and modern neutronic simulation of fast reactors with accompanying fuel cycle activities such as transportation, recycling, storage, and criticality safety.

  2. The application of CANDU neutron economy for the annihilation of the minor actinides

    International Nuclear Information System (INIS)

    Dastur, Adi; Gagnon, Nathalie

    1995-01-01

    A strategically indispensable role, comparable to the one of operating with natural uranium, is proposed for CANDU as an incentive to ensure future CANDU sales in an environment where enrichment and reprocessing technology are globally available. Because of their high neutron economy, CANDU reactors can operate with minimal fissile content and consequently at high neutron flux. This is especially so in the absence of uranium, i.e. when transuranic actinides are used as fuel. The low fissile requirement and the on-power refuelling capability of CANDU can be exploited to achieve a once-through cycle for actinide annihilation. This avoids recycling and refabrication costs and provides relatively high annihilation rates. In addition, CANDUs ability to operate without uranium and extract energy from the minor actinides makes it the ultimate resource conserver and gives it a unique role in sustainable energy growth. (author)

  3. Comparative Study of the Reactor Burner Efficiency for Transmutation of Minor Actinides

    Energy Technology Data Exchange (ETDEWEB)

    Gulevich, A.; Zemskov, E. [Institute of Physics and Power Engineering, Bondarenko sq. 1, Obninsk, Kaluga region, 249020 (Russian Federation); Degtyarev, A.; Kalugin, A.; Ponomarev, L. [Russian Research Center ' Kurchatov Institute' , Kurchatov sq. 1, Moscow, 123182 (Russian Federation); Konev, V.; Seliverstov, V. [Institute of Theoretical and Experimental Physics, ul. B. Cheremushinskaya 25, Moscow, 117259 (Russian Federation)

    2009-06-15

    Transmutation of minor actinides (MA) in the closed nuclear fuel cycle (NFC) is a one of the most important problem for future nuclear energetic. There are several approaches for MA transmutation but there are no common criteria for the comparison of their efficiency. In paper [1] we turned out the attention to the importance of taking into account the duration of the closed NFC in addition to a usual criterion of the neutron economy. In accordance with these criteria the transmutation efficiency are compared of two fast reactors (sodium and lead cooled) and three types of ADS-burners: LBE-cooled reactors (fast neutron spectrum), molten-salt reactor (intermediate spectrum) and heavy water reactor (thermal spectrum). It is shown that the time of transmutation of loaded MA in the closed nuclear fuel cycle is more than 50 years. References: A. Gulevich, A. Kalugin, L. Ponomarev, V. Seliverstov, M. Seregin, 'Comparative Study of ADS for Minor Actinides Transmutation', Progress in Nuclear Energy, 50, March-August, p. 358, 2008. (authors)

  4. Hybrid KED/XRF measurement of minor actinides in reprocessing plants

    International Nuclear Information System (INIS)

    Hsue, S.T.; Collins, M.L.

    1996-01-01

    Minor actinides have received considerable attention recently in the nuclear power industry. Because of their potential value as recycle fuels in thermal and breeder reactors, reprocessing plants may have an economic incentive to extract Np, Am, and Cm from their waste streams. This report discusses the technique of hybrid densitometry and its potential to measure Np and Am in reprocessing plants. Precision estimates are made for the hybrid analysis of Np and Am in two types of dissolver solutions

  5. Denaturing of plutonium by transmutation of minor-actinides for enhancement of proliferation resistance

    International Nuclear Information System (INIS)

    Sagara, Hiroshi; Saito, Masaki; Peryoga, Yoga; Ezoubtchenko, Alexey; Takivayev, Alan

    2005-01-01

    Feasibility study for the plutonium denaturing by utilizing minor-actinide transmutation in light water reactors has been performed. And the intrinsic feature of proliferation resistance of plutonium has been discussed based on IAEA's publication and Kessler's proposal. The analytical results show that not only 238 Pu but also other plutonium isotopes with even-mass-number have very important role for denaturing of plutonium due to their relatively large critical mass and noticeably high spontaneous fission neutron generation. With the change of the minor-actinide doping ratio in U-Pu mix oxide fuel and moderator to fuel ratio, it is found that the reactor-grade plutonium from conventional light water reactors can be denatured to satisfy the proliferation resistance criterion based on the Kessler's proposal but not to be sufficient for the criterion based on IAEA's publication. It has been also confirmed that all the safety coefficients take negative value throughout the irradiation. (author)

  6. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Timothy A. Hyde

    2012-06-01

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  7. Plutonium and minor actinides management in the nuclear fuel cycle: assessing and controlling the inventory

    International Nuclear Information System (INIS)

    Mouney, H.

    2002-01-01

    The mastering of the plutonium and minor actinides inventory in the French Nuclear Cycle is based on a progressive approach from the present status, dealing with the partial reprocessing of spent fuels and the recycling of Pu in the MOX assemblies loaded in the 20 licensed PWRs. This strategy keeps the door open long-term, for example, for the eventual multi-recycling of excess Pu in dedicated new assemblies, such as APA or CORAIL in order to stabilize the Pu inventory in the fuel cycle or allow its utilization in new types of fast reactors. Presently, in the framework of 1991 law, scenario studies relying on present and/or innovative technologies are carried out in order to transmute both Pu and minor actinides, thus minimising the quantities to be for disposal. (author)

  8. Summary Report of Third Research Coordination Meeting on Minor Actinide Nuclear Reaction Data (MANREAD)

    International Nuclear Information System (INIS)

    Gunsing, Frank; Otsuka, Naohiko

    2010-12-01

    The Third Research Co-ordination Meeting of the MANREAD (Minor Actinides Neutron Reaction Data) Coordinated Research Project (CRP) was held at IAEA Headquarters in Vienna from 19 to 22 October 2010. A summary of the presentation, and the discussions which took place during the meeting, are reported here. In addition, a task assignment list of the experimental data assessment activities was agreed, and is provided together with the plan for future CRP activities. The Third Research Co-ordination Meeting of the MANREAD (Minor Actinides Neutron Reaction Data) Coordinated Research Project (CRP) was held at IAEA Headquarters in Vienna from 19 to 22 October 2010. A summary of the presentation, and the discussions which took place during the meeting, are reported here. In addition, a task assignment list of the experimental data assessment activities was agreed, and is provided together with the plan for future CRP activities. (author)

  9. LEMA facility and equipments for minor actinides compounds fabrication and characterisation

    Energy Technology Data Exchange (ETDEWEB)

    Donnet, L. [Commissariat a l' Energie Atomique - CEA, CEA/DEN/VRH/DTEC/SDTC/LEMA (France)

    2008-07-01

    The LEMA (Actinide based materials study laboratory) is mainly involved in minor actinides materials development and fabrication, from raw materials choice and synthesis to finished products including pin assembly. The aim of the technological analyses is to establish choices of raw materials and manufacturing techniques. The LEMA is located in the ATALANTE facility in Marcoule. It consists in two shielded chains (one specific for neutrons) and three hot laboratories. The laboratory has various apparatuses in hot cells such as: ball mills, press, dilatometer, TGA (thermo-gravimetry analyser), calcination and sintering furnaces (2000 deg. C). The laboratory has also characterisation apparatuses such as XRD and SEM (scanning electron microscopy) dedicated to structural and microstructural studies. Thanks to the diversity of its equipment, the LEMA has well established worldwide collaborations and takes part in international fuels/target fabrication and irradiation experiments. (author)

  10. LEMA facility and equipments for minor actinides compounds fabrication and characterisation

    International Nuclear Information System (INIS)

    Donnet, L.

    2008-01-01

    The LEMA (Actinide based materials study laboratory) is mainly involved in minor actinides materials development and fabrication, from raw materials choice and synthesis to finished products including pin assembly. The aim of the technological analyses is to establish choices of raw materials and manufacturing techniques. The LEMA is located in the ATALANTE facility in Marcoule. It consists in two shielded chains (one specific for neutrons) and three hot laboratories. The laboratory has various apparatuses in hot cells such as: ball mills, press, dilatometer, TGA (thermo-gravimetry analyser), calcination and sintering furnaces (2000 deg. C). The laboratory has also characterisation apparatuses such as XRD and SEM (scanning electron microscopy) dedicated to structural and microstructural studies. Thanks to the diversity of its equipment, the LEMA has well established worldwide collaborations and takes part in international fuels/target fabrication and irradiation experiments. (author)

  11. Numerical simulation of minor actinide recovery behaviour in batch processing of spent metallic fuel by electrorefining

    Energy Technology Data Exchange (ETDEWEB)

    Nawada, H P; Bhat, N P [Metallurgy Division, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Balasubramanian, G R [Atomic Energy Commission, Mumbai (India)

    1994-06-01

    Numerical simulation of electro-transport of fuel actinides (FAs), minor actinides (MAs) and rare earths (REs) in the electro-refiner (ER) for pyrochemical reprocessing of a typical spent IFR metallic fuel has been attempted based on improved thermo-chemical model developed for application to multi-component system in the ER. Optimization of MA recovery and decontamination factors (DFs) for MAs and REs in batch processing is presented. (author). 7 refs., 4 figs., 1 tab.

  12. A study on the feasibility of minor actinides in BWR

    International Nuclear Information System (INIS)

    Abdul Waris; Budiono

    2008-01-01

    Preliminary study on the feasibility of actinides minor (MA) recycling without mixing them with plutonium in boiling water reactor (BWR) has been carried out. The results show that increasing of fissile MA content in mixed oxide fuel (MOX) and/or reducing void fraction can enlarge the effective multiplication factor at the beginning of cycle, but the reactor still can not obtain its criticality condition. Furthermore, dropping the void fraction results in higher reactivity swing and therefore plummeting the safety factor of the reactor. (author)

  13. Optimization of plutonium and minor actinide transmutation in an AP1000 fuel assembly via a genetic search algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Washington, J., E-mail: jwashing@gmail.com; King, J., E-mail: kingjc@mines.edu

    2017-01-15

    Highlights: • We model a modified AP1000 fuel assembly in SCALE6.1. • We couple the NEWT module of SCALE to the MOGA module of DAKOTA. • Transmutation is optimized based on choice of coating and fuel. • Greatest transmutation achieved with PuZrO{sub 2}MgO fuel pins coated with Lu{sub 2}O{sub 3}. - Abstract: The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, which contains approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are the preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a near-term solution. Previous simulation work demonstrated the potential to transmute transuranic elements in a modified light water reactor fuel pin. This study optimizes a quarter-assembly containing target fuels coated with spectral shift absorbers for the transmutation of plutonium and minor actinides in light water reactors. The spectral shift absorber coating on the target fuel pin tunes the neutron energy spectrum experienced by the target fuel. A coupled model developed using the NEWT module from SCALE 6.1 and a genetic algorithm module from the DAKOTA optimization toolbox provided performance data for the burnup of the target fuel pins in the present study. The optimization with the coupled NEWT/DAKOTA model proceeded in three stages. The first stage optimized a single-target fuel pin per quarter-assembly adjacent to the central instrumentation channel. The second stage evaluated a variety of quarter-assemblies with multiple target fuel pins from the first stage and the third stage re-optimized the pins in the optimal second stage quarter-assembly. An 8 wt% PuZrO{sub 2}MgO inert matrix fuel pin with a 1.44 mm radius and a 0.06 mm Lu{sub 2}O{sub 3} coating in a five target fuel pin per quarter-assembly configuration represents the optimal combination for the

  14. Comparison of different options for minor actinide transmutation in the frame of the French law for waste management

    International Nuclear Information System (INIS)

    Chabert, Christine; Leudet, Alain; Saturnin, Anne

    2011-01-01

    In the frame of the French Act for waste management which has been passed by French Parliament on June 28th, 2006, it is requested to obtain in 2012 an assessment of industrial perspectives of partitioning and transmutation of long-lived elements. These studies must be carried out in tight connection with GENIV systems development. The expected results must include the evaluation of technical and economic scenarios taking into account the optimization options between the minor actinide transmutation processes, their interim storage and geological disposal, including an analysis of several criteria. In this perspective, the CEA has established a working group named 'GT TES' (Working Group on Technical and Economic Scenarios) involving EDF and AREVA to define scenarios, the various criteria to evaluate them, to conduct these evaluations and then to highlight the key results. The group also relied on ANDRA for the geological storage studies. The scenarios evaluations take place in the French context. The nuclear energy production is supposed to remain constant during the scenarios and equal to 430 TWhe/year in accordance with the current French nuclear power installed capacity of 60 GW(e). The deployment of the first Sodium-cooled Fast Reactor (SFR) starts in 2040, considering that at this date the SFR technology should be mature. Several management schemes of minor actinides have been studied: Plutonium recycling in SFR (minor actinides are sent to the waste). Plutonium recycling and minor actinide (or Am alone) transmutation in SFR and in homogeneous mode ('Hom.'). Plutonium recycling and minor actinide (or Am alone) transmutation in SFR and in heterogeneous mode ('Het.'). Plutonium recycling in SFR and minor actinide transmutation in Accelerator-Driven-System (ADS). The criteria used to analyze these different scenarios, should take into account the viewpoint of scientists, industrials, administrations, and the general public. They are listed below: Inventories and

  15. Actinide Separation Demonstration Facility, Tarapur

    International Nuclear Information System (INIS)

    Vishwaraj, I.

    2017-01-01

    Partitioning of minor actinide from high level waste could have a substantial impact in lowering the radio toxicity associated with high level waste as well as it will reduce the burden on geological repository. In Indian context, the partitioned minor actinide could be routed into the fast breeder reactor systems scheduled for commissioning in the near period. The technological breakthrough in solvent development has catalyzed the partitioning programme in India, leading to the setting up and hot commissioning of the Actinide Separation Demonstration Facility (ASDF) at BARC, Tarapur. The engineering scale Actinide Separation Demonstration Facility (ASDF) has been retrofitted in an available radiological hot cell situated adjacent to the Advanced Vitrification Facility (AVS). This location advantage ensures an uninterrupted supply of high-level waste and facilitates the vitrification of the high-level waste after separation of minor actinides

  16. Neutron Capture Measuremetns on Minor Actinides at the n_TOF Facility at CERN: Past, Present and Future

    CERN Document Server

    Cano-Ott, D; Eleftheriadis, C; Leeb, H; Calvino, F; Herrera-Martinez, A; Savvidis, I; Vlachoudis, V; Haas, B; Abbondanno, U; Vannini, G; Oshima, M; Gramegna, F; Wiescher, M; Pigni, M T; Wiendler, H; Mengoni, A; Quesada, J; Becvar, F; Rosetti, M; Cennini, P; Mosconi, M; Duran, I; Rauscher, T; Ketlerov, V; Couture, A; Capote, R; Sarchiapone, L; Vlastou, R; Domingo-Pardo, C; Pavlopoulos, P; Karamanis, D; Krticka, M; Griesmayer, E; Jericha, E; Ferrari, A; Martinez, T; Oberhummer, H; Karadimos, D; Plompen, A; Mendoza, E; Terlizzi, R; Cortes, G; Cox, J; Voss, F; Pretel, C; Colonna, N; Berthoumieux, E; Dolfini, R; Vaz, P; Heil, M; Lopes, I; Lampoudis, C; Walter, S; Calviani, M; Gonzalez-Romero, E; Stephan, C; Tain, J L; Belloni, F; Igashira, M; Papachristodoulou, C; Aerts, G; Tavora, L; Milazzo, P M; Rudolf, G; Andrzejewski, J; Villamarin, D; Ferreira-Marques, R; Meaze, M H; O'Brien, S; Gunsing, F; Reifarth, R; Perrot, L; Lindote, A; Neves, F; Poch, A; Konovalov, V; Kerveno, M; Marques, L; Rubbia, C; Koehler, P; Dahlfors, M; Wisshak, K; Fujii, K; De Albornoz, A C; Salgado, J; Dridi, W; Ventura, A; Andriamonje, S; Dillman, I; Assimakopoulos, P; Ferrant, L; Lozano, M; Patronis, N; Chiaveri, E; Guerrero, C; Kadi, Y; Vicente, M C; Praena, J; Baumann, P; Moreau, C; Kappeler, F; Rullhusen, P; Furman, W; David, S; Marrone, S; Paradela, C; Audouin, L; Tassan-Got, L; Alvarez-Velarde, F; Massimi, C; Mastinu, P; Isaev, S; Pancin, J; Papadopoulos, C; Tagliente, G; Alvarez, H; Haight, R; Goverdovski, A; Chepel, V; Plag, R; Kossionides, E; Badurek, G; Marganiec, J; Lukic, S; Frais-Koelbl, H; Pavlik, A; Goncalves, I

    2011-01-01

    The successful development of advanced nuclear systems for sustainable energy production and nuclear waste management depends on high quality nuclear data libraries. Recent sensitivity studies and reports {[}1-3] have identified the need for substantially improving the accuracy of neutron cross-section data for minor actinides. The n\\_TOF collaboration has initiated an ambitious experimental program for the measurement of neutron capture cross sections of minor actinides. Two experimental setups have been constructed for this purpose: a Total Absorption Calorimeter (TAC) {[}4] for measuring neutron capture cross-sections of low-mass and/or radioactive samples and a set of two low neutron sensitivity C(6)D(6) detectors for the less radioactive materials.

  17. On the use of moderating material to enhance the feedback coefficients in SFR cores with high minor actinide content

    International Nuclear Information System (INIS)

    Merk, B.; Weiss, F. P.

    2012-01-01

    The use of fine distributed moderating material to enhance the feedback effects and to reduce the sodium void effecting sodium cooled fast reactor cores is described. The influence of the moderating material on the neutron spectrum, the power distribution, and the burnup distribution is shown. The consequences of the use of fine distributed moderating material into fuel assemblies with fuel configurations foreseen for minor actinide transmutation is analyzed and the transmutation efficiency is compared. The degradation of the feedback effects due to the insertion of minor actinides and the compensation by the use of moderating materials is discussed. (authors)

  18. On the use of moderating material to enhance the feedback coefficients in SFR cores with high minor actinide content

    Energy Technology Data Exchange (ETDEWEB)

    Merk, B. [Helmholtz-Zentrum Dresden-Rossendorf, Institut fuer Sicherheitsforschung, Postfach 51 01 19, 01314 Dresden (Germany); Weiss, F. P. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS MbH Forschungszentrum, Boltzmannstr. 14, 85748 Garching (Germany)

    2012-07-01

    The use of fine distributed moderating material to enhance the feedback effects and to reduce the sodium void effecting sodium cooled fast reactor cores is described. The influence of the moderating material on the neutron spectrum, the power distribution, and the burnup distribution is shown. The consequences of the use of fine distributed moderating material into fuel assemblies with fuel configurations foreseen for minor actinide transmutation is analyzed and the transmutation efficiency is compared. The degradation of the feedback effects due to the insertion of minor actinides and the compensation by the use of moderating materials is discussed. (authors)

  19. Demonstration of Minor Actinide separation from a genuine PUREX raffinate by TODGA/TBP and SANEX reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Magnusson, D. [European Commission, Joint Research Center, Institute for Transuranium Elements, Postfach 2340 D-76125 Karlsruhe (Germany); Chalmers University of Technology, Nuclear Chemistry, Deparment of Chemical and Biological Engineering, Gothenburg (Sweden); Christiansen, B.; Glatz, J.P.; Malmbeck, R.; Serrano-Purroy, D. [European Commission, Joint Research Center, Institute for Transuranium Elements, Postfach 2340 D-76125 Karlsruhe (Germany); Modolo, G. [Forschungszentrum Juelich, Institute for Energy Research, Safety Research and Reactor Technology, D-52425 Juelich (Germany); Sorel, C. [Commissariat a l' Energie Atomique Valrho (CEA), DRCP/SCPS, BP17171, 30207 Bagnols-sur-Ceze (France)

    2008-07-01

    A genuine High Active Raffinate was produced from small scale Purex reprocessing of a UO{sub 2} spent fuel solution and used as feed for a subsequent TODGA/TBP process. In this process, efficient recovery of the trivalent Minor Actinides (MA) actinides could be demonstrated using a hot cell set-up of 32 centrifugal contactor stages. The feed decontamination factors obtained for Am and Cm were in the range of 4.10{sup 4} which corresponds to a recovery of more than 99.99 % in the product fraction. Trivalent lanthanides and Y were co-extracted, otherwise only a small part of the Ru ended up in the product. The collected actinide/lanthanide fraction was later used as feed for a Sanex (separation of actinides from lanthanides) process based on the CyMe{sub 4}-BTBP ligand. Preliminary results show recoveries of more than 99.9 % of Am, Cm and less than 0.1 % of the major lanthanides in the product. (authors)

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  1. Comparison of fission and capture cross sections of minor actinides

    International Nuclear Information System (INIS)

    Nakagawa, Tsuneo; Iwamoto, Osamu

    2003-01-01

    The fission and capture cross sections of minor actinides given in JENDL-3.3 are compared with other evaluated data and experimental data. The comparison was made for 32 nuclides of Th-227, 228, 229, 230, 233, 234, Pa-231, 232, 233, U-232, 234, 236, 237, Np-236, 237, 238, Pu-236, 237, 238, 242, 244, Am-241, 242, 242m, 243, Cm-242, 243, 244, 245, 246, 247 and 248. Given in the present report are figures of these cross sections and tables of cross sections at 0.0253 eV and resonance integrals. (author)

  2. Managing Zirconium Chemistry and Phase Compatibility in Combined Process Separations for Minor Actinide Partitioning

    Energy Technology Data Exchange (ETDEWEB)

    Wall, Nathalie [Washington State Univ., Pullman, WA (United States); Nash, Ken [Washington State Univ., Pullman, WA (United States); Martin, Leigh [Washington State Univ., Pullman, WA (United States)

    2017-03-17

    In response to the NEUP Program Supporting Fuel Cycle R&D Separations and Waste Forms call DEFOA- 0000799, this report describes the results of an R&D project focusing on streamlining separation processes for advanced fuel cycles. An example of such a process relevant to the U.S. DOE FCR&D program would be one combining the functions of the TRUEX process for partitioning of lanthanides and minor actinides from PUREX(UREX) raffinates with that of the TALSPEAK process for separating transplutonium actinides from fission product lanthanides. A fully-developed PUREX(UREX)/TRUEX/TALSPEAK suite would generate actinides as product(s) for reuse (or transmutation) and fission products as waste. As standalone, consecutive unit-operations, TRUEX and TALSPEAK employ different extractant solutions (solvating (CMPO, octyl(phenyl)-N,Ndiisobutylcarbamoylmethylphosphine oxide) vs. cation exchanging (HDEHP, di-2(ethyl)hexylphosphoric acid) extractants), and distinct aqueous phases (2-4 M HNO3 vs. concentrated pH 3.5 carboxylic acid buffers containing actinide selective chelating agents). The separate processes may also operate with different phase transfer kinetic constraints. Experience teaches (and it has been demonstrated at the lab scale) that, with proper control, multiple process separation systems can operate successfully. However, it is also recognized that considerable economies of scale could be achieved if multiple operations could be merged into a single process based on a combined extractant solvent. The task of accountability of nuclear materials through the process(es) also becomes more robust with fewer steps, providing that the processes can be accurately modeled. Work is underway in the U.S. and Europe on developing several new options for combined processes (TRUSPEAK, ALSEP, SANEX, GANEX, ExAm are examples). There are unique challenges associated with the operation of such processes, some relating to organic phase chemistry, others arising from the

  3. Sigma Team for Minor Actinide Separation: PNNL FY 2011 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Lumetta, Gregg J.; Braley, Jenifer C.; Sinkov, Sergey I.; Levitskaia, Tatiana G.; Carter, Jennifer C.; Warner, Marvin G.; Pittman, Jonathan W.

    2011-08-13

    This report summarizes work conducted in FY 2011 at PNNL to investigate new methods of separating the minor actinide elements (Am and Cm) from the trivalent lanthanide elements, and separation of Am from Cm. For the former, work focused on a solvent extraction system combining an acidic extractant (HDEHP) with a neutral extractant (CMPO) to form a hybrid solvent extraction system referred to as TRUSPEAK (combining the TRUEX and TALSPEAK processes). For the latter, ligands that strongly bing uranyl ion were investigated for stabilizing corresponding americyl ion.

  4. Methods of fabricating cermet materials and methods of utilizing same

    Science.gov (United States)

    Kong, Peter C.

    2006-04-04

    Methods of fabricating cermet materials and methods of utilizing the same such as in filtering particulate and gaseous pollutants from internal combustion engines having intermetallic and ceramic phases. The cermet material may be made from a transition metal aluminide phase and an aluminia phase. The mixture may be pressed to form a green compact body and then heated in a nitrogen-containing atmosphere so as to melt aluminum particles and form the cermet. Filler materials may be added to increase the porosity or tailor the catalytic properties of the cermet material. Additionally, the cermet material may be reinforced with fibers or screens. The cermet material may also be formed so as to pass an electrical current therethrough to heat the material during use.

  5. Methods of producing cermet materials and methods of utilizing same

    Science.gov (United States)

    Kong, Peter C [Idaho Falls, ID

    2008-12-30

    Methods of fabricating cermet materials and methods of utilizing the same such as in filtering particulate and gaseous pollutants from internal combustion engines having intermetallic and ceramic phases. The cermet material may be made from a transition metal aluminide phase and an alumina phase. The mixture may be pressed to form a green compact body and then heated in a nitrogen-containing atmosphere so as to melt aluminum particles and form the cermet. Filler materials may be added to increase the porosity or tailor the catalytic properties of the cermet material. Additionally, the cermet material may be reinforced with fibers or screens. The cermet material may also be formed so as to pass an electrical current therethrough to heat the material during use.

  6. Calculated investigation of actinide transmutation in the BOR-60 reactor

    International Nuclear Information System (INIS)

    Zhemkov, I.Yu.; Ishunina, O.V.; Yakovleva, I.V.

    2000-01-01

    One of the prospective actinide burner reactor type is the fast reactor with a 'hard' spectrum and small breeding factor, which is the BOR-60. The calculated investigations demonstrate that Loading up to 40% of minor-actinides to the BOR-60 reactor did not lead to the considerable change of neutron-physical characteristics. The performed calculations show that the BOR- 60 reactor possesses a high efficiency of the minor-actinide and plutonium bum-up (up to 37 kg/(TW · h)) hat is comparable with properties of the actinide burner-reactors under design. The BOR-60 reactor can provide a homogeneous minor-actinide Loading (minor-actinide addition to the standard fuel) to the core and heterogeneous Loading (as separate assemblies-targets with a high minor-actinide fraction) to the first rows of a radial blanket that allows the optimum usage of the reactor and its characteristics. (authors)

  7. Summary report of 2. research coordination meeting on Minor Actinide Neutron Reaction Data (MANREAD)

    International Nuclear Information System (INIS)

    Nagai, Y.; Mengoni, A.

    2009-07-01

    The second Research Co-ordination Meeting of the MANREAD (Minor Actinides Neutron Reaction Data) was held at the IAEA Headquarters in Vienna from 31 March to 3 April 2009. A summary of the discussion which took place at the meeting is reported here together with a list of the main outcomes and recommendations produced by the RCM participants. (author)

  8. Monazite-type ceramics for the immobilization of minor actinides plutonium

    International Nuclear Information System (INIS)

    Heuser, Julia Maria

    2015-01-01

    The safe disposal of radioactive waste in deep geological formations is a challenging task of present and future generations. Innovative strategies as the conditioning of radionuclides in ceramic matrices can make a contribution here. This work points out monazite-type ceramics as potential waste forms for minor actinides and Pu. Several aspects concerning nuclear disposal as well as fundamental structural information were investigated. Lanthanide phosphate endmembers (LnPO 4 ) within the stability field of monazite (Ln = La-Gd) were synthesised within the scope of this work. To extend the knowledge of monazite phases, monoclinic TbPO 4 - and DyPO 4 -phases were prepared and characterised. Tb- and Dy-phosphates are situated in the xenotime stability field close to that of monazite. They can exist as metastable monazite phases. Structural characterisations of long- and short-range order were performed by X-ray diffraction, infrared (IR) and Raman spectroscopy. Structural data could be complemented, enhanced and gaps of knowledge could be filled by the first systematic consideration of the complete Ln-monazite-series (Ln = La-Dy). Furthermore, this work focuses on Sm-monazite phases. Samarium with an atomic number of 62 is located in the middle part of the lanthanides showing the monazite structure. Accordingly, it has a mean cationic radius within the Ln-monazite-series and hence shows a relative high flexibility regarding the incorporation of radionuclides with different radii. Sintering densities of SmPO 4 ceramics were optimised by varying process parameters like pressure and number of pressing steps. An irregular texture as well as densities of 94% of the theoretical value could be achieved. The resistance of Sm-monazite against ionising radiation were examined. Radiation damages caused by the α-decay of radionuclides incorporated in a ceramic matrix were simulated by computer calculations and experimentally by heavy ion bombardment of SmPO 4 . Thin layers of

  9. COMPORTAMENTO A CORROSIONE E TRIBOCORROSIONE DI RIVESTIMENTI CERMET E CERMET/ SUPERLATTICE

    OpenAIRE

    Monticelli, C.; Zucchi, F.

    2009-01-01

    È stato studiato il comportamento a corrosione e tribocorrosione di riporti cermet e cermet/superlattice,applicati su campioni di acciaio. I riporti cermet consistono in riporti termici HVOF a spessore,di tipo WC-12Co o Cr3C2-37WC-18Me. I doppi riporti cermet/superlattice sono ottenuti sovrapponendoai depositi cermet citati un superlattice a base di nitruri, in cui si alternano strati di CrN e di NbN. Unasoluzione al 3.5 % di NaCl costituisce l’ambiente aggressivo. Le condizioni di tribocorro...

  10. Analysis of the minority actinides transmutation in a sodium fast reactor with uniform load pattern by the MCNPX-CINDER code

    International Nuclear Information System (INIS)

    Ochoa Valero, R.; Garcia-Herranz, N.; Aragones, J. M.

    2010-01-01

    The aim of this study is to evaluate the minority actinides transmutation in sodium fast reactors (SFR) assuming a uniform load pattern. It is determined the isotopic evolution of the actinides along burn, and the evolution of the reactivity and the reactivity coefficients. For that, it is used the MCNPX neutron transport code coupled with the inventory code CINDER90.

  11. Process for denitrating waste solutions containing nitric acid actinides simultaneously separating the actinides

    International Nuclear Information System (INIS)

    Gompper, K.

    1984-01-01

    The invention should reduce the acid and nitrate content of waste solutions containing nitric acid as much as possible, should reduce the total salt content of the waste solution, remove the actinides contained in it by precipitation and reduce the α radio-activity in the remaining solution, without having to worry about strong reactions or an increase in the volume of the waste solution. The invention achieves this by mixing the waste solution with diethyl oxalate at room temperature and heating the mixture to at least 80 0 C. (orig.) [de

  12. Cermet materials, self-cleaning cermet filters, apparatus and systems employing same

    Science.gov (United States)

    Kong, Peter C.

    2005-07-19

    A self-cleaning porous cermet material, filter and system utilizing the same may be used in filtering particulate and gaseous pollutants from internal combustion engines having intermetallic and ceramic phases. The porous cermet filter may be made from a transition metal aluminide phase and an alumina phase. Filler materials may be added to increase the porosity or tailor the catalytic properties of the cermet material. Additionally, the cermet material may be reinforced with fibers or screens. The porous filter may also be electrically conductive so that a current may be passed therethrough to heat the filter during use. Further, a heating element may be incorporated into the porous cermet filter during manufacture. This heating element can be coated with a ceramic material to electrically insulate the heating element. An external heating element may also be provided to heat the cermet filter during use.

  13. Transmutation of minor actinides in a spherical torus tokamak fusion reactor, FDTR

    International Nuclear Information System (INIS)

    Feng, K.M.; Zhang, G.S.; Deng, M.G.

    2003-01-01

    In this paper, a concept for the transmutation of minor actinide (MA) nuclear wastes based on a spherical torus (ST) tokamak reactor, FDTR, is put forward. A set of plasma parameters suitable for the transmutation blanket was chosen. The 2-D neutron transport code TWODANT, the 3-D Monte Carlo code MCNP/4B, the 1-D neutron transport and burn-up calculation code BISON3.0 and their associated data libraries were used to calculate the transmutation rate, the energy multiplication factor and the tritium breeding ratio of the transmutation blanket. The calculation results for the system parameters and the actinide series isotopes for different operation times are presented. The engineering feasibility of the center-post (CP) of FDTR has been investigated and the results are also given. A preliminary neutronics calculation based on an ST transmutation blanket shows that the proposed system has a high transmutation capability for MA wastes. (author)

  14. Neutron nuclear data evaluation for actinide nucleic

    International Nuclear Information System (INIS)

    Chen Guochang; Yu Baosheng; Duan Junfeng; Ge Zhigang; Cao Wentian; Tang Guoyou; Shi Zhaomin; Zou Yubin

    2010-01-01

    The nuclear data with high accuracy for minor actinides are playing an important role in nuclear technology applications, including reactor design and operation, fuel cycle concepts, estimation of the amount of minor actinides in high burn-up reactors and the minor actinides transmutation. Through describe the class of nuclear data and nuclear date library, and introduce the procedure of neutron nuclear data evaluation. 234 U(n, f) and 237 Np(n, 2n) reaction experimental data evaluation was evaluated. The fission nuclear data are updated and improved. (authors)

  15. A review of the demonstration of innovative solvent extraction processes for the recovery of trivalent minor actinides from PUREX raffinate

    International Nuclear Information System (INIS)

    Modolo, G.; Wilden, A.; Geist, A.; Magnusson, D.; Malmbeck, R.

    2012-01-01

    The selective partitioning (P) of long-lived minor actinides from highly active waste solutions and their transmutation (T) to short-lived or stable isotopes by nuclear reactions will reduce the long-term hazard of the high-level waste and significantly shorten the time needed to ensure their safe confinement in a repository. The present paper summarizes the on-going research activities at Forschungszentrum Juelich (FZJ), Karlsruher Institut fuer Technologie (KIT) and Institute for Transuranium Elements (ITU) in the field of actinide partitioning using innovative solvent extraction processes. European research over the last few decades, i.e. in the NEWPART, PARTNEW and EUROPART programmes, has resulted in the development of multi-cycle processes for minor actinide partitioning. These multi-cycle processes are based on the co-separation of trivalent actinides and lanthanides (e.g. by the DIAMEX process), followed by the subsequent actinide(III)/lanthanide(III) group separation in the SANEX process. The current direction of research for the development of innovative processes within the recent European ACSEPT project is discussed additionally. This paper is focused on the development of flow-sheets for recovery of americium and curium from highly active waste solutions. The flow-sheets are verified by demonstration processes, in centrifugal contactors, using synthetic or genuine fuel solutions. The feasibility of the processes is also discussed. (orig.)

  16. Enhancing BWR proliferation resistance fuel with minor actinides

    Science.gov (United States)

    Chang, Gray S.

    2009-03-01

    To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides ( 237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO 2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm 3) to the top (0.35 g/cm 3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides ( 237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in

  17. Possibility of plutonium burning out and minor actinides transmutation in CANDU type reactor

    International Nuclear Information System (INIS)

    Gerasimov, A.S.; Kiselev, G.V.; Myrtsymova, L.A.

    2000-01-01

    The possibility of power or weapon-grade plutonium use as nuclear fuel in CANDU type reactor with simultaneous minor actinides burn-out is studied. Total thermal power is 1900 MW. The fuel lifetime makes 0.24 years, neutron flux density 10 14 neutr/cm 2 s. About 40-45 % of plutonium is incinerated during fuel lifetime. If weapon-grade plutonium is used in fuel channels instead of power one, its consumption is 40% lower. (author)

  18. Mechanical behaviour of U3O8-Al cermets

    International Nuclear Information System (INIS)

    Figueredo, A.M. de; Ferreira, I.

    1981-01-01

    Homogeneous, high density U 3 O 8 -Al cermets, containing between 5 W% and 55 Wt% of U 3 ω 8 were fabricated using hot swaging and powder metallurgy technics. Tensile tests were performed at room temperature on specimens obtained from the cermets fabricated. The results show that the ultimate tensile strength (UTS) and elongation to fracture decrease with increasing U 3 O 8 in the cermet. The UTS is shown to be proportional to the minimum matrix load bearing cross-sectional area. The main influence of an increase in the content of U 3 O 8 in the cermet appears to be the decrease in the minimum matrix, load bearing cross-section. (Author) [pt

  19. Solid solution cermet: (Ti,Nb)(CN)-Ni cermet.

    Science.gov (United States)

    Kwon, Hanjung; Jung, Sun-A

    2014-11-01

    Solid solution powders without W, (Ti,Nb)(CN) powders with a B1 structure (NaCl like), were synthesized by high energy milling and carbothermal reduction in nitrogen. The range of molar ratios of Ti/Nb for forming complete (Ti,Nb)(CN) phase was broader than that of Ti/W for the (Ti,W)(CN) phase because carbide or carbonitride of Nb had a B1 crystal structure identical to Ti(CN) while WC had a hexagonal crystal structure. The results revealed that the hardness of (Ti,Nb)(CN)-Ni cermets was higher than that of (Ti,W)(CN)-Ni cermets. The lower density of the (Ti,Nb)(CN) powder contributed to the higher hardness compared to (Ti,W)(CN) because the volumetric ratio of (Ti,Nb)(CN) in the (Ti,Nb)(CN)-Ni cermets was higher than that of (Ti,Nb)(CN) in the (Ti,W)(CN)-Ni cermets at the same weight ratio of Ni. Additionally, it was assumed that intrinsic the properties of (Ti,Nb)(CN) could also be the cause for the high hardness of the (Ti,Nb)(CN)-Ni cermets.

  20. Nuclear fuel activity with minor actinides after their useful life in a BWR

    International Nuclear Information System (INIS)

    Martinez C, E.; Ramirez S, J. R.; Alonso V, G.

    2016-09-01

    Nuclear fuel used in nuclear power reactors has a life cycle, in which it provides energy, at the end of this cycle is withdrawn from the reactor core. This used fuel is known as spent nuclear fuel, a strong problem with this fuel is that when the fuel was irradiated in a nuclear reactor it leaves with an activity of approximately 1.229 x 10 15 Bq. The aim of the transmutation of actinides from spent nuclear fuel is to reduce the activity of high level waste that must be stored in geological repositories and the lifetime of high level waste; these two achievements would reduce the number of necessary repositories, as well as the duration of storage. The present work is aimed at evaluating the activity of a nuclear fuel in which radioactive actinides could be recycled to remove most of the radioactive material, first establishing a reference of actinides production in the standard nuclear fuel of uranium at end of its burning in a BWR, and a fuel rod design containing 6% of actinides in an uranium matrix from the enrichment tails is proposed, then 4 standard uranium fuel rods are replaced by 4 actinide bars to evaluate the production and transmutation of the same, finally the reduction of actinide activity in the fuel is evaluated. (Author)

  1. Use of plutonium and minor actinides as fuel in high temperature pebble bed reactors for waste minimization

    International Nuclear Information System (INIS)

    Meier, Astrid; Bernnat, Wolfgang; Lohnert, Guenther

    2009-01-01

    Energy production by nuclear fission gives rise to longlived radionuclides, such as plutonium and americium. The ''PuMA'' (Plutonium and Minor Actinides Waste Management) research project within the 6th Framework Program of the European Union serves to minimize waste arisings and transmute plutonium and minor actinides from spent LWR fuel elements by means of modular high-temperature reactors (HTR). Coating the fuel, which consists of kernels approx. 250 μm in radius and surrounded by graphite as the moderator material, allows very high operating and accident temperatures and very high burnups. One point examined is whether the inherent safety characteristics known for uranium oxide also exist for (PuO 2 + MAO 2 ) fuel. On the basis of a reference reactor similar to the South African PBMR-400, various loading strategies at maximum burnup are considered with a view to the inherent safety of the HTR. (orig.)

  2. Summary report of first research coordination meeting on Minor Actinide Nuclear Reaction Data (MANREAD)

    International Nuclear Information System (INIS)

    Kaeppeler, F.; Mengoni, A.

    2008-09-01

    The first Research Co-ordination Meeting of the MANREAD (Minor Actinides Neutron Reaction Data) was held at the IAEA Headquarters in Vienna from 19 to 23 November 2007. A summary of the discussion which took place at the meeting is reported here. In addition, a task assignment list of the experimental data assessment activities was agreed, and is provided together with the plan for future CRP activities. (author)

  3. Measurement of fast neutron induced fission cross section of minor-actinide

    International Nuclear Information System (INIS)

    Hirakawa, Naohiro

    2000-06-01

    In fuel cycles with recycled actinide, core characteristics are largely influenced by minor actinide (MA: Np, Am, Cm). Accurate nuclear data of MA such as fission cross section are required to estimate the effect of MA with high accuracy. In this study, fast neutron induced fission cross section of MA is measured using Dynamitron accelerator in Tohoku University. The followings were performed in this fiscal year; (1) Research of nuclear data of MA, (2) Sample preparation and sample mass assay, (3) Investigation of neutron sources with the energy of several 10 keV, (4) Preliminary measurement of fission cross section using Dynamitron accelerator. As the result, four 237 Np samples were prepared and the sample mass were measured using alpha-spectrometry with the accuracy of 1.2%. Then, it was confirmed that a neutron source via 7 Li(p,n) 7 Be reaction using a Li-thick target is suitable for measuring fission cross section of MA in the energy region of several 10 keV. Furthermore, it was verified by the preliminary measurement that the measurement of fission cross section of MA is available using a fission chamber and electronics developed in this study. (author)

  4. Incineration of actinide targets in a pressurized water reactor spin project

    International Nuclear Information System (INIS)

    Puill, A.; Bergeron, J.

    1993-01-01

    The ability of Pressurized Water Reactors (PWR) with uranium fuel to limit the inventory growth of minor actinides (237 neptunium, and americium) produced by the French nuclear powerplants is studied. Targets containing an actinide oxide mixed to an inert matrix are loaded in some reactors. After being irradiated along with the fuel, the target is specially reprocessed. The remaining actinide and the plutonium which is produced, added to fresh actinide, are recycled in new targets. The radiotoxicity balance, with and without incineration, is examined considering that only the losses coming from the target reprocessing treated as waste. A scenario arbitrarily based on 18 years of operation results in a reduction of the radiotoxicity of the waste by a factor between 10 and 20, depending on the actinide considered. 6 refs., 6 figs., 6 tabs

  5. U-Zr-RE Fuel Alloy with Minor Actinides

    Energy Technology Data Exchange (ETDEWEB)

    Song, Hoon; Kim, Jong Hwan; Ko, Young Mo; Kim, Ki Hwan; Park, Jeong Yong; Lee, Chan Bock [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Metallic fuels, such as the 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 the amount of highly radioactive spent nuclear fuels since the 1980s. Metallic fuels fit well with such a concept owing to their high thermal conductivity, high thermal expansion, compatibility with a pyro-metallurgical reprocessing scheme, and their demonstrated fabrication at engineering scale in a remote hot cell environment. To increase the productivity and efficiency of the fuel fabrication process waste streams must be minimized and fuel losses quantified and reduced to lower levels. In this study, U-Zr alloy system fuel slugs were fabricated by an injection casting method. After casting a considerable number of fuel slugs in the casting furnaces, the fuel loss in the melting chamber, the crucible, and the molds have been evaluated quantitatively.

  6. Irradiation Degradation of Adsorbents for Minor Actinides Recovery

    International Nuclear Information System (INIS)

    Watanabe, S.; Sano, Y.; Kofuji, H.; Takeuchi, M.; Koizumi, T.

    2015-01-01

    Extraction chromatography is one of the promising technologies for minor actinides (MA: Am and Cm) recovery from high-level liquid waste. The degradation behaviour of the organic species in the adsorbents under radiation exposure is important to discuss the safety and durability of the adsorbent in the extraction chromatography process. In this study, gamma-ray irradiation experiments on TODGA/SiO 2 -P adsorbent were carried out to investigate the degradation products from radiolysis of the adsorbent. The degraded organic species eluted from the adsorbent and those remaining inside the adsorbent were thoroughly identified by GC/MS, FT-IR and NMR analyses. The species suspected as hydrolysis products of TODGA were mainly detected from the analyses. Since some radicals such as.H or.OH are generated by the gamma-ray irradiation on water molecules, it was discussed that the radicals products from radiolysis of HNO 3 solution are related to the degradation reaction of the extractants. (authors)

  7. Electrical and thermal properties of niobium-base cermets

    International Nuclear Information System (INIS)

    Skidan, B.S.; Vlasov, A.S.; Alekseev, V.A.; Myl'nikova, T.S.; Ryzhkov, Yu.F.

    1979-01-01

    Behaviour of corundum-niobium cermets containing 16-70 vol.% metal was studied at low temperatures. It is found that the given materials are superconductors with Tsub(k) 6-7 K but their resistivity before their transfer into superconducting state is determined by the metal concentration and is found within 10 -2 -10 -4 Ohm. cm. The cermet heat conductivity is found to increase with the metal content

  8. Molybdenum-UO2 cermet irradiation at 1145 K.

    Science.gov (United States)

    Mcdonald, G.

    1971-01-01

    Two molybdenum-uranium dioxide cermet fuel pins with molybdenum clad were fission-heated in a forced-convection helium coolant for sufficient time to achieve 5.3% burnup. The cermet core contained 20 wt % of 93.2% enriched uranium dioxide. The results were as follows: there was no visible change in the appearance of the molybdenum clad during irradiation; the maximum increase in diameter of the fuel pins was 0.8%; there was no migration of uranium dioxide along grain boundaries and no evident interaction between molybdenum and uranium dioxide; and, finally, approximately 12% of the fission gas formed was released from the cermet core into the gas plenum.

  9. Uranium-Based Cermet Alloys; Cermets a base d'uranium; Metallokeramicheskie splavy na osnove urana; Cermets a base de uranio

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, V. E.; Zelenskij, V. F.; Voloshchuk, A. I.; Grishok, V. N. [Fiziko-Tekhnicheskij Institut an USSR, Khar' kov, SSSR (Russian Federation)

    1963-11-15

    The paper describes certain features of dispersion-hardened uranium-based cermets. As possible hardening materials, consideration was given to UO{sub 2}, UC, Al{sub 2}O{sub 3}, MgO and UBe{sub 13}. Data were obtained on the behaviour of uranium alloys containing the above-mentioned admixtures during creep tests, short-term strength tests and cyclic thermal treatment. The corrosion resistance o f UBe{sub 13}-based uranium alloys was also studied. )author) [French] Les auteurs decrivent certaines proprietes de cermets a base d'uranium, dont la resistance a ete accrue a l'aide de particules dispersees. Les materiaux utilises a cette fin sont notamment: UO{sub 2}, UC, Al{sub 2}O{sub 3}, MgO et UBe{sub 13}. Les auteurs indiquent les donnees obtenues sur le comportement des cermets a l'uranium; durant les essais de fluage, les essais de resistance a court terme et le traitement thermique cyclique, en mentionnant les substances ajoutees. Ils etudient enfin la resistance a la corrosion des cermets d'uranium et UBe{sub 13}. (author) [Spanish] Los autores describen algunas propiedades de los cermets a base de uranio, reforzados por particulas de diversos compuestos en dispersion. En calidad de posibles materiales de refuerzo, ensayaron el UO{sub 2}, el UC, el Al{sub 2}O{sub 3}, el MgO y el UBe{sub 13}. Obtuvieron datos sobre el comportamiento de esas aleaciones en ensayos de fluencia, ensayoe rapidos de resistencia y tratamiento termico ciclico. Por ultimo, estudiaron la resistencia a la corrosion de las aleaciones de uranio a base de UBe{sub 13}. (author) [Russian] Daetsya opisanie nekotorykh svojstv metallokeramicheskikh splavov urana, uprochnennykh dispersionnymi chastitsami. V kachestve vozmozhnykh uprochnyayushchikh materialov izuchalis' UO{sub 2}, UC, Al{sub 2}O{sub 3} , MgO i UBe{sub 13}. Polucheny dannye o povedenii splavov urana s ukazannymi primesyami pri kripovykh ispytaniyakh, pri kratkovremennykh prochnostnykh ispytaniyakh i pri tsiklicheskoj termoobrabotke

  10. Polymer Inclusion Membrane Containing a Tripodal Diglycolamide Ligand: Actinide Ion Uptake and Transport Studies

    NARCIS (Netherlands)

    Mahanty, B.; Mohapatra, P.K.; Raut, D.R.; Das, D.K.; Behere, P.G.; Afzal, M.; Verboom, Willem

    2016-01-01

    A cellulose triacetate (CTA)-based polymer inclusion membrane (PIM) containing a C-pivot tripodal diglycolamide (T-DGA) as the carrier extractant and 2-nitrophenyl octyl ether (NPOE) as the plasticizer shows potential for the uptake of actinides from acidic feed solutions. The uptake of actinides

  11. Cermet anode compositions with high content alloy phase

    Science.gov (United States)

    Marschman, Steven C.; Davis, Norman C.

    1989-01-01

    Cermet electrode compositions comprising NiO-NiFe.sub.2 O.sub.4 -Cu-Ni, and methods for making, are disclosed. Addition of nickel metal prior to formation and densification of a base mixture into the cermet allows for an increase in the total amount of copper and nickel that can be contained in the NiO-NiFe.sub.2 O.sub.4 oxide system. Nickel is present in a base mixture weight concentration of from 0.1% to 10%. Copper is present in the alloy phase in a weight concentration of from 10% to 30% of the densified composition. Such cermet electrodes can be formed to have electrical conductivities well in excess of 100 ohm.sup.-1 cm.sup.-1. Other alloy and oxide system cermets having high content metal phases are also expected to be manufacturable in accordance with the invention.

  12. Minor actinides incineration by loading moderated targets in fast reactor

    International Nuclear Information System (INIS)

    Wu Hongchun; Sato, Daisuke; Takeda, Toshikazu

    2000-01-01

    The effect of hydrogen concentration and loaded mass of minor actinides (MAs) in the target on the core performance and MAs transmutation rate was analyzed in this paper. An optimum core was proposed which has 96 MAs target assemblies of which MAs fuel pins per assembly is 38 with the composition ratio U/MA/Zr/H of 1/4/10/50. This optimized core offers good core performance and can transmute MAs very effectively, the transmutation rate was about 67% (939 kg) and the incinerate (transmute by fission) rate was about 35% (489 kg) through 3 years of reactor operation. It is about 2-3 times larger than current transmutation method that MAs are loaded homogeneously in the PWR and fast reactor core. (author)

  13. State-of-art technology of fuels for burning minor actinides. An OECD/NEA study

    International Nuclear Information System (INIS)

    Ogawa, Toru; Konings, R.J.M.; Pillon, S.; Schram, R.P.C.; Verwerft, M.; Wallenius, J.

    2005-01-01

    At OECD/NEA, Working Party on Scientific Issues in Partitioning and Transmutation was formed for 2000-2004, which studied the status and trends of scientific issues in Partitioning and Transmutation (P and T). The study included the scientific and technical issues of fuels and materials, which are related to dedicated systems for transmutation. This paper summarizes the state-of-art technology of the fuels for burning minor actinides (neptunium, americium and curium). (author)

  14. Demonstration of innovative partitioning processes for minor actinide recycling from high active waste solutions

    International Nuclear Information System (INIS)

    Modolo, G.; Wilden, A.; Geist, A.; Malmbeck, R.; Taylor, R.

    2014-01-01

    The recycling of the minor actinides (MA) using the Partitioning and Transmutation strategy (P and T) could contribute significantly to reducing the volume of high level waste in a geological repository and to decreasing the waste's longterm hazards originating from the long half-life of the actinides. Several extraction processes have been developed worldwide for the separation and recovery of MA from highly active raffinates (HAR, e.g. the PUREX raffinate). A multi-cycle separation strategy has been developed within the framework of European collaborative projects. The multi-cycle processes, on the one hand, make use of different extractants for every single process. Within the recent FP7 European research project ACSEPT (Actinide reCycling by SEParation and Transmutation), the development of new innovative separation processes with a reduced number of cycles was envisaged. In the so-called 'innovative SANEX' concept, the trivalent actinides and lanthanides are co-extracted from the PUREX raffinate by a DIAMEX like process (e.g. TODGA). Then, the loaded solvent is subjected to several stripping steps. The first one concerns selectively stripping the actinides(III) with selective water-soluble ligands (SO3-Ph-BTB), followed by the subsequent stripping of trivalent lanthanides. A more challenging route studied also within our laboratories is the direct actinide(III) separation from a PUREX-type raffinate using a mixture of CyMe 4 BTBP and TODGA as extractants, the so-called One cycle SANEX process. A new approach, which was also studied within the ACSEPT project, is the GANEX (Grouped ActiNide EXtraction) concept addressing the simultaneous partitioning of all transuranium (TRU) elements for their homogeneous recycling in advanced generation IV reactor systems. Bulk uranium is removed in the GANEX 1st cycle, e.g. using a monoamide extractant and the GANEX 2nd cycle then separates the TRU. A solvent composed of TODGA + DMDOHEMA in kerosene has been shown to

  15. Cermets and method for making same

    Science.gov (United States)

    Aaron, W. Scott; Kinser, Donald L.; Quinby, Thomas C.

    1983-01-01

    The present invention is directed to a method for making a wide variety of general-purpose cermets and for radioactive waste disposal from ceramic powders prepared from urea-dispersed solutions containing various metal values. The powders are formed into a compact and subjected to a rapid temperature increase in a reducing atmosphere. During this reduction, one or more of the more readily reducible oxides in the compact is reduced to a selected substoichiometric state at a temperature below the eutectic phase for that particular oxide or oxides and then raised to a temperature greater than the eutectic temperature to provide a liquid phase in the compact prior to the reduction of the liquid phase forming oxide to solid metal. This liquid phase forms at a temperature below the melting temperature of the metal and bonds together the remaining particulates in the cermet to form a solid polycrystalline cermet.

  16. Minor Actinide Transmutation Physics for Low Conversion Ratio Sodium Fast Reactors

    International Nuclear Information System (INIS)

    Mehdi Asgari; Samuel E. Bays; Benoit Forget; Rodolfo Ferrer

    2007-01-01

    The effects of varying the reprocessing strategy used in the closed cycle of a Sodium Fast Reactor (SNF) prototype are presented in this paper. The isotopic vector from the aqueous separation of transuranic (TRU) elements in Light Water Reactor (LWR) spent nuclear fuel (SNF) is assumed to also vary according to the reprocessing strategy of the closed fuel cycle. The decay heat, gamma energy, and neutron emission of the fuel discharge at equilibrium are found to vary depending on the separation strategy. The SFR core used in this study corresponds to a burner configuration with a conversion ratio of ∼0.5 based on the Super-PRISM design. The reprocessing strategies stemming from the choice of either metal or oxide fuel for the SFR are found to have a large impact on the equilibrium discharge decay heat, gamma energy, and neutron emission. Specifically, metal fuel SFR with pyroprocessing of the discharge produces the largest amount of TRU consumption (166 kg per Effective Full Power Year or EFPY), but also the highest decay heat, gamma energy, and neutron emission. On the other hand, an oxide fuel SFR with PUREX reprocessing minimizes the decay heat and related parameters of interest to a minimum, even when compared to thermal Mixed Oxide (MOX) or Inert Matrix Fuel (IMF) on a per mass basis. On an assembly basis, however, the metal SFR discharge has a lower decay heat than an equivalent oxide SFR assembly for similar minor actinide consumptions (∼160 kg/EFPY.) Another disadvantage in the oxide PUREX reprocessing scenario is that there is no consumption of americium and curium, since PUREX reprocessing separates these minor actinides (MA) and requires them to be disposed of externally

  17. Oxidation-resistant cermet

    Science.gov (United States)

    Phillips, W. M.

    1977-01-01

    Chromium metal alloys and chromium oxide ceramic are combined to produce cermets with oxidation-resistant properties. Application of cermets includes use in hot corrosive environments requiring strong resistive materials.

  18. Measurement of fast neutron induced fission cross section of minor-actinide

    International Nuclear Information System (INIS)

    Hirakawa, Naohiro

    1997-03-01

    In fuel cycles with recycled actinide, core characteristics are largely influenced by minor actinide (MA: Np, Am, Cm). Accurate nuclear data of MA such as fission cross section are required to estimate the effect of MA with high accuracy. In this study, fast neutron induced fission cross section of MA is measured using Dynamitron Accelerator in Tohoku University. The experimental method and the samples, which were developed or introduced during the last year, were improved in this fiscal year: (1) Development of a sealed fission chamber, (2) Intensification of Li neutron target, (3) Improvement of time-resolution of Time-of-Flight (TOF) electronic circuit, (4) Introduction of Np237 samples with large sample mass and (5) Introduction of a U235 sample with high purity. Using these improved tools and samples, the fission cross section ratio of Np237 relative to U235 was measured between 5 to 100 keV, and the fission cross section of Np237 was deduced. On the other hand, samples of Am241 and Am243 were obtained from Japan Atomic Energy Research Institute (JAERI) after investigating fission cross section of two americium isotopes (Am241 and Am 243) which are important for core physics calculation of fast reactors. (author)

  19. Cermets from molten metal infiltration processing

    Science.gov (United States)

    Landingham, Richard Lee

    2012-09-18

    New cermets with improved properties and applications are provided. These new cermets have lower density and/or higher hardness than B4C cermet. By incorporating other new ceramics into B4C powders or as a substitute for B4C, lower densities and/or higher hardness cermets result. The ceramic powders have much finer particle size than those previously used which significantly reduces grain size of the cermet microstructure and improves the cermet properties.

  20. Plutonium and minor actinides recycle in equilibrium fuel cycles of pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Waris, A.; Sekimoto, H. [Research Lab. for Nuclear Reactors, Tokyo Institute of Technology, Tokyo (Japan)

    2001-07-01

    A study on plutonium and minor actinides (MA) recycle in equilibrium fuel cycles of pressurized water reactors (PWR) has been performed. The calculation results showed that the enrichment and the required amount of natural uranium decrease significantly with increasing number of confined plutonium and MA when uranium is discharged from the reactor. However, when uranium is totally confined, the enrichment becomes extremely high. The recycle of plutonium and MA together with discharging uranium can reduce the radio-toxicity of discharged heavy metal (HM) waste to become less than that of loaded uranium. (author)

  1. Superconducting cermets

    International Nuclear Information System (INIS)

    Goyal, A.; Funkenbusch, P.D.; Chang, G.C.S.; Burns, S.J.

    1988-01-01

    Two distant classes of superconducting cermets can be distinguished, depending on whether or not a fully superconducting skeleton is established. Both types of cermets have been successfully fabricated using non-noble metals, with as high as 60wt% of the metal phase. The electrical, magnetic and mechanical behavior of these composites is discussed

  2. Use of fast reactors for actinide transmutation

    International Nuclear Information System (INIS)

    1993-03-01

    The management of radioactive waste is one of the key issues in today's discussions on nuclear energy, especially the long term disposal of high level radioactive wastes. The recycling of plutonium in liquid metal fast breeder reactors (LMFBRs) would allow 'burning' of the associated extremely long life transuranic waste, particularly actinides, thus reducing the required isolation time for high level waste from tens of thousands of years to hundreds of years for fission products only. The International Working Group on Fast Reactors (IWGFR) decided to include the topic of actinide transmutation in liquid metal fast breeder reactors in its programme. The IAEA organized the Specialists Meeting on Use of Fast Breeder Reactors for Actinide Transmutation in Obninsk, Russian Federation, from 22 to 24 September 1992. The specialists agree that future progress in solving transmutation problems could be achieved by improvements in: Radiochemical partitioning and extraction of the actinides from the spent fuel (at least 98% for Np and Cm and 99.9% for Pu and Am isotopes); technological research and development on the design, fabrication and irradiation of the minor actinides (MAs) containing fuels; nuclear constants measurement and evaluation (selective cross-sections, fission fragments yields, delayed neutron parameters) especially for MA burners; demonstration of the feasibility of the safe and economic MA burner cores; knowledge of the impact of maximum tolerable amount of rare earths in americium containing fuels. Refs, figs and tabs

  3. Partnew - New solvent extraction processes for minor actinides - final report

    International Nuclear Information System (INIS)

    Madic, C.; Testard, F.; Hudson, M.J.; Liljenzin, J.O.; Christiansen, B.; Ferrando, M.; Facchini, A.; Geist, A.; Modolo, G.; Gonzalez-Espartero, A.; Mendoza, J. de

    2004-01-01

    The objectives of the European project PARTNEW were to define solvent extraction processes for the partitioning of the minor actinides, Am and Cm, from the aqueous high active raffinate or high active concentrate issuing the reprocessing of nuclear spent fuels by the PUREX process. Eleven laboratories participated to the research: 1/ CEA-DEN (Marcoule), 2/ CEA-DSM (Saclay), 3/ UREAD (U.K.), 4/ CTU (Sweden), 5/ ITU (Germany), 6/ ENEA (Italy), 7/ PoliMi (Italy), 8/ FZK-INE (Germany), 9/ FZJ-ISR (Germany), 10/ CIEMAT (Spain) and 11/ UAM (Spain). The research was organised into eight work packages (WP): Basic and applied DIAMEX studies, using diamide extractants for the co-extraction of actinides(III) (An(III)) and lanthanides(III) (Ln(III)) nitrates (WP1 and WP2), Basic and applied SANEX studies based on the use of polydentate N-ligands for the An(III)/Ln(III) separation (WP3 and WP4), Basic and applied SANEX studies based on the use of synergistic mixtures made of bis-(chloro-phenyl)-di-thio-phosphinic acid + neutral O-bearing ligand, (WP5 and WP6), Basic SANEX studies for the An(III)/Ln(III) separation, based on the use of new S-bearing ligands, Basic and applied studies for the Am(III)/Cm(III) separation. The work done in the fundamental and applied domains was very fruitful. Several processes have been successfully tested with genuine high active raffinates and concentrate. (authors)

  4. Remote micro-encapsulation of curium-gold cermets

    International Nuclear Information System (INIS)

    Coops, M.S.; Voegele, A.L.; Hayes, W.N.; Sisson, D.H.

    1980-01-01

    A technique is described for fabricating minature, high-density capsules of curium-244 oxide contained in three concentric jackets of metallic gold (or silver), with the outer surface being free of alpha contamination. The completed capsules are right circular cylinders 0.2500-inch diameter and 0.125-inch tall, with each level of containment soldered (or brazed) closed. A typical capsule would contain approx. 70 mg of 244 Cm (5.7 Ci) mixed with 120 mg of gold powder in the form of a cermet wafer clad in three concentric, 0.010-inch thick, liquid tight jackets. This method of fabrication eliminates voids between the jackets and produces a minimum size, maximum density capsule. Cermet densities of 11.5 g/cc were obtained, with an overall density of 17.3 g/cc for the finished capsule

  5. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.

    1987-01-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. The concept evolved in the 1960's with the objective of developing a reactor design which could be used for a wide range of mobile power generation systems including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests and in-reactor irradiation tests using cermet fuel were carried out by General Electric in the 1960's as part of the 710 Development Program and by Argonne National laboratory in a subsequent activity. Cermet fuel development programs are currently underway at Argonne National laboratory and Pacific Northwest Laboratory as part of the Multi-Megawatt Space Power Program. Key features of the cermet fueled reactor design are 1) the ability to achieve very high coolant exit temperatures, and 2) thermal shock resistance during rapid power changes, and 3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, there is a potential for achieving a long operating life because of 1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and 2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core

  6. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

  7. Minor actinide burning in dedicated lead-bismuth cooled fast reactors

    International Nuclear Information System (INIS)

    Hejzlar, P.; Driscoll, M.J.; Kazimi, M.S.; Todreas, N.E.

    2001-01-01

    The destruction of minor actinides (MA) in dedicated burners is of contemporary interest in Europe and Japan because it requires the deployment of smaller number of special transmutation facilities. A major fraction of Pu from spent LWR fuel can be then burned in PWRs (or fast reactors) using dedicated fertile-free fuel assemblies. However, the design of MA burning fast spectrum cores poses significant challenges because of deterioration of key safety parameters, in particular of the coolant void coefficient. This study proposes the concept of an lead-bismuth eutectic (LBE)-cooled dedicated MA burner having metallic fuel (MA-Pu-Zr) and streaming assemblies to attain acceptable coolant void worth performance. It is shown that a large 1800 MWth fertile-free core containing 37 wt% TRU with very high fraction of MA(59 wt%) from LWR spent fuel can be burned in a first cycle for 700 EFPDs with a very small reactivity swing: less than β eff . Moreover, the reactivity void worth is negative for a fully voided core when all surrounding coolant is kept at reference density. However, the core reactivity increases as coolant density falls from the reference value of 10.25 to 6 g/cm 3 . Because its coolant density coefficient value is less than that of a sodium cooled IFR, the concept provides good potential for the achievement of self-regulation characteristics in unprotected events, provided that small negative fuel temperature feedback can be maintained. (authors)

  8. New strategy for minor actinides partitioning preliminary results on the electrovolatilization of ruthenium and on the stabilization of Am(IV) in nitric acid with phosphotunsgstate ligand

    International Nuclear Information System (INIS)

    Adnet, J.M.; Madic, C.

    1989-01-01

    On problems related to the long term storage in deep geological repositories of high active wastes (H.A.W.) is due to the presence of minor actinide isotopes. Thus after the decay of the fission products (≅ 300 years) the toxicity of these H.A.W. is mainly due to the minor actinides. One solution is based on actinide partitioning followed by transmutation into fission products with short half-lives. A simpler processes than those developed previously, can be based on the possible oxidation of minor actinides to the + IV or + VI oxidation states and their selective extraction. The first step to study is the elimination of the ruthenium (whose presence would be detrimental to oxidize minor actinides) which can be done by electrovolatilization of Ru on the RuO 4 form. The rate of electrovolatilization can be increased by the use of the following electronic mediators, AgI/AgII(1); CeIII/Ce(2), and CoII/CoIII(3), the efficiency of which decreases in the order: 1 > 2 > 3. The effectiveness of that process has been proven when treating real H.A.W solution produced during the study of the reprocessing of a MOX fuel irradiated to as burn-up of 52 GWd/t in a LWR: complete Ru removal was obtained. The second part of the study concerns the electrochemical oxidation of AmIII in nitric acid solutions in the presence of a strong complexing agent: P 2 W 17 O 61 K 10 (P.W.).Total americium oxidation to AmIV can be obtained in nitric acid solution with a concentration up to 8 M. No particular drawback was induced by the presence of an amount of lanthanide III (NdIII) in 6 fold excess vs P.W. The stability of AmIV was studied. The other actinides will be present in these solutions, after the electrochemical oxidation step, in the + VI or+ IV oxidation states, thus a selective extraction (vs fission products) could be performed. A possible way to extract actinide IV/P.W complexes is to use dodecylamine nitrate as extractant

  9. Effect of Mo2C content on the properties of TiC/TiB2 base cermets

    International Nuclear Information System (INIS)

    Takagi, Ken-ichi; Osada, Ken; Koike, Wataru; Fujima, Takuya

    2009-01-01

    The effects of Mo 2 C content on the microstructure and mechanical properties of TiC/TiB 2 base cermets were studied using the model cermets with the compositions of TiC/TiB 2 -(11-17)Mo 2 C-24Ni (mass%). TiC and TiB 2 ratio is set to molar ratio of 59:41 that is near quasi-eutectic composition. As a result, both transverse rupture strength and hardness of the cermets showed maxima for the cermet containing 13% Mo 2 C. The cermet achieved remarkable microstructural refinement and still maintained characteristic core-rim structure of the TiC base cermets. TiC/TiB 2 cermets, in addition to TiCN base cermets, are a good alternative material to cemented carbides.

  10. Influence of irradiation on electrical properties of cermet composition

    International Nuclear Information System (INIS)

    Iskhakov, V.M.; Avanesyan, R.R.; Daukeev, D.K.; Nedorezov, V.G.; Chormonov, N.T.; Chormonov, T.Kh.; Shevelev, G.A.

    1986-01-01

    Cermet composition radiation stability and also possibility of directed change of the composition properties during radiation treatment were studied. Investigations were carried out using cermet composition containing 40 mass % of conducting phase (RuO 2 +Nb 2 O 5 additions) and 60 mass % of alumoborosilicate glass. Composition and organic binder mixture was applied to a dielectric substrate with land by stenciling, then was calcinated in the travelling furnace at 850 deg C for 15 min

  11. Cermet high level waste forms: a pregress report

    International Nuclear Information System (INIS)

    Aaron, W.S.; Quinby, T.C.; Kobisk, E.H.

    1978-06-01

    The fixation of high level radioactive waste from both commercial and DOE defense sources as cermets is currently under study. This waste form consists of a continuous iron-nickel base metal matrix containing small particles of fission product oxides. Preliminary evaluations of cermets fabricated from a variety of simulated wastes indicate they possess properties providing advantages over other waste forms presently being considered, namely thermal conductivity, waste loading levels, and leach resistance. This report describes the progress of this effort, to date, since its initiation in 1977

  12. Enhancing VVER annular proliferation resistance fuel with minor actinides

    International Nuclear Information System (INIS)

    Chang, G. S.

    2007-01-01

    Key aspects of the Global Nuclear Energy Partnership (GNEP) are to significantly advance the science and technology of nuclear energy systems and the Advanced Fuel Cycle (AFC) program. It consists of both innovative nuclear reactors and innovative research in separation and transmutation. To accomplish these goals, international cooperation is very important and public acceptance is crucial. The merits of nuclear energy are high-density energy, with low environmental impacts (i.e. almost zero greenhouse gas emission). Planned efforts involve near term and intermediate-term improvements in fuel utilization and recycling in current light water reactors (LWRs) as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The challenges are solving the energy needs of the world, protection against nuclear proliferation, the problem of nuclear waste, and the global environmental problem. To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 2 38Pu and 2 40Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides ( 2 37Np and 2 41Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 2 38Pu /Pu. For future advanced nuclear systems, the minor actinides (MA) are viewed more as a resource to be recycled, or transmuted to less hazardous and possibly more useful forms, rather than simply as a waste stream to be disposed of in expensive repository facilities. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors

  13. Fabrication and characterisation of composite targets for the transmutation of actinides

    International Nuclear Information System (INIS)

    Naestren, C.; Haas, D.; Fernandez, A.; Somers, J.

    2006-01-01

    Transmutation of transuranic elements separated from spent fuel is a way to reduce the toxicity of long-lived nuclides in the waste before disposal. Plutonium and the minor actinides (MA) are reintroduced into the fuel cycle for further irradiation and incineration. Currently CERMET fuel forms, in which a ceramic actinide is dispersed in a matrix, are considered for MA transmutation. In a first step, PuO 2 beads are produced by a sol gel method in which a Pu nitrate solution is converted to solid, dust-free, particles. These porous beads are then infiltrated with an americium nitrate solution to the incipient wetness point and calcined to give the (PuAm)O 2 beads, which are blended with a metal matrix and compacted and sintered to form the final fuel pellet. The matrix used is molybdenum due to its high thermal conductivity and low neutron capture cross section, if it is enriched in 92 Mo. In this work, optimization of the bead porosity is investigated to achieve a higher Am content by infiltration. Addition of carbon to the mother solution in the sol gel step increases the bead porosity but it also changes both bead and final fuel pellet microstructure. A surrogate fuel, with cerium simulating the actinides has been fabricated and its mechanical stability and bead characteristics investigated as a function of carbon content and thermal treatment. The characterization of the surrogate fuel by ceramography, density, porosity, bead-quality, etc., is a necessary step in the process optimization, to be transferred to the production of the actinide samples. This process is now at an advanced stage and is being used for the production of fuels for irradiation tests in the Phenix (Futurix) and HFR-Petten (HELIOS) reactors. In parallel, studies on the dissolution of the fuel pellets, with the aim of dissolving the Mo-matrix while keeping the CeO 2 beads intact, have been initiated. Thus, Mo can be recycled for further fuel fabrication either from production scraps or from

  14. Process for denitrating waste solutions containing nitrates and actinides with simultaneous separation of the actinides

    International Nuclear Information System (INIS)

    Gompper, K.

    1986-01-01

    The invention is intended to reduce the acid and nitrate content of nitrate waste solutions, to reduce the total salt content of the waste solution, to remove the actinides contained in it by precipitation, without any danger of violent reactions or an increase in the volume of the waste solution. The invention achieves this by mixing the waste solution with diethyl oxalate at room temperature and heating the mixture to at least 80 0 C. (orig./PW) [de

  15. Comparative study for minor actinide transmutation in various fast reactor core concepts

    International Nuclear Information System (INIS)

    Ohki, S.

    2001-01-01

    A comparative evaluation of minor actinide (MA) transmutation property was performed for various fast reactor core concepts. The differences of MA transmutation property were classified by the variations of fuel type (oxide, nitride, metal), coolant type (sodium, lead, carbon dioxide) and design philosophy. Both nitride and metal fuels bring about 10% larger MA transmutation amount compared with oxide fuel. The MA transmutation amount is almost unchanged by the difference between sodium and lead coolants, while carbon dioxide causes a reduction by about 10% compared with those. The changes of MA transmutation property by fuel and coolant types are comparatively small. The effects caused by the difference of core design are rather significant. (author)

  16. Direct metal brazing to cermet feedthroughs

    International Nuclear Information System (INIS)

    Hopper, A.C. Jr.

    1984-01-01

    An improved method for brazing metallic components to a cermet surface in an alumina substrate eliminates the prior art metallized layer over the cermet via and adjoining alumina surfaces. Instead, a nickel layer is applied over the cermet surface only and metallic components are brazed directly to this nickel coated cermet surface. As a result, heretofore unachievable tensile strength joints are produced. In addition, cermet vias with their brazed metal components can be spaced more closely in the alumina substrate because of the elimination of the prior art metallized alumina surfaces

  17. Quantification of the Partitioning Ratio of Minor Actinide Surrogates between Zirconolite and Glass in Glass-Ceramic for Nuclear Waste Disposal.

    Science.gov (United States)

    Liao, Chang-Zhong; Liu, Chengshuai; Su, Minhua; Shih, Kaimin

    2017-08-21

    Zirconolite-based glass-ceramic is considered a promising wasteform for conditioning minor actinide-rich nuclear wastes. Recent studies on this wasteform have sought to enhance the partitioning ratio (PR) of minor actinides in zirconolite crystal. To optimize the PR in the SiO 2 -Al 2 O 3 -CaO-TiO 2 -ZrO 2 system, a novel conceptual approach, which can be derived from the chemical composition and quantity of zirconolite crystal in glass-ceramic, was introduced based on the results of Rietveld quantitative X-ray diffraction analysis and transmission electron microscopy energy dispersive X-ray spectroscopy. To verify this new conceptual approach, the influences of the crystallization temperature, the concentration of additives, and ionic radii on the PR of various surrogates (Ce, Nd, Gd, and Yb) in zirconolite were examined. The results reveal that the PR of Nd 3+ in zirconolite can be as high as 41%, but it decreases as the crystallization temperature increases. The quantities of all phases (including crystalline and amorphous) remained nearly constant when increasing the loading of Nd 2 O 3 in glass-ceramic products crystallized at 1050 °C for 2 h. Correspondingly, the PR of Nd 3+ decreases in a linear fashion with the loading contents of Nd 2 O 3 . The radius of ions also has a great influence on the PR, and an increase in the ionic radius leads to a decrease in the PR. This new approach will be an important tool to facilitate the exploration of a glass-ceramic matrix for the disposal of minor actinide-rich nuclear wastes.

  18. Microstructural characterization of cermet-steel interface in rock drilling tool

    International Nuclear Information System (INIS)

    Ybarra, L.A.C.; Molisani, A.L.; Yoshimura, H.N.

    2010-01-01

    Rock drilling tools basically present a WC cermet bonded to a steel shank. The interface cermet-steel plays fundamental role during drilling operation, since the fracture of this interface is the main failure mode of the tools. In this work, the microstructure of this interface in crown samples (type A), prepared in an industrial like process, was evaluated. In this process, a WC-containing powder was infiltrated with a copper alloy at 1100 deg C in a graphite mold previously mounted with a 1020 steel tube. The powder was characterized by XRD analysis and the cross-section microstructure of cermet-steel was analyzed using SEM-EDS. It was observed that Ni and small amount of Cu from cermet matrix diffused into the superficial region of the steel, and the Cu alloy dissolved and penetrated along the steel grain boundaries, resulting in good metallurgical bonding of the interface.(author)

  19. Cermet sintering on the oase of molybdenum, nickel, aluminium oxide in dry and wet hydrogen medium

    International Nuclear Information System (INIS)

    Fedotov, A.V.; Lutskaya, E.Eh.

    1985-01-01

    Cermet sintering on the base of molybdenum, nickel and aluminium oxide in dry and wer hydrogen medium is studied. It is stated that presence of water vapours permits to decrease sintering temperature of molybdenum containing cermets and to prepare dense nickeliferous cermets. Cermet density can he rather high at final stages of sintering that is probably conditioned by decrease of growth rate of corundum crystals. Pressing pressure activates cermet siptering at intermediate stages and it is low effective at finite stages of condensation. Constancy of relative reduction of void volume is preserved only at final stages of sintering

  20. Development of Separation Process for Minor Actinides Using TDdDGA and New Extractants

    International Nuclear Information System (INIS)

    Matsumura, T.; Tsubata, Y.

    2015-01-01

    Full text of publication follows: Separation process for minor actinides (MA = Am, Cm and Np) has been developed at Japan Atomic Energy Agency using new innovative extractants to improve the partitioning process from the viewpoints of the economy and the reduction of secondary wastes. Phosphorus-free compounds consisting of carbon, hydrogen, oxygen and nitrogen (CHON principle) were applied to the separation steps for MA. At the first step, MA and lanthanide elements (Ln) are recovered from high-level liquid waste by solvent extraction with N,N,N',N'-tetra-dodecyl-diglycolamide (TDdDGA). Trivalent actinides Am and Cm, are separated from RE at the next step by solvent extraction using podand type soft-donor extractant such as N,N,N',N'- tetrakis(pyridin-2-ylmethyl)- decane-1,2-diamine (TPDN) or hybrid type extractant such as N-octyl-N-(ptolyl)- 1,10-phenanthroline-2-carboxamide (OctTolPTA). This paper presents the current status of the research and development programme. This study is carried out under the Innovative Nuclear Research and Development Programme by the Ministry of Education, Culture, Sports, Science and Technology of Japan. (authors)

  1. Burn of actinides in MOX fuel cells

    International Nuclear Information System (INIS)

    Martinez C, E.; Ramirez S, J. R.; Alonso V, G.

    2017-09-01

    The spent fuel from nuclear reactors is stored temporarily in dry repositories in many countries of the world. However, the main problem of spent fuel, which is its high radio-toxicity in the long term, is not solved. A new strategy is required to close the nuclear fuel cycle and for the sustain ability of nuclear power generation, this strategy could be the recycling of plutonium to obtain more energy and recycle the actinides generated during the irradiation of the fuel to transmute them in less radioactive radionuclides. In this work we evaluate the quantities of actinides generated in different fuels and the quantities of actinides that are generated after their recycling in a thermal reactor. First, we make a reference calculation with a regular enriched uranium fuel, and then is changed to a MOX fuel, varying the plutonium concentrations and determining the quantities of actinides generated. Finally, different amounts of actinides are introduced into a new fuel and the amount of actinides generated at the end of the fuel burn is calculated, in order to determine the reduction of minor actinides obtained. The results show that if the concentration of plutonium in the fuel is high, then the production of minor actinides is also high. The calculations were made using the cell code CASMO-4 and the results obtained are shown in section 6 of this work. (Author)

  2. Process for fabrication of cermets

    Science.gov (United States)

    Landingham, Richard L [Livermore, CA

    2011-02-01

    Cermet comprising ceramic and metal components and a molten metal infiltration method and process for fabrication thereof. The light weight cermets having improved porosity, strength, durability, toughness, elasticity fabricated from presintered ceramic powder infiltrated with a molten metal or metal alloy. Alumina titanium cermets biocompatible with the human body suitable for bone and joint replacements.

  3. Flexible waste management to increase the effectiveness of minor actinide PT technology

    Energy Technology Data Exchange (ETDEWEB)

    Fukasawa, T. [Hitachi-GE Nuclear Energy, Ltd., 3-1-1 Saiwai, Hitachi 317-0073 (Japan); Inagaki, Y.; Arima, T. [Kyshu University, 744 Motooka, Nishi, Fukuoka 819-0395 (Japan); Sato, S. [Fukushima National College of Technology, 30 Aza-Nagao, Tairakamiarakawa, Iwaki 970-8034 (Japan)

    2016-07-01

    Partitioning and transmutation (PT) technologies have been developed for minor actinides (MA) to reduce the high level waste (HLW) volume and long-term radiotoxicity. Although the MA PT can reduce the potential radiotoxicity effectively by 1-3 orders of magnitude, the actual operation of PT requires several tens of years for developing elemental technologies of nuclide separation, MA containing fuel fabrication, transmutation and their practical systematization. The high level liquid waste (HLLW) containing MA is presently vitrified immediately after spent fuel reprocessing, stored about 50 years at surface facility and will be disposed of at deep geological repository. Vitrified HLW form works as an excellent artificial barrier against nuclides release during storage and disposal. On the other hand, it is difficult to recover MA from the form. So the present waste management scheme has an issue of MA PT technology application until its deployment, which will produce much amount of vitrified HLW including long-lived MA without PT application. Thus the authors proposed the flexible waste management method to increase the effectiveness of the MA PT. The system adopts the HLLW calcination instead of the vitrification to produce granule for its dry storage of about 50 years until the MA PT technology will be applicable. The granule should be easily dissolved by the nitric acid solution to apply the typical aqueous MA partitioning technologies to be developed. This paper reports the purpose of the study, the feasibility evaluation results for the calcined granule storage and the evaluation results for the environmental burden reduction effect. (authors)

  4. Safe actinide disposition in molten salt reactors

    International Nuclear Information System (INIS)

    Gat, U.

    1997-01-01

    Safe molten salt reactors (MSR) can readily accommodate the burning of all fissile actinides. Only minor compromises associated with plutonium are required. The MSRs can dispose safely of actinides and long lived isotopes to result in safer and simpler waste. Disposing of actinides in MSRs does increase the source term of a safety optimized MSR. It is concluded that the burning and transmutation of actinides in MSRs can be done in a safe manner. Development is needed for the processing to handle and separate the actinides. Calculations are needed to establish the neutron economy and the fuel management. 9 refs

  5. For cermet inert anode containing oxide and metal phases useful for the electrolytic production of metals

    Science.gov (United States)

    Ray, Siba P.; Liu, Xinghua; Weirauch, Douglas A.

    2002-01-01

    A cermet inert anode for the electrolytic production of metals such as aluminum is disclosed. The inert anode comprises a ceramic phase including an oxide of Ni, Fe and M, where M is at least one metal selected from Zn, Co, Al, Li, Cu, Ti, V, Cr, Zr, Nb, Ta, W, Mo, Hf and rare earths, preferably Zn and/or Co. Preferred ceramic compositions comprise Fe.sub.2 O.sub.3, NiO and ZnO or CoO. The cermet inert anode also comprises a metal phase such as Cu, Ag, Pd, Pt, Au, Rh, Ru, Ir and/or Os. A preferred metal phase comprises Cu and Ag. The cermet inert anodes may be used in electrolytic reduction cells for the production of commercial purity aluminum as well as other metals.

  6. Thermodynamics of carbothermic synthesis of actinide mononitrides

    International Nuclear Information System (INIS)

    Ogawa, T.; Shirasu, Y.; Minato, K.; Serizawa, H.

    1997-01-01

    Carbothermic synthesis will be further applied to the fabrication of nitride fuels containing minor actinides (MA) such as neptunium, americium and curium. A thorough understanding of the carbothermic synthesis of UN will be beneficial in the development of the MA-containing fuels. Thermodynamic analysis was carried out for conditions of practical interest in order to better understand the recent fabrication experiences. Two types of solution phases, oxynitride and carbonitride phases, were taken into account. The Pu-N-O ternary isotherm was assessed for the modelling of M(C, N, O). With the understanding of the UN synthesis, the fabrication problems of Am-containing nitrides are discussed. (orig.)

  7. Minor Actinide Recycle in Sodium Cooled Fast Reactors Using Heterogeneous Targets

    International Nuclear Information System (INIS)

    Bays, Samuel; Medvedev, Pavel; Pope, Michael; Ferrer, Rodolfo; Forget, Benoit; Asgari, Mehdi

    2009-01-01

    This paper investigates the plausible design of transmutation target assemblies for minor actinides (MA) in Sodium Fast Reactors (SFR). A heterogeneous recycling strategy is investigated, whereby after each reactor pass, un-burned MAs from the targets are blended with MAs produced by the driver fuel and additional MAs from Spent Nuclear Fuel (SNF). A design iteration methodology was adopted for customizing the core design, target assembly design and matrix composition design. The overall design was constrained against allowable peak or maximum in-core performances. While respecting these criteria, the overall design was adjusted to reduce the total number of assemblies fabricated per refueling cycle. It was found that an inert metal-hydride MA-Zr-Hx target matrix gave the highest transmutation efficiency, thus allowing for the least number of targets to be fabricated per reactor cycle.

  8. Properties of U sub 3 O sub 8 -aluminum cermet fuel

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, H.B.

    1989-10-01

    Nuclear fuel elements containing U{sub 3}O{sub 8} dispersed in an aluminum matrix have been used in research and test reactors for about 30 years. These elements, sometimes called cermet fuel, are made by powder metallurgical methods (PM) and can accommodate up to approximately 50 wt % uranium in the core section of extruded tubes. Cermet fuel elements have been fabricated and irradiated at the Savannah River Site (SRS). Irradiation behavior is excellent. Extruded tubes with up to 50 wt % uranium have been successfully irradiated to fission densities of about 2 {times} 10{sup 21} fissions per cc of core. Physical, mechanical, and chemical properties of cermet fuels are assembled into a reference document. Results will be used by Argonne National Laboratory to design cermet fuel elements for possible use in the New Production Reactor at SRS. 57 refs., 33 figs., 12 tabs.

  9. Basic actinide chemistry and physics research in close cooperation with hot laboratories: ACTILAB

    International Nuclear Information System (INIS)

    Minato, K; Konashi, K; Fujii, T; Uehara, A; Nagasaki, S; Ohtori, N; Tokunaga, Y; Kambe, S

    2010-01-01

    Basic research in actinide chemistry and physics is indispensable to maintain sustainable development of innovative nuclear technology. Actinides, especially minor actinides of americium and curium, need to be handled in special facilities with containment and radiation shields. To promote and facilitate actinide research, close cooperation with the facilities and sharing of technical and scientific information must be very important and effective. A three-year-program B asic actinide chemistry and physics research in close cooperation with hot laboratories , ACTILAB, was started to form the basis of sustainable development of innovative nuclear technology. In this program, research on actinide solid-state physics, solution chemistry and solid-liquid interface chemistry is made using four main facilities in Japan in close cooperation with each other, where basic experiments with transuranium elements can be made. The 17 O-NMR measurements were performed on (Pu 0.91 Am 0.09 )O 2 to study the electronic state and the chemical behaviour of Am and Cm ions in electrolyte solutions was studied by distribution experiments.

  10. Criteria for achieving actinide reduction goals

    International Nuclear Information System (INIS)

    Liljenzin, J.O.

    1996-01-01

    In order to discuss various criteria for achieving actinide reduction goals, the goals for actinide reduction must be defined themselves. In this context the term actinides is interpreted to mean plutonium and the so called ''minor actinides'' neptunium, americium and curium, but also protactinium. Some possible goals and the reasons behind these will be presented. On the basis of the suggested goals it is possible to analyze various types of devices for production of nuclear energy from uranium or thorium, such as thermal or fast reactors and accelerator driven system, with their associated fuel cycles with regard to their ability to reach the actinide reduction goals. The relation between necessary single cycle burn-up values, fuel cycle processing losses and losses to waste will be defined and discussed. Finally, an attempt is made to arrange the possible systems on order of performance with regard to their potential to reduce the actinide inventory and the actinide losses to wastes. (author). 3 refs, 3 figs, 2 tabs

  11. Cermet-fueled reactors for advanced space applications

    International Nuclear Information System (INIS)

    Cowan, C.L.; Palmer, R.S.; Taylor, I.N.; Vaidyanathan, S.; Bhattacharyya, S.K.; Barner, J.O.

    1987-12-01

    Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel were carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper

  12. The compaction and sintering of UO_2-Zr cermet pellets

    International Nuclear Information System (INIS)

    Tri Yulianto; Meniek Rachmawati; Etty Mutiara

    2013-01-01

    An innovative fuel pellet of UO_2-Zr cermet has been developed to improve thermal conductivity of UO_2 pellet by adding small amount Zr metal in to UO_2 matrix below 10 % weight. Zirconium powder will serve for the creation of bridges or web structure during compaction and will effectively reduce contact between of UO_2 particles. Based on the theory of phase equilibrium of metals-metal oxides-ceramic, this fabrication technique may produce UO_2 pellets containing continuous metal channel on the grain boundary of UO_2 through sintering in a reduction atmosphere. The fabrication was done by varying process parameters of mixing and compaction. Characterisation of UO_2-Zr cermet pellet involved visual test, dimensional and density measurement, and ceramography test. This advanced cermet fabrication technology may address common issue with cermet fuels such as microstructure with continuous metal channel structure in the UO_2 matrix, which is more effectively than the commonly accepted microstructure involving fraction of UO_2 pellet by standard fabrication route. (author)

  13. Solar Absorptance of Cermet Coatings Evaluated

    Science.gov (United States)

    Jaworske, Donald A.

    2004-01-01

    Cermet coatings, molecular mixtures of metal and ceramic, are being considered for the heat inlet surface of solar Stirling convertors. In this application, the key role of the cermet coating is to absorb as much of the incident solar energy as possible. To achieve this objective, the cermet coating has a high solar absorptance value. Cermet coatings are manufactured utilizing sputter deposition, and many different metal and ceramic combinations can be created. The ability to mix metal and ceramic at the atomic level offers the opportunity to tailor the composition, and hence, the optical properties of these coatings. The NASA Glenn Research Center has prepared and characterized a wide variety of cermet coatings utilizing different metals deposited in an aluminum oxide ceramic matrix. In addition, the atomic oxygen durability of these coatings has been evaluated.

  14. Transmutation of minor actinide using BWR fueled mixed oxide

    International Nuclear Information System (INIS)

    Susilo, Jati

    2000-01-01

    Nuclear spent fuel recycle has a strategic importance in the aspect of nuclear fuel economy and prevention of its spread-out. One among other application of recycle is to produce mixed oxide fuel (Mo) namely mixed Plutonium and uranium oxide. As for decreasing the burden of nuclear high level waste (HLW) treatment, transmutation of minor actinide (MA) that has very long half life will be carried out by conversion technique in nuclear reactor. The purpose of this study was to know influence of transition fuel cell regarding the percent weight of transmutation MA in the BWR fueled MOX. Calculation of cell BWR was used SRAC computer code, with assume that the reactor in equilibrium. The percent weight of transmutation MA to be optimum by increasing the discharge burn-up of nuclear fuel, raising ratio of moderator to fuel volume (Vm/Vf), and loading MA with percent weight about 3%-6% and also reducing amount of percent weight Pu in MOX fuel. For mixed fuel standard reactor, reactivity value were obtained between about -50pcm ∼ -230pcm for void coefficient and -1.8pcm ∼ -2.6pcm for fuel temperature coefficient

  15. The concept of electro-nuclear facility for useful power generation and minor actinides transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Bergelson, B.R.; Balyuk, S.A. [ITEP, Moscow (Russian Federation)

    1995-10-01

    The possibility is shown to design in principle the double-purpose liquid fuel electro nuclear facility for useful power generation and minor actinides transmutation in U-Pu fuel cycle conditions. D{sub 2}O and a melt of fluorine salts are considered as a working media for liquid fuel. Such facility replenished with depicted or natural uranium only makes it possible to generate power of 900 MW (c) for external consumers and serve 20 WWER-1000 reactors for transmutation of MA. The facility could be thought as an alternative to fast reactors since appr. 30% of the total power confined in uranium is utilized in it.

  16. Structural study and properties of peraluminous formulations for the fission products and minor actinides confinement

    International Nuclear Information System (INIS)

    Gasnier, E.

    2013-01-01

    In this work, peraluminous glasses (lack of alkaline and alkaline earth ions regarding aluminum) are under study to assess the potentiality of these matrices to confine fission products and minor actinides (FPA) at higher rate than current R7T7 glass (18,5 wt % FPA). The first part of this work aims at studying the physical and chemical properties of complex peraluminous glasses containing increasing FPA rate (18.5 to 32 wt %) to compare them with the specifications. The very low crystallization tendency of complex glasses containing up to 22.5 wt % as well as the very good chemical durability observed are major assets. The other part focuses on the lanthanides incorporation in simplified glass compositions in the SiO 2 -B 2 O 3 -Al 2 O 3 -Na 2 O-CaO-Ln 2 O 3 system (Ln = Nd or La). The glass homogeneity and devitrification tendency are investigated at different scales by XRD, SEM, TEM and structural techniques such as NMR (MAS, MQMAS, REDOR, HMQC, DHMQC) and neodymium optical spectroscopy that appear very powerful to determine the lanthanides structural role regarding aluminum and describe more precisely the structural organization of peraluminous network, as still unknown in such systems. The glass homogeneity was demonstrated in a large composition domain and new structural data were put in evidence at high lanthanides content. (author) [fr

  17. Transmutation of minor actinides in a Candu thorium borner

    International Nuclear Information System (INIS)

    Sahin, S.; Sahin, H. M.; Acir, A.; Yalcin, S.; Yildiz, K.; Sahin, N.; Altinok, T.; Alkan, M.

    2007-01-01

    latter is used for denaturized the new 2 33U fuel with 2 38U. The temporal variation of the criticality k ∞ and the burn-up values of the reactor have been calculated by full power operation for a period of 20 years. The criticality starts by k ∞ = ∼ 1.48 for both fuel compositions. A sharp decrease of the criticality has been observed in the first year as a consequence of rapid plutonium burnout. The criticality becomes quasi constant after the 2nd year and remains above k ∞ > 1.06 for ∼ 20 years. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner. Nuclear waste actinides can also be used as a booster fissile fuel material in form of mixed fuel with thorium in a CANDU reactor in order to assure the initial criticality at startup. In the third phase, two different fuel compositions have been found useful to provide sufficient reactor criticality over a long operation period: 1) 95% thoria (ThO 2 ) + 5% minor actinides MAO 2 and 2) 95% ThO 2 + 5% MAO 2 + 5% UO 2 . The latter allows a higher degree of nuclear safeguarding thorough denaturing the new 2 33U fuel with 2 38U. The temporal variation of the criticality k ∞ and the burn-up values of the reactor have been calculated by full power operation for a period of 10 years. The criticality starts by k ∞ > 1.3 for both fuel compositions. A sharp decrease of the criticality has been observed in the first year as a consequence of rapid plutonium burnout in the actinide fuel. The criticality becomes quasi constant after the 2nd year and remains close to k ∞ =∼1.06 for ∼ 10 years. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner. Finally, in the fourth phase, a CANDU reactor fueled with a mixed fuel made of thoria (ThO 2 ) and the totality of nuclear waste actinides has been investigated. The mixed fuel composition has been varied in radial direction to achieve a uniform power distribution and fuel burn up in the fuel bundle. The

  18. Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit

    International Nuclear Information System (INIS)

    Wagner, J.C.

    2001-01-01

    This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses

  19. Examination of minor actinide annihilation by BWR core

    International Nuclear Information System (INIS)

    Hida, Kazuki

    1995-01-01

    From the viewpoint of reducing burden for disposing high level waste generated from spent fuel, the examination of recycling minor actinide (MA) to reactors and reducing its accumulation has been advanced. In this study, the possibility of annihilation in the case of recycling it to a BWR was examined. The main MAs are 237 Np, 241 Am, 243 Am, 242 Cm, and 244 Cm. However, as for Cm isotopes, the half life is short, the amount of generation is small, and the rate of neutron emission is high, therefore, those are disposed as waste, and 237 Np, 241 Am and 243 Am were taken as the objects of recycling. In order to grasp the basic characteristics in the case of recycling MAs to a BWR, MAs were added to UO 2 fuel, MOX fuel and HCR fuel and burned, and the nuclear conversion characteristics were examined. As the result, it was found that they were converted to short half life nuclides, and as the neutron spectra were softer, the rate of annihilation was higher. In the case of recycling MAs by concentrating to a specific reactor, reactivity loss, the degree of uranium enrichment required for compensating reactivity, and the rate of MA annihilation were calculated. Based on these data, the MA recycling system was set up, and the rate of MA annihilation was evaluated. This is reported. (K.I.)

  20. Actinides transmutation - a comparison of results for PWR benchmark

    International Nuclear Information System (INIS)

    Claro, Luiz H.

    2009-01-01

    The physical aspects involved in the Partitioning and Transmutation (P and T) of minor actinides (MA) and fission products (FP) generated by reactors PWR are of great interest in the nuclear industry. Besides these the reduction in the storage of radioactive wastes are related with the acceptability of the nuclear electric power. From the several concepts for partitioning and transmutation suggested in literature, one of them involves PWR reactors to burn the fuel containing plutonium and minor actinides reprocessed of UO 2 used in previous stages. In this work are presented the results of the calculations of a benchmark in P and T carried with WIMSD5B program using its new cross sections library generated from the ENDF-B-VII and the comparison with the results published in literature by other calculations. For comparison, was used the benchmark transmutation concept based in a typical PWR cell and the analyzed results were the k∞ and the atomic density of the isotopes Np-239, Pu-241, Pu-242 and Am-242m, as function of burnup considering discharge of 50 GWd/tHM. (author)

  1. Cryogenic gamma detectors enable direct detection of 236U and minor actinides for non-destructive assay

    International Nuclear Information System (INIS)

    Miguel Velazquez; Jonathan Dreyer; Drury, O.B.; Friedrich, Stephan; Saleem Salaymeh

    2016-01-01

    We demonstrate the utility of a superconducting transition edge sensor (TES) γ-ray detector with high energy resolution and low Compton background for nondestructive assay (NDA) of a uranium sample from reprocessed nuclear fuel. We show that TES γ-detectors can separate low-energy actinide γ-emissions from the background and nearby lines, even from minor isotopes whose signals are often obscured in NDA with conventional Ge detectors. Superconducting γ-detectors may therefore bridge the gap between high-accuracy destructive assay (DA) and easier-to-use NDA. (author)

  2. Recent progress on R and D of innovative extractants and adsorbents for partitioning of minor actinides at JAEA

    International Nuclear Information System (INIS)

    Kimura, Takaumi; Morita, Yasuji; Koma, Yoshikazu

    2010-01-01

    The R and D effort on partitioning of minor actinides (MA) at the Japan Atomic Energy Agency (JAEA) has been concentrated on development and improvement of innovative extractants and adsorbents as the fundamental studies and of MA recovery process as the advanced aqueous reprocessing system in fast reactor cycle technology development (FaCT) project. This paper reviews current status and prospects of the R and D activities on the partitioning of MA at JAEA. (authors)

  3. Transmutation of actinides in power reactors.

    Science.gov (United States)

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

  4. Actinide recycle potential in the integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1993-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. In the IFR pyroprocessing, minor actinides accompany plutonium product stream, and therefore, actinide recycle occurs naturally. The fast neutron spectrum of the IFR makes it an ideal actinide burner, as well. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and potential implications on long-term waste management

  5. Spent fuel reprocessing and minor actinide partitioning safety related research at the UK National Nuclear Laboratory

    International Nuclear Information System (INIS)

    Carrott, Michael; Flint, Lauren; Gregson, Colin; Griffiths, Tamara; Hodgson, Zara; Maher, Chris; Mason, Chris; McLachlan, Fiona; Orr, Robin; Reilly, Stacey; Rhodes, Chris; Sarsfield, Mark; Sims, Howard; Shepherd, Daniel; Taylor, Robin; Webb, Kevin; Woodall, Sean; Woodhead, David

    2015-01-01

    The development of advanced separation processes for spent nuclear fuel reprocessing and minor actinide recycling is an essential component of international R and D programmes aimed at closing the nuclear fuel cycle around the middle of this century. While both aqueous and pyrochemical processes are under consideration internationally, neither option will gain broad acceptance without significant advances in process safety, waste minimisation, environmental impact and proliferation resistance; at least when compared to current reprocessing technologies. The UK National Nuclear Laboratory (NNL) is developing flowsheets for innovative aqueous separation processes. These include advanced PUREX options (i.e. processes using tributyl phosphate as the extractant for uranium, plutonium and possibly neptunium recovery) and GANEX (grouped actinide extraction) type processes that use diglycolamide based extractants to co-extract all transuranic actinides. At NNL, development of the flowsheets is closely linked to research on process safety, since this is essential for assessing prospects for future industrialisation and deployment. Within this context, NNL is part of European 7. Framework projects 'ASGARD' and 'SACSESS'. Key topics under investigation include: hydrogen generation from aqueous and solvent phases; decomposition of aqueous phase ligands used in separations prior to product finishing and recycle of nitric acid; dissolution of carbide fuels including management of organics generated. Additionally, there is a strong focus on use of predictive process modelling to assess flowsheet sensitivities as well as engineering design and global hazard assessment of these new processes. (authors)

  6. Neutron-Induced Fission Cross Section of Uranium, Americium and Curium Isotopes. Progress report - Research Contract 14485, Coordinated Research Project on Minor Actinide Neutron Reaction Data (MANREAD)

    International Nuclear Information System (INIS)

    Alekseev, A.A.; Bergman, A.A.; Berlev, A.I.; Koptelov, E.A.; Samylin, B.F.; Trufanov, A.M.; Fursov, B.I.; Shorin, V.S.

    2009-12-01

    This report contains brief description of the Lead Slowing Down Spectrometer and results of measurements of neutron-induced fission cross sections for 236 U, 242m Am, 243 Cm, 244 Cm, 245 Cm and 246 Cm done at this spectrometer. The work was partially supported through the IAEA research contract RC-14485-RD in the framework of the IAEA Coordinated Research Project 'Minor Actinide Neutron Reaction Data (MANREAD)'. The detailed description of the experimental set up, measurements procedure and data treatment can be found in the JIA-1182 (2007) and JIA-1212 (2009) reports from the Institute of Nuclear Research of the Russian Academy of Science published in Russian. Part 1 contains the first year report of the research contract and part 2 the second year report. (author)

  7. Research for actinides extractants from various wastes

    International Nuclear Information System (INIS)

    Musikas, C.; Cuillerdier, C.; Condamines, N.

    1990-01-01

    This paper is an overview of the actinides solvent extraction research undertaken in Fontenay-aux-Roses. Two kinds of extractants are investigated; those usable for the improvement of the nowadays nuclear fuels reprocessing and those necessary for advanced fuels cycles which include the minor actinides (Np, Am) recovery for a further elimination through nuclear reactions. In the first class the mono and diamides, alternative to the organophosphorus extractants, TBP and polyfunctional phosphonates, showed promising properties. The main results are discussed. For the future efficient extractants for trivalent actinides-lanthanides group separations are suitable. The point about the actinides (III) - lanthanides (III) group separation chemistry and the development of some of these extractants are given

  8. Study on remain actinides recovery in pyro reprocessing

    International Nuclear Information System (INIS)

    Suharto, Bambang

    1996-01-01

    The spent fuel reprocessing by dry process called pyro reprocessing have been studied. Most of U, Pu and MA (minor actinides) from the spent fuel will be recovered and be fed back to the reactor as new fuel. Accumulation of remain actinides will be separated by extraction process with liquid cadmium solvent. The research was conducted by computer simulation to calculate the stage number required. The calculation's results showed on the 20 stages extractor more than 99% actinides can be separated. (author)

  9. Measurements of the neutron capture cross sections and incineration potentials of minor-actinides in high thermal neutron fluxes: Impact on the transmutation of nuclear wastes

    International Nuclear Information System (INIS)

    Bringer, O.

    2007-10-01

    This thesis comes within the framework of minor-actinide nuclear transmutation studies. First of all, we have evaluated the impact of minor actinide nuclear data uncertainties within the cases of 241 Am and 237 Np incineration in three different reactor spectra: EFR (fast), GT-MHR (epithermal) and HI-HWR (thermal). The nuclear parameters which give the highest uncertainties were thus highlighted. As a result of fact, we have tried to reduce data uncertainties, in the thermal energy region, for one part of them through experimental campaigns in the moderated high intensity neutron fluxes of ILL reactor (Grenoble). These measurements were focused onto the incineration and transmutation of the americium-241, the curium-244 and the californium-249 isotopes. Finally, the values of 12 different cross sections and the 241 Am isomeric branching ratio were precisely measured at thermal energy point. (author)

  10. The effect of Co particle structures on the mechanical properties and microstructure of TiCN-based cermets

    International Nuclear Information System (INIS)

    Deng, Y.; Jiang, X.Q.; Zhang, Y.H.; Chen, H.; Tu, M.J.; Deng, L.; Zou, J.P.

    2016-01-01

    Ti(C,N) based cermets are composite materials composed of a hard phase and a binder phase structure. Cubic-structured Co particles are the best choice for the binder phase of Ti(C,N) based cermets due to their excellent toughness performance. However, the application of β-Co particles in cermets has not been reported in the literature so far. In this pioneer study, ultrafine Ti(C,N) based cermet samples were prepared by separately using Co particles of different structures as the binder phase, and the effect of the Co particle structures on the mechanical properties and microstructure of the cermets were studied: First, the Empirical Electron Theory was used to calculate the difference in the interface density (∆ρ) for different crystals, and the interface combined strength between the hard phase of different structures containing Co particles were evaluated. Second, we systematically investigated the evolution of the microstructures of the two cermets during the sintering process, and evaluated the characteristics of the microstructure (which determines the properties of the cermets). Finally, the mechanical properties of the samples were tested, and the performances of the Co structures were evaluated. The results show that β-Co particles can optimize the cermet microstructure, which leads to excellent mechanical performance.

  11. The effect of Co particle structures on the mechanical properties and microstructure of TiCN-based cermets

    Energy Technology Data Exchange (ETDEWEB)

    Deng, Y. [Chongqing University of Arts and Science, Chongqing 402160 (China); State Key Laboratory of Powder Metallurgy, Central South University, Changsha 410083 (China); Jiang, X.Q. [Southwest University, Chongqing Academy Science and Technology, Chongqing 4100715 (China); Zhang, Y.H.; Chen, H.; Tu, M.J. [Chongqing University of Arts and Science, Chongqing 402160 (China); Deng, L., E-mail: dengying.163@163.com [Chengdu Chengliang Tool Group Co., Ltd., Chengdu 610056 (China); Zou, J.P., E-mail: 1042551842@qq.com [State Key Laboratory of Powder Metallurgy, Central South University, Changsha 410083 (China)

    2016-10-15

    Ti(C,N) based cermets are composite materials composed of a hard phase and a binder phase structure. Cubic-structured Co particles are the best choice for the binder phase of Ti(C,N) based cermets due to their excellent toughness performance. However, the application of β-Co particles in cermets has not been reported in the literature so far. In this pioneer study, ultrafine Ti(C,N) based cermet samples were prepared by separately using Co particles of different structures as the binder phase, and the effect of the Co particle structures on the mechanical properties and microstructure of the cermets were studied: First, the Empirical Electron Theory was used to calculate the difference in the interface density (∆ρ) for different crystals, and the interface combined strength between the hard phase of different structures containing Co particles were evaluated. Second, we systematically investigated the evolution of the microstructures of the two cermets during the sintering process, and evaluated the characteristics of the microstructure (which determines the properties of the cermets). Finally, the mechanical properties of the samples were tested, and the performances of the Co structures were evaluated. The results show that β-Co particles can optimize the cermet microstructure, which leads to excellent mechanical performance.

  12. Cermets based on rhenium and rare earth element oxides

    International Nuclear Information System (INIS)

    Varfolomeev, M.B.; Velichko, A.V.; Zajtseva, L.L.; Shishkov, N.V.

    1977-01-01

    The reduction of perrhenates of rare earth elements and of yttrium by hydrogen and the subsequent sintering have yielded cermets based on rhenium and rare earth element oxides inherent in which are more disperse and homogeneous structures than those of the ''molecular'' rare earth element-Tc cermets. The dispersity of cermets increases in the rare earth elements series from La to Lu. The microhardness of the Re phase in cermets is 490 kgf/mm 2 ; the total microhardness of a cermet is substantially higher

  13. EC-FP7 ARCAS: technical and economical comparison of Fast Reactors and Accelerator Driven Systems for transmutation of Minor Actinides

    International Nuclear Information System (INIS)

    Van den Eynde, G.; Romanello, V.; Heek, A. van; Martin-Fuertes, F.; Zimmerman, C.; Lewin, B.

    2015-01-01

    The ARCAS project aims to compare, on a technological and economical basis, Accelerator Driven Systems and Fast Reactors as Minor Actinide burners. It is split in five work packages: the reference scenario definition, the fast reactor system definition, the accelerator driven system definition, the fuel reprocessing and fabrication facilities definition and the economical comparison. This paper summarizes the status of the project and its five work packages. (author)

  14. Properties of Fission-Product decay heat from Minor-Actinide fissioning systems

    International Nuclear Information System (INIS)

    Oyamatsu, Kazuhiro; Mori, Hideki

    2000-01-01

    The aggregate Fission-Product (FP) decay heat after a pulse fission is examined for Minor Actinide (MA) fissiles 237 Np, 241 Am, 243 Am, 242 Cm and 244 Cm. We find that the MA decay heat is comparable but smaller than that of 235 U except for cooling times at about 10 8 s (approx. = 3 y). At these cooling times, either the β or γ component of the FP decay heat for these MA's is substantially larger than the one for 235 U. This difference is found to originate from the cumulative fission yield of 106 Ru (T 1/2 = 3.2x10 7 s). This nuclide is the parent of 106 Rh (T 1/2 = 29.8 s) which is the dominant source of the decay heat at 10 8 s (approx. = 3 y). The fission yield is nearly an increasing function of the fissile mass number so that the FP decay heat is the largest for 244 Cm among the MA's at the cooling time. (author)

  15. Study on neutron spectrum for effective transmutation of minor actinides in thermal reactors

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Yokoyama, Kenji

    1997-01-01

    The transmutation of minor actinides (MAs) has been investigated in thermal reactor cells using mixed oxide fuel with MAs. The effect of neutron spectra on transmutation is studied by changing the neutron spectra. Five transmutation rates are compared: direct fission incineration rate, capture transmutation rate, consumption rate, overall fission incineration rate and inventory difference transmutation rate. The relations between these transmutation rates and their dependence on the neutron spectrum were investigated. To effectively incinerate MAs it is necessary to maximize the overall fission incineration rate and the inventory difference transmutation rate. These transmutation rates become maximum when the fraction of neutrons below 1 eV is about 8% for the case where the MA addition is 1-3%. When the MA addition is over 5%, the transmutation rates become maximum for very hard neutron spectrum. (Author)

  16. Synthesis and characterization of novel lanthanide- and actinide-containing titanates and zircono-titanates; relevance to nuclear waste disposal

    International Nuclear Information System (INIS)

    Shoup, S.L.S.

    1995-08-01

    Before experiments using actinide elements are performed, synthetic routes are tested using lanthanides of comparable ionic radii as surrogates. Compound and solid solution formation in several lanthanide-containing titanate and zircono-titanate systems have been established using X-ray diffraction (XRD) analysis, which helped to define interesting and novel experiments, some of which have been performed and are discussed, for selected actinide elements. The aqueous solubilities of several lanthanide- and actinide-containing compounds, representative of the systems studied, were tested in several leachants, including the WIPP open-quotes Aclose quotes brine, following modified Materials Characterization Center procedures (MCC-3). The WIPP open-quotes Aclose quotes brine is a synthetic substitute for that found in nature at the Waste Isolation Pilot Plant (WIPP) in New Mexico. The concentrations of cerium, used as a surrogate for plutonium, leached by the WIPP open-quotes Aclose quotes brine from all the cerium-containing compounds and solid solutions tested were below the Inductively Coupled Plasma (ICP) atomic emission spectrometry limit of detection (10 ppm) established for cerium in this brine. The concentrations of plutonium leached from the two plutonium-containing solid solutions were less than 1 ppm as determined by gross alpha counting and alpha pulse height analysis. Concentrations of strontium leached by the WIPP brine from stable strontium containing titanate compounds, studied as possible immobilizers of both 90 Sr and actinide elements, were also quite low. These compound and solid solution formation investigations and the aqueous solubility studies suggest that the types of titanate and zircono-titanate compounds and solid solutions studied in this work appear to be useful as host matrices for nuclear waste immobilization

  17. Enhancing charge transfer kinetics by nanoscale catalytic cermet interlayer.

    Science.gov (United States)

    An, Jihwan; Kim, Young-Beom; Gür, Turgut M; Prinz, Fritz B

    2012-12-01

    Enhancing the density of catalytic sites is crucial for improving the performance of energy conversion devices. This work demonstrates the kinetic role of 2 nm thin YSZ/Pt cermet layers on enhancing the oxygen reduction kinetics for low temperature solid oxide fuel cells. Cermet layers were deposited between the porous Pt cathode and the dense YSZ electrolyte wafer using atomic layer deposition (ALD). Not only the catalytic role of the cermet layer itself but the mixing effect in the cermet was explored. For cells with unmixed and fully mixed cermet interlayers, the maximum power density was enhanced by a factor of 1.5 and 1.8 at 400 °C, and by 2.3 and 2.7 at 450 °C, respectively, when compared to control cells with no cermet interlayer. The observed enhancement in cell performance is believed to be due to the increased triple phase boundary (TPB) density in the cermet interlayer. We also believe that the sustained kinetics for the fully mixed cermet layer sample stems from better thermal stability of Pt islands separated by the ALD YSZ matrix, which helped to maintain the high-density TPBs even at elevated temperature.

  18. Cermet coatings for magnetic fusion reactors

    International Nuclear Information System (INIS)

    Smith, M.F.; Whitley, J.B.; McDonald, J.M.

    1984-01-01

    Cermet coatings consisting of SiC particles in an aluminum matrix were produced by a low pressure chamber plasma spray process. Properties of these coatings are being investigated to evaluate their suitability for use in the next generation of magnetic confinement fusion reactors. Although this preliminary study has focused primarily upon SiC-Al cermets, the deposition process can be adapted to other ceramic-metal combinations. Potential applications for cermet coatings in magnetic fusion devices are presented along with experimental results from thermal tests of candidate coatings. (Auth.)

  19. Actinide recycle

    Energy Technology Data Exchange (ETDEWEB)

    Till, C; Chang, Y [Argonne National Laboratory, Argonne, IL (United States)

    1990-07-01

    A multitude of studies and assessments of actinide partitioning and transmutation were carried out in the late 1970s and early 1980s. Probably the most comprehensive of these was a study coordinated by Oak Ridge National Laboratory. The conclusions of this study were that only rather weak economic and safety incentives existed for partitioning and transmuting the actinides for waste management purposes, due to the facts that (1) partitioning processes were complicated and expensive, and (2) the geologic repository was assumed to contain actinides for hundreds of thousands of years. Much has changed in the few years since then. A variety of developments now combine to warrant a renewed assessment of the actinide recycle. First of all, it has become increasingly difficult to provide to all parties the necessary assurance that the repository will contain essentially all radioactive materials until they have decayed. Assurance can almost certainly be provided to regulatory agencies by sound technical arguments, but it is difficult to convince the general public that the behavior of wastes stored in the ground can be modeled and predicted for even a few thousand years. From this point of view alone there would seem to be a clear benefit in reducing the long-term toxicity of the high-level wastes placed in the repository.

  20. Actinide recycle

    International Nuclear Information System (INIS)

    Till, C.; Chang, Y.

    1990-01-01

    A multitude of studies and assessments of actinide partitioning and transmutation were carried out in the late 1970s and early 1980s. Probably the most comprehensive of these was a study coordinated by Oak Ridge National Laboratory. The conclusions of this study were that only rather weak economic and safety incentives existed for partitioning and transmuting the actinides for waste management purposes, due to the facts that (1) partitioning processes were complicated and expensive, and (2) the geologic repository was assumed to contain actinides for hundreds of thousands of years. Much has changed in the few years since then. A variety of developments now combine to warrant a renewed assessment of the actinide recycle. First of all, it has become increasingly difficult to provide to all parties the necessary assurance that the repository will contain essentially all radioactive materials until they have decayed. Assurance can almost certainly be provided to regulatory agencies by sound technical arguments, but it is difficult to convince the general public that the behavior of wastes stored in the ground can be modeled and predicted for even a few thousand years. From this point of view alone there would seem to be a clear benefit in reducing the long-term toxicity of the high-level wastes placed in the repository

  1. PRODUCTION OF ACTINIDE METAL

    Science.gov (United States)

    Knighton, J.B.

    1963-11-01

    A process of reducing actinide oxide to the metal with magnesium-zinc alloy in a flux of 5 mole% of magnesium fluoride and 95 mole% of magnesium chloride plus lithium, sodium, potassium, calcium, strontium, or barium chloride is presented. The flux contains at least 14 mole% of magnesium cation at 600-- 900 deg C in air. The formed magnesium-zinc-actinide alloy is separated from the magnesium-oxide-containing flux. (AEC)

  2. IAEA Activities on Assessment of Partitioning Processes for Transmutation of Actinides

    International Nuclear Information System (INIS)

    Basak, Uddharan; Dyck, Gary R.

    2010-01-01

    In these days of nuclear renaissance, appropriate management of radioactive materials arising from the nuclear fuel cycle back end is one of the most important issues related to the long term sustainability of nuclear energy. The present practice in the back end of the closed fuel cycle involves the recovery of uranium and plutonium from spent fuel by the aqueous based PUREX process for reuse in reactors and the conditioning of reprocessing waste into a form suitable for long term storage. The waste contains mainly fission products and transuranium elements immobilized in glass matrix. However, advanced fuel cycles incorporating partitioning of actinides along with minor actinides and their subsequent transmutation (P and T) in a fast neutron energy spectrum could be proliferation resistant and at the same time reduce the waste radiotoxicity by many orders of magnitude. Considering the importance of P and T on long term sustainability, the International Atomic Energy Agency has initiated many collaborative research programs in this area as part of our advanced fuel cycle activities. This paper presents the current and future activities on advanced partitioning methods, highlighting the challenges associated with these processes, fuel manufacturing techniques suitable for integration with reprocessing facility and the IAEA's minor actinide data base (MADB), as a part of integrated nuclear fuel cycle information system (iNFCIS). (authors)

  3. Fissile fuel breeding and minor actinide transmutation in the life engine

    International Nuclear Information System (INIS)

    Sahin, Suemer; Khan, Mohammad Javed; Ahmed, Rizwan

    2011-01-01

    zone (50 cm), containing MA as fissionable fuel. A 3rd ODS layer (2 cm) separates the molten salt zone on the right side from the graphite reflector (30 cm). Calculations have been conducted for a fusion driver power of 500 MW th in S 8 -P 3 approximation using 238-neutron groups. Minor actinides (MA) out of the nuclear waste of LWRs are used as fissile carbide fuel in TRISO particles with volume fractions of 0, 2, 3, 4 and 5% have been dispersed homogenously in the Flibe coolant. For these cases, tritium breeding at startup is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. In the course of plant operation, TBR and fissile neutron multiplication factor decrease gradually. For a self-sustained reactor, TBR > 1.05 can be kept for all cases over 8 years. Higher fissionable fuel content in the molten salt leads also to higher blanket energy multiplication, namely M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5% TRISO volume fraction at start up, respectively. For all investigated cases, fissile burn up exceeds 400 000 MW D/MT. Major damage mechanisms have been calculated as DPA = 50 and He = 176 appm per year. This implies a replacement of the first wall every 3 years.

  4. General review on cermets

    International Nuclear Information System (INIS)

    Nouvel, G.

    1975-01-01

    After an attempt to classify the basic constituents of cermets in function of their resistivity and microhardness, the history of ceramic-metal composites is reviewed. The two main methods of production, sintering and impregnation, are examined from the point of view of the problems posed and solutions required to obtain optimum mechanical properties (resistance to creep and thermal shock...). A summary table of the chief cermets and their applications is proposed with precise details in some particular cases [fr

  5. Scenarios for minor actinides transmutation in the framework of the French Act on Waste Management

    International Nuclear Information System (INIS)

    Coquelet-Pascal, C.; Meyer, M.; Tiphine, M.; Girieud, R.; Eschbach, R.; Chabert, C.; Garzenne, C.; Barbrault, P.; Van Den Durpel, L.; Caron-Charles, M.; Favet, D.; Arslan, M.; Caron-Charles, M.; Carlier, B.; Lefevre, J.C.

    2013-01-01

    In the framework of the French Act on Waste Management, options of minor actinides (MA) transmutation are studied, based on several scenarios of sodium fast reactor deployment. Basically, one of these scenarios considers the deployment of a 60 GWe SFR fleet in two steps (20 GWe from 2040 to 2050 and 40 GWe, as well as, from 2080 to 2100). For this scenario, the advantages and drawbacks of different transmutation options are evaluated: - transmutation of all minor actinides or only of americium; - transmutation in homogeneous mode (MA bearing fuel in all the core or just in the outer core) or in heterogeneous mode (MA bearing radial blankets). Scenarios have been optimised to limit the impacts of MA transmutation on the cycle: - reduction of the initial MA content in the core in the case of transmutation in homogeneous mode to reduce the impact on reactivity coefficients; - reduction of the number of rows of blankets and fuel decay heat in the case of transmutation in heterogeneous mode. The sensitivity of transmutation options to cycle parameters such as the fuel cooling time before transportation is also assessed. Thus, the transmutation of only americium in one row of radial blankets containing initially 10 pc % Am and irradiated during the same duration as the standard fuel assemblies appears to be a suitable solution to limit the transmutation impacts on fuel cycle and facilities. A comparison of results obtained with MA transmutation in dedicated systems is also presented with a symbiotic scenario considering ADS (accelerator-driven system) deployment to transmute MA together with a SFR fleet to produce energy. The MA inventory within the cycle is higher in the case of transmutation in ADS than in the case of transmutation in SFR. Considering the industrial feasibility of MA transmutation, it appears important to study 'independently' SFR deployment and MA transmutation. Consequently, scenarios of progressive introduction of MA options are assessed

  6. Stochastic Computer Simulation of Cermet Coatings Formation

    OpenAIRE

    Solonenko, Oleg P.; Jordan, Vladimir I.; Blednov, Vitaly A.

    2015-01-01

    An approach to the modeling of the process of the formation of thermal coatings lamellar structure, including plasma coatings, at the spraying of cermet powders is proposed. The approach based on the theoretical fundamentals developed which could be used for rapid and sufficiently accurate prediction of thickness and diameter of cermet splats as well as temperature at interface “flattening quasi-liquid cermet particle-substrate” depending on the key physical parameters (KPPs): temperature, ve...

  7. Partitioning of minor actinides: research at Juelich and Karlsruhe Research Centres

    International Nuclear Information System (INIS)

    Geist, A.; Weigl, M.; Gompper, K.; Modolo, G.

    2007-01-01

    Full text of publication follows. The work on minor actinide (MA) partitioning carried out at Karlsruhe and Juelich is integrated in the EC FP6 programme, EUROPART. Studies include the DIAMEX process (co-extraction of MA and lanthanides from PUREX raffinate) and the SANEX process (separation of MA from lanthanides). Aspects ranging from developing and improving highly selective and efficient extraction reagents, to fundamental structural studies, to process development and testing are covered. SANEX is a challenge in separation chemistry because of the chemical similarity of trivalent actinides and lanthanides. The extracting agents 2,6-di(5,6-di-propyl-1,2,4-triazine-3-yl)pyridine (n-Pr-BTP), developed at Karlsruhe, and the synergetic mixture of di(chloro-phenyl)di-thio-phosphinic acid (R2PSSH) with tri-n-octyl-phosphine oxide (TOPO), developed at Juelich, are considered a breakthrough because of their high separation efficiency in acidic systems. Separation factors for americium over lanthanides of more than 30 (R2PSSH+TOPO) and 130 (n-Pr-BTP) are achieved. To gain understanding of these selectivities, comparative investigations on the structures of curium and europium complexed with these SANEX ligands were performed at Karlsruhe. Extended X-ray absorption fine structure (EXAFS) analysis revealed distinct structural differences between curium and europium complexed with R2PSSH + TOPO, though no such differences were found for n-Pr-BTP. These investigations were therefore complemented by time-resolved laser fluorescence spectroscopic investigations (TRLFS), showing complex stabilities and speciation to differ between n-Pr-BTP complexes of curium and europium. Kinetics of mass transfer was studied for both R2PSSH+TOPO and n-Pr-BTP systems. For the R2PSSH + TOPO system, diffusion was identified to control extraction rates. For the n-Pr-BTP system, a slow chemical reaction was identified as the rate-controlling process. These results were implemented into computer

  8. Relationship between Magnetic and Mechanical Properties of Cermet Tools

    International Nuclear Information System (INIS)

    Ahn, Dong Gil; Lee, Jeong Hee

    2000-01-01

    The commercial cermet cutting tools consist of multi-carbide and a binder metal of iron group, such as cobalt and nickel which are ferromagnetic. In this paper, a new approach to evaluate the mechanical properties of TiCN based cermet by magnetic properties were studied in relation to binder content and sintering conditions. The experimental cermet was prepared using commercial composition with the other binder contents by PM process. It was found that the magnetic properties of the sintered cermets remarkably depended on the microstructure and the total carbon content. The magnetic saturation was proportional to increment of coercive force. At high carbon content in sintered cermet, the magnetic saturation was increased by decreasing the concentration of solutes such as W, Mo, Ti in Co-Ni binder. As the coercive force increases, the hardness usually increases. The strength and toughness of the cermet also increased with increasing the magnetic saturation. The measurement of magnetic properties made it possible to evaluate the mechanical properties in the cermet cutting tools

  9. Heterogeneous all actinide recycling in LWR all actinide cycle closure concept

    International Nuclear Information System (INIS)

    Tondinelli, Luciano

    1980-01-01

    A project for the elimination of transuranium elements (Waste Actinides, WA) by neutron transmutation is developed in a commercial BWR with U-Pu (Fuel Actinides, FA) recycle. The project is based on the All Actinide Cycle Closure concept: 1) closure of the 'back end' of the fuel cycle, U-Pu coprocessing, 2) waste actinide disposal by neutron transmutation. The reactor core consists of Pu-island fuel assemblies containing WAs in target pins. Two parallel reprocessing lines for FAs and WAs are provided. Mass balance, hazard measure, spontaneous activity during 10 recycles are calculated. Conclusions are: the reduction in All Actinide inventory achieved by Heterogeneous All Actinide Recycling is on the order of 83% after 10 recycles. The U235 enrichment needed for a constant end of cycle reactivity decreases for increasing number of recycles after the 4th recycle. A diffusion-burnup calculation of the pin power peak factors in the fuel assembly shows that design limits can be satisfied. A strong effort should be devoted to the solution of the problems related to high values of spontaneous emission by the target pins

  10. Solid solution lithium alloy cermet anodes

    Science.gov (United States)

    Richardson, Thomas J.

    2013-07-09

    A metal-ceramic composite ("cermet") has been produced by a chemical reaction between a lithium compound and another metal. The cermet has advantageous physical properties, high surface area relative to lithium metal or its alloys, and is easily formed into a desired shape. An example is the formation of a lithium-magnesium nitride cermet by reaction of lithium nitride with magnesium. The reaction results in magnesium nitride grains coated with a layer of lithium. The nitride is inert when used in a battery. It supports the metal in a high surface area form, while stabilizing the electrode with respect to dendrite formation. By using an excess of magnesium metal in the reaction process, a cermet of magnesium nitride is produced, coated with a lithium-magnesium alloy of any desired composition. This alloy inhibits dendrite formation by causing lithium deposited on its surface to diffuse under a chemical potential into the bulk of the alloy.

  11. Optimization of SFR Reactor design with recycling or minor actinides; Optimizacion del diseno de reactor SFR con reciclado de actinidos minoritarios

    Energy Technology Data Exchange (ETDEWEB)

    Martin-Fuertes, F.; Vazquez, M.; Alvarez, F.

    2012-07-01

    In this paper we show results of the design features and ESFR optimized in three configurations: the reference, load the minority actinides homogeneous throughout the reactor and the high content of AM on a radial mantle. Was calculated reactivity evolution in five cycles burned (2050 days) to recharge One approach. To do this, we have employed EVOLCODE2 a development tool of CIEMAT own coupling MCNPX and ORIGEN.

  12. Evaluation of the possibility of plutonium and minor actinides transmutation in HWR

    International Nuclear Information System (INIS)

    Ghitescu, P.; Ghizdeanu, N. B.

    2008-01-01

    Partitioning and Transmutation (P and T) techniques could contribute to reduce the radioactive inventory and its associated radio-toxicity. Until now, for this purpose were studied ADS and/or FBR, but not HWR. There are several developed computer codes that analyze the inventory of the radio-nuclides in spent fuel before and after transmutation. WIMSD code is a deterministic lattice spectrum code, which can analyze the reactor neutronic behaviour It also has the capacity to generate burn up and can calculate the inventory of the radio-nuclides of the spent fuel. The advantage of WIMSD code is the variety of the created geometries, together with the big amount of calculated information (K-infinite, macroscopic cross-sections, burnable material radioactive inventory etc). Starting from WIMSD code, the paper presents a model, which simulates the possibility of fuel transmutation in PHWRs. First step was to propose a model, which simulates a CANDU reactor lattice and calculate the radionuclides inventory in an irradiated CANDU fuel bundle. The results were compared with the existing experimental data from CANDU reactors and the calculated parameters were found to be in good agreement with them. After the validation, several simulations were made for PHWRs in order to establish the optimal parameters, related to the efficiency of the transmutation process. Therefore, the code was used for a new type of fuel, containing Plutonium and minor actinides that could be transmuted. The new radioactive inventories were calculated. The simulations showed that Pu content decreases up to 8% in a CANDU reactor and 25% in an ACR. Thus, ACR can reduce the Plutonium inventory from MOX fuel and could be a transmutation solution. (authors)

  13. Corrosion Resistance of Murataite-Based Ceramics Containing Simulated Actinide/Rare Earth Fraction of High Level Waste

    International Nuclear Information System (INIS)

    Stefanovsky, S.V.; Varlakova, G.A.; Burlaka, O.A.; Stefanovsky, O.I.; Nikonov, B.S.; Yudintsev, S.V.

    2009-01-01

    Two samples of murataite-based ceramics containing simulated Actinide/Rare Earth (An/RE) fraction of high level waste (HLW) produced by a cold crucible inductive melting (CCIM) were tested using a single-pass-flow-through (SPFT) procedure. As-prepared and leached samples were examined by X-ray diffraction (XRD) and scanning electron microscopy with energy dispersive system (SEM/EDS). The as-prepared ceramics were composed of murataite, perovskite and crichtonite as well as minor zirconolite and rutile (in one sample). Elemental concentrations at pH=2 and T=90 deg. C were measured and leach rates were calculated. Perovskite concentrating Ca and Ce-group REs (La, Ce, Pr, Nd) was found to be the lowest durable phase. Leach rates of Ca and Ce-group REs (Ce, Nd) from the sample with higher perovskite content were found to be higher than those of U and Zr by one to three orders of magnitude. Elemental leach rates from the ceramic with lower perovskite content are lower by up to 10 times. (authors)

  14. Effect of graphite content on magnetic and mechanical properties of TiC-TiN-Mo-Ni cermets

    Science.gov (United States)

    Zhang, Man; Yang, Qingqing; Xiong, Weihao; Huang, Bin; Ruan, Linji; Mao, Qiao; Li, Shengtao

    2018-04-01

    TiC-10TiN-6Mo-xGr-yNi (mol%, Gr represents graphite, x = 0, 2, 4, 6, 8, and y = 15, 30) cermets were prepared by powder metallurgy method, in order to inverstigate the effect of Gr content on magnetic and mechanical properties of TiC-TiN-Mo-Ni cermets. Room-temperature (RT) saturation magnetization (Ms) and remanence (Mr) of cermets increased with increasing x. This was mainly attributed to that the total content of non-ferromagnetic carbonitride-forming elements Ti and Mo in Ni-based binder phase decreased with increasing x. At the same x, cermets for y = 15 had lower RT Ms and Mr than those for y = 30. Cermets containing more than 2 mol% Gr became ferromagnetic at RT. Bending strength of cermets first increased and then decreased with increasing x. It reached the maximum at x = 2, mainly due to high total content of solutes Ti and Mo in Ni-based binder phase, and moderate thickness of outer rim of Ti(C,N) ceramic grains. Hardness of cermets was not significantly affected by x, mainly due to the combined action of the decrease of the total content of Ti and Mo in binder phase and the increase of the volume fraction of ceramic grains. At the same x, cermets for y = 15 had lower bending strength and higher hardness than those for y = 30.

  15. European Europart integrated project on actinide partitioning

    International Nuclear Information System (INIS)

    Madic, C.; Hudson, M.J.

    2005-01-01

    This poster presents the objectives of EUROPART, a scientific integrated project between 24 European partners, mostly funded by the European Community within the FP6. EUROPART aims at developing chemical partitioning processes for the so-called minor actinides (MA) contained in nuclear wastes, i.e. from Am to Cf. In the case of dedicated spent fuels or targets, the actinides to be separated also include U, Pu and Np. The techniques considered for the separation of these radionuclides belong to the fields of hydrometallurgy and pyrometallurgy, as in the previous FP5 programs named PARTNEW and PYROREP. The two main axes of research within EUROPART will be: The partitioning of MA (from Am to Cf) from high burn-up UO x fuels and multi-recycled MOx fuels; the partitioning of the whole actinide family for recycling, as an option for advanced dedicated fuel cycles (and in connection with the studies to be performed in the EUROTRANS integrated project). In hydrometallurgy, the research is organised into five Work Packages (WP). Four WP are dedicated to the study of partitioning methods mainly based on the use of solvent extraction methods, one WP is dedicated to the development of actinide co-conversion methods for fuel or target preparation. The research in pyrometallurgy is organized into four WP, listed hereafter: development of actinide partitioning methods, study of the basic chemistry of trans-curium elements in molten salts, study of the conditioning of the wastes, some system studies. Moreover, a strong management team will be concerned not only with the technical and financial issues arising from EUROPART, but also with information, communication and benefits for Europe. Training and education of young researchers will also pertain to the project. EUROPART has also established collaboration with US DOE and Japanese CRIEPI. (authors)

  16. Significance of actinide chemistry for the long-term safety of waste disposal

    International Nuclear Information System (INIS)

    Kim, Jae Il

    2006-01-01

    A geochemical approach to the long-term safety of waste disposal is discussed in connection with the significance of actinides, which shall deliver the major radioactivity inventory subsequent to the relatively short-term decay of fission products. Every power reactor generates transuranic (TRU) elements: plutonium and minor actinides (Np, Am, Cm), which consist chiefly of long-lived nuclides emitting alpha radiation. The amount of TRU actinides generated in a fuel life period is found to be relatively small (about 1 wt% or less in spent fuel) but their radioactivity persists many hundred thousands years. Geological confinement of waste containing TRU actinides demands, as a result, fundamental knowledge on the geochemical behavior of actinides in the repository environment for a long period of time. Appraisal of the scientific progress in this subject area is the main objective of the present paper. Following the introductory discussion on natural radioactivities, the nuclear fuel cycle is briefly brought up with reference to actinide generation and waste disposal. As the long-term disposal safety concerns inevitably with actinides, the significance of the aquatic actinide chemistry is summarized in two parts: the fundamental properties relevant to their aquatic behavior and the geochemical reactions in nanoscopic scale. The constrained space of writing allows discussion on some examples only, for which topics of the primary concern are selected, e.g. apparent solubility and colloid generation, colloid-facilitated migration, notable speciation of such processes, etc. Discussion is summed up to end with how to make a geochemical approach available for the long-term disposal safety of nuclear waste or for the Performance Assessment (PA) as known generally

  17. Detailed investigation of neutron emitters in the transmutation of Minor Actinides

    Energy Technology Data Exchange (ETDEWEB)

    Letourneau, A.; Bringer, O.; Dupont, E.; Panebianco, S.; Veyssiere, Ch. [CEA/Saclay/DSM/IRFU - Gif-sur-Yvette (France); Al Mahamid, I. [Wadsworth Center, New York State Department of Health, Albany, NY 12201 (United States); Chartier, F. [CEA/Saclay/DEN/DPC/SECR - Gif-sur-Yvette (France); Mutti, P. [Institut Laue-Langevin, Grenoble (France); Oriol, L. [CEA/Cadarache/DEN/DER/SPEX - Saint-Paul-lez-Durances (France)

    2008-07-01

    The production of neutron emitters during the incineration process of minor actinides could be very penalizing for the reprocessing of the targets when transmuted in heterogeneous mode, either in dedicated systems (ADS) or in generation IV reactors. Therefore their production has to be carefully evaluated. The reliability of such evaluation really depends on nuclear data (capture and fission cross sections) and their accuracy. In this paper we present a work we have done to investigate the production of neutron emitters in the incineration of {sup 237}Np and {sup 241}Am targets in fast and thermal nuclear reactor concepts. The impact of nuclear data uncertainties on the production of those neutron-emitters was evaluated by sensitivity calculations. The reduction for some of these uncertainties in the thermal energy region was done by measuring more precisely the {sup 244}Cm(n,gamma){sup 245}Cm, {sup 245}Cm(n,f) and {sup 249}Cf(n,gamma){sup 250}Cf capture cross sections at the Laue-Langevin Institute (ILL). It amounts to (15.6+-2.4) b for the first one, (1923+-49) b for the second and (389+-10) b for the third one. (authors)

  18. Minor Actinide Burn in Thermal Spectrum with Enhanced Moderation

    International Nuclear Information System (INIS)

    Petrovic, B.; Huang, L. M.

    2010-01-01

    Resolving the issue of spent nuclear fuel and nuclear waste management is the necessary condition for long-term sustainability of nuclear power, and requires addressing plutonium, minor actinides (MA) and fission products. Various strategies from once-through homogeneous burn to partitioning and transmutation, and from thermal to fast systems, are being considered. The optimum system-level performance will likely require advanced critical or subcritical systems with a range of neutron spectra. Thermal systems, while not optimum, may be deployed sooner, and may provide mid-term amelioration of the issue. This paper examines burn of MA in thermal systems. One specific concern in this case is deterioration of safety parameters due to a high thermal absorption cross section of MA. Enhanced moderation has potential to at least partly remedy this concern. Therefore, we have evaluated adopting the IRIS neutronic design to MA burn. The IRIS reactor design offers enhanced safety margin, due to its fully passive safety systems and safety-by-design approach. Also, in addition to the standard UO 2 fuel (reference IRIS design), an alternative core with enhanced moderation fuel was considered. These two features (safety margin, enhanced moderation) provide a good starting point for MA burn in a thermal system. Further modifications to accommodate MA-bearing rods will be discussed. The paper will examine the benefit of the enhanced moderation in comparison to homogeneous MA burn in a typical PWR reactor.(author).

  19. Flammability Analysis For Actinide Oxides Packaged In 9975 Shipping Containers

    Energy Technology Data Exchange (ETDEWEB)

    Laurinat, James E.; Askew, Neal M.; Hensel, Steve J.

    2013-03-21

    Packaging options are evaluated for compliance with safety requirements for shipment of mixed actinide oxides packaged in a 9975 Primary Containment Vessel (PCV). Radiolytic gas generation rates, PCV internal gas pressures, and shipping windows (times to reach unacceptable gas compositions or pressures after closure of the PCV) are calculated for shipment of a 9975 PCV containing a plastic bottle filled with plutonium and uranium oxides with a selected isotopic composition. G-values for radiolytic hydrogen generation from adsorbed moisture are estimated from the results of gas generation tests for plutonium oxide and uranium oxide doped with curium-244. The radiolytic generation of hydrogen from the plastic bottle is calculated using a geometric model for alpha particle deposition in the bottle wall. The temperature of the PCV during shipment is estimated from the results of finite element heat transfer analyses.

  20. High level waste fixation in cermet form

    International Nuclear Information System (INIS)

    Kobisk, E.H.; Aaron, W.S.; Quinby, T.C.; Ramey, D.W.

    1981-01-01

    Commercial and defense high level waste fixation in cermet form is being studied by personnel of the Isotopes Research Materials Laboratory, Solid State Division (ORNL). As a corollary to earlier research and development in forming high density ceramic and cermet rods, disks, and other shapes using separated isotopes, similar chemical and physical processing methods have been applied to synthetic and real waste fixation. Generally, experimental products resulting from this approach have shown physical and chemical characteristics which are deemed suitable for long-term storage, shipping, corrosive environments, high temperature environments, high waste loading, decay heat dissipation, and radiation damage. Although leach tests are not conclusive, what little comparative data are available show cermet to withstand hydrothermal conditions in water and brine solutions. The Soxhlet leach test, using radioactive cesium as a tracer, showed that leaching of cermet was about X100 less than that of 78 to 68 glass. Using essentially uncooled, untreated waste, cermet fixation was found to accommodate up to 75% waste loading and yet, because of its high thermal conductivity, a monolith of 0.6 m diameter and 3.3 m-length would have only a maximum centerline temperature of 29 K above the ambient value

  1. Resource recovery of WC-Co cermet using hydrothermal oxidation technique

    International Nuclear Information System (INIS)

    Gao Ningfeng; Inagaki, F.; Sasai, R.; Itoh, H.; Watari, K.

    2005-01-01

    WC-Co cermet is widely used in industrial applications such as cutting tools, dies, wear parts and so on. It is of great importance to establish the recycling process for the precious metal resources contained in WC-Co cermet, because all these metals used in Japan are imported. In this paper we reported a hydrothermal oxidation technique using nitric acid for the reclamation of WC and Co. The WC-Co cermet specimens with various WC particle sizes and Co contents were hydrothermally treated in HNO 3 aqueous solutions at temperatures of 110-200 C for durations of 6-240 h. The Co was preferentially leached out into the acidic solution, while the WC was oxidized to insoluble WO 3 hydrate which was subsequently separated by filtration. The hydrothermal treatment parameters such as solvent concentrations, treatment temperatures, holding time were optimized in respect to different kinds of WC-Co cermets. A hydrothermal oxidation treatment in 3M HNO 3 aqueous solution at 150 C for 24 h was capable of fully disintegrating the cermet chip composed of coarse WC grains of 1-5 μm in size with 20 wt% of Co as binder. While the more oxidation resistant specimen composed of fine WC grains of 0.5-1.0 μm in size with 13 wt% of Co, was completely disintegrated by a treatment in 7 M HNO 3 aqueous solution at 170 C for 24 h. The filtered solid residues were composed of fine WO 3 .0.33H 2 O powder and a small amount of WO 3 . The recovered WO 3 .0.33H 2 O powder can be easily returned to the industrial process for the synthesis of WC powder so that the overall recycling cost can be possibly lowered. (orig.)

  2. Minor Actinide Separations Using Ion Exchangers Or Ionic Liquids

    International Nuclear Information System (INIS)

    Hobbs, D.; Visser, A.; Bridges, N.

    2011-01-01

    This project seeks to determine if (1) inorganic-based ion exchange materials or (2) electrochemical methods in ionic liquids can be exploited to provide effective Am and Cm separations. Specifically, we seek to understand the fundamental structural and chemical factors responsible for the selectivity of inorganic-based ion-exchange materials for actinide and lanthanide ions. Furthermore, we seek to determine whether ionic liquids can serve as the electrolyte that would enable formation of higher oxidation states of Am and other actinides. Experiments indicated that pH, presence of complexants and Am oxidation state exhibit significant influence on the uptake of actinides and lanthanides by layered sodium titanate and hybrid zirconium and tin phosphonate ion exchangers. The affinity of the ion exchangers increased with increasing pH. Greater selectivity among Ln(III) ions with sodium titanate materials occurs at a pH close to the isoelectric potential of the ion exchanger. The addition of DTPA decreased uptake of Am and Ln, whereas the addition of TPEN generally increases uptake of Am and Ln ions by sodium titanate. Testing confirmed two different methods for producing Am(IV) by oxidation of Am(III) in ionic liquids (ILs). Experimental results suggest that the unique coordination environment of ionic liquids inhibits the direct electrochemical oxidation of Am(III). The non-coordinating environment increases the oxidation potential to a higher value, while making it difficult to remove the inner coordination of water. Both confirmed cases of Am(IV) were from the in-situ formation of strong chemical oxidizers.

  3. Selective extraction of actinides from high level liquid wastes. Study of the possibilities offered by the Redox properties of actinides

    International Nuclear Information System (INIS)

    Adnet, J.M.

    1991-07-01

    Partitioning of high level liquid wastes coming from nuclear fuel reprocessing by the PUREX process, consists in the elimination of minor actinides (Np, Am, and traces of Pu and U). Among the possible processes, the selective extraction of actinides with oxidation states higher than three is studied. First part of this work deals with a preliminary step; the elimination of the ruthenium from fission products solutions using the electrovolatilization of the RuO4 compound. The second part of this work concerns the complexation and oxidation reactions of the elements U, Np, Pu and Am in presence of a compound belonging to the insaturated polyanions family: the potassium phosphotungstate. For actinide ions with oxidation state (IV) complexed with phosphotungstate anion the extraction mechanism by dioctylamine was studied and the use of a chromatographic extraction technic permitted successful separations between tetravalents actinides and trivalents actinides. Finally, in accordance with the obtained results, the basis of a separation scheme for the management of fission products solutions is proposed

  4. Heterogeneous composite bodies with isolated lenticular shaped cermet regions

    Science.gov (United States)

    Sherman, Andrew J [Cirtland Hills, OH

    2009-12-22

    A heterogeneous body having ceramic rich cermet regions in a more ductile metal matrix. The heterogeneous bodies are formed by thermal spray operations on metal substrates. The thermal spray operations apply heat to a cermet powder and project it onto a solid substrate. The cermet powder is composed of complex composite particles in which a complex ceramic-metallic core particle is coated with a matrix precursor. The cermet regions are generally comprised of complex ceramic-metallic composites that correspond approximately to the core particles. The cermet regions are approximately lenticular shaped with an average width that is at least approximately twice the average thickness. The cermet regions are imbedded within the matrix phase and generally isolated from one another. They have obverse and reverse surfaces. The matrix phase is formed from the matrix precursor coating on the core particles. The amount of heat applied during the formation of the heterogeneous body is controlled so that the core particles soften but do not become so fluid that they disperse throughout the matrix phase. The force of the impact on the surface of the substrate tends to flatten them. The flattened cermet regions tend to be approximately aligned with one another in the body.

  5. Actinides reduction by recycling in a thermal reactor

    International Nuclear Information System (INIS)

    Ramirez S, J. R.; Martinez C, E.; Balboa L, H.

    2014-10-01

    This work is directed towards the evaluation of an advanced nuclear fuel cycle in which radioactive actinides could be recycled to remove most of the radioactive material; firstly a production reference of actinides in standard nuclear fuel of uranium at the end of its burning in a BWR reactor is established, after a fuel containing plutonium is modeled to also calculate the actinides production in MOX fuel type. Also it proposes a design of fuel rod containing 6% of actinides in a matrix of uranium from the tails of enrichment, then four standard uranium fuel rods are replaced by actinides rods to evaluate the production and transmutation thereof, the same procedure was performed in the fuel type MOX and the end actinide reduction in the fuel was evaluated. (Author)

  6. Wear resistance of TiB/sub 2/-Fe cermets

    International Nuclear Information System (INIS)

    Champagne, B.; Dallaire, S.

    1985-01-01

    A material which consists of TiB/sub 2/ dispersed in an iron matrix was synthesized by the exothermic reaction of ferrotitanium and boron. The as-reacted products were hot isostatically pressed to produce TiB/sub 2/-Fe cermets. The influence of HIP variables on the density and total fractional porosity of specimens is presented. Density above 95% is obtained by HIPping at temperatures below 1300 0 C. Increasing the temperature and the time of HIPping enhance the mechanical properties and wear resistance of TiB/sub 2/-Fe cermets by reducing their residual porosity. Relations obtained by regression analysis showed that the porosity strongly affects the properties of parts. Regression analysis point out that the wear loss of a 5% porosity TiB/sub 2/-Fe cermet is 270% higher than a dense HIPped cermet. Low stress and high stress abrasion resistance tests utilizing various abrasive media were carried out on dense HIPped cermets and results were compared with those obtained from WC-Co cermets and 1020 steel

  7. Analytical determination of thermal conductivity of W-UO2 and W-UN CERMET nuclear fuels

    Science.gov (United States)

    Webb, Jonathan A.; Charit, Indrajit

    2012-08-01

    The thermal conductivity of tungsten based CERMET fuels containing UO2 and UN fuel particles are determined as a function of particle geometry, stabilizer fraction and fuel-volume fraction, by using a combination of an analytical approach and experimental data collected from literature. Thermal conductivity is estimated using the Bruggeman-Fricke model. This study demonstrates that thermal conductivities of various CERMET fuels can be analytically predicted to values that are very close to the experimentally determined ones.

  8. Partnew - New solvent extraction processes for minor actinides - final report; Partnew - Nouveaux procedes d'extraction par solvant pour les actinides mineurs - rapport final

    Energy Technology Data Exchange (ETDEWEB)

    Madic, C.; Testard, F.; Hudson, M.J.; Liljenzin, J.O.; Christiansen, B.; Ferrando, M.; Facchini, A.; Geist, A.; Modolo, G.; Gonzalez-Espartero, A.; Mendoza, J. de

    2004-07-01

    The objectives of the European project PARTNEW were to define solvent extraction processes for the partitioning of the minor actinides, Am and Cm, from the aqueous high active raffinate or high active concentrate issuing the reprocessing of nuclear spent fuels by the PUREX process. Eleven laboratories participated to the research: 1/ CEA-DEN (Marcoule), 2/ CEA-DSM (Saclay), 3/ UREAD (U.K.), 4/ CTU (Sweden), 5/ ITU (Germany), 6/ ENEA (Italy), 7/ PoliMi (Italy), 8/ FZK-INE (Germany), 9/ FZJ-ISR (Germany), 10/ CIEMAT (Spain) and 11/ UAM (Spain). The research was organised into eight work packages (WP): Basic and applied DIAMEX studies, using diamide extractants for the co-extraction of actinides(III) (An(III)) and lanthanides(III) (Ln(III)) nitrates (WP1 and WP2), Basic and applied SANEX studies based on the use of polydentate N-ligands for the An(III)/Ln(III) separation (WP3 and WP4), Basic and applied SANEX studies based on the use of synergistic mixtures made of bis-(chloro-phenyl)-di-thio-phosphinic acid + neutral O-bearing ligand, (WP5 and WP6), Basic SANEX studies for the An(III)/Ln(III) separation, based on the use of new S-bearing ligands, Basic and applied studies for the Am(III)/Cm(III) separation. The work done in the fundamental and applied domains was very fruitful. Several processes have been successfully tested with genuine high active raffinates and concentrate. (authors)

  9. Surface properties of copper based cermet materials

    International Nuclear Information System (INIS)

    Voinea, M.; Vladuta, C.; Bogatu, C.; Duta, A.

    2008-01-01

    The paper presents the characterization of the surface properties of copper based cermets obtained by two different techniques: spray pyrolysis deposition (SPD) and electrodeposition. Copper acetate was used as precursor of Cu/CuO x cermet. The surface morphology was tailored by adding copolymers of maleic anhydride with controlled hydrophobia. The films morphology of Cu/CuO x was assessed using contact angle measurements and AFM analysis. The porous structures obtained via SPD lead to higher liquid adsorption rate than the electrodeposited films. A highly polar liquid - water is recommended as testing liquid in contact angle measurements, for estimating the porosity of copper based cermets, while glycerol can be used to distinguish among ionic and metal predominant structures. Thus, contact angle measurements can be used for a primary evaluation of the films morphology and, on the other hand, of the ratio between the cermet components

  10. Manufacture of annular cermet articles

    Science.gov (United States)

    Forsberg, Charles W.; Sikka, Vinod K.

    2004-11-02

    A method to produce annular-shaped, metal-clad cermet components directly produces the form and avoids multiple fabrication steps such as rolling and welding. The method includes the steps of: providing an annular hollow form with inner and outer side walls; filling the form with a particulate mixture of ceramic and metal; closing, evacuating, and hermetically sealing the form; heating the form to an appropriate temperature; and applying force to consolidate the particulate mixture into solid cermet.

  11. Update on the FUTURIX-FTA Experiment in PHENIX

    International Nuclear Information System (INIS)

    Jaecki, P.; Pillon, S.; Warin, D.; Donnet, L.; Jorion, F.; Drin, N.; Hayes, S.L.; Kennedy, J.R.; Pasamehmetoglu, K.; Voit, S.L.; Haas, D.; Fernandez, A.

    2006-01-01

    Europe and the USA are following similar R and D partitioning and transmutation strategies to manage long lived waste, especially minor actinides. Actinide transmutation and recycle is presently being considered in dedicated systems such as Accelerator-Driven Systems (ADS), in Generation IV fast reactors, and in current light water reactors. In ADS, this leads to specific fuel formulations, highly enriched in Minor Actinides. Fertile uranium is excluded from ADS fuels to avoid the production of plutonium ('non-fertile or U-free fuels'). For low-fertile fuels, Uranium content is adjusted to achieve a conversion ratio of 0.25 or less and plutonium production is therefore limited. Very little information is available about the irradiation behaviour of the resulting fuel compositions. The FUTURIX-FTA experiment objective is to compare in similar and representative conditions the behaviour of fuels proposed for TRU burning. The FUTURIX-FTA program began in January 2003 and the irradiation in Phenix will start at the beginning of 2007. This progress report focuses on the latest results, R and D and irradiation test design which are almost complete. The paper has the following contents: I. Introduction; II. Fuel compositions; III. CERCER Fuels; 1. Design and safety considerations; 2. R and D on CERCER fuel fabrication; Preparation of (Pu,Am)O 2 by co-conversion; Fabrication of composites; Oxygen (O/M) stoichiometry, density and crystal structures of the actinide compounds; (Pu 0.2 Am 0.8 )O 2-x MgO Cercer Matrix; Microstructure and Porosity of the Compounds; IV. CERMET Fuels; 1. Design and safety considerations; 2. R and D on CERMET fuel fabrication; V. Metallic fuels; 1. Design and safety considerations; 2. R and D on CERMET fuel fabrication; VI. Nitride fuels; 1. Design and safety considerations;. R and D on NITRIDE fuel fabrication; II. Irradiation conditions; III. Conclusion. To summarize, preliminary design work is complete, and the final design will be fixed in 2006

  12. Actinide nitride ceramic transmutation fuels for the Futurix-FTA irradiation experiment

    International Nuclear Information System (INIS)

    Voit, St.; McClellan, K.; Stanek, Ch.; Maloy, St.

    2007-01-01

    Full text of publication follows. The transmutation of plutonium and other minor actinides is an important component of an advanced nuclear fuel cycle. The Advanced Fuel Cycle Initiative (AFCI) is currently considering mono-nitrides as potential transmutation fuel material on account of the mutual solubility of actinide mono-nitrides as well as their desirable thermal characteristics. The feedstock is most commonly produced by a carbothermic reduction/nitridisation process, as it is for this programme. Fuel pellet fabrication is accomplished via a cold press/sinter approach. In order to allow for easier investigation of the synthesis and fabrication processes, surrogate material studies are used to compliment the actinide activities. Fuel compositions of particular interest denoted as low fertile (i.e. containing uranium) and non-fertile (i.e. not containing uranium) are (PuAmNp) 0.5 U 0.5 N and (PuAm) 0.42 Zr 0.58 N, respectively. The AFCI programme is investigating the validity of these fuel forms via Advanced Test Reactor (ATR) and Phenix irradiations. Here, we report on the recent progress of actinide-nitride transmutation fuel development and production for the Futurix-FTA irradiation experiment. Furthermore, we highlight specific cases where the complimentary approach of surrogate studies and actinide development aid in the understanding complex material issues. In order to allow for easier investigation of the fundamental materials properties, surrogate materials have been used. The amount of surrogate in each compound was determined by comparing both molar concentration and lattice parameter mismatch via Vegard Law. Cerium was chosen to simultaneously substitute for Pu, Am and Np, while depleted U was chosen to substitute for enriched U. Another goal of this work was the optimisation of added graphite during carbothermic reduction in order to minimise the duration of the carbon removal step (i.e. heat treatment under H 2 containing gas). One proposed

  13. The Aqueous Electrochemical Response of TiC–Stainless Steel Cermets

    Directory of Open Access Journals (Sweden)

    Chukwuma Onuoha

    2018-05-01

    Full Text Available A family of TiC–stainless steel ceramic–metal composites, or cermets, has been developed in the present study, using steel grades of 304 L, 316 L, or 410 L as the binder phase. Melt infiltration was used to prepare the cermets, with the steel binder contents varying between 10–30 vol. %. The corrosion behaviour was evaluated using a range of electrochemical techniques in an aqueous solution containing 3.5 wt. % NaCl. The test methods included potentiodynamic, cyclic, and potentiostatic polarisation. The corroded samples were subsequently characterised using scanning electron microscopy (SEM and energy dispersive X-ray spectroscopy (EDS, while the post-corrosion solutions were analysed using inductively coupled plasma optical emission spectroscopy (ICP-OES to determine the residual ionic and particulate material removed from the cermets during electrochemical testing. It was demonstrated that the corrosion resistance was enhanced through decreasing the steel binder content, which arises due to the preferential dissolution of the binder phase, while the TiC ceramic remains largely unaffected. Increasing corrosion resistance was observed in the sequence TiC-304 L > TiC-316 L > TiC-410 L.

  14. Inverted amorphous silicon solar cell utilizing cermet layers

    Science.gov (United States)

    Hanak, Joseph J.

    1979-01-01

    An amorphous silicon solar cell incorporating a transparent high work function metal cermet incident to solar radiation and a thick film cermet contacting the amorphous silicon opposite to said incident surface.

  15. Biocompatibility assessment of spark plasma-sintered alumina-titanium cermets.

    Science.gov (United States)

    Guzman, Rodrigo; Fernandez-García, Elisa; Gutierrez-Gonzalez, Carlos F; Fernandez, Adolfo; Lopez-Lacomba, Jose Luis; Lopez-Esteban, Sonia

    2016-01-01

    Alumina-titanium materials (cermets) of enhanced mechanical properties have been lately developed. In this work, physical properties such as electrical conductivity and the crystalline phases in the bulk material are evaluated. As these new cermets manufactured by spark plasma sintering may have potential application for hard tissue replacements, their biocompatibility needs to be evaluated. Thus, this research aims to study the cytocompatibility of a novel alumina-titanium (25 vol. % Ti) cermet compared to its pure counterpart, the spark plasma sintered alumina. The influence of the particular surface properties (chemical composition, roughness and wettability) on the pre-osteoblastic cell response is also analyzed. The material electrical resistance revealed that this cermet may be machined to any shape by electroerosion. The investigated specimens had a slightly undulated topography, with a roughness pattern that had similar morphology in all orientations (isotropic roughness) and a sub-micrometric average roughness. Differences in skewness that implied valley-like structures in the cermet and predominance of peaks in alumina were found. The cermet presented a higher surface hydrophilicity than alumina. Any cytotoxicity risk associated with the new materials or with the innovative manufacturing methodology was rejected. Proliferation and early-differentiation stages of osteoblasts were statistically improved on the composite. Thus, our results suggest that this new multifunctional cermet could improve current alumina-based biomedical devices for applications such as hip joint replacements. © The Author(s) 2015.

  16. Actinide-Aluminate Speciation in Alkaline Radioactive Waste

    International Nuclear Information System (INIS)

    Clark, David L.; Fedosseev, Alexander M.

    2001-01-01

    Investigation of behavior of actinides in alkaline media containing AL(III) showed that no aluminate complexes of actinides in oxidation states (IIII-VIII) were formed in alkaline solutions. At alkaline precipitation IPH (10-14) of actinides in presence of AL(III) formation of aluminate compounds is not observed. However, in precipitates contained actinides (IIV)<(VI), and to a lesser degree actinides (III), some interference of components takes place that is reflected in change of solid phase properties in comparison with pure components or their mechanical mixture. The interference decreases with rise of precipitation PH and at PH 14 is exhibited very feebly. In the case of NP(VII) the individual compound with AL(III) is obtained, however it is not aluminate of neptunium(VII), but neptunate of aluminium(III) similar to neptunates of other metals obtained earlier

  17. Microstructural development and mechanical properties of iron based cermets processed by pressureless and spark plasma sintering

    International Nuclear Information System (INIS)

    Alvaredo, P.; Gordo, E.; Van der Biest, O.; Vanmeensel, K.

    2012-01-01

    Highlights: ► Processing of Fe-based cermets by pressureless sintering and spark plasma sintering. ► Influence of carbon content on the sintering mechanism and hardness. ► The cermet phase diagram was calculated and permits to explain the microstructure. ► SPS provides ferritic matrix and different carbide distribution than CPS samples. ► Pressureless sintered samples contain retained austenite at room temperature. - Abstract: Iron-based cermets are an interesting class of metal-ceramic composites in which properties and the factors influencing them are to be explored. In this work the metal matrix contains Cr, W, Mo and V as alloying elements, and the hard phase is constituted by 50 vol% of titanium carbonitride (TiCN) particles. The work studies the influence of the C content and the processing method on the sinterability, microstructure and hardness of the developed cermet materials. For that purpose, cermet samples with different C content in the matrix (0 wt%, 0.25 wt%, 0.5 wt%, 1.0 wt%) were prepared by conventional pressureless sintering (CPS) and, in order to achieve finer microstructures and to reduce the sintering time, by spark plasma sintering (SPS). The density and hardness (HV30) of the processed materials was evaluated, while their phase composition and microstructure was characterised by X-ray diffraction (XRD) and scanning electron microscopy (SEM), respectively. The equilibrium phase diagram of the composite material was calculated by ThermoCalc software in order to elucidate the influence of the carbon content on the obtained phases and developed microstructures.

  18. Actinide burning in the integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1993-01-01

    During the past few years, Argonne National Laboratory has been developing the integral fast reactor (IFR), an advanced liquid-metal reactor concept. In the IFR, the inherent properties of liquid-metal cooling are combined with a new metallic fuel and a radically different refining process to allow breakthroughs in passive safety, fuel cycle economics, and waste management. A key feature of the IFR concept is its unique pyroprocessing. Pyroprocessing has the potential to radically improve long-term waste management strategies by exploiting the following attributes: 1. Minor actinides accompany plutonium product stream; therefore, actinide recycling occurs naturally. Actinides, the primary source of long-term radiological toxicity, are removed from the waste stream and returned to the reactor for in situ burning, generating useful energy. 2. High-level waste volume from pyroprocessing call be reduced substantially as compared with direct disposal of spent fuel. 3. Decay heat loading in the repository can be reduced by a large factor, especially for the long-term burden. 4. Low-level waste generation is minimal. 5. Troublesome fission products, such as 99 Tc, 129 I, and 14 C, are contained and immobilized. Singly or in combination, the foregoing attributes provide important improvements in long-term waste management in terms of the ease in meeting technical performance requirements (perhaps even the feasibility of demonstrating that technical performance requirements can be met) and perhaps also in ultimate public acceptance. Actinide recycling, if successfully developed, could well help the current repository program by providing an opportunity to enhance capacity utilization and by deferring the need for future repositories. It also represents a viable technical backup option in the event unforeseen difficulties arise in the repository licensing process

  19. Novel boride base cermets with very high strength

    International Nuclear Information System (INIS)

    Ken-ichi Takagi; Mari Yonetsu; Yuji Yamasaki

    2001-01-01

    Mo 2 NiB 2 boride base cermets consist of a Mo 2 NiB 2 type complex boride as a hard phase and a Ni base binder. The addition of Cr and V to the cermets changed the boride structure from orthorhombic to tetragonal and resulted in the improvement of mechanical properties and microstructural refinement. The tetragonal Mo 2 NiB 2 was formed through the orthorhombic Mo 2 NiB 2 by the solid state reaction during sintering and not formed directly from the raw material powders. Ni-4.5B-46.9Mo-12.5V-xMn (wt.%) model cermets with five levels of Mn content from 0 to 10 wt.% were prepared to investigate the effects of Mn on the mechanical properties and microstructure Of Mo 2 NiB 2 base cermets. The transverse rupture strength (TRS) of the cermets depended strongly on the microstructure, which varied significantly with Mn content. The maximum TRS obtained at 2.5 wt.%Mn were 3.5 Gpa with hardness of 87 R A . (author)

  20. Measurements of thermal fission and capture cross sections of minor actinides within the Mini-INCA project

    Energy Technology Data Exchange (ETDEWEB)

    Bringer, O.; Chabod, S.; Dupont, E.; Letourneau, A.; Panebianco, S.; Veyssiere, Ch. [CEA Saclay, Dept. d' Astrophysique de Physique des Particules, de Physique Nucleaire et de l' Instrumentation Associee, 91- Gif sur Yvette (France); Oriol, L. [CEA Cadarache, Dept. d' Etudes des Reacteurs, 13 - Saint Paul lez Durance (France); Chartier, F. [CEA Saclay, Dept. de Physico-Chimie, 91 - Gif sur Yvette (France); Mutti, P. [Institut Laue Langevin, 38 - Grenoble, (France); AlMahamid, I. [Wadsworth Center, New York State Dept. of Health, Albany, NY (United States)

    2008-07-01

    In the framework of nuclear waste transmutation studies, the Mini-INCA project has been initiated at Cea/DSM to determine optimal conditions for transmutation and incineration of Minor Actinides in high intensity neutron fluxes in the thermal region. Our experimental tool is based on alpha- and gamma-spectroscopy of irradiated samples and microscopic fission-chambers. It can provide both microscopic information on nuclear reactions (total and partial cross sections for neutron capture and/or fission reactions) and macroscopic information on transmutation and incineration potentials. The {sup 232}Th, {sup 237}Np, {sup 241}Am, and {sup 244}Cm transmutation chains have been explored in details, showing some discrepancies in comparison with evaluated data libraries but in overall good agreement with recent experimental data. (authors)

  1. Measurements of thermal fission and capture cross sections of minor actinides within the Mini-INCA project

    International Nuclear Information System (INIS)

    Bringer, O.; Chabod, S.; Dupont, E.; Letourneau, A.; Panebianco, S.; Veyssiere, Ch.; Oriol, L.; Chartier, F.; Mutti, P.; AlMahamid, I.

    2008-01-01

    In the framework of nuclear waste transmutation studies, the Mini-INCA project has been initiated at Cea/DSM to determine optimal conditions for transmutation and incineration of Minor Actinides in high intensity neutron fluxes in the thermal region. Our experimental tool is based on alpha- and gamma-spectroscopy of irradiated samples and microscopic fission-chambers. It can provide both microscopic information on nuclear reactions (total and partial cross sections for neutron capture and/or fission reactions) and macroscopic information on transmutation and incineration potentials. The 232 Th, 237 Np, 241 Am, and 244 Cm transmutation chains have been explored in details, showing some discrepancies in comparison with evaluated data libraries but in overall good agreement with recent experimental data. (authors)

  2. Cutting performance of TiCN–HSS cermet in dry machining

    OpenAIRE

    Canteli Fernández, José Antonio; Cantero Guisández, José Luis; Marín, N.C.; Gómez, B.; Gordo Odériz, Elena; Miguélez, Henar

    2010-01-01

    This work is focused on the cutting performance of a new cermet based on high-speed steel (HSS) matrix with hard phase TiCN. The processing route to manufacture the cermet M2+ 50 vol.% TiCN is described. Orthogonal cutting tests, carried out in a lathe showed the ability of the new cermet to achieve turning operations, showing reasonably wear resistance performing dry cutting operations. Tool life was significantly increased, when the cermet was compared with the reference materia...

  3. Nonlinear oxidation kinetics of nickel cermets

    International Nuclear Information System (INIS)

    Galinski, Henning; Bieberle-Huetter, Anja; Rupp, Jennifer L.M.; Gauckler, Ludwig J.

    2011-01-01

    The oxidation of a cermet of screen-printed nickel (Ni) and gadolinia-doped ceria (CGO) with an approximate median porosity of 50 vol.% has been studied via in situ X-ray diffraction and focused ion beam nanotomography in the temperature range 773-848 K. The oxidation kinetics of Ni to NiO is found to be highly nonlinear with an apparent activation energy of 2.8(2) eV in this temperature range. The nonlinear oxidation kinetics found is in good agreement with theoretical works on oxide growth driven by nonlinear inbuilt fields. Stress-induced Kirkendall void formation has been identified as the physical process that enhances the oxidation of Ni/CGO cermets. Compressive stresses within the Ni matrix result from the thermal expansion mismatch of Ni and CGO and cause plastic deformation as they exceed the yield stress of the Ni matrix. The pore size distribution of Kirkendall voids formed has been measured by FIB nanotomography and shows a significant temperature dependence. It is shown that even one cycle of reoxidation changes irreversibly the microstructure of the cermet which can be interpreted as the onset and main contribution to the mechanical degradation of the cermet.

  4. Multi-axial response of idealized cermets

    International Nuclear Information System (INIS)

    Pickering, E.G.; Bele, E.; Deshpande, V.S.

    2016-01-01

    The yield response of two idealized cermets comprising mono and bi-disperse steel spheres in a Sn/Pb solder matrix has been investigated for a range of axisymmetric stress states. Proportional stress path experiments are reported, from which are extracted the initial yield surfaces and their evolution with increasing plastic strain. The initial yield strength is nearly independent of the hydrostatic pressure but the strain hardening rate increases with stress triaxiality up to a critical value. For higher triaxialities, the responses are independent of hydrostatic pressure. Multi-axial measurements along with X-ray tomography were used to demonstrate that the deformation of these idealized cermets occurs by two competing mechanisms: (i) a granular flow mechanism that operates at low levels of triaxiality, where volumetric dilation occurs under compressive stress states, and (ii) a plastically incompressible mechanism that operates at high stress triaxialities. A phenomenological viscoplastic constitutive model that incorporates both deformation mechanisms is presented. While such multi-axial measurements are difficult for commercial cermets with yield strengths on the order of a few GPa, the form of their constitutive relation is expected to be similar to that of the idealized cermets presented here.

  5. Fabrication and thermal conductivity of boron carbide/copper cermet

    International Nuclear Information System (INIS)

    Maruyama, Tadashi; Onose, Shoji

    1999-01-01

    Studies on fabrication and thermal conductivity of B 4 C/Cu cermet were made to obtain high performance neutron absorber materials for Liquid Metal-cooled Fast Breeder Reactor (LMFBR). A mixed powder of B 4 C and Cu was mechanically blended at high speed thereby a coating layer of Cu was formed on the surface of B 4 C powder. Then the B 4 C powder with Cu coating was hot pressed at temperatures from 950 to 1,050degC to form a B 4 C cermet. A high density B 4 C/Cu cermet with 70 vol% of B 4 C and relative density higher than 90% was successfully fabricated. In spite of the low volume fraction of Cu, the B 4 C/Cu cermet exhibited high thermal conductivity which originated from the existence of continuous metallic phase Cu in B 4 C/Cu cermet. (author)

  6. Actinide partitioning from high level liquid waste using the Diamex process

    International Nuclear Information System (INIS)

    Madic, C.; Blanc, P.; Condamines, N.; Baron, P.; Berthon, L.; Nicol, C.; Pozo, C.; Lecomte, M.; Philippe, M.; Masson, M.; Hequet, C.

    1994-01-01

    The removal of long-lived radionuclides, which belong to the so-called minor actinides elements, neptunium, americium and curium, from the high level nuclear wastes separated during the reprocessing of the irradiated nuclear fuels in order to transmute them into short-lived nuclides, can substantially decrease the potential hazards associated with the management of these nuclear wastes. In order to separate minor actinides from high-level liquid wastes (HLLW), a liquid-liquid extraction process was considered, based on the use of diamide molecules, which display the property of being totally burnable, thus they do not generate secondary solid wastes. The main extracting properties of dimethyldibutyltetradecylmalonamide (DMDBTDMA), the diamide selected for the development of the DIAMEX process, are briefly described in this paper. Hot tests of the DIAMEX process (using DMDBTDMA) related to the treatment of an mixed oxide fuels (MOX) type HLLW, were successfully performed. The minor actinide decontamination factors of the HLLW obtained were encouraging. The main results of these tests are presented and discussed in this paper. (authors). 9 refs., 2 figs., 7 tabs

  7. Actinides compounds for the transmutation: scientific contributions of american and japanese collaborations; Composes d'actinides pour la transmutation: apports scientifiques de collaborations americaines et japonaises

    Energy Technology Data Exchange (ETDEWEB)

    Raison, Ph.; Albiot, T

    2000-07-01

    This paper deals with the minor actinides transmutation and the scientific contribution of the ORNL and the JAERI. It presents researches on the Am-Zr-Y-O system in the framework of the heterogeneous reprocessing, the curium and pyrochlore structures, with the ORNL contribution and phase diagrams, data of Thermodynamics, actinides nitrides, with the JAERI. (A.L.B.)

  8. Molybdenum-base cermet fuel development

    International Nuclear Information System (INIS)

    Gurwell, W.E.; Moss, R.W.; Pilger, J.P.; White, G.D.

    1987-07-01

    Development of a multimegawatt (MMW) space nuclear power system requires identification and resolution of several technical feasibility issues before selecting one or more promising system concepts. Demonstration of reactor fuel fabrication technology is required for cermet-fueled reactor concepts. MMW reactor fuel development activity at Pacific Northwest Laboratory (PNL) is focused on producing a molybdenum-matrix uranium-nitride (UN) fueled cermet. This cermet is to have a high matrix density (≥95%) for high strength and high thermal conductance coupled with a high particle (UN) porosity (∼25%) for retention of released fission gas at high burnup. Fabrication process development involves the use of porous TiN microspheres as surrogate fuel material until porous UN microspheres become available. Process development has been conducted in the areas of microsphere synthesis, particle sealing/coating, and high-energy-rate forming (HERF) and vacuum hot press consolidation techniques. This paper summarizes the status of these activities

  9. On the influence of the americium isotopic vector on the cooling time of minor actinides bearing blankets in fast reactors

    Directory of Open Access Journals (Sweden)

    Kooyman Timothée

    2018-01-01

    Full Text Available In the heterogeneous minor actinides transmutation approach, the nuclei to be transmuted are loaded in dedicated targets often located at the core periphery, so that long-lived heavy nuclides are turned into shorter-lived fission products by fission. To compensate for low flux level at the core periphery, the minor actinides content in the targets is set relatively high (around 20 at.%, which has a negative impact on the reprocessing of the targets due to their important decay heat level. After a complete analysis of the main contributors to the heat load of the irradiated targets, it is shown here that the choice of the reprocessing order of the various feeds of americium from the fuel cycle depends on the actual limit for fuel reprocessing. If reprocessing of hot targets is possible, it is more interesting to reprocess first the americium feed with a high 243Am content in order to limit the total cooling time of the targets, while if reprocessing of targets is limited by their decay heat, it is more interesting to wait for an increase in the 241Am content before loading the americium in the core. An optimization of the reprocessing order appears to lead to a decrease of the total cooling time by 15 years compared to a situation where all the americium feeds are mixed together when two feeds from SFR are considered with a high reprocessing limit.

  10. Development and characterization of cermet forms for radioactive waste

    International Nuclear Information System (INIS)

    Aaron, W.S.; Quinby, T.C.; Kobisk, E.H.

    1979-01-01

    Cermets designed to isolate high-level wastes in a solid form are a composite consisting of various ceramic phase particles uniformly dispersed in and microencapsulated by an iron-nickel base alloy matrix. The metal matrix provides this waste form with many advantageous features including excellent thermal conductivity and mechanical strength. These cermets are formed by first dissolving the waste in molten urea, precipitating and calcining all the constituents, compacting the calcine, and sintering and reduction to form the final product. The exact formulation of cermets through additions to the waste is designed to fix most of the fission products in stable, leach resistant ceramic phases which are subsequently microencapsulated by an alloy matrix. The alloy matrix, which is derived primarily from the waste itself and includes the reducible fission and activation products from the waste, can be compositionally adjusted through additions to optimize its corrosion resistance under conditions existing in various disposal environments. The processes by which cermets are formed include several new and unique materials preparation options that are being developed to permit engineering scale-up and to be compatible with remote operations. Cermets formed by alternate processing methods are being characterized. Initially, cermet samples were prepared using a laboratory scale, batch process developed for the preparation of special ceramics having high compositional uniformity and excellent sinterability. The modification of this batch process to one suitable for scale-up and remote operation is the subject of this paper. Cermet characterization is also discussed

  11. Experimental studies of actinides in molten salts

    Energy Technology Data Exchange (ETDEWEB)

    Reavis, J.G.

    1985-06-01

    This review stresses techniques used in studies of molten salts containing multigram amounts of actinides exhibiting intense alpha activity but little or no penetrating gamma radiation. The preponderance of studies have used halides because oxygen-containing actinide compounds (other than oxides) are generally unstable at high temperatures. Topics discussed here include special enclosures, materials problems, preparation and purification of actinide elements and compounds, and measurements of various properties of the molten volts. Property measurements discussed are phase relationships, vapor pressure, density, viscosity, absorption spectra, electromotive force, and conductance. 188 refs., 17 figs., 6 tabs.

  12. Experimental studies of actinides in molten salts

    International Nuclear Information System (INIS)

    Reavis, J.G.

    1985-06-01

    This review stresses techniques used in studies of molten salts containing multigram amounts of actinides exhibiting intense alpha activity but little or no penetrating gamma radiation. The preponderance of studies have used halides because oxygen-containing actinide compounds (other than oxides) are generally unstable at high temperatures. Topics discussed here include special enclosures, materials problems, preparation and purification of actinide elements and compounds, and measurements of various properties of the molten volts. Property measurements discussed are phase relationships, vapor pressure, density, viscosity, absorption spectra, electromotive force, and conductance. 188 refs., 17 figs., 6 tabs

  13. New solvent extraction processes for minor actinides: CIEMAT contribution to the partnew project: EU contract n. FIKW-CT2000-0087: first semestral period 2001 september 2000-february 2001

    International Nuclear Information System (INIS)

    2002-01-01

    This report includes the work developed at CIEMAT into the partnew project: '' New solvent extraction processes for minor actinides, during the first semestral period (september 2000 to february 2001), CIEMAT is involved in the following task: the study of the actinides (AN) and lanthanides (LN) extracting properties of new compounds with chemical structure based on two malonamide groups linked to an aromatic platform. The study of new-bearing extractants with chemical structure similar to malonamides aforementioned, changing the 0 atoms by s atoms, and the determination of the selectivity of these new thiomalonamides for AN(III) extraction. (Author)

  14. Adsorption of actinides by chelating agents containing benzene rings, fixed on charcoal

    International Nuclear Information System (INIS)

    Valentini Ganzerli, M.T.; Crespi Caramella, V.; Maggi, L.

    1999-01-01

    The focus of this paper is on the analysis of the actinides in the hydrosphere to study their environmental dispersion. The 8-hydroxyquinoline family and the benzohydroxamic acid have a complexing ability towards the actinides, even if in different oxidation states. Taking advantage of this ability, their salts with some organic acids or bases were prepared. In this way compounds were obtained easily incorporated into active charcoal. Only a small amount of the prepared adsorber may be equilibrated with large sample volumes. Subsequently it can be recovered by filtration. The adsorbed ions may be then re-dissolved with a small volume of the appropriate eluting solution. The 8-hydroxy-quinolines and the 8-hydroxyquinoline produced salts with the benzilic acid. These compounds similarly behave and show wide adsorption coefficients for solutions of pH higher than 3. The adsorption takes place by means of the formation of a complex of the actinide ion with the hydroxyquinoline moiety and also with the benzilic anion. Provided that the active extracting agent is not dissolved in a medium but fixed into a solid phase, the whole adsorption process may be regarded as a solvent extraction reaction. The benzohydroxamic acid was treated with the diphenylamine or with the tribenzylamine to obtain salts, later adsorbed into the charcoal. The adsorption of actinide ions seems to take place by means of a precipitation mechanism of the actinide ions with the hydroxamate ions for solution of pH higher than 3.5. Also in this case high values were obtained for the distribution coefficients. The actinide ions act similarly in the +4 or +6 oxidation state towards the prepared adsorber series. Therefore, it is possible to use only one adsorber to concentrate all actinides. Methods of analysis of actinides in the environment may be suitably set up and the concentration step based on these new prepared adsorber may improve the whole procedure. (authors)

  15. Development of cermets for high-level radioactive waste fixation

    International Nuclear Information System (INIS)

    Aaron, W.S.; Quinby, T.C.; Kobisk, E.H.

    1979-01-01

    A method is currently under development for the solidification and fixation of commercial and defense high-level radioactive wastes in the form of ceramic particles encapsulated by metal, i.e., a cermet. The chemical and physical processing techniques which have been developed and the properties of the resulting cermet bodies are described in this paper. These cermets have the advantages of high thermal conductivity and low leach rates

  16. Actinide separative chemistry

    International Nuclear Information System (INIS)

    Boullis, B.

    2004-01-01

    Actinide separative chemistry has focused very heavy work during the last decades. The main was nuclear spent fuel reprocessing: solvent extraction processes appeared quickly a suitable, an efficient way to recover major actinides (uranium and plutonium), and an extensive research, concerning both process chemistry and chemical engineering technologies, allowed the industrial development in this field. We can observe for about half a century a succession of Purex plants which, if based on the same initial discovery (i.e. the outstanding properties of a molecule, the famous TBP), present huge improvements at each step, for a large part due to an increased mastery of the mechanisms involved. And actinide separation should still focus R and D in the near future: there is a real, an important need for this, even if reprocessing may appear as a mature industry. We can present three main reasons for this. First, actinide recycling appear as a key-issue for future nuclear fuel cycles, both for waste management optimization and for conservation of natural resource; and the need concerns not only major actinide but also so-called minor ones, thus enlarging the scope of the investigation. Second, extraction processes are not well mastered at microscopic scale: there is a real, great lack in fundamental knowledge, useful or even necessary for process optimization (for instance, how to design the best extracting molecule, taken into account the several notifications and constraints, from selectivity to radiolytic resistivity?); and such a need for a real optimization is to be more accurate with the search of always cheaper, cleaner processes. And then, there is room too for exploratory research, on new concepts-perhaps for processing quite new fuels- which could appear attractive and justify further developments to be properly assessed: pyro-processes first, but also others, like chemistry in 'extreme' or 'unusual' conditions (supercritical solvents, sono-chemistry, could be

  17. Cermet-fueled reactors for multimegawatt space power applications

    International Nuclear Information System (INIS)

    Cowan, C.L.; Armijo, J.S.; Kruger, G.B.; Palmer, R.S.; Van Hoomisson, J.E.

    1988-01-01

    The cermet-fueled reactor has evolved as a potential power source for a broad range of multimegawatt space applications. In particular, the fast spectrum reactor concept can be used to deliver 10s of megawatts of electric power for continuous, long term, unattended operation, and 100s of megawatts of electric power for times exceeding several hundred seconds. The system can also be utilized with either a gas coolant in a Brayton power conversion cycle, or a liquid metal coolant in a Rankine power conversion cycle. Extensive testing of the cermet fuel element has demonstrated that the fuel is capable of operating at very high temperatures under repeated thermal cycling conditions, including transient conditions which approach the multimegawatt burst power requirements. The cermet fuel test performance is reviewed and an advanced cermet-fueled multimegawatt nuclear reactor is described in this paper

  18. A cermet fuel reactor for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    Kruger, G.

    1991-01-01

    Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk

  19. Inherent protection of plutonium by doping minor actinide in thermal neutron spectra

    International Nuclear Information System (INIS)

    Peryoga, Yoga; Sagara, Hiroshi; Saito, Masaki; Ezoubtchenko, Alexey

    2005-01-01

    The present study focuses on the exploration of the effect of minor actinide (MA) addition into uranium oxide fuels of different enrichment (5% 235 U and 20% 235 U) as ways of increasing fraction of even-mass-number plutonium isotopes. Among plutonium isotopes, 238 Pu, 240 Pu and 242 Pu have the characteristics of relatively high decay heat and spontaneous fission neutron rate that can improve proliferation-resistant properties of a plutonium composition. Two doping options were proposed, i.e. doping of all MA elements (Np, Am and Cm) and doping of only Np to observe their effect on plutonium proliferation-resistant properties. Pressurized water reactor geometry has been chosen for fuels irradiation environment where irradiation has been extended beyond critical to explore the subcritical system potential. Results indicate that a large amount of MA doping within subcritical operation highly improves the proliferation-resistant properties of the plutonium with high total plutonium production. Doping of 1% MA or Np into 5% 235 U enriched uranium fuel appears possible for critical operation of the current commercial light water reactor with reasonable improvement in the plutonium proliferation-resistant properties. (author)

  20. Cermet fuels for space power systems

    International Nuclear Information System (INIS)

    Barner, J.O.; Coomes, E.P.; Williford, R.E.; Neimark, L.A.

    1986-01-01

    A refractory-metal matrix, UN-fueled cermet is a very promising fuel candidate for a wide range of multi-megawatt space reactor systems, e.g., steady-state, flexible duty-cycle, or bimodal, single- or two-phase liquid-metal cooled reactors, or thermionic reactors. Cermet fuel is especially promising for reactor designs that require operational strategies which incorporate rapid power changes because of its anticipated capability to withstand thermal shock

  1. The removal of actinide metals from solution

    International Nuclear Information System (INIS)

    Hancock, R.D.; Howell, I.V.

    1980-01-01

    A process is specified for removing actinide metals (e.g. uranium) from solutions. The solution is contacted with a substrate comprising the product obtained by reacting an inorganic solid containing surface hydroxyl groups (e.g. silica gel) with a compound containing a silane grouping, a nitrogen-containing group (e.g. an amine) and other specified radicals. After treatment, the actinide metal is recovered from the substrate. (U.K.)

  2. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    International Nuclear Information System (INIS)

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-01-01

    The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: (1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs; (2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs; (3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs; and (4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs

  3. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10

    The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

  4. [Comparative studies on fissure sealing: composite versus Cermet cement].

    Science.gov (United States)

    Hickel, R; Voss, A

    1989-06-01

    Fifty two molars sealed with either composite or Cermet cement were compared. The composite sealant was applied after enamel etching using a rubber dam. Before sealing with Cermet cement the enamel was only cleaned with pumice powder and sodium hypochlorie and the material was applied without enamel etching. After an average follow-up of 1.6 years composite sealants proved to be significantly more reliable. Cermet cement sealings showed defects more frequently.

  5. Stochastic Computer Simulation of Cermet Coatings Formation

    Directory of Open Access Journals (Sweden)

    Oleg P. Solonenko

    2015-01-01

    Full Text Available An approach to the modeling of the process of the formation of thermal coatings lamellar structure, including plasma coatings, at the spraying of cermet powders is proposed. The approach based on the theoretical fundamentals developed which could be used for rapid and sufficiently accurate prediction of thickness and diameter of cermet splats as well as temperature at interface “flattening quasi-liquid cermet particle-substrate” depending on the key physical parameters (KPPs: temperature, velocity and size of particle, substrate temperature, and concentration of finely dispersed solid inclusions uniformly distributed in liquid metal binder. The results are presented, which concern the development of the computational algorithm and the program complex for modeling the process of laying the splats in the coating with regard to the topology of its surface, which varies dynamically at the spraying, as well as the formation of lamellar structure and porosity of the coating. The results of numerical experiments are presented through the example of thermal spraying the cermet TiC-30 vol.% NiCr powder, illustrating the performance of the developed computational technology.

  6. Cermet electrode

    Science.gov (United States)

    Maskalick, Nicholas J.

    1988-08-30

    Disclosed is a cermet electrode consisting of metal particles of nickel, cobalt, iron, or alloys or mixtures thereof immobilized by zirconia stabilized in cubic form which contains discrete deposits of about 0.1 to about 5% by weight of praseodymium, dysprosium, terbium, or a mixture thereof. The solid oxide electrode can be made by covering a substrate with particles of nickel, cobalt, iron, or mixtures thereof, growing a stabilized zirconia solid oxide skeleton around the particles thereby immobilizing them, contacting the skeleton with a compound of praseodymium, dysprosium, terbium, or a mixture thereof, and heating the skeleton to a temperature of at least 500.degree. C. The electrode can also be made by preparing a slurry of nickel, cobalt, iron, or mixture and a compound of praseodymium, dysprosium, terbium, or a mixture thereof, depositing the slurry on a substrate, heating the slurry to dryness, and growing a stabilized zirconia skeleton around the metal particles.

  7. A cermet fuel reactor for nuclear thermal propulsion

    Science.gov (United States)

    Kruger, Gordon

    1991-01-01

    Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that the work done in the 1960's has demonstrated that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk.

  8. Iron-iron boride cermets - new P/M matrix composites

    International Nuclear Information System (INIS)

    Klimek, L.; Nowacki, J.

    1993-01-01

    Possibilities of producing Fe-Fe 2 B cermets as a result of sintering pure elements Fe and B in a vacuum have been analysed. Attempts of sintering in the solid phase and with the participation of the liquid phase - the Fe-Fe 2 B eutectic have been made. Various investigation of the cermets allowed determination of their structure as well as description of the kinetics of quantitative changes in phase proportions while sintering them. It has been found that its structure varies widely depending on sintering parameters and composition of the sinter. Measurements of Fe-Fe 2 B cermet hardness and wear during dry friction have shown distinct advantages of the cermets under investigation as constructional materials. (author). 10 refs, 6 figs

  9. Iron-iron boride cermets - new P/M matrix composites

    Energy Technology Data Exchange (ETDEWEB)

    Klimek, L.; Nowacki, J. [Politechnika Lodzka, Lodz (Poland)

    1993-12-31

    Possibilities of producing Fe-Fe{sub 2}B cermets as a result of sintering pure elements Fe and B in a vacuum have been analysed. Attempts of sintering in the solid phase and with the participation of the liquid phase - the Fe-Fe{sub 2}B eutectic have been made. Various investigation of the cermets allowed determination of their structure as well as description of the kinetics of quantitative changes in phase proportions while sintering them. It has been found that its structure varies widely depending on sintering parameters and composition of the sinter. Measurements of Fe-Fe{sub 2}B cermet hardness and wear during dry friction have shown distinct advantages of the cermets under investigation as constructional materials. (author). 10 refs, 6 figs.

  10. Calculational study for criticality safety data of fissionable actinides

    International Nuclear Information System (INIS)

    Nojiri, Ichiro; Fukasaku, Yasuhiro.

    1997-01-01

    This study has been carried out to obtain basic criticality safety characteristics of minor actinides nuclides. Criticality safety data of minor actinides nuclides have been surveyed through public literatures. Critical mass of seven nuclides, Np-237, Am-241, Am-242m, Am-243, Cm-243, Cm-244 and Cm-245, have been calculated by using two code systems of criticality safety analysis, SCALE-4 and MCNP4A, under some material and reflector conditions. Some applicable cross-section libraries have been used for each code systems. Calculated data have been compared with each other and with published data. The results of this comparison shows that there is no discrepancy within the computational codes and the calculated data is strongly depend on the cross-section library. (author)

  11. High flux transmutation of fission products and actinides

    International Nuclear Information System (INIS)

    Gerasimov, A.; Kiselev, G.; Myrtsymova, L.

    2001-01-01

    Long-lived fission products and minor actinides accumulated in spent nuclear fuel of power reactors comprise the major part of high level radwaste. Their incineration is important from the point of view of radwaste management. Transmutation of these nuclides by means of neutron irradiation can be performed either in conventional nuclear reactors, or in specialized transmutation reactors, or in ADS facilities with subcritical reactor and neutron source with application of proton accelerator. Different types of transmutation nuclear facilities can be used in order to insure optimal incineration conditions for radwaste. The choice of facility type for optimal transmutation should be based on the fundamental data in the physics of nuclide transformations. Transmutation of minor actinides leads to the increase of radiotoxicity during irradiation. It takes significant time compared to the lifetime of reactor facility to achieve equilibrium without effective transmutation. High flux nuclear facilities allow to minimize these draw-backs of conventional facilities with both thermal and fast neutron spectrum. They provide fast approach to equilibrium and low level of equilibrium mass and radiotoxicity of transmuted actinides. High flux facilities are advantageous also for transmutation of long-lived fission products as they provide short incineration time

  12. Final report on cermet high-level waste forms

    International Nuclear Information System (INIS)

    Kobisk, E.H.; Quinby, T.C.; Aaron, W.S.

    1981-08-01

    Cermets are being developed as an alternate method for the fixation of defense and commercial high level radioactive waste in a terminal disposal form. Following initial feasibility assessments of this waste form, consisting of ceramic particles dispersed in an iron-nickel base alloy, significantly improved processing methods were developed. The characterization of cermets has continued through property determinations on samples prepared by various methods from a variety of simulated and actual high-level wastes. This report describes the status of development of the cermet waste form as it has evolved since 1977. 6 tables, 18 figures

  13. Burn of actinides in MOX fuel cells; Quemado de actinidos en celdas de combustible MOX

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.; Ramirez S, J. R.; Alonso V, G., E-mail: eduardo.martinez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2017-09-15

    The spent fuel from nuclear reactors is stored temporarily in dry repositories in many countries of the world. However, the main problem of spent fuel, which is its high radio-toxicity in the long term, is not solved. A new strategy is required to close the nuclear fuel cycle and for the sustain ability of nuclear power generation, this strategy could be the recycling of plutonium to obtain more energy and recycle the actinides generated during the irradiation of the fuel to transmute them in less radioactive radionuclides. In this work we evaluate the quantities of actinides generated in different fuels and the quantities of actinides that are generated after their recycling in a thermal reactor. First, we make a reference calculation with a regular enriched uranium fuel, and then is changed to a MOX fuel, varying the plutonium concentrations and determining the quantities of actinides generated. Finally, different amounts of actinides are introduced into a new fuel and the amount of actinides generated at the end of the fuel burn is calculated, in order to determine the reduction of minor actinides obtained. The results show that if the concentration of plutonium in the fuel is high, then the production of minor actinides is also high. The calculations were made using the cell code CASMO-4 and the results obtained are shown in section 6 of this work. (Author)

  14. Duo_2-Steel cermet manufacturing technology for PWR Spent Nuclear Fuel (SNF) casks

    International Nuclear Information System (INIS)

    Siti Alimah; Budiarto

    2005-01-01

    Assessment of DUO_2-Steel cermet manufacturing technology for PWR SNF casks has been done. DUO_2-Steel cermet consisting of DUO_2 particulates and other particulates, embedded in a steel matrix. Cermet SNF casks have the potential for superior performance compared with casks constructed of other materials. The addition of DUO_2 ceramic particulates can increase SNF cask capacity, improve of repository performance and disposal of excess depleted uranium as potential waste. Two sets of cermet manufacturing technologies are casting and powder metallurgy. Three casting methods are infusion casting, traditional casting and centrifugal casting. While for powder metallurgy methods there are traditional method and new method. DUO_2-Steel cermet have traditionally been produced by powder metallurgy methods. The production of a cask, however, presents special requirements: the manufacture of an annular object with weights up to 100 tons, and methods are being not to manufacture a cermet of this size and geometry. A new powder metallurgy method, is a method for manufacturing cermet for PWR SNF cask. This powder metallurgy techniques have potentials low costs and provides greater freedom In the design of the cermet cask by allowing variable cermet properties. (author)

  15. Development and application of high strength ternary boride base cermets

    International Nuclear Information System (INIS)

    Takagi, Ken-ichi

    2006-01-01

    Reaction boronizing sintering is a novel strategy to form a ternary boride coexisting with a metal matrix in a cermet during liquid phase sintering. This new sintering technique has successfully developed world first ternary boride base cermets with excellent mechanical properties such as Mo 2 FeB 2 , Mo 2 NiB 2 and WCoB base ones. In these cermets Mo 2 FeB 2 and Mo 2 NiB 2 base ones consist of a tetragonal M 3 B 2 (M: metal)-type complex boride as a hard phase and a transition metal base matrix. The cermets have already been applied to wear resistant applications such as injection molding machine parts, can making tools, and hot copper extruding dies, etc. This paper focuses on the characteristics, effects of the additional elements on the mechanical properties and structure, and practical applications of the ternary boride base cermets. - Graphical abstract: TRS and hardness of Ni-5B-51Mo-17.5Cr and Ni-5B-51Mo-12.5Cr-5V-xMn mass% cermets as functions of Mn content (Fig. 17)

  16. Cermet cements.

    Science.gov (United States)

    McLean, J W

    1990-01-01

    Cermet ionomer cements are sintered metal/glass powders, which can be made to react with poly(acids). These new cements are significantly more resistant to abrasion than regular glass ionomer cements and are widely accepted as core build-up materials and lining cements. They can strengthen teeth and provide the clinician with an opportunity to treat early dental caries.

  17. Actinides compounds for the transmutation: scientific contributions of american and japanese collaborations

    International Nuclear Information System (INIS)

    Raison, Ph.; Albiot, T.

    2000-01-01

    This paper deals with the minor actinides transmutation and the scientific contribution of the ORNL and the JAERI. It presents researches on the Am-Zr-Y-O system in the framework of the heterogeneous reprocessing, the curium and pyrochlore structures, with the ORNL contribution and phase diagrams, data of Thermodynamics, actinides nitrides, with the JAERI. (A.L.B.)

  18. Evaluation on transmutation performance of minor actinides with high-flux BWR

    International Nuclear Information System (INIS)

    Setiawan, M.B.; Kitamoto, A.; Taniguchi, A.

    2001-01-01

    The performance of high-flux BWR (HFBWR) for burning and/or transmutation (B/T) treatment of minor actinides (MA) and long-lived fission products (LLFP) was discussed herein for estimating an advanced waste disposal with partitioning and transmutation (P and T). The concept of high-flux B/T reactor was based on a current 33 GWt-BWR, to transmute the mass of long-lived transuranium (TRU) to short-lived fission products (SLFP). The nuclide selected for B/T treatment was MA (Np-237, Am-241, and Am-243) included in the discharged fuel of LWR. The performance of B/T treatment of MA was evaluated by a new function, i.e. [F/T ratio], defined by the ratio of the fission rate to the transmutation rate in the core, at an arbitrary burn-up, due to all MA nuclides. According to the results, HFBWR could burn and/or transmute MA nuclides with higher fission rate than BWR, but the fission rate did not increase proportionally to the flux increment, due to the higher rate of neutron adsorption. The higher B/T fraction of MA would result in the higher B/T capacity, and will reduce the units of HFBWR needed for the treatment of a constant mass of MA. In addition, HFBWR had a merit of higher mass transmutation compared to the reference BWR, under the same mass loading of MA

  19. Plutonium and minor actinides recycling in PWRs with new APA concepts

    International Nuclear Information System (INIS)

    Golfier, H.; Rohart, M.; Aniel, S.; Bergeron, J.; Deffain, J.P.

    2001-01-01

    In the frame of the studies required by the French law of 1991, CEA have launched a wide range of assessments on waste management for different reactors (PWR, FBR). Considerable R and D work has already been performed in order to improve the use of Plutonium (Pu) in PWRs. In this context, the Advanced Plutonium Assembly (APA) aims to improve the use of Plutonium (Pu) in PWRs while minimizing Minor Actinides (MA) production, with only slight modifications of the core design. From a neutronic point of view, the overall studied cases lead to the stabilization of the Pu inventory with approximately 30% of the park refueled with APA assemblies in full APA cores. Multi-recycling could satisfy the stabilization of Pu+ (Am+Cm) inventory by the implementation of approximately 40% APA reactors in a conventional PWRs park. After 7 or 8 recycles, the equilibrium is reached. The Pu inventory in the fuel cycle ranges from 210 tons to 270 tons for Pu multi-recycling, and from 240 tons to 290 tons for Pu+(Am+Cm) multi-recycling. The saving in Natural Uranium and Separative Work Units (SWU) due to the use of APA reactors would be between 30% and 15% in comparison with the UO 2 open cycle. This paper presents a selection of the main preliminary Pu recycling results of the joint study program COGEMA-CEA. (author)

  20. Composite fuels. Defining a strategy for better Pu utilization in PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Porta, Jacques; Baldi, Stefano [DRN/DER/SIS CE Cadarache Bat. 211, 13108 St. Paul lez Durance CEDEX (France); Puill, Andre; Aillaud, Cecile

    1999-07-01

    After a description of the French context dominated by the accumulation of plutonium, and the international context in which the tendency is towards the disappearance of plutonium reserves and the destruction of minor actinides, a few solutions selected in the framework of the Innovating Fuels programme are presented. The aim is to burn the surplus Pu by means of fuel in inert matrices since, in addition to the increased burnup, the absence of plutonium generating conversion must be considered. The experiments intended to qualify selected CERCERs and CERMETs are presented and discussed, the feasibility of CERMET UO2 and MOX cores is established. Things are more complicated for PuO2 CERMET and we show that, in order to optimize loading, a dedicated heterogeneous assembly has to be defined. The Advanced Plutonium Assembly (APA) is presented. The last part of this article is devoted to the first fabrication tests on these very particular fuels and to the definition of reference accidents liable to affect 'cold' fuels such as CERMETs. (author)

  1. Composite fuels. Defining a strategy for better Pu utilization in PWRs

    International Nuclear Information System (INIS)

    Porta, Jacques; Baldi, Stefano; Puill, Andre; Aillaud, Cecile

    1999-01-01

    After a description of the French context dominated by the accumulation of plutonium, and the international context in which the tendency is towards the disappearance of plutonium reserves and the destruction of minor actinides, a few solutions selected in the framework of the Innovating Fuels programme are presented. The aim is to burn the surplus Pu by means of fuel in inert matrices since, in addition to the increased burnup, the absence of plutonium generating conversion must be considered. The experiments intended to qualify selected CERCERs and CERMETs are presented and discussed, the feasibility of CERMET UO2 and MOX cores is established. Things are more complicated for PuO2 CERMET and we show that, in order to optimize loading, a dedicated heterogeneous assembly has to be defined. The Advanced Plutonium Assembly (APA) is presented. The last part of this article is devoted to the first fabrication tests on these very particular fuels and to the definition of reference accidents liable to affect 'cold' fuels such as CERMETs. (author)

  2. The effect of corrosion product colloids on actinide transport

    International Nuclear Information System (INIS)

    Gardiner, M.P.; Smith, A.J.; Williams, S.J.

    1992-01-01

    The near field of the proposed UK repository for ILW/LLW will contain containers of conditioned waste in contact with a cementious backfill. It will contain significant quantities of iron and steel, Magnox and Zircaloy. Colloids deriving from their corrosion products may possess significant sorption capacity for radioelements. If the colloids are mobile in the groundwater flow, they could act as a significant vector for activity transport into the far field. The desorption of plutonium and americium from colloidal corrosion products of iron and zirconium has been studied under chemical conditions representing the transition from the near field to the far field. Desorption R d values of ≥ 5 x 10 6 ml g -1 were measured for both actinides on these oxides and hydroxides when actinide sorption took place under the near-field conditions and desorption took place under the far-field conditions. Desorption of the actinides occurred slowly from the colloids under far-field conditions when the colloids had low loadings of actinide and more quickly at high loadings of actinide. Desorbed actinide was lost to the walls of the experimental vessel. (author)

  3. The promise and challenges of cermet fueled nuclear thermal propulsion reactors

    International Nuclear Information System (INIS)

    Brengle, R.G.; Harty, R.B.; Bhattacharyya, S.K.

    1993-06-01

    The use of cermet fuels in nuclear thermal propulsion systems was examined and the characteristics of systems using these fuel forms is discussed in terms of current mission and safety requirements. For use at high temperatures cermet fueled reactors utilize ceramic fuels with refractory metals as the matrix material. Cermet fueled reactors tend to be heavy when compared to concepts that utilize graphite as the fuel matrix because of the high density of the refractory metal matrix which makes up 20-40 percent of the total volume. On the positive side the metal matrix is strong and more resistant to loads from either the launch or flow induced vibration. The compatibility of the tungsten cermet with hydrogen is excellent and lifetimes of several hours is certainly achievable. Probably the biggest drawback to cermet nuclear thermal propulsion concepts is that the amount of actual data to support the theoretical conclusions is small. In fact there is no data under representative conditions of temperature, propellant and flux for the required fuel burnup. Although cermet systems appear to be attractive, the lack of fuel data at representative conditions does not allow reliable comparisons of cermet systems to systems where fuel data is available. 10 refs

  4. Minor actinides transmutation performance in a fast reactor

    International Nuclear Information System (INIS)

    Takeda, Toshikazu

    2016-01-01

    Highlights: • A method for calculating MA transmutation for individual nuclides has been proposed by introducing two formulas of the MA transmutation. One corresponds to the difference of MA amounts, and the other corresponds to the sum of the fission amounts and the plutonium production amounts. • Using the method the MA transmutation was calculated for Np-237 and Am-241 in a fast reactor. The burnup period was changed from 1 year to 12 year. • For the 1 year burnup a large amount of Am-242m, Cm-242 are produced from Am-241. The total MA transmutation amount increases with burnup time, but its gradient with respect to burnup time decreases after 9 years, and the transmutation amount by overall fission increases almost linearly with burnup time. • However, after the 6 year burnup the fission contribution became large because of the large production of Pu isotopes from the original Am-241. • In addition to the homogeneous loading of the MA nuclides into the cores, a heterogeneous loading of Am-241 to the blanket region was considered. - Abstract: Results obtained in the project named “Study on Minor Actinides Transmutation using Monju Data”, which has been sponsored by the Ministry of Education, Culture, Sports, Science and Technology in Japan (MEXT) are described. In order to physically understand transmutation of individual MA nuclides in fast reactors, a new method was developed in which the MAs transmutation is interpreted by two formulas. One corresponds to the difference of individual MA nuclides amounts before and after a burnup period, and the other is the sum of amount of fission of a relevant MA nuclide and the net plutonium production from the MA nuclide during a burnup period. The method has been applied to two fast reactors with MA fuels loaded in cores homogeneously and in a blanket region heterogeneously. Numerical results of MA transmutation for the two reactors are shown.

  5. Cermet Coatings for Solar Stirling Space Power

    Science.gov (United States)

    Jaworske, Donald A.; Raack, Taylor

    2004-01-01

    Cermet coatings, molecular mixtures of metal and ceramic are being considered for the heat inlet surface of a solar Stirling space power converter. This paper will discuss the solar absorption characteristics of as-deposited cermet coatings as well as the solar absorption characteristics of the coatings after heating. The role of diffusion and island formation, during the deposition process and during heating will also be discussed.

  6. Fission cross-section measurements on 233U and minor actinides at the CERN n-TOF facility

    International Nuclear Information System (INIS)

    Calviani, M.; Cennini, P.; Chiaveri, E.; Dahlfors, M.; Ferrari, A.; Herrera-Martinez, A.; Kadi, Y.; Sarchiapone, L.; Vlachoudis, V.; Colonna, N.; Terlizzi, R.; Abbondanno, U.; Marrone, S.; Belloni, F.; Fujii, K.; Moreau, C.; Aerts, G.; Andriamonje, S.; Berthoumieux, E.; Dridi, W.; Gunsing, F.; Pancin, J.; Perrot, L.; Plukis, A.; Alvarez, H.; Duran, I.; Paradela, C.; Alvarez-Velarde, F.; Cano-Ott, D.; Embid-Sesura, M.; Gonzalez-Romero, E.; Guerrero, C.; Martinez, T.; Vincente, M. C.; Andrzejewski, J.; Assimakopoulos, P.; Audouin, L.; David, S.; Ferrant, L.; Stephan, C.; Tassan-Got, L.; Badurek, G.; Jericha, E.; Leeb, H.; Oberhummer, H.; Pigni, M. T.; Baumann, P.; Kerveno, M.; Lukic, S.; Rudolf, G.; Becvar, F.; Calvino, F.; Capote, R.; Carrapico, C.; Chepel, V.; Ferreira-Marques, R.; Goncalves, I.; Lindote, A.; Lopes, I.; Neves, F.; Cortes, G.; Poch, A.; Pretel, C.; Couture, A.; Cox, J.; O'Brien, S.; Wiescher, M.; Dillmann, I.; Heil, M.; Kaeppeler, F.; Mosconi, M.; Plag, R.; Walter, S.; Wisshak, K.; Domingo-Pardo, C.; Eleftheriadis, C.; Furman, W.; Goverdovski, A.; Gramegna, F.; Mastinu, P.; Praena, J.; Haas, B.; Haight, R.; Igashira, M.; Karadimos, D.; Karamanis, D.; Ketlerov, V.; Koehler, P.; Konovalov, V.; Kossionides, E.; Krticka, M.; Lampoudis, C.; Lozano, M.; Marganiec, J.; Massimi, C.; Mengoni, A.; Milazzo, P. M.; Papachristodoulou, C.; Papadopoulos, C.; Patronis, N.; Pavlik, A.; Pavlopoulos, P.; Plompen, A.; Quesada, J.; Rauscher, T.; Reifarth, R.; Rubbia, C.; Rullhusen, P.; Salgado, J.; Santos, C.; Savvidis, I.; Tagliente, G.; Tain, J. L.; Tavora, L.; Vannini, G.; Vaz, P.; Ventura, A.; Villamarin, D.; Vlastou, R.; Voss, F.

    2010-01-01

    Neutron-induced fission cross-sections of minor actinides have been measured at the white neutron source n-TOF at CERN, Geneva. The studied isotopes include 233 U, interesting for Th/U based nuclear fuel cycles, 241, 243 Am and 245 Cm, relevant for transmutation and waste reduction studies in new generation fast reactors (Gen-IV) or Accelerator Driven Systems. The measurements take advantage of the unique features of the n-TOF facility, namely the wide energy range, the high instantaneous neutron flux and the low background. Results for the involved isotopes are reported from ∼30 meV to around 1 MeV neutron energy. The measurements have been performed with a dedicated Fission Ionization Chamber (FIC), relative to the standard cross-section of the 235 U fission reaction, measured simultaneously with the same detector. Results are here reported. (authors)

  7. Solubility of actinides and surrogates in nuclear glasses

    International Nuclear Information System (INIS)

    Lopez, Ch.

    2003-01-01

    The nuclear wastes are currently incorporated in borosilicate glass matrices. The resulting glass must be perfectly homogeneous. The work discussed here is a study of actinide (thorium and plutonium) solubility in borosilicate glass, undertaken to assess the extent of actinide solubility in the glass and to understand the mechanisms controlling actinide solubilization. Glass specimens containing; actinide surrogates were used to prepare and optimize the fabrication of radioactive glass samples. These preliminary studies revealed that actinide Surrogates solubility in the glass was enhanced by controlling the processing temperature, the dissolution kinetic of the surrogate precursors, the glass composition and the oxidizing versus reducing conditions. The actinide solubility was investigated in the borosilicate glass. The evolution of thorium solubility in borosilicate glass was determined for temperatures ranging from 1200 deg C to 1400 deg C.Borosilicate glass specimens containing plutonium were fabricated. The experimental result showed that the plutonium solubility limit ranged from 1 to 2.5 wt% PuO 2 at 1200 deg C. A structural approach based on the determination of the local structure around actinides and their surrogates by EXAFS spectroscopy was used to determine their structural role in the glass and the nature of their bonding with the vitreous network. This approach revealed a correlation between the length of these bonds and the solubility of the actinides and their surrogates. (author)

  8. Fabrication of High Temperature Cermet Materials for Nuclear Thermal Propulsion

    Science.gov (United States)

    Hickman, Robert; Panda, Binayak; Shah, Sandeep

    2005-01-01

    Processing techniques are being developed to fabricate refractory metal and ceramic cermet materials for Nuclear Thermal Propulsion (NTP). Significant advances have been made in the area of high-temperature cermet fuel processing since RoverNERVA. Cermet materials offer several advantages such as retention of fission products and fuels, thermal shock resistance, hydrogen compatibility, high conductivity, and high strength. Recent NASA h d e d research has demonstrated the net shape fabrication of W-Re-HfC and other refractory metal and ceramic components that are similar to UN/W-Re cermet fuels. This effort is focused on basic research and characterization to identify the most promising compositions and processing techniques. A particular emphasis is being placed on low cost processes to fabricate near net shape parts of practical size. Several processing methods including Vacuum Plasma Spray (VPS) and conventional PM processes are being evaluated to fabricate material property samples and components. Surrogate W-Re/ZrN cermet fuel materials are being used to develop processing techniques for both coated and uncoated ceramic particles. After process optimization, depleted uranium-based cermets will be fabricated and tested to evaluate mechanical, thermal, and hot H2 erosion properties. This paper provides details on the current results of the project.

  9. ACTINET: a European Network for Actinide Sciences

    International Nuclear Information System (INIS)

    Bernard Boullis; Pascal Chaix

    2006-01-01

    Full text of publication follows: The research in Actinide sciences appear as a strategic issue for the future of nuclear systems. Sustainability issues are clearly in connection with the way actinide elements are managed (either addressing saving natural resource, or decreasing the radiotoxicity of the waste). The recent developments in the field of minor actinide P and T offer convincing indications of what could be possible options, possible future processes for the selective recovery of minor actinides. But they point out, too, some lacks in the basic understanding of key-issues (such as for instance the control An versus Ln selectivity, or solvation phenomena in organic phases). Such lacks could be real obstacles for an optimization of future processes, with new fuel compounds and facing new recycling strategies. This is why a large and sustainable work appears necessary, here in the field of basic actinide separative chemistry. And similar examples could be taken from other aspects of An science, for various applications (nuclear fuel or transmutation targets design, or migration issues,): future developments need a strong, enlarged, scientific basis. The Network ACTINET, established with the support of the European Commission, has the following objectives: - significantly improve the accessibility of the major actinide facilities to the European scientific community, and form a set of pooled facilities, as the corner-stone of a progressive integration process, - improve mobility between the member organisations, in particular between Academic Institutions and National Laboratories holding the pooled facilities, - merge part of the research programs conducted by the member institutions, and optimise the research programs and infrastructure policy via joint management procedures, - strengthen European excellence through a selection process of joint proposals, and reduce the fragmentation of the community by putting critical mass of resources and expertise on

  10. Innovative SANEX process for trivalent actinides separation from PUREX raffinate

    International Nuclear Information System (INIS)

    Sypula, Michal

    2013-01-01

    Recycling of nuclear spent fuel and reduction of its radiotoxicity by separation of long-lived radionuclides would definitely help to close the nuclear fuel cycle ensuring sustainability of the nuclear energy. Partitioning of the main radiotoxicity contributors followed by their conversion into short-lived radioisotopes is known as partitioning and transmutation strategy. To ensure efficient transmutation of the separated elements (minor actinides) the content of lanthanides in the irradiation targets has to be minimised. This objective can be attained by solvent extraction using highly selective ligands that are able to separate these two groups of elements from each other. The objective of this study was to develop a novel process allowing co-separation of minor actinides and lanthanides from a high active acidic feed solution with subsequent actinide recovery using just one cycle, so-called innovative SANEX process. The conditions of each step of the process were optimised to ensure high actinide separation efficiency. Additionally, screening tests of several novel lipophilic and hydrophilic ligands provided by University of Twente were performed. These tests were aiming in better understanding the influence of the extractant structural modifications onto An(III)/Ln(III) selectivity and complexation properties. Optimal conditions for minor actinides separation were found and a flow-sheet of a new innovative SANEX process was proposed. Tests using a single centrifugal contactor confirmed high Eu(III)/Am(III) separation factor of 15 while the lowest SF Ln/Am obtained was 6,5 (for neodymium). In addition, a new masking agent for zirconium was found as a substitution for oxalic acid. This new masking agent (CDTA) was also able to mask palladium without any negative influence on An(III)/Ln(III). Additional tests showed no influence of CDTA on plutonium present in the feed solution unlike oxalic acid which causes Pu precipitation. Therefore, CDTA was proposed as a Zr

  11. Innovative SANEX process for trivalent actinides separation from PUREX raffinate

    Energy Technology Data Exchange (ETDEWEB)

    Sypula, Michal

    2013-07-01

    Recycling of nuclear spent fuel and reduction of its radiotoxicity by separation of long-lived radionuclides would definitely help to close the nuclear fuel cycle ensuring sustainability of the nuclear energy. Partitioning of the main radiotoxicity contributors followed by their conversion into short-lived radioisotopes is known as partitioning and transmutation strategy. To ensure efficient transmutation of the separated elements (minor actinides) the content of lanthanides in the irradiation targets has to be minimised. This objective can be attained by solvent extraction using highly selective ligands that are able to separate these two groups of elements from each other. The objective of this study was to develop a novel process allowing co-separation of minor actinides and lanthanides from a high active acidic feed solution with subsequent actinide recovery using just one cycle, so-called innovative SANEX process. The conditions of each step of the process were optimised to ensure high actinide separation efficiency. Additionally, screening tests of several novel lipophilic and hydrophilic ligands provided by University of Twente were performed. These tests were aiming in better understanding the influence of the extractant structural modifications onto An(III)/Ln(III) selectivity and complexation properties. Optimal conditions for minor actinides separation were found and a flow-sheet of a new innovative SANEX process was proposed. Tests using a single centrifugal contactor confirmed high Eu(III)/Am(III) separation factor of 15 while the lowest SF{sub Ln/Am} obtained was 6,5 (for neodymium). In addition, a new masking agent for zirconium was found as a substitution for oxalic acid. This new masking agent (CDTA) was also able to mask palladium without any negative influence on An(III)/Ln(III). Additional tests showed no influence of CDTA on plutonium present in the feed solution unlike oxalic acid which causes Pu precipitation. Therefore, CDTA was proposed as

  12. Ultratrace analysis of transuranic actinides by laser-induced fluorescence

    Science.gov (United States)

    Miller, S.M.

    1983-10-31

    Ultratrace quantities of transuranic actinides are detected indirectly by their effect on the fluorescent emissions of a preselected fluorescent species. Transuranic actinides in a sample are coprecipitated with a host lattice material containing at least one preselected fluorescent species. The actinide either quenches or enhances the laser-induced fluorescence of the preselected fluorescent species. The degree of enhancement or quenching is quantitatively related to the concentration of actinide in the sample.

  13. An Advanced TALSPEAK Concept for Separating Minor Actinides. Part 2. Flowsheet Test with Actinide-spiked Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Wilden, Andreas [Forschungszentrum Jülich GmbH, Institut für Energie – und Klimaforschung – Nukleare Entsorgung und Reaktorsicherheit (IEK-6), Jülich, Germany; Lumetta, Gregg J. [Nuclear Science and Engineering Group, Pacific Northwest National Laboratory, Richland, DC, USA; Sadowski, Fabian [Forschungszentrum Jülich GmbH, Institut für Energie – und Klimaforschung – Nukleare Entsorgung und Reaktorsicherheit (IEK-6), Jülich, Germany; Schmidt, Holger [Forschungszentrum Jülich GmbH, Institut für Energie – und Klimaforschung – Nukleare Entsorgung und Reaktorsicherheit (IEK-6), Jülich, Germany; Schneider, Dimitri [Forschungszentrum Jülich GmbH, Institut für Energie – und Klimaforschung – Nukleare Entsorgung und Reaktorsicherheit (IEK-6), Jülich, Germany; Gerdes, Markus [Forschungszentrum Jülich GmbH, Institut für Energie – und Klimaforschung – Nukleare Entsorgung und Reaktorsicherheit (IEK-6), Jülich, Germany; Law, Jack D. [Aqueous Separations and Radiochemistry Department, Idaho National Laboratory, Idaho Falls, ID, USA; Geist, Andreas [Karlsruhe Institute of Technology (KIT), Institute for Nuclear Waste Disposal (INE), Karlsruhe, Germany; Bosbach, Dirk [Forschungszentrum Jülich GmbH, Institut für Energie – und Klimaforschung – Nukleare Entsorgung und Reaktorsicherheit (IEK-6), Jülich, Germany; Modolo, Giuseppe [Forschungszentrum Jülich GmbH, Institut für Energie – und Klimaforschung – Nukleare Entsorgung und Reaktorsicherheit (IEK-6), Jülich, Germany

    2017-08-17

    A solvent extraction system has been developed for separating trivalent actinides from lanthanides. This “Advanced TALSPEAK” system uses 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester to extract the lanthanides into a n-dodecane-based solvent phase, while the actinides are retained in a citrate-buffered aqueous phase by complexation to N-(2-hydroxyethyl)ethylenediamine-N,N',N'-triacetic acid. Batch distribution measurements indicate that the separation of americium from the light lanthanides decreases as the pH decreases. For example, the separation factor between La and Am increases from 2.5 at pH 2.0 to 19.3 at pH 3.0. However, previous investigations indicated that the extraction rates for the heavier lanthanides decrease with increasing pH. So, a balance between these two competing effects is required. An aqueous phase in which the pH was set at 2.6 was chosen for further process development because this offered optimal separation, with a minimum separation factor of ~8.4, based on the separation between La and Am. Centrifugal contactor single-stage efficiencies were measured to characterize the performance of the system under flow conditions.

  14. Corrosion of MTR type fuel plates containing U3O8-Al cermet cores

    International Nuclear Information System (INIS)

    Durazzo, M.

    1985-01-01

    The fuel plate samples containing U 3 O 8 -Al cermet cores with concentrations from 10 to 90% of U 3 O 8 weight were fabricated. Samples with 58% of U 3 O 8 eight were fabricated using compacts with densities from 75 to 95% of theoretical density. The influences of U 3 O 8 concentration and porosity of compacted core on porosity and uniformity of core thickness are discussed. The U 3 O 8 -Al cores were submitted to corrosion tests and exposed to deionized water at temperatures of 30, 50, 70 and 90 0 C by cladding deffect produced artificially. The results shown that core corrosion is accompanied by hydrogen release. The total volum of released hydrogen and the time interval to observe the initiation of hydrogen releasing (incubation time) are depending on core pososity and absolute temperature. A mechanism for U 3 O 8 -Al core corrosion process is proposed and discussed. The cladding of fuel plate samples was submitted to corrosion tests under similar conditons of the IAE-R1 reactor operating at 2, 5 and 10 MW. (Author) [pt

  15. CERMET fuel behavior and properties in ADS reactors

    International Nuclear Information System (INIS)

    Haas, D.; Fernandez, A.; Staicu, D.; Somers, J.; Maschek, W.; Liu, P.; Chen, X.

    2008-01-01

    Within the EUROTRANS Integrated Project, Forschungszentrum Karlsruhe (FZK) and the Institute for Transuranium Elements (ITU) are joining their efforts to study the behavior of Mo-based CERMET non-uranium fuel for the ADS. Contributions include core safety calculations, and fuel property measurements and irradiation experiments. Safety studies for optimized EFIT core designs have concluded that, for the new low power cores of EFIT with a power class of ∼400 MWth and a fuel power density of ∼250 MW/m 3 , the CERMET-loaded cores behave favorably and the design limits of the fuels were not violated. Mo-based CERMET fuel pellets and pins loaded with Pu and Am were fabricated for irradiation programmes which will start by mid-2007 in PHENIX (France) and HFR-Petten (The Netherlands). The thermal diffusivity and specific heat of the CERMET fuels (loaded with Pu and Am) were the main properties measured, and the thermal conductivity was deduced. The results were used to prepare the safety report for the irradiation experiments

  16. Neutronics design for lead-bismuth cooled accelerator-driven system for transmutation of minor actinide

    International Nuclear Information System (INIS)

    Tsujimoto, Kazufumi; Sasa, Toshinobu; Nishihara, Kenji; Oigawa, Hiroyuki; Takano, Hideki

    2004-01-01

    Neutronics design study was performed for lead-bismuth cooled accelerator-driven system (ADS) to transmute minor actinides. Early study for ADS indicated two problems: a large burnup reactivity swing and a significant peaking factor. To solve these problems, effect of design parameters on neutronics characteristics were searched. The design parameters were initial plutonium loading, buffer region between spallation target and core, and zone fuel loading. Parametric survey calculations were performed considering fuel cycle consisting of burnup and recycle. The results showed that burnup reactivity swing depends on the plutonium fraction in the initial fuel loading, and the lead-bismuth buffer region and the two-zone loading were effective for solving the problems. Moreover, an optimum value for the effective multiplication factor was also evaluated using reactivity coefficients. From the result, the maximum allowable value of the effective multiplication factor for a practical ADS can be set at 0.97. Consequently, a new core concept combining the buffer region and the two-zone loading was proposed base on the results of the parametric survey. (author)

  17. The OSMOSE program for the qualification of integral cross sections of actinides: Preliminary results in a PWR-UOx spectrum

    Energy Technology Data Exchange (ETDEWEB)

    Hudelot, J. P. [CEA Cadarache, DEN/DER, 13108 Saint Paul lez Durance (France); Klann, R. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Antony, M.; Bernard, D.; Fougeras, P. [CEA Cadarache, DEN/DER, 13108 Saint Paul lez Durance (France); Jorion, F.; Drin, N.; Donnet, L.; Leorier, C. [CEA VALRHO, DEN/DRCP, BP171, 30207 Bagnols-sur-Ceze Cedex (France); Zhong, Z. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2006-07-01

    The need for improved nuclear data for minor actinides has been stressed by various organizations throughout the world - especially for studies relating to plutonium management, waste incineration, transmutation of waste, and Pu burning in future nuclear concepts. Several international programs have indicated a strong desire to obtain accurate integral reaction rate data for improving the major and minor actinides cross sections. Data on major actinides (i.e. {sup 235}U, {sup 236}U, {sup 238}U, {sup 239}Pu, {sup 240}Pu, {sup 241}Pu, {sup 242}Pu and {sup 241}Am) are reasonably well-known and available in the Evaluated Nuclear Data Files (JEFF, JENDL, ENDF-BX However information on the minor actinides (i.e. {sup 232}Th, {sup 233}U, {sup 237}Np, {sup 238}Pu, {sup 242}Am, {sup 243}Am, {sup 243}Cm, {sup 235}Cm, {sup 244}Cm, {sup 245}Cm, {sup 246}Cm and {sup 247}Cm) is less well-known and considered to be relatively poor in some cases, having to rely on model and extrapolation of few data points. In this framework, the ambitious OSMOSE program between the Commissariat a l'Energie Atomique (CEA), Electricite de France (EDF) and the U.S. Dept. of Energy (DOE) has been undertaken with the aim of measuring the integral absorption rate parameters of actinides in the MINERVE experimental facility located at the CEA Cadarache Research Center. The OSMOSE Program (Oscillation in Minerve of isotopes in 'Eupraxic' Spectra) includes a complete analytical program associated with the experimental measurement program and aims at understanding and resolving potential discrepancies between calculated and measured values. In the OSMOSE program, the reactivity worth of samples containing separated actinides are measured in different neutron spectra using an oscillation technique with an overall expected accuracy better than 3%. Reactivity effects of less than 10 pcm (0.0001 or approximately 1.5 cents) are measured and compared with calibrations to determine the differential

  18. Actinide recycling in reactors

    International Nuclear Information System (INIS)

    Kuesters, H.; Wiese, H.W.; Krieg, B.

    1995-01-01

    The objective is an assessment of the transmutation of long-lived actinides and fission products and the incineration of plutonium for reducing the risk potential of radioactive waste from reactors in comparison to direct waste disposal. The contribution gives an interim account on homogeneous and heterogeneous recycling of 'risk nuclides' in thermal and fast reactors. Important results: - A homogeneous 5 percent admixture of minor actinides (MA) from N4-PWRs to EFR fuel would allow a transmutation not only of the EFR MA, but in addition of the MA from 5 or 6 PWRs of equal power. However, the incineration is restricted by safety considerations. - LWR have only a very low MA incineration potential, due to their disadvantageous neutron capture/fission ratio. - In order to keep the Cm inventory at a low level, it is advantageous to concentrate the Am heterogeneously in particular fuel elements or rods. (orig./HP)

  19. Corrosion of cermet cores of fuel plates for nuclear research reactor

    International Nuclear Information System (INIS)

    Durazzo, M.; Ramanathan, L.V.

    1984-01-01

    Materials Testing Reactor (MTR) type fuel plates containing U 3 O 8 -Al cores and clad with Al are used in various research reactor. Preliminary investigations, where in the cladding of samples was drilled to simulate conditions of rupture due to pitting attack, revealed that considerable quantities of H 2 was evolved upon exposure of the core to water. The corrosion of cermets cores of different densities was characterized as a function of H 2 evolution that revealed 3 stages. A first stage consisting of an incubation period followed by initiation of H 2 evolution, a second stage with a constant rate of H 2 evolution and a third stage with a low rate of H 2 evolution. All 3 stages were found to vary as a function of cermet density and water temperature. (Author) [pt

  20. ACTRAN: a code for depletion calculations in PWR cores aiming the production of minor actinide for using in ADS

    International Nuclear Information System (INIS)

    Santos, Rubens Souza dos

    2009-01-01

    Despite of the renewed willing to accept nuclear power as a mean of mitigate the climate changing, to deal with the long lived waste still cause some concerning in relation to maintain in safety condition, during so many years. A technological solution to overcome this leg of time is to use a facility that burn these waste, besides to generate electricity. This is the idea built in the accelerator driven systems (ADS). This technology is being though to use some minor actinides (MAs) as fuel. This work presents a program to assess actinide concentrations, aiming a fertile-free fuel to be used in the future ADS technology. For that, use was made of a numerical code to solve the steady-state multigroup diffusion equation 3D to calculate the neutron fluxes, coupled it with a new code to solve, also numerically, depletion equations, named ACTRAN code. This paper shows the simulation of a PWR core during the residence time of the nuclear fuel, for three years, and after, for almost four hundred years, to assess the MAs production. The results show some insight in the best management to get a minimum amount of some MAs to use in the future generations of ADS. (author)

  1. High temperature oxidation resistance of (Ti,Ta)(C,N)-based cermets

    International Nuclear Information System (INIS)

    Chicardi, E.; Córdoba, J.M.; Gotor, F.J.

    2016-01-01

    Highlights: • Cermets based on (Ti,Ta)(C,N) were oxidized in air between 800 and 1100 °C for 48 h. • The substitution of Ti by Ta resulted in a high resistance to oxidation. • A protective layer of cobalt titanates at the surface of cermets was observed. • A rutile phase in which some Ti"4"+ are replaced by Ta"5"+ was detected. • This replacement decelerated the oxygen diffusion into the cermets. - Abstract: Cermets based on titanium–tantalum carbonitride were oxidized in static air between 800 °C and 1100 °C for 48 h. The thermogravimetric and microstructural study showed an outstanding reduction in the oxidation of more than 90% when the Ta content was increased. In cermets with low Ta content, the formation of a thin CoO/Co_3O_4 outer layer tends to disappear by reacting with the underlying rutile phase, which emerges at the surface. However, in cermets with higher Ta content, the formation of an external titanate layer, observed even at a low temperature, appears to prevent the oxygen diffusion and the oxidation progression.

  2. Measurements of the neutron capture cross sections and incineration potentials of minor-actinides in high thermal neutron fluxes: Impact on the transmutation of nuclear wastes; Mesures des sections efficaces de capture et potentiels d'incineration des actinides mineurs dans les hauts flux de neutrons: Impact sur la transmutation des dechets

    Energy Technology Data Exchange (ETDEWEB)

    Bringer, O

    2007-10-15

    This thesis comes within the framework of minor-actinide nuclear transmutation studies. First of all, we have evaluated the impact of minor actinide nuclear data uncertainties within the cases of {sup 241}Am and {sup 237}Np incineration in three different reactor spectra: EFR (fast), GT-MHR (epithermal) and HI-HWR (thermal). The nuclear parameters which give the highest uncertainties were thus highlighted. As a result of fact, we have tried to reduce data uncertainties, in the thermal energy region, for one part of them through experimental campaigns in the moderated high intensity neutron fluxes of ILL reactor (Grenoble). These measurements were focused onto the incineration and transmutation of the americium-241, the curium-244 and the californium-249 isotopes. Finally, the values of 12 different cross sections and the {sup 241}Am isomeric branching ratio were precisely measured at thermal energy point. (author)

  3. Magnesium Cermets and Magnesium-Beryllium Alloys; Cermets au magnesium et au magnesium-beryllium; Metallokeramicheskie magnievye i magnievo-berillievye splavy; Cermets de magnesio y aleaciones de magnesio y berillio

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, V. E.; Zelenskij, V. F.; Fajfer, S. I.; Zhdanov, S. M.; Maksimenko, V. I.; Savchenko, V. I. [Fiziko-Tekhnicheskij Institut an USSR, Khar' kov, SSSR (Russian Federation)

    1963-11-15

    The paper describes some results of work on the development of magnesium-magnesium oxide cermets and of super heat-resistant magnesiumberyllium alloys produced by powder metallurgical methods. The introduction of even a minute quantity of finely dispersed magnesium oxide into magnesium results in a strengthening of the material, the degree of which increases with increased magnesium oxide concentration, although variation of this concentration within the limits of 0.3 to 5 wt.% has a comparatively slight effect on the corresponding variation in the short-term strength over the whole range of temperatures investigated. At 20{sup o}C, in the case of the cermets, {sigma}{sub {beta}} = 28 to 31 kg/mm{sup 2} and {delta} = 3 .5 to 4.5%; at 500{sup o}C {sigma}{sub {beta}} = 2.6 to 3.2 kg/mm{sup 2} and {delta} =30 to 40%. The positive effect of the finely dispersed oxide phase is particularly evident in protracted tests. For magnesium cermets, {sigma} (300)/100 = 2.2 kg/mm{sup 2}. Characteristic of the mixtures is the high thermal stability of the strength properties, linked chiefly with the thermodynamic stability of the strength-giving oxide phase in the metal matrix. The use of powder metallurgical methods has yielded super heat-resistant magnesium-beryllium alloys containing heightened concentrations of beryllium (PMB alloys). In their strength characteristics PMB alloys are close to Mg-MgO cermets, but the magnesium-beryllium alloys have a degree and duration of resistance to high temperature oxidation which exceeds the corresponding qualities of the magnesium alloys at present known. Thus, in air of 580{sup o}C, PMB alloys with 2 to 5% beryllium maintain a high resistance to oxidation for a period of over 12000 to 14000 h. This long-term heat resistance is chiefly a result of the amount of beryllium in the alloy, and increases with increasing beryllium content. PMB alloys are also marked by high resistance to short bursts of overheating. Magnesium cermets and

  4. Sintering of cermets on the base of corundum and molybdenum

    International Nuclear Information System (INIS)

    Fedotov, A.V.

    1987-01-01

    Liquid-phase sintering of cermets has been studied to develop rational technology allowing to produce a dense material at lower temperatures. Molybdenum of the MPCh mark with the specific surface ranged from 1900 to 4000 cm 2 /g and the corundum powder of the VK-94-1 mark with the specific surface of 6000 cm 2 /g containing upto 10% of the glass-phase have been used as initial materials. It is shown that application of the VK-94-1 ceramics powder for molybdenum content cermets allows to decrease the temperature of dense material production (∼ upto 100 deg C). To produce dense materials, it is necessary to restrict the initial porosity of compaction and to correspond it to the sintering conditions. The increase of molybdenum dispersion allows to produce material with the more homogeneous structure, higher density and strength. Molybdenum presence decreases recrystallization of corundum crystals and causes structure production resistant to high-temperature heating

  5. Micro-scale mechanical characterization of Inconel cermet coatings deposited by laser cladding

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Ch.; Verdi, D.; Garrido, M.A.; Ruiz-Hervias, J.

    2016-07-01

    In this study, an Inconel 625-Cr3C2 cermet coating was deposited on a steel alloy by laser cladding. The elastic and plastic mechanical properties of the cermet matrix were studied by the depth sensing indentation (DSI) in the micro scale. These results were compared with those obtained from an Inconel 600 bulk specimen. The values of Young's modulus and hardness of cermet matrix were higher than those of an Inconel 600 bulk specimen. Meanwhile, the indentation stress–strain curve of the cermet matrix showed a strain hardening value which was more than twice the one obtained for the Inconel 600 bulk. Additionally, the mechanical properties of unmelted Cr3C2 ceramic particles, embedded in the cermet matrix were also evaluated by DSI using a spherical indenter. (Author)

  6. Micro-scale mechanical characterization of Inconel cermet coatings deposited by laser cladding

    Directory of Open Access Journals (Sweden)

    Chao Chang

    2016-07-01

    Full Text Available In this study, an Inconel 625-Cr3C2 cermet coating was deposited on a steel alloy by laser cladding. The elastic and plastic mechanical properties of the cermet matrix were studied by the depth sensing indentation (DSI in the micro scale. These results were compared with those obtained from an Inconel 600 bulk specimen. The values of Young's modulus and hardness of cermet matrix were higher than those of an Inconel 600 bulk specimen. Meanwhile, the indentation stress–strain curve of the cermet matrix showed a strain hardening value which was more than twice the one obtained for the Inconel 600 bulk. Additionally, the mechanical properties of unmelted Cr3C2 ceramic particles, embedded in the cermet matrix were also evaluated by DSI using a spherical indenter.

  7. Light modulation in phase change disordered metamaterial - A smart cermet concept

    OpenAIRE

    Kumar , Sunil; Maury , Francis; Bahlawane , Naoufal

    2017-01-01

    International audience; Cermet coatings are popular solar selective absorbers as they allow capturing most of the solar energywhile minimising radiative losses. Embedded metallic nanoparticles in dielectric matrices promotemultiple internal reflection of light and provide an overall low emissivity. VO2 in the metamaterial stateis regarded in this study as a responsive mixed phase comprising metallic rutile VO2 inclusions insemiconducting monoclinic VO2 phase mimicking cermet. The smart cermet...

  8. Saturation of cermets based on titanium carbide and diboride by metal melts

    International Nuclear Information System (INIS)

    Kitsaj, A.A.; Tsyganova, T.V.; Ordan'yan, S.S.

    1985-01-01

    Different sintered composites - TiC-Ni(Mo), TiC-Fe (Ni), TiB 2 -Fe (Mo) are studied for their interaction in contact with metal melts at the temperature of liquid phase existence in the cermet. Due to structural and physicochemical similarity of cermets the processes occuring with contact interaction are identical: additional quantity of liquid is imbibed into the cermet resulting in reconstruction of the solid phase frame and volumetric growth of the specimen. Elongation of the specimens permits concluding that the intensity of the solid phase (frame) reconstruction process in the cermet TiC-Fe (Ni) is lower than in TiC-Ni (Mo) and TiB 2 -Fe (Mo) systems. In the TiC-Fe (Ni) cermet it causes prevalence of the processes of diffusional levellng for compositions of the metal-binder and contacting metal over the process of laminar flow of the melt into the specimen. Choosing the composite components it is possible to control intensity of the cermet saturation by the additional quantity of the melt and distribution of the liquid phase in the article volume

  9. Cermet insert high voltage holdoff for ceramic/metal vacuum devices

    Science.gov (United States)

    Ierna, William F.

    1987-01-01

    An improved metal-to-ceramic seal is provided wherein the ceramic body of the seal contains an integral region of cermet material in electrical contact with the metallic member, e.g., an electrode, of the seal. The seal is useful in high voltage vacuum devices, e.g., vacuum switches, and increases the high-voltage holdoff capabilities of such devices. A method of fabricating such seals is also provided.

  10. Nuclear waste forms for actinides

    Science.gov (United States)

    Ewing, Rodney C.

    1999-01-01

    The disposition of actinides, most recently 239Pu from dismantled nuclear weapons, requires effective containment of waste generated by the nuclear fuel cycle. Because actinides (e.g., 239Pu and 237Np) are long-lived, they have a major impact on risk assessments of geologic repositories. Thus, demonstrable, long-term chemical and mechanical durability are essential properties of waste forms for the immobilization of actinides. Mineralogic and geologic studies provide excellent candidate phases for immobilization and a unique database that cannot be duplicated by a purely materials science approach. The “mineralogic approach” is illustrated by a discussion of zircon as a phase for the immobilization of excess weapons plutonium. PMID:10097054

  11. Optical and electrochromic properties of Sn:WO3 cermets

    International Nuclear Information System (INIS)

    Ashrit, P.V.; Bader, G.; Girouard, F.E.; Truong, V.V.

    1989-01-01

    This paper discusses optical and electrochromic properties of Sn:WO 3 cermets deposited by alternate layer thermal deposition. These cermets exhibit electrical and optical behavior in the as deposited state. The inclusion of Sn in the WO 3 matrix enhances the Electrical conductivity of the system and renders them fairly transparent in the visible region. The electrochromic behavior of such systems is studied under both proton and Li + ion injection. The good conductivity and good transmission combined with good electrochromic characteristics of these systems indicate the possibility of utilizing this type of cermet for the dual role of transparent conductor (TC) and electrochromic (EC) layer

  12. Glass-ionomer-silver-cermet interim Class I restorations for permanent teeth.

    Science.gov (United States)

    Croll, T P; Killian, C M

    1992-11-01

    Glass-ionomer-silver-cermet cement has proved to be a worthy alternative to silver amalgam for restoring certain Class I lesions in primary teeth. Such restorations are now known to last up to 8 years without need for repair or replacement. Cermet cement has also been used for interim restoration of permanent teeth in special cases, with ideal results. The procedure for placing a glass-ionomer-silver-cermet cement Class I restoration is described.

  13. Material attractiveness of plutonium composition on doping minor actinide of large FBR

    International Nuclear Information System (INIS)

    Permana, Sidik; Suzuki, Mitsutoshi; Kuno, Yusuke

    2011-01-01

    Material attractiveness analysis on isotopic plutonium compositions of fast breeder reactors (FBR) has been investigated based on figure of merit (FOM) formulas as key parameters as well as decay heat (DH) and spontaneous fission neutron (SFN) compositions. Increasing minor actinide (MA) doping gives the significant effect to increase Pu-238 composition. However, the compositions of Pu-240 and Pu-242 become less with increasing MA doping. DH and SFN compositions in the core regions similar to the DH and SFN compositions of MOX-grade. Material attractiveness based on FOM1 formula shows all isotopic plutonium compositions in the blanket regions as well as in the core regions are categorized as high attractive material. Adopted FOM2 formula can distinguishes the material attractiveness levels which show the plutonium compositions in blanket regions as high attractiveness level and its composition in the core regions as low level of material attractiveness. MA doping is effective to reduce the material attractiveness level of blanket regions from high to medium and it requires much more MA doping rate to achieve low level of attractiveness (FOM<1) based on adopted FOM1 formula. Low material attractiveness level can be obtained by 4 % or more doping MA based on adopted FOM2 formula which considers not only DH composition effect, but also SFN composition effect that gives relatively higher contribution to material barrier of plutonium isotopes. (author)

  14. Protected plutonium breeding by transmutation of minor actinides in fast breeder reactor

    International Nuclear Information System (INIS)

    Meiliza, Yoshitalia; Saito, Masaki; Sagara, Hiroshi

    2008-01-01

    The improvement of proliferation resistance properties of Pu and the burnup characteristics of fast breeder reactor (FBR) had been studied by utilizing minor actinides (MAs) to produce more 238 Pu from 237 Np and 241 Am through neutron capture reaction. The higher the 238 Pu content in the fuel, the higher the proliferation resistance of the fuel would be owing to the natural characteristics of 238 Pu with high decay heat and high neutron production. The present paper deals with the assessment of passive measure against nuclear material proliferation by focusing on improving the inherent proliferation barrier of discharged Pu from an FBR. Results showed that 5% MA doping to the blanket of an FBR gives as high as 17-19% 238 Pu, which could be seen as a significant improvement of the proliferation properties of Pu. Moreover, additional 5% ZrH 2 , together with 5% MA doping to the blanket, could enhance the 238 Pu fraction much more (22-24%). With an assumption of protected Pu whose 238 Pu isotopic fraction is more than 12%, the present paper revealed that protected Pu could be produced more than the Pu consumed (protected Pu breeding) through incineration in an FBR with doping of a minimum 3% MAs or (2% MAs+5% ZrH 2 ) to the blanket. (author)

  15. Advanced core concepts with enhanced proliferation resistance by transmutation of minor actinides

    International Nuclear Information System (INIS)

    Saito, Masaki

    2005-01-01

    ''Protected Plutonium Production (P 3 )'' has been proposed to establish high burn-up cores and to produce protected with high proliferation resistance due to high decay heat and large number of spontaneous fission neutron of 238 Pu by the transmutation of Minor Actinides (MAs) which is presently treated as high-level waste. The burn-up calculations have shown that the advanced fuel with UO 2 (11-13% enrichment of 235 U) by doping 237 Np to produce 238 Pu in the commercialized large LWRs burn up to 100 GWd/t with 238 Pu to Pu ratio of about 20% which means the fuel is highly protected from proliferation. It was also predicted that medium or small size LWR cores with 15-17% enrichment, liquid metal cooled cores, and gas cooled cores added by 1-2% Np could achieve 100 GWd/t burning with bearing high proliferation resistance. The 237 Np mass balance calculations have revealed that more than 20 nuclear P 3 plants of 300 MWe could be supplied with enough 237 Np from the Japanese commercial plants in equilibrium fuel cycles. From the present studies, it is confirmed that MAs are treated as burnable and fertile materials not only to extend the core life but also to improve plutonium proliferation resistance of the future nuclear energy systems instead of their geological disposal or just their burning through fission. (author)

  16. Behavior of actinides in the Integral Fast Reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

    1994-06-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors` confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  17. Actinide behavior in the Integral Fast Reactor. Final project report

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C.

    1994-11-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  18. Actinide behavior in the Integral Fast Reactor. Final project report

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1994-11-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ( 237 Np, 240 Pu, 241 Am, and 243 Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

  19. Behavior of actinides in the Integral Fast Reactor fuel cycle

    International Nuclear Information System (INIS)

    Courtney, J.C.; Lineberry, M.J.

    1994-01-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ( 237 Np, 240 Pu, 241 Am, and 243 Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors' confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

  20. Use of fast reactors for actinide transmutation. Proceedings of a specialists meeting held in Obninsk, Russian Federation, 22-24 September 1992

    Energy Technology Data Exchange (ETDEWEB)

    1993-03-15

    The management of radioactive waste is one of the key issues in today`s discussions on nuclear energy, especially the long term disposal of high level radioactive wastes. The recycling of plutonium in liquid metal fast breeder reactors (LMFBRs) would allow `burning` of the associated extremely long life transuranic waste, particularly actinides, thus reducing the required isolation time for high level waste from tens of thousands of years to hundreds of years for fission products only. The International Working Group on Fast Reactors (IWGFR) decided to include the topic of actinide transmutation in liquid metal fast breeder reactors in its programme. The IAEA organized the Specialists Meeting on Use of Fast Breeder Reactors for Actinide Transmutation in Obninsk, Russian Federation, from 22 to 24 September 1992. The specialists agree that future progress in solving transmutation problems could be achieved by improvements in: Radiochemical partitioning and extraction of the actinides from the spent fuel (at least 98% for Np and Cm and 99.9% for Pu and Am isotopes); technological research and development on the design, fabrication and irradiation of the minor actinides (MAs) containing fuels; nuclear constants measurement and evaluation (selective cross-sections, fission fragments yields, delayed neutron parameters) especially for MA burners; demonstration of the feasibility of the safe and economic MA burner cores; knowledge of the impact of maximum tolerable amount of rare earths in americium containing fuels. Refs, figs and tabs.

  1. Study of the radiotoxicity of actinides recycling in boiling water reactors fuel

    International Nuclear Information System (INIS)

    Francois, J.L.; Guzman, J.R.; Martin-del-Campo, C.

    2009-01-01

    In this paper the production and destruction, as well as the radiotoxicity of plutonium and minor actinides (MA) obtained from the multi-recycling of boiling water reactors (BWR) fuel are analyzed. A BWR MOX fuel assembly, with uranium (from enrichment tails), plutonium and minor actinides is designed and studied using the HELIOS code. The actinides mass and the radiotoxicity of the spent fuel are compared with those of the once-through or direct cycle. Other type of fuel assembly is also analyzed: an assembly with enriched uranium and minor actinides; without plutonium. For this study, the fuel remains in the reactor for four cycles, where each cycle is 18 months length, with a discharge burnup of 48 MWd/kg. After this time, the fuel is placed in the spent fuel pool to be cooled during 5 years. Afterwards, the fuel is recycled for the next fuel cycle; 2 years are considered for recycle and fuel fabrication. Two recycles are taken into account in this study. Regarding radiotoxicity, results show that in the period from the spent fuel discharge until 1000 years, the highest reduction in the radiotoxicity related to the direct cycle is obtained with a fuel composed of MA and enriched uranium. However, in the period after few thousands of years, the lowest radiotoxicity is obtained using the fuel with plutonium and MA. The reduction in the radiotoxicity of the spent fuel after one or two recycling in a BWR is however very small for the studied MOX assemblies, reaching a maximum reduction factor of 2.

  2. The reprocessing-recycling of spent nuclear fuel. Actinides separation - Application to wastes management; Le traitement-recyclage du combustible nucleaire use. La separation des actinides - Application a la gestion des dechets

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-07-01

    After its use in the reactor, the spent fuel still contains lot of recoverable material for an energetic use (uranium, plutonium), but also fission products and minor actinides which represent the residues of nuclear reactions. The reprocessing-recycling of the spent fuel, as it is performed in France, implies the chemical separation of these materials. The development and the industrial implementation of this separation process represent a major contribution of the French science and technology. The reprocessing-recycling allows a good management of nuclear wastes and a significant saving of fissile materials. With the recent spectacular rise of uranium prices, this process will become indispensable with the development of the next generation of fast neutron reactors. This book takes stock of the present and future variants of the chemical process used for the reprocessing of spent fuels. It describes the researches in progress and presents the stakes and recent results obtained by the CEA. content: the separation of actinides, a key factor for a sustainable nuclear energy; the actinides, a discovery of the 20. century; the radionuclides in nuclear fuels; the aquo ions of actinides; some redox properties of actinides; some complexing properties of actinide cations; general considerations about treatment processes; some characteristics of nuclear fuels in relation with their reprocessing; technical goals and specific constraints of the PUREX process; front-end operations of the PUREX process; separation and purification operations of the PUREX process; elaboration of finite products in the framework of the PUREX process; management and treatment of liquid effluents; solid wastes of the PUREX process; towards a joint management of uranium and plutonium: the COEX{sup TM} process; technical options of treatment and recycling techniques; the fuels of generation IV reactors; front-end treatment processes of advanced fuels; hydrometallurgical processes for future fuel

  3. Cermet coatings for solar Stirling space power

    International Nuclear Information System (INIS)

    Jaworske, Donald A.; Raack, Taylor

    2004-01-01

    Cermet coatings, molecular mixtures of metal and ceramic, are being considered for the heat inlet surface of a solar Stirling space power convertor. The role of the cermet coating is to absorb as much of the incident solar energy as possible. The ability to mix metal and ceramic at the atomic level offers the opportunity to tailor the composition and the solar absorptance of these coatings. Several candidate cermet coatings were created and their solar absorptance was characterized as-manufactured and after exposure to elevated temperatures. Coating composition was purposely varied through the thickness of the coating. As a consequence of changing composition, islands of metal are thought to form in the ceramic matrix. Computer modeling indicated that diffusion of the metal atoms played an important role in island formation while the ceramic was important in locking the islands in place. Much of the solar spectrum is absorbed as it passes through this labyrinth

  4. The design of cermet fuel phase fraction and fuel particle diameter

    International Nuclear Information System (INIS)

    Tian Sheng.

    1986-01-01

    UO 2 -Zr-2 is an ideal cermet fuel. As an exemplification with this fuel, this paper emphatically elucidates the irradiation theory of cermet fuel and its application in the design of cermet fuel phase fraction and of fuel particle diameter. From the point of view of the irradiation theory and the consideration for sandwich rolling, the suitable volume fraction of UO 2 phase of 25% and diameter of UO 2 particle of 100 +- 15 μm are selected

  5. Investigations on cermet electrodes for thermionic emitters

    International Nuclear Information System (INIS)

    Schmidt, D.; Nazare, S.

    1975-01-01

    Unstable Ba 2 CaWO 6 -W with their own supply of Ba, as well as stable UO 2 -Mo-emitter cermets that have to be operated with an external Ba-source, have been prepared by axial hot pressing. The relevant properties of these cermets such as electrical resistivity and thermal expansion are reported and compared with theoretical predictions. The electron emission of these materials is discussed on the basis of the surface films formed. It provides the basis for optimising the behavior of these materials

  6. Current Progress in Solution Precursor Plasma Spraying of Cermets: A Review

    Directory of Open Access Journals (Sweden)

    Romnick Unabia

    2018-06-01

    Full Text Available Ceramic and metal composites, known also as cermets, may considerably improve many material properties with regards to that of initial components. Hence, cermets are frequently applied in many technological fields. Among many processes which can be employed for cermet manufacturing, thermal spraying is one of the most frequently used. Conventional plasma spraying of powders is a popular and cost-effective manufacturing process. One of its most recent innovations, called solution precursor plasma spraying (SPPS, is an emerging coating deposition method which uses homogeneously mixed solution precursors as a feedstock. The technique enables a single-step deposition avoiding the powder preparation procedures. The nanostructured coatings developed by SPPS increasingly find a place in the field of surface engineering. The present review shows the recent progress in the fabrication of cermets using SPPS. The influence of starting solution precursors, such as their chemistry, concentration, and solvents used, to the micro-structural characteristics of cermet coatings is discussed. The effect of the operational plasma spray process parameters such as solution injection mode to the deposition process and coatings’ microstructure is also presented. Moreover, the advantages of the SPPS process and its drawbacks compared to the conventional powder plasma spraying process are discussed. Finally, some applications of SPPS cermet coatings are presented to understand the potential of the process.

  7. The reprocessing-recycling of spent nuclear fuel. Actinides separation - Application to wastes management

    International Nuclear Information System (INIS)

    2008-01-01

    After its use in the reactor, the spent fuel still contains lot of recoverable material for an energetic use (uranium, plutonium), but also fission products and minor actinides which represent the residues of nuclear reactions. The reprocessing-recycling of the spent fuel, as it is performed in France, implies the chemical separation of these materials. The development and the industrial implementation of this separation process represent a major contribution of the French science and technology. The reprocessing-recycling allows a good management of nuclear wastes and a significant saving of fissile materials. With the recent spectacular rise of uranium prices, this process will become indispensable with the development of the next generation of fast neutron reactors. This book takes stock of the present and future variants of the chemical process used for the reprocessing of spent fuels. It describes the researches in progress and presents the stakes and recent results obtained by the CEA. content: the separation of actinides, a key factor for a sustainable nuclear energy; the actinides, a discovery of the 20. century; the radionuclides in nuclear fuels; the aquo ions of actinides; some redox properties of actinides; some complexing properties of actinide cations; general considerations about treatment processes; some characteristics of nuclear fuels in relation with their reprocessing; technical goals and specific constraints of the PUREX process; front-end operations of the PUREX process; separation and purification operations of the PUREX process; elaboration of finite products in the framework of the PUREX process; management and treatment of liquid effluents; solid wastes of the PUREX process; towards a joint management of uranium and plutonium: the COEX TM process; technical options of treatment and recycling techniques; the fuels of generation IV reactors; front-end treatment processes of advanced fuels; hydrometallurgical processes for future fuel cycles

  8. EUROPART: an European integrated project on actinide partitioning

    International Nuclear Information System (INIS)

    Madic, C.; Baron, P.; Hudson, M.J.

    2006-01-01

    Full text of publication follows: The EUROPART project is a scientific integrated project between 24 European partners, from 10 countries, mostly funded by the European Community within the FP6, together with CRIEPI from Japan and ANSTO from Australia. EUROPART aims at developing chemical partitioning processes for the so-called minor actinides (MA) contained in nuclear wastes, i.e. from Am to Cf. In the case of the treatment of dedicated spent fuels or targets, the actinides to be separated also include U, Pu and Np. The techniques considered for the separation of these radionuclides belong to the fields of hydrometallurgy and pyrometallurgy, as in the previous European FP5 programs named PARTNEW, CALIXPART and PYROREP, respectively. The two main axes of research within EUROPART are: 1/ the partitioning of MA (from Am to Cf) from wastes issuing from the reprocessing of high burn-up UOX fuels and multi-recycled MOX fuels, 2/ the partitioning of the whole actinide family of elements for recycling, as an option for advanced dedicated fuel cycles (this work will be connected to the studies to be performed within the EUROTRANS European integrated project). In hydrometallurgy, the research is organized in five Work Packages (WP). Four are dedicated to the study of partitioning methods mainly based on the use of solvent extraction methods and of solid extractants, one WP is dedicated to the development of actinide co-conversion methods for fuel or target preparations. The research in pyrometallurgy is organized into four WPs, listed hereafter: (i) study of the basic chemistry of transuranium elements and of some fission products in molten salts (chlorides, fluorides), (ii) development of actinide partitioning methods, (iii) study of the conditioning of the salt wastes, (iv) system studies. Moreover, a strong management team is concerned not only with the technical and financial issues arising from EUROPART, but also with information, communication and benefits for Europe

  9. Aspects of fretting wear of sprayed cermet coatings

    International Nuclear Information System (INIS)

    Chivers, T.C.

    1985-01-01

    Two experimental fretting programmes which investigated aspects of fretting wear of sprayed cermet coatings are reviewed. These programmes were conducted in support of components used in the advanced gas-cooled reactor. It is speculated that the results from these programmes are compatible with a simple two-stage wear model. This model assumes that an initial wear process occurs which is dominated by an interlocking and removal of asperities. Such a phase will be dependent on the superficial contact areas and possibly the interfacial load, but the latter aspect is not considered. This initial wear is of very short duration and is followed by a mild, oxidative, wear mode. Coatings data are also compared with those for structural steels. In short-term low temperature tests it appears that structural steels have comparable performance with the cermet coatings but it is argued that this is an artefact of the wear process. However, at high temperatures (600 0 C) wear of stainless steel could not be determined, the specimens showing a net weight gain. It is concluded that for in-reactor fretting applications cermet coatings will have advantages over structural steels at low temperatures. Even in high temperature regions some operation at low temperatures is experienced and consequently cermet coatings may be useful here also. (orig.)

  10. Characteristics and fabrication of cermet spent nuclear fuel casks: ceramic particles embedded in steel

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Swaney, P.M.; Tiegs, T.N.

    2004-01-01

    Cermets are being investigated as an advanced material of construction for casks that can be used for storage, transport, or disposal of spent nuclear fuel (SNF). Cermets, which consist of ceramic particles embedded in steel, are a method to incorporate brittle ceramics with highly desirable properties into a strong ductile metal matrix with a high thermal conductivity, thus combining the best properties of both materials. Traditional applications of cermets include tank armor, vault armor, drill bits, and nuclear test-reactor fuel. Cermets with different ceramics (DUO 2 , Al 2 O 3 , Gd 2 O 3 , etc.) are being investigated for the manufacture of SNF casks. Cermet casks offer four potential benefits: greater capacity (more SNF assemblies) for the same gross weight cask, greater capacity (more SNF assemblies) for the same external dimensions, improved resistance to assault, and superior repository performance. These benefits are achieved by varying the composition, volume fraction, and particulate size of the ceramic particles in the cermet with position in the cask body. Addition of depleted uranium dioxide (DUO 2 ) to the cermet increases shielding density, improves shielding effectiveness, and increases cask capacity for a given cask weight or size. Addition of low-density aluminium oxide (Al 2 O 3 ) to the outer top and bottom sections of the cermet cask, where the radiation levels are lower, can lower cask weight without compromising shielding. The use of Al2O3 and other oxides, in appropriate locations, can increase resistance to assault. Repository performance may be improved by compositional control of the cask body to (1) create a local geochemical environment that slows the long-term degradation of the SNF and (2) enables the use of DUO 2 for longterm criticality control. While the benefits of using cermets follow directly from their known properties, the primary challenge is to develop low-cost methods to fabricate casks with variable cermet compositions

  11. Actinide transmutation using inert matrix fuels versus recycle in a low conversion fast burner reactor

    Energy Technology Data Exchange (ETDEWEB)

    Deinert, M.R.; Schneider, E.A.; Recktenwald, G.; Cady, K.B. [The Department of Mechanical Engineering, The University of Texas at Austin, 1 University Station, C2200, Austin, 78712 (United States)

    2009-06-15

    Reducing the disposal burden of the long lived radioisotopes that are contained within spent uranium oxide fuel is essential for ensuring the sustainability of nuclear power. Because of their non-fertile matrices, inert matrix fuels (IMFs) could allow light-water reactors to achieve a significant burn down of plutonium and minor actinides that are that are currently produced as a byproduct of operating light-water reactors. However, the extent to which this is possible is not yet fully understood. We consider a ZrO{sub 2} based IMF with a high transuranic loading and show that the neutron fluence (and the subsequent fuel residence time required to achieve it) present a practical limit for the achievable actinide burnup. The accumulation of transuranics in spent uranium oxide fuel is a major obstacle for the sustainability of nuclear power. While commercial light-water reactors (LWR's) produce these isotopes, they can be used to transmute them. At present, the only viable option for doing this is to partly fuel reactors with mixed oxide fuel (MOX) made using recycled plutonium. However, because of parasitic neutron capture in the uranium matrix of MOX, considerable plutonium and minor actinides are also bred as the fuel is burned. A better option is to entrain the recycled isotopes in a non-fertile matrix such as ZrO{sub 2}. Inert matrices such as these were originally envisioned for burning plutonium from dismantled nuclear weapons [1]. However, because they achieve a conversion ratio of zero, they have also been considered as a better alternative to MOX [2-6]. Plutonium and minor actinides dominate the long term heat and radiological outputs from spent nuclear fuel. Recent work has shown that that IMFs can be used to reduce these outputs by at least a factor of four, on a per unit of energy generated basis [6]. The degree of reduction is strongly dependent on IMF burnup. In principle, complete transmutation of the transuranics could be achieved though this

  12. Results of experimental investigations for substantiation of WWER cermet fuel pin performance

    International Nuclear Information System (INIS)

    Popov, V.V.; Karpin, A.D.; Isupov, I.A.; Rumyantsev, V.N.; Troyanov, V.M.; Subonyaev, V.N.; Melnichenko, N.A.

    1997-01-01

    The out-of-pile experiment results on interaction of the cladding and matrix materials and uranium dioxide at cermet fuel temperature for normal operating conditions of the WWER-440 reactor are analyzed. Cermet fuel element behaviour under the maximum designed damage of the WWER-440 reactor is considered. In the AM reactor loop a fission product output from the unsealed cermet fuel elements have been studied. (author). 6 figs, 3 tabs

  13. In-Situ Optical Studies of Oxidation/Reduction Kinetics on SOFC Cermet Anodes

    Science.gov (United States)

    2010-12-28

    DATES COVERED (From - To) 1/29/10-9/30/10 4. TITLE AND SUBTITLE In situ optical studies of oxidation/reduction kinetics on SOFC cermet anodes 5a...0572 In-situ Optical Studies of Oxidation/Reduction Kinetics on SOFC Cermet Anodes Department of Chemistry and Biochemistry Montana State University...of Research In-situ Optical Studies of Oxidation/Reduction Kinetics on SOFC Cermet Anodes Principal Investigator Robert Walker Organization

  14. Cermet insert high voltage holdoff improvement for ceramic/metal vacuum devices

    Science.gov (United States)

    Ierna, W.F.

    1986-03-11

    An improved metal-to-ceramic seal is provided wherein the ceramic body of the seal contains an integral region of cermet material in electrical contact with the metallic member, e.g., an electrode, of the seal. The seal is useful in high voltage vacuum devices, e.g., vacuum switches, and increases the high-voltage holdoff capabilities of such devices. A method of fabricating such seals is also provided.

  15. The OSMOSE Experimental Program for the qualification of integral cross sections of actinides

    Energy Technology Data Exchange (ETDEWEB)

    Antony, Muriel; Hudelot, Jean-Pascal [CEA, Centre de Cadarache, F-13108 Saint Paul lez Durance (France); Klann, Raymond [Nuclear Engineering Division, Argonne. National Laboratory, 9700 South Cass Ave., Argonne, IL 60439-4814 (United States)

    2006-07-01

    The need of better nuclear data on minor actinides has been stressed by various organizations throughout the world. It especially deals with the studies on plutonium management and waste incineration in existing systems and transmutation of waste or Pu burning in future nuclear concepts. To address this issue, a Working Party of the OECD has been concerned with identifying these needs and has produced a detailed High Priority Request List for Nuclear Data. The first step in obtaining better nuclear data consists in measuring accurate integral data and comparing them to integrated energy dependent data: this comparison provides a direct assessment of the effect of deficiencies in the differential data. Several international programs have indicated a strong desire to obtain accurate integral reaction rate data for improving the major and minor actinides cross sections. Data on major actinides (i.e. {sup 235}U, {sup 236}U, {sup 238}U, {sup 239}Pu, {sup 240}Pu, {sup 241}Pu, {sup 242}Pu and {sup 241}Am) are reasonably well-known and available in the Evaluated Nuclear Data Files (JEFF, JENDL, ENDF-B). However information on the minor actinides (i.e. {sup 232}Th, {sup 233}U, {sup 237}Np, {sup 238}Pu, {sup 242}Am, {sup 243}Am, {sup 242}Cm, {sup 243}Cm, {sup 244}Cm, {sup 245}Cm, {sup 246}Cm and {sup 247}Cm) is less well-known and considered to be relatively poor in some cases, having to rely on model and extrapolation of few data points. In this framework, the ambitious OSMOSE program between the Commissariat a l'Energie Atomique (CEA), Electricite de France (EDF) and the U.S. Department of Energy (DOE) has been undertaken with the aim of measuring the integral absorption rate parameters of actinides in the MINERVE experimental facility located at the CEA Cadarache Research Center. The OSMOSE Program (Oscillation in Minerve of isOtopes in 'Eupraxic' Spectra) includes a complete analytical program associated with the experimental measurement program and aims

  16. A thermionic energy converter with a molybdenum-alumina cermet emitter

    NARCIS (Netherlands)

    Gubbels, G.H.M.; Wolff, L.R.; Metselaar, R.

    1990-01-01

    A study is made of the properties of cermets as electrode materials for thermionic energy converters. For thermodynamic reasons it is expected that all cermets composed of pure Mo and refractory oxides have the same bare work function. From data on the work function of Mo in an oxygen atmosphere

  17. Neutronics analysis of minor actinides transmutation in a fusion-driven subcritical system

    International Nuclear Information System (INIS)

    Yang, Chao; Cao, Liangzhi; Wu, Hongchun; Zheng, Youqi; Zu, Tiejun

    2013-01-01

    Highlights: • A fusion fission hybrid system for MA transmutation is proposed. • The analysis of neutronics effects on the transmutation is performed. • The transmutation rate of MA reaches 86.5% by 25 times of recycling. -- Abstract: The minor actinides (MAs) transmutation in a fusion-driven subcritical system is analyzed in this paper. The subcritical reactor is driven by a tokamak D-T fusion device with relatively easily achieved plasma parameters and tokamak technologies. The MAs discharged from the light water reactor (LWR) are loaded in transmutation zone. Sodium is used as the coolant. The mass percentage of the reprocessed plutonium (Pu) in the fuel is raised from 0 to 48% and stepped by 12% to determine its effect on the MAs transmutation. The lesser the Pu is loaded, the larger the MAs transmutation rate is, but the smaller the energy multiplication factor is. The neutronics analysis of two loading patterns is performed and compared. The loading pattern where the mass percentage of Pu in two regions is 15% and 32.9% respectively is conducive to the improvement of the transmutation fraction within the limits of burn-up. The final transmutation fraction of MAs can reach 17.8% after five years of irradiation. The multiple recycling is investigated. The transmutation fraction of MAs can reach about 61.8% after six times of recycling, and goes up to about 86.5% after 25

  18. Numerical Simulation of Brazing TiC Cermet to Iron with TiZrNiCu Filler Metal

    Institute of Scientific and Technical Information of China (English)

    Lixia ZHANG; Jicai FENG

    2004-01-01

    The maximum thermal stress and stress concentration zones of iron/TiC cermet joint during cooling were studied in this paper. The results showed that the shear stress on iron/TiC cermet joint concentrates on the interface tip and the maximum shear stress appears on the left tip of iron/TiZrNiCu interlace. Positive tensile stress on TiC cermet undersurface concentrates on both sides of TiC cermet and its value decreases during cooling. Negative tensile stress on TiC cermet undersurface concentrates on the center of TiC cermet and its value increases during cooling. Brazing temperature has little effect on the development and maximum thermal stress.

  19. Flexibility of ADS for minor actinides transmutation in different two-stage PWR-ADS fuel cycle scenarios

    International Nuclear Information System (INIS)

    Zhou, Shengcheng; Wu, Hongchun; Zheng, Youqi

    2018-01-01

    Highlights: •ADS reloading scheme is optimized to raise discharge burnup and lower reactivity loss. •ADS is flexible to be combined with various pyro-chemical reprocessing technologies. •ADS is flexible to transmute MAs from different spent PWR fuels. -- Abstract: A two-stage Pressurized Water Reactor (PWR)-Accelerator Driven System (ADS) fuel cycle is proposed as an option to transmute minor actinides (MAs) recovered from the spent PWR fuels in the ADS system. At the second stage, the spent fuels discharged from ADS are reprocessed by the pyro-chemical process and the recovered actinides are mixed with the top-up MAs recovered from the spent PWR fuels to fabricate the new fuels used in ADS. In order to lower the amount of nuclear wastes sent to the geological repository, an optimized scattered reloading scheme for ADS is proposed to maximize the discharge burnup and lower the burnup reactivity loss. Then the flexibility of ADS for MA transmutation is evaluated in this research. Three aspects are discussed, including: different cooling time of spent ADS fuels before reprocessing, different reprocessing loss of spent ADS fuels, and different top-up MAs recovered from different kinds of spent PWR fuels. The ADS system is flexible to be combined with various pyro-chemical reprocessing technologies with specific spent fuels cooling time and unique reprocessing loss. The reduction magnitudes of the long-term decay heat and radiotoxicity of MAs by transmutation depend on the reprocessing loss. The ADS system is flexible to transmute MAs recovered from different kinds of spent PWR fuels, regardless of UOX or MOX fuels. The reduction magnitudes of the long-term decay heat and radiotoxicity of different MAs by transmutation stay on the same order.

  20. Characteristics and fabrication of cermet spent nuclear fuel casks: ceramic particles embedded in steel

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Swaney, P.M.; Tiegs, T.N. [Oak Ridge National Lab., Oak Ridge, TN (United States)

    2004-07-01

    Cermets are being investigated as an advanced material of construction for casks that can be used for storage, transport, or disposal of spent nuclear fuel (SNF). Cermets, which consist of ceramic particles embedded in steel, are a method to incorporate brittle ceramics with highly desirable properties into a strong ductile metal matrix with a high thermal conductivity, thus combining the best properties of both materials. Traditional applications of cermets include tank armor, vault armor, drill bits, and nuclear test-reactor fuel. Cermets with different ceramics (DUO{sub 2}, Al{sub 2}O{sub 3}, Gd{sub 2}O{sub 3}, etc.) are being investigated for the manufacture of SNF casks. Cermet casks offer four potential benefits: greater capacity (more SNF assemblies) for the same gross weight cask, greater capacity (more SNF assemblies) for the same external dimensions, improved resistance to assault, and superior repository performance. These benefits are achieved by varying the composition, volume fraction, and particulate size of the ceramic particles in the cermet with position in the cask body. Addition of depleted uranium dioxide (DUO{sub 2}) to the cermet increases shielding density, improves shielding effectiveness, and increases cask capacity for a given cask weight or size. Addition of low-density aluminium oxide (Al{sub 2}O{sub 3}) to the outer top and bottom sections of the cermet cask, where the radiation levels are lower, can lower cask weight without compromising shielding. The use of Al2O3 and other oxides, in appropriate locations, can increase resistance to assault. Repository performance may be improved by compositional control of the cask body to (1) create a local geochemical environment that slows the long-term degradation of the SNF and (2) enables the use of DUO{sub 2} for longterm criticality control. While the benefits of using cermets follow directly from their known properties, the primary challenge is to develop low-cost methods to fabricate

  1. System and safety studies of accelerator driven systems for transmutation. Annual report 2007

    International Nuclear Information System (INIS)

    Arzhanov, Vasily; Fokau, Andrei; Persson, Calle; Runevall, Odd; Sandberg, Nils; Tesinsky, Milan; Wallenius, Janne; Youpeng Zhang

    2008-05-01

    Within the project 'System and safety studies of accelerator driven systems for transmutation', research on design and safety of sub-critical reactors for recycling of minor actinides is performed. During 2007, the reactor physics division at KTH has calculated safety parameters for EFIT-400 with cermet fuel, permitting to start the transient safety analysis. The accuracy of different reactivity meters applied to the YALINA facility was assessed and neutron detection studies were performed. A model to address deviations from point kinetic behaviour was developed. Studies of basic radiation damage physics included calculations of vacancy formation and activation enthalpies in bcc niobium. In order to predict the oxygen potential of inert matrix fuels, a thermo-chemical model for mixed actinide oxides was implemented in a phase equilibrium code

  2. A Science-Based Understanding of Cermet Processing

    Energy Technology Data Exchange (ETDEWEB)

    Cesarano, III, Joseph [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Roach, Robert Allen [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kilgo, Alice C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Susan, Donald Francis [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Van Ornum, David J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Stuecker, John N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Shollenberger, Kimberly A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2006-04-01

    This report is a summary of the work completed in FY01 for science-based characterization of the processes used to fabricate 1) cermet vias in source feedthrus using slurry and paste-filling techniques and 2) cermet powder for dry pressing. Common defects found in cermet vias were characterized based on the ability of subsequent processing techniques (isopressing and firing) to remove the defects. Non-aqueous spray drying and mist granulation techniques were explored as alternative methods of creating CND50, the powder commonly used for dry pressed parts. Compaction and flow characteristics of these techniques were analyzed and compared to standard dry-ball-milled CND50. Due to processing changes, changes in microstructure can occur. A microstructure characterization technique was developed to numerically describe cermet microstructure. Machining and electrical properties of dry pressed parts were also analyzed and related to microstructure using this analytical technique.3 Executive SummaryThis report outlines accomplishments in the science-based understanding of cermet processing up to fiscal year 2002 for Sandia National Laboratories. The three main areas of work are centered on 1) increasing production yields of slurry-filled cermets, 2) evaluating the viability of high-solids-loading pastes for the same cermet components, and 3) optimizing cermet powder used in pressing processes (CND50). An additional development that was created as a result of the effort to fully understand the impacts of alternative processing techniques is the use of analytical methods to relate microstructure to physical properties. Recommendations are suggested at the end of this report. Summaries of these four efforts are as follows:1.Increase Production Yields of Slurry-Filled Cermet Vias Finalized slurry filling criteria were determined based on three designs of experiments where the following factors were analyzed: vacuum time, solids loading, pressure drop across the filter paper

  3. Fuels and targets for incineration and transmutation of actinides: the ITU programme

    International Nuclear Information System (INIS)

    Fernandez, A.; Glatz, J.P.; Haas, D.; Konings, R.J.M.; Somers, J.; Toscano, E.; Walker, C.T.; Wegen, D.

    2000-01-01

    The ITU programme for the development of fuels and targets for transmutation of actinides is presented. The fabrication of various types of oxide fuels/targets by dust-free processes is described. Selected results of post-irradiation examinations of irradiation experiments (SUPERFACT, TRABANT-1, EFTTRA-T4) are presented to demonstrate the irradiation behaviour of these fuels/targets. Finally, the future developments at ITU in this field are described, including the new shielded facility (the MA lab) for fabrication of minor actinide fuels. (authors)

  4. Fuels and targets for incineration and transmutation of actinides: the ITU programme

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, A.; Glatz, J.P.; Haas, D.; Konings, R.J.M.; Somers, J.; Toscano, E.; Walker, C.T.; Wegen, D. [Eurpean Commission, Joint Research Centre, Institute for Transuranium Elements, Kurlsruhe (Germany)

    2000-07-01

    The ITU programme for the development of fuels and targets for transmutation of actinides is presented. The fabrication of various types of oxide fuels/targets by dust-free processes is described. Selected results of post-irradiation examinations of irradiation experiments (SUPERFACT, TRABANT-1, EFTTRA-T4) are presented to demonstrate the irradiation behaviour of these fuels/targets. Finally, the future developments at ITU in this field are described, including the new shielded facility (the MA lab) for fabrication of minor actinide fuels. (authors)

  5. Numerical analysis on reduction of radioactive actinides by recycling of nuclear fuel

    International Nuclear Information System (INIS)

    Balboa L, H. E.

    2014-01-01

    Worldwide, human growth has reached unparalleled levels historically, this implies a need for more energy, and just in 2007 was consumed in the USA 4157 x 10 9 kWh of electricity and there were 6 x 10 9 metric tons of carbon dioxide, which causes a devastating effect on our environment. To this problem, a solution to the demand for non-fossil energy is nuclear energy, which is one of the least polluting and the cheapest among non-fossil energy; however, a problem remains unresolved the waste generation of nuclear fuels. In this work the option of a possible transmutation of actinides in a nuclear reactor of BWR was analyzed, an example of this are the nuclear reactors at the Laguna Verde nuclear power plant, which have generated spent fuel stored in pools awaiting a decision for final disposal or any other existing alternative. Assuming that the spent fuel was reprocessed to separate useful materials and actinides such as plutonium and uranium remaining, could take these actinides and to recycle them inside the same reactor that produced them, so il will be reduced the radiotoxicity of spent fuel. The main idea of this paper is to evaluate by means of numeric simulation (using the Core Management System (CMS)) the reduction of minor actinides in the case of being recycled in fresh fuel of the type BWR. The actinides were introduced hypothetically in the fuel pellets to 6% by weight, and then use a burned in the range of 0-65 G Wd/Tm, in order to have a better panorama of their behavior and thus know which it is the best choice for maximum reduction of actinides. Several cases were studied, that is to say were used as fuels; the UO 2 and MOX. Six different cases were also studied to see the behavior of actinides in different situations. The CMS platform calculation was used for the analysis of the cases presented. Favorable results were obtained, having decreased from a range of 35% to 65% of minor actinides initially introduced in the fuel rods, reducing the

  6. Characterization of Cr-O cermet solar selective coatings deposited by using direct-current magnetron sputtering technology

    International Nuclear Information System (INIS)

    Lee, Kil Dong

    2006-01-01

    Cr-O (Cr-CrO) cermet solar selective coatings with a double cermet layer film structure were prepared by using a special direct-current (dc) magnetron sputtering technology. The typical film structure from the surface to the bottom substrate was an Al 2 O 3 anti-reflection layer on a double Cr-O cermet layer on an Al metal infrared reflection layer. The deposited Cr-O cermet solar selective coating had an absorptance of α = 0.93 - 0.95 and an emittance of ε = 0.09 - 0.10(100 .deg. C). The absorption layers of the Cr-O cermet coatings deposited on glass and silicon substrates were identified as being amorphous by using X-ray diffraction (XRD). Atomic force microscopy (AFM) showed that Cr-O cermet layers were very smooth and that their grain sizes were very small. The result of thermal stability test showed that the Cr-O cermet solar selective coating was stable for use at temperatures of under 400 .deg. C.

  7. Monazite-type ceramics for the immobilization of minor actinides plutonium; Keramiken des Monazit-Typs zur Immobilisierung von minoren Actinoiden und Plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Heuser, Julia Maria

    2015-07-01

    The safe disposal of radioactive waste in deep geological formations is a challenging task of present and future generations. Innovative strategies as the conditioning of radionuclides in ceramic matrices can make a contribution here. This work points out monazite-type ceramics as potential waste forms for minor actinides and Pu. Several aspects concerning nuclear disposal as well as fundamental structural information were investigated. Lanthanide phosphate endmembers (LnPO{sub 4}) within the stability field of monazite (Ln = La-Gd) were synthesised within the scope of this work. To extend the knowledge of monazite phases, monoclinic TbPO{sub 4}- and DyPO{sub 4}-phases were prepared and characterised. Tb- and Dy-phosphates are situated in the xenotime stability field close to that of monazite. They can exist as metastable monazite phases. Structural characterisations of long- and short-range order were performed by X-ray diffraction, infrared (IR) and Raman spectroscopy. Structural data could be complemented, enhanced and gaps of knowledge could be filled by the first systematic consideration of the complete Ln-monazite-series (Ln = La-Dy). Furthermore, this work focuses on Sm-monazite phases. Samarium with an atomic number of 62 is located in the middle part of the lanthanides showing the monazite structure. Accordingly, it has a mean cationic radius within the Ln-monazite-series and hence shows a relative high flexibility regarding the incorporation of radionuclides with different radii. Sintering densities of SmPO{sub 4} ceramics were optimised by varying process parameters like pressure and number of pressing steps. An irregular texture as well as densities of 94% of the theoretical value could be achieved. The resistance of Sm-monazite against ionising radiation were examined. Radiation damages caused by the α-decay of radionuclides incorporated in a ceramic matrix were simulated by computer calculations and experimentally by heavy ion bombardment of Sm

  8. Teknologi Pembuatan Cermet Du0¬2 - Steel Untuk Wadah Limbah Bahan Bakar Bekas Pwr

    OpenAIRE

    Alimah, Siti; Budiarto, Budiarto

    2005-01-01

    DUO­2-STEEL CERMET MANUFACTURING TECHNOLOGY FOR PWR Spent Nuclear Fuel (SNF) CASKS. Assessment of DU02 - Steel cermet manufacturing technology for PWR SNF casks has been done. DU02 - Steel cermet consisting of DU02 particulates and other particulates, embedded in a steel matrix. Cermet SNF casks have the potential for superior performance compared with casks constructed of other materials. The addition of DU02 ceramic particulates can increase SNF cask capacity, improve of repository performa...

  9. Solid state synthesis and sintering of monazite-type ceramics: application to minor actinides conditioning

    International Nuclear Information System (INIS)

    Bregiroux, D.

    2005-11-01

    In the framework of the French law of 1991 concerning the nuclear waste management, several studies are undertaken to develop specific crystalline conditioning matrices. Monazite, a rare earth (TR 3+ ) orthophosphate with a general formula TR 3+ PO 4 , is a natural mineral containing significant amount of thorium and uranium. Monazite has been proposed as a host matrix for the minor actinides (Np, Am and Cm) specific conditioning, thanks to its high resistance to self irradiation and its low solubility. Its is now of prime importance to check the conservation of these properties on synthesized materials, which implies to master all the stages of the elaboration process, from the powder synthesis to the sintering of controlled microstructure pellets. This work can be divided into two main parts: The first part deals with the synthesis by high temperature solid state route of TR 3+ PO 4 powders (with TR 3+ = La 3+ to Gd 3+ , Pu 3+ and Am 3+ ). The chemical reactions occurring during the firing of starting reagents are described in the case of monazite with only one or several cations. From these results, a protocol of synthesis is described. The incorporation of tetravalent cations (Ce 4+ , U 4+ , Pu 4+ ) in the monazite structure was also studied. The second part of the present work deals with the elaboration of controlled density and microstructure monazite pellets and their related mechanical and thermal properties. The study of crushing and sintering is presented. For the first time, experimental results are confronted with theoretical models in order to deduce the densification and grain growth mechanisms. By the comprehension of the various physicochemical phenomena occurring during the various stages of the monazite pellets elaboration process (powder synthesis, crushing, sintering...), this work allowed the development of a protocol of elaboration of controlled microstructure monazite TR 3+ PO 4 pellets. The determination of some mechanical and thermal

  10. Complexation of f elements by tripodal ligands containing aromatic nitrogens. Application to the selective extraction of actinides(III)

    International Nuclear Information System (INIS)

    Wietzke, Raphael

    1999-01-01

    This work initiates a research project, whose aim is the actinides(lll)/lanthanides(III) separation by liquid-liquid extraction. We were interested in the study of the coordination chemistry of lanthanides(III) and uranium(III) (uranium(III) as model for the actinides(III)), with the aim to show differences between the two families and to better understand the coordination properties involved in the extraction process. We studied the lanthanide(III) and uranium(III) complexation with tripodal ligands containing aromatic nitrogens. Several tripodal ligands were synthesized varying the type and the number of the donor atoms. The lanthanide(III) complexes have been characterized in the solid state and in solution (by several techniques: "1H NMR, ESMS, luminescence, UV spectrophotometry, conductometry). Differences in the coordination were found depending on the nature of the donor atoms. The new ligands, tris(2-pyrazinylmethyl)amine (tpza) et tris(N,N-diethyl-2-carbamoyl-6- pyridylmethyl)amine (tpaa), have shown a selectivity for the actinides(III) with promising results in liquid-liquid extraction. The comparison between the lanthanum(III) and uranium(III) complexes with the ligand tpza showed differences in the bonding nature, which could be attributed to a covalent contribution to the metal-ligand bond. (author) [fr

  11. Status of the French research programme for actinides and fission products partitioning and transmutation

    International Nuclear Information System (INIS)

    Warin, D.

    2003-01-01

    The paper focus on separation and transmutation research and development programme and main results over these ten last years. The massive research programme on enhanced separation, conducted by CEA and supported by broad international cooperation, has recently achieved some vital progress. Based on real solutions derived from the La Hague process, the CEA demonstrated the lab-scale feasibility of extracting minor actinides and some fission products (I, Cs and Tc) using an hydrometallurgical process that can be extrapolated on the industrial scale. The CEA also conducted programmes proving the technical feasibility of the elimination of minor actinides and fission products by transmutation: fabrication of specific targets and fuels for transmutation tests in the HFR and Phenix reactors, neutronics and technology studies for ADS developments in order to support the MEGAPIE, TRADE and MYRRHA experiments and the future 100 MW international ADS demonstrator. Scenarios studies aimed at stabilizing the inventory with long-lived radionuclides, plutonium, minor actinides and certain long-lived fission products in different nuclear power plant parks and to verify the feasibility at the level of the cycle facilities and fuels involved in those scenarios. Three French Research Groups CEA-CNRS carry out partitioning (PRACTIS) and transmutation (NOMADE and GEDEON) more basic studies. (author)

  12. Micro-scale mechanical characterization of Inconel cermet coatings deposited by laser cladding

    OpenAIRE

    Chao Chang; Davide Verdi; Miguel Angel Garrido; Jesus Ruiz-Hervias

    2016-01-01

    In this study, an Inconel 625-Cr3C2 cermet coating was deposited on a steel alloy by laser cladding. The elastic and plastic mechanical properties of the cermet matrix were studied by the depth sensing indentation (DSI) in the micro scale. These results were compared with those obtained from an Inconel 600 bulk specimen. The values of Young's modulus and hardness of cermet matrix were higher than those of an Inconel 600 bulk specimen. Meanwhile, the indentation stress–strain curve of the cerm...

  13. Solubility of actinides and surrogates in nuclear glasses; Solubilite des actinides et de leurs simulants dans les verres nucleaires. Limites d'incorporation et comprehension des mecanismes

    Energy Technology Data Exchange (ETDEWEB)

    Lopez, Ch

    2003-07-01

    The nuclear wastes are currently incorporated in borosilicate glass matrices. The resulting glass must be perfectly homogeneous. The work discussed here is a study of actinide (thorium and plutonium) solubility in borosilicate glass, undertaken to assess the extent of actinide solubility in the glass and to understand the mechanisms controlling actinide solubilization. Glass specimens containing; actinide surrogates were used to prepare and optimize the fabrication of radioactive glass samples. These preliminary studies revealed that actinide Surrogates solubility in the glass was enhanced by controlling the processing temperature, the dissolution kinetic of the surrogate precursors, the glass composition and the oxidizing versus reducing conditions. The actinide solubility was investigated in the borosilicate glass. The evolution of thorium solubility in borosilicate glass was determined for temperatures ranging from 1200 deg C to 1400 deg C.Borosilicate glass specimens containing plutonium were fabricated. The experimental result showed that the plutonium solubility limit ranged from 1 to 2.5 wt% PuO{sub 2} at 1200 deg C. A structural approach based on the determination of the local structure around actinides and their surrogates by EXAFS spectroscopy was used to determine their structural role in the glass and the nature of their bonding with the vitreous network. This approach revealed a correlation between the length of these bonds and the solubility of the actinides and their surrogates. (author)

  14. Cermet based solar selective absorbers : further selectivity improvement and developing new fabrication technique

    OpenAIRE

    Nejati, Mohammadreza

    2008-01-01

    Spectral selectivity of cermet based selective absorbers were increased by inducing surface roughness on the surface of the cermet layer using a roughening technique (deposition on hot substrates) or by micro-structuring the metallic substrates before deposition of the absorber coating using laser and imprint structuring techniques. Cu-Al2O3 cermet absorbers with very rough surfaces and excellent selectivity were obtained by employing a roughness template layer under the infrared reflective l...

  15. High temperature resistant cermet and ceramic compositions

    Science.gov (United States)

    Phillips, W. M. (Inventor)

    1978-01-01

    Cermet compositions having high temperature oxidation resistance, high hardness and high abrasion and wear resistance, and particularly adapted for production of high temperature resistant cermet insulator bodies are presented. The compositions are comprised of a sintered body of particles of a high temperature resistant metal or metal alloy, preferably molybdenum or tungsten particles, dispersed in and bonded to a solid solution formed of aluminum oxide and silicon nitride, and particularly a ternary solid solution formed of a mixture of aluminum oxide, silicon nitride and aluminum nitride. Also disclosed are novel ceramic compositions comprising a sintered solid solution of aluminum oxide, silicon nitride and aluminum nitride.

  16. Effect of Mo and C additions on magnetic properties of TiC–TiN–Ni cermets

    International Nuclear Information System (INIS)

    Zhang, Man; Yang, Qingqing; Xiong, Weihao; Zheng, Liyun; Huang, Bin; Chen, Shan; Yao, Zhenhua

    2015-01-01

    The effect of 2–8 mol.% Mo and 4 mol.% C additions on magnetic properties of TiC–10TiN–30Ni (mol.%) cermet was investigated. Saturation magnetization M_s, remanence M_r and Curie temperature T_c of as-sintered cermets (1420 °C, 1 h) decreased with increasing Mo. This was mainly attributed to that the total content of non-magnetic alloying elements Mo and Ti in Ni-based binder phase increased with increasing Mo in cermets, leading to the weakening of magnetic exchange interaction among Ni atoms in binder phase. The further addition of 4 mol.% C inversely increased M_s, M_r and T_c of cermets, which was mainly attributed to that it decreased the total content of Mo and Ti in binder phase, leading to the strengthening of magnetic exchange interaction among Ni atoms in binder phase. T_c of cermets without C addition was about 250 K at 6 mol.% Mo and 115 K at 8 mol.% Mo, respectively, and that of cermets with 4 mol.% C addition was about 194 K at 8 mol.% Mo. - Highlights: • M_s, M_r and T_c of TiC–10TiN–30Ni–xMo cermets decreased with the increase of Mo content, x. • Further addition of 4 mol.% C inversely increased M_s, M_r and T_c of cermets at the same Mo content. • T_c of cermets without C addition was about 250 K at x = 6 and 115 K at x = 8, respectively. • T_c of cermets with 4 mol.% C addition was about 194 K at x = 8.

  17. Preparation and Mechanical Properties of TiC-Fe Cermets and TiC-Fe/Fe Bilayer Composites

    Science.gov (United States)

    Zheng, Yong; Zhou, Yang; Li, Runfeng; Wang, Jiaqi; Chen, Lulu; Li, Shibo

    2017-10-01

    TiC-Fe cermets and TiC-Fe/Fe bilayer composites consisting of a pure Fe layer and a TiC-Fe cermets layer were fabricated by hot-pressing sintering. The pure Fe layer contributes to the toughness of composites, and the TiC-Fe cermets layer endows the composites with an improved tensile strength and hardness. The effect of TiC contents (30-60 vol.%) on the mechanical properties of TiC-Fe cermets and TiC-Fe/Fe bilayer composites was investigated. Among the TiC-Fe cermets, the 40 vol.% TiC-Fe cermets possessed the highest tensile strength of 581 MPa and Vickers hardness of 5.1 GPa. The maximum fracture toughness of 17.0 MPa m1/2 was achieved for the TiC-Fe cermets with 30 vol.% TiC. For the TiC-Fe/Fe bilayer composites, the 40 vol.% TiC-Fe/Fe bilayer composite owns the maximum tensile strength of 588 MPa, which is higher than that of 40 vol.% TiC-Fe cermets. In addition, the 33.5% increment of tensile strength of 30 vol.% TiC-Fe/Fe bilayer composite comparing with the 30 vol.% TiC-Fe cermets, which is attributed to the 30 vol.% TiC-Fe/Fe bilayer composite exhibited the largest interlaminar shear strength of 335 MPa. The bilayer composites are expected to be used as wear resistance components in some heavy wear conditions.

  18. Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydin, E-mail: karahan@mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, 77 Massachusetts Avenue, Cambridge MA 24-204 (United States)

    2011-07-15

    Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

  19. Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

    Science.gov (United States)

    Karahan, Aydın

    2011-07-01

    Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

  20. Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

    International Nuclear Information System (INIS)

    Karahan, Aydin

    2011-01-01

    Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

  1. A Historical Review of Cermet Fuel Development and the Engine Performance Implications

    Science.gov (United States)

    Stewart, Mark E.

    2015-01-01

    To better understand Cermet engine performance, examined historical material development reports two issues: High vaporization rate of UO2, High temperature chemical stability of UO2. Cladding and chemical stabilizers each result in large, order of magnitude improvements in high temperature performance. Few samples were tested above 2770 K. Results above 2770 K are ambiguous. Contemporary testing may clarify performance. Cermet sample testing during the NERVA Rover era. Important properties, melting temperature, vaporization rate, strength, Brittle-to-Ductile Transition, cermet sample test results, engine performance, location, peak temperature.

  2. Characterization of boride-based powders and detonation gun sprayed cermet coatings

    International Nuclear Information System (INIS)

    Keraenen, J.; Stenberg, T.; Maentylae, T.

    1995-01-01

    Detonation gun sprayed (DGS) cermet coatings containing complex ternary transition metal boride as hard particles dispersed in a stainless steel or nickel based superalloy matrix have been characterized. Microstructure of the coatings, as well as powders, were studied with optical microscopy, scanning electron microscopy (SEM), X-ray diffraction (XRD) and analytical transmission electron microscopy (AEM). X-ray microanalysis of the coatings were carried out using energy dispersive X-ray spectrometer (EDS) attached to the SEM and AEM. Moreover, abrasion wear resistance of the coatings was evaluated with a rubber wheel abrasion test equipment. The general microstructure of studied coatings appeared to be heterogeneous in the terms of the distribution, size and crystallographic nature of the phases. Nonetheless, very low porosities were obtained and in the coatings the oxide phase as well as the unmelted particles and the formation of oxide phase were avoided by optimization of DGS parameters. So far the abrasive wear resistance of boride-based cermet coatings is not so good as that of the WC-12Co coatings

  3. Actinide metal processing

    International Nuclear Information System (INIS)

    Sauer, N.N.; Watkin, J.G.

    1992-01-01

    A process for converting an actinide metal such as thorium, uranium, or plutonium to an actinide oxide material by admixing the actinide metal in an aqueous medium with a hypochlorite as an oxidizing agent for sufficient time to form the actinide oxide material and recovering the actinide oxide material is described together with a low temperature process for preparing an actinide oxide nitrate such as uranyl nitrate. Additionally, a composition of matter comprising the reaction product of uranium metal and sodium hypochlorite is provided, the reaction product being an essentially insoluble uranium oxide material suitable for disposal or long term storage

  4. Robust membrane systems for actinide separations

    International Nuclear Information System (INIS)

    Jarvinen, Gordon D.; McCleskey, T. Mark; Bluhm, Elizabeth A.; Abney, Kent D.; Ehler, Deborah S.; Bauer, Eve; Le, Quyen T.; Young, Jennifer S.; Ford, Doris K.; Pesiri, David R.; Dye, Robert C.; Robison, Thomas W.; Jorgensen, Betty S.; Redondo, Antonio; Pratt, Lawrence R.; Rempe, Susan L.

    2000-01-01

    Our objective in this project is to develop very stable thin membrane structures containing ionic recognition sites that facilitate the selective transport of target metal ions, especially the actinides

  5. Study and choice of main characteristics of fast reactor - Effective minor actinide burner

    International Nuclear Information System (INIS)

    Krivitski, I.Yu.; Matveev, V.I.; Poplavski, V.M.

    1996-01-01

    This paper presents the principal design and performance data of advanced fast power reactor core for plutonium and actinides burning. Some information concerning the Russian programme of plutonium utilization are also presented. (author). 2 refs, 4 figs, 5 tabs

  6. Microstructure and properties of multiphase sintered cermets Fe-Fe2B

    International Nuclear Information System (INIS)

    Nowacki, J.; Klimek, L.

    1998-01-01

    The process of multiphase sintering of iron in the vacuum has been analysed. As a result of the process iron-iron boride cermets have been produced. Fe-Fe 2 B cermets were obtained as a result of sintering of the Fe and B pure elements in the vacuum. Attemps at sintering in the solid phase and with the participation of the liquid phase, the Fe-Fe 2 B eutectic, have been made. Metallographic qualitative and quantitative studies, X-ray structural qualitative and qauantitative analysis allowed to determine the structure of Fe 2 B cermets, as well as a description of the kinetics of quantitative changes in phase proportions in the course of sintering. It has been found that their structure varies widely depending on sintering parameters and the composition of the sinters. Measurements of the Fe-Fe 2 B cermets hardness and measurements on wear during dry friction by the pin-on-disc method have shown distinct advantages of the cermets as a modern constructional materials. The hardness of Fe-Fe 2 B cermets, depending on their chemical composition and sintering parameters, ranges widely from 150 to 1500 HV, and their resistance to wear is comparable to that of diffusively boronized steels. FeFe 2 B cermets are a composite material in which iron boride, Fe 2 B, with a hardness of about 1800 HV plays the role of the reinforcement,while iron-iron boride, Fe-Fe 2 B, with a hardness of about 500 HV plays the role of matrix. The eutectic in the spaces between iron boride grains is composed of boron solid solution plates in iron with a hardness of arround 250 HV, and iron boride, Fe 2 B, plates with a hardness of approximaly 1800 HV. The combination of such different materials, a hard reinforcement and a relatively plastic matrix produces favourable properties of the cermet thus produced high hardness (1500 HV) constant over whole cross section of the material, resistance of abrasive wear and acceptable ductility. The properties mentioned above, resulting from the cermet

  7. Corrosion stability of cermets on the base of titanium nitride

    International Nuclear Information System (INIS)

    Kajdash, O.N.; Marinich, M.A.; Kuzenkova, M.A.; Manzheleev, I.V.

    1991-01-01

    Corrosion resistance of titanium nitride and its cermets in 5% of HCl, 7% of HNO 3 , 10% of H 2 SO 4 is studied. It is established that alloys TiN-Ni-Mo alloyed with chromium (from 10 to 15%) possess the highest corrosion resistance. Cermet TiN-Cr has the higher stability than titanium nitride due to formation of binary nitride (Ti, Cr)N

  8. Processing and microstructural characterization of B4C-Al cermets

    International Nuclear Information System (INIS)

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

    1985-01-01

    Reaction thermodynamics and wetting studies were employed to evaluate boron carbide-aluminum cermets. Wetting phonomenon and interfacial reactions are characterized using ''macroscale'' and ''microscale'' techniques. Macroscale evaluation involved aluminium sessile drop studies on boron carbide substrates. Microscale evaluation involved the fabrication of actural cermet microstructures and their characterization through sem, x-ray diffraction, metallography, and electron microprobe. Contact-angle measurements and interfacial-reaction products are reported

  9. Synthesis of tetravalent actinide chlorides. Versatile compounds for actinide chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Maerz, Juliane [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Div. Chemistry of the F-Elements

    2016-07-01

    Anhydrous actinide tetrachlorides (AnCl{sub 4}) were synthesized under mild conditions to provide versatile compounds for actinide chemistry. They enable a direct access to actinide complexes with organic and inorganic ligands.

  10. Effect of Mo and C additions on magnetic properties of TiC–TiN–Ni cermets

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Man [State Key Laboratory of Material Processing and Die & Mould Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); Yang, Qingqing, E-mail: yqqah@sina.com [State Key Laboratory of Material Processing and Die & Mould Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); Xiong, Weihao [State Key Laboratory of Material Processing and Die & Mould Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); Zheng, Liyun [School of Equipment Manufacture, Hebei University of Engineering, Handan 056038 (China); Huang, Bin; Chen, Shan; Yao, Zhenhua [State Key Laboratory of Material Processing and Die & Mould Technology, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2015-11-25

    The effect of 2–8 mol.% Mo and 4 mol.% C additions on magnetic properties of TiC–10TiN–30Ni (mol.%) cermet was investigated. Saturation magnetization M{sub s}, remanence M{sub r} and Curie temperature T{sub c} of as-sintered cermets (1420 °C, 1 h) decreased with increasing Mo. This was mainly attributed to that the total content of non-magnetic alloying elements Mo and Ti in Ni-based binder phase increased with increasing Mo in cermets, leading to the weakening of magnetic exchange interaction among Ni atoms in binder phase. The further addition of 4 mol.% C inversely increased M{sub s}, M{sub r} and T{sub c} of cermets, which was mainly attributed to that it decreased the total content of Mo and Ti in binder phase, leading to the strengthening of magnetic exchange interaction among Ni atoms in binder phase. T{sub c} of cermets without C addition was about 250 K at 6 mol.% Mo and 115 K at 8 mol.% Mo, respectively, and that of cermets with 4 mol.% C addition was about 194 K at 8 mol.% Mo. - Highlights: • M{sub s}, M{sub r} and T{sub c} of TiC–10TiN–30Ni–xMo cermets decreased with the increase of Mo content, x. • Further addition of 4 mol.% C inversely increased M{sub s}, M{sub r} and T{sub c} of cermets at the same Mo content. • T{sub c} of cermets without C addition was about 250 K at x = 6 and 115 K at x = 8, respectively. • T{sub c} of cermets with 4 mol.% C addition was about 194 K at x = 8.

  11. Optimizing analysis of W-AlN cermet solar absorbing coatings

    International Nuclear Information System (INIS)

    Zhang Qichu

    2001-01-01

    The layer thickness and tungsten metal volume fraction of W-AlN cermet solar selective absorbing coatings on a W, Cu or Al infrared reflector with a surface aluminium oxynitride (AlON) or Al 2 O 3 ceramic anti-reflector layer were optimized using physical modelling calculations. Due to limited published data for the refractive index of AlN, and likely oxygen contamination during reactive sputtering of AlN ceramic materials, AlON was used as the ceramic component and the published value of its refractive index was employed. The dielectric function and then the complex refractive index of W-AlON cermet materials were calculated using the Ping Sheng approximation. The downhill simplex method in multi-dimensions was used in the numerical calculation to achieve maximum photo-thermal conversion efficiency at 350 0 C under a concentration factor of 30 for a solar collector tube. Optimization calculation results show that the initial graded (ten-step layers) cermet films all converge to something close to a three-layer film structure, which consists of a low metal volume fraction cermet layer on a high metal volume fraction cermet layer on a metallic infrared reflector with a surface ceramic anti-reflection layer. The optimized three-layer solar coatings have a high solar absorptance of 0.95 for AlON and 0.96 for the Al 2 O 3 anti-reflection layer, and a low hemispherical emittance of 0.073 at 350 deg. C. For the optimized three-layer films the solar radiation is efficiently absorbed internally and by phase interference. Thermal loss is very low for optimized three-layer films due to high reflectance values in the thermal infrared wavelength range and a very sharp edge between low solar reflectance and high thermal infrared reflectance. The high metal volume fraction cermet layer has a metal-like optical behaviour in the thermal infrared wavelength range and makes the largest contribution to the increase of emittance compared with that of the metal infrared reflector

  12. Optimizing analysis of W-AlN cermet solar absorbing coatings

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Qichu [School of Physics, University of Sydney, NSW (Australia)

    2001-11-07

    The layer thickness and tungsten metal volume fraction of W-AlN cermet solar selective absorbing coatings on a W, Cu or Al infrared reflector with a surface aluminium oxynitride (AlON) or Al{sub 2}O{sub 3} ceramic anti-reflector layer were optimized using physical modelling calculations. Due to limited published data for the refractive index of AlN, and likely oxygen contamination during reactive sputtering of AlN ceramic materials, AlON was used as the ceramic component and the published value of its refractive index was employed. The dielectric function and then the complex refractive index of W-AlON cermet materials were calculated using the Ping Sheng approximation. The downhill simplex method in multi-dimensions was used in the numerical calculation to achieve maximum photo-thermal conversion efficiency at 350{sup 0}C under a concentration factor of 30 for a solar collector tube. Optimization calculation results show that the initial graded (ten-step layers) cermet films all converge to something close to a three-layer film structure, which consists of a low metal volume fraction cermet layer on a high metal volume fraction cermet layer on a metallic infrared reflector with a surface ceramic anti-reflection layer. The optimized three-layer solar coatings have a high solar absorptance of 0.95 for AlON and 0.96 for the Al{sub 2}O{sub 3} anti-reflection layer, and a low hemispherical emittance of 0.073 at 350 deg. C. For the optimized three-layer films the solar radiation is efficiently absorbed internally and by phase interference. Thermal loss is very low for optimized three-layer films due to high reflectance values in the thermal infrared wavelength range and a very sharp edge between low solar reflectance and high thermal infrared reflectance. The high metal volume fraction cermet layer has a metal-like optical behaviour in the thermal infrared wavelength range and makes the largest contribution to the increase of emittance compared with that of the metal

  13. Optimizing analysis of W-AlN cermet solar absorbing coatings

    Energy Technology Data Exchange (ETDEWEB)

    Qi-Chu Zhang [University of Sydney, NSW (Australia). School of Physics

    2001-11-07

    The layer thickness and tungsten metal volume fraction of W-AlN cermet solar selective absorbing coatings on a W, Cu or Al infrared reflector with a surface aluminium oxynitride (AlON) or Al{sub 2}O{sub 3} ceramic anti-reflector layer were optimized using physical modelling calculations. Due to limited published data for the refractive index of AlN, and likely oxygen contamination during reactive sputtering of AlN ceramic materials, AlON was used as the ceramic component and the published value of its refractive index was employed. The dielectric function and then the complex refractive index of W-AlON cermet materials were calculated using the Ping Sheng approximation. The downhill simplex method in multi-dimensions was used in the numerical calculation to achieve maximum photo-thermal conversion efficiency at 350{sup o}C under a concentration factor of 30 for a solar collector tube. Optimization calculation results show that the initial graded (ten-step layers) cermet films all converge to something close to a three-layer film structure, which consists of a low metal volume fraction cermet layer on a high metal volume fraction cermet layer on a metallic infrared reflector with a surface ceramic anti-reflection layer. The optimized three-layer solar coatings have a high solar absorptance of 0.95 for AlON and 0.96 for the Al{sub 2}O{sub 3} anti-reflection layer, and a low hemispherical emittance of 0.073 at 350{sup o}C. For the optimized three-layer films the solar radiation is efficiently absorbed internally and by phase interference. Thermal loss is very low for optimized three-layer films due to high reflectance values in the thermal infrared wavelength range and a very sharp edge between low solar reflectance and high thermal infrared reflectance. The high metal volume fraction cermet layer has a metal-like optical behaviour in the thermal infrared wavelength range and makes the largest contribution to the increase of emittance compared with that of the metal

  14. High performance W-AIN cermet solar coatings designed by modelling calculations and deposited by DC magnetron sputtering

    Energy Technology Data Exchange (ETDEWEB)

    Qi-Chu Zhang [The University of Sydney (Australia). School of Physics; Shen, Y.G. [City University of Hong Kong (Hong Kong). Department of Manufacturing Engineering and Engineering Management

    2004-01-25

    High solar performance W-AIN cermet solar coatings were designed using a numerical computer model and deposited experimentally. In the numerical calculations aluminium oxynitride (AlON) was used as ceramic component. The dielectric functions and then complex refractive index of W-AlON cermet materials were calculated using the Sheng's approximation. The layer thickness and W metal volume fraction were optimised to achieve maximum photo-thermal conversion efficiency for W-AlON cermet solar coatings on an Al reflector with a surface AlON ceramic anti-reflection layer. Optimisation calculations show that the W-AlON cermet solar coatings with two and three cermet layers have nearly identical solar absorptance, emittance and photo-thermal conversion efficiency that are much better than those for films with one cermet layer. The optimised calculated AlON/W-AlON/Al solar coating film with two cermet layers has a high solar absorptance of 0.953 and a low hemispherical emittance of 0.051 at 80{sup o}C for a concentration factor of 2. The AlN/W-AlN/Al solar selective coatings with two cermet layers were deposited using two metal target direct current magnetron sputtering technology. During the deposition of W-AlN cermet layer, both Al and W targets were run simultaneously in a gas mixture of argon and nitrogen. By substrate rotation a multi-sub-layer system consisting of alternating AlN ceramic and W metallic sub-layers was deposited that can be considered as a macro-homogeneous W-AlN cermet layer. A solar absorptance of 0.955 and nearly normal emittance of 0.056 at 80{sup o}C have been achieved for deposited W-AlN cermet solar coatings. (author)

  15. Cermet based metamaterials for multi band absorbers over NIR to LWIR frequencies

    International Nuclear Information System (INIS)

    Pradhan, Jitendra K; Behera, Gangadhar; Anantha Ramakrishna, S; Agarwal, Amit K; Ghosh, Amitava

    2017-01-01

    Cermets or ceramic-metals are known for their use in solar thermal technologies for their absorption across the solar band. Use of cermet layers in a metamaterial perfect absorber allows for flexible control of infra-red absorption over the short wave infra-red, to long wave infra-red bands, while keeping the visible/near infra-red absorption properties constant. We design multilayered metamaterials consisting of a conducting ground plane, a low metal volume fraction cermet/ZnS as dielectric spacer layers, and a top structured layer of an array of circular discs of metal/high volume metal fraction cermet that give rise to specified absorption bands in the near-infra-red (NIR) frequencies, as well as any specified band at SWIR–LWIR frequencies. Thus, a complete decoupling of the absorption at optical/NIR frequencies and the infra-red absorption behaviour of a structured metamaterial is demonstrated. (paper)

  16. Microstructure analysis and wear behavior of titanium cermet femoral head with hard TiC layer.

    Science.gov (United States)

    Luo, Yong; Ge, Shirong; Liu, Hongtao; Jin, Zhongmin

    2009-12-11

    Titanium cermet was successfully synthesized and formed a thin gradient titanium carbide coating on the surface of Ti6Al4V alloy by using a novel sequential carburization under high temperature, while the titanium cermet femoral head was produced. The titanium cermet phase and surface topography were characterized with X-ray diffraction (XRD) and backscattered electron imaging (BSE). And then the wear behavior of titanium cermet femoral head was investigated by using CUMT II artificial joint hip simulator. The surface characterization indicates that carbon effectively diffused into the titanium alloys and formed a hard TiC layer on the Ti6Al4V alloys surface with a micro-porous structure. The artificial hip joint experimental results show that titanium cermet femoral head could not only improve the wear resistance of artificial femoral head, but also decrease the wear of UHMWPE joint cup. In addition, the carburized titanium alloy femoral head could effectively control the UHMWPE debris distribution, and increase the size of UHMWPE debris. All of the results suggest that titanium cermet is a prospective femoral head material in artificial joint.

  17. The effect of actinides on the microstructural development in a metallic high-level nuclear waste form

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D. D., Jr.; Sinkler, W.; Abraham, D. P.; Richardson, J. W., Jr.; McDeavitt, S. M.

    1999-10-25

    Waste forms to contain material residual from an electrometallurgical treatment of spent nuclear fuel have been developed by Argonne National Laboratory. One of these waste forms contains waste stainless steel (SS), fission products that are noble to the process (e.g., Tc, Ru, Pd, Rh), Zr, and actinides. The baseline composition of this metallic waste form is SS-15wt.% Zr. The metallurgy of this baseline alloy has been well characterized. On the other hand, the effects of actinides on the alloy microstructure are not well understood. As a result, SS-Zr alloys with added U, Pu, and/or Np have been cast and then characterized, using scanning electron microscopy, transmission electron microscopy, and neutron diffraction, to investigate the microstructural development in SS-Zr alloys that contain actinides. Actinides were found to congregate non-uniformally in a Zr(Fe,Cr,Ni){sub 2+x} phase. Apparently, the actinides were contained in varying amounts in the different polytypes (C14, C15, and C36) of the Zr(Fe,Cr,Ni){sub 2+x} phase. Heat treatment of an actinide-containing SS-15 wt.% Zr alloy showed the observed microstructure to be stable.

  18. Liquid phase sintered superconducting cermet

    International Nuclear Information System (INIS)

    Ray, S.P.

    1990-01-01

    This patent describes a method of making a superconducting cermet having superconducting properties with improved bulk density, low porosity and in situ stabilization. It comprises: forming a structure of a superconducting ceramic material having the formula RM 2 Cu 3 O (6.5 + x) wherein R is one or more rare earth elements capable of reacting to form a superconducting ceramic, M is one or more alkaline earth metal elements selected from barium and strontium capable of reacting to form a superconducting ceramic, x is greater than 0 and less than 0.5; and a precious metal compound in solid form selected from the class consisting of oxides, sulfides and halides of silver; and liquid phase sintering the mixture at a temperature wherein the precious metal of the precious metal compound is molten and below the melting point of the ceramic material. The liquid phase sintering is carried out for a time less than 36 hours but sufficient to improve the bulk density of the cermet

  19. In vitro caries-inhibitory properties of a silver cermet.

    Science.gov (United States)

    Swift, E J

    1989-06-01

    Recurrent caries is one of the primary causes of failure of dental restorations. One method for reducing the frequency and severity of this problem is the use of fluoride-releasing restorative materials. The glass-ionomer cements are a type of fluoride-releasing material. They have been used extensively in recent years for a variety of clinical applications. However, in comparison with other restorative materials such as amalgam and composite resins, glass ionomers have relatively poor physical properties. Sintering of silver particles to glass-ionomer powder is a means of improving these physical properties. The sintered material is called a silver-glass ionomer or silver cermet. This study examined the in vitro caries-inhibitory potential of a silver cement by means of two methods. First, long-term fluoride release was measured. Second, an artificial caries system was used for evaluation of caries inhibition by cerment restorations in extracted teeth. In comparison with a standard glass-ionomer restorative material, fluoride release from the cermet material was significantly less throughout a 12-month period. The results from the artificial caries system indicated that this decreased fluoride release corresponded with a lesser degree of caries inhibition. Lesions around cermet restorations in both enamel and root surfaces were significantly more severe than those around conventional glass-ionomer restorations. However, in comparison with amalgam and composite resin restorations, the cermet did have some cariostatic activity.

  20. Ab Initio Enhanced calphad Modeling of Actinide-Rich Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, Dane [Univ. of Wisconsin, Madison, WI (United States); Yang, Yong Austin [Univ. of Wisconsin, Madison, WI (United States)

    2013-10-28

    The process of fuel recycling is central to the Advanced Fuel Cycle Initiative (AFCI), where plutonium and the minor actinides (MA) Am, Np, and Cm are extracted from spent fuel and fabricated into new fuel for a fast reactor. Metallic alloys of U-Pu-Zr-MA are leading candidates for fast reactor fuels and are the current basis for fast spectrum metal fuels in a fully recycled closed fuel cycle. Safe and optimal use of these fuels will require knowledge of their multicomponent phase stability and thermodynamics (Gibbs free energies). In additional to their use as nuclear fuels, U-Pu-Zr-MA contain elements and alloy phases that pose fundamental questions about electronic structure and energetics at the forefront of modern many-body electron theory. This project will validate state-of-the-art electronic structure approaches for these alloys and use the resulting energetics to model U-Pu-Zr-MA phase stability. In order to keep the work scope practical, researchers will focus on only U-Pu-Zr-{Np,Am}, leaving Cm for later study. The overall objectives of this project are to: Provide a thermodynamic model for U-Pu-Zr-MA for improving and controlling reactor fuels; and, Develop and validate an ab initio approach for predicting actinide alloy energetics for thermodynamic modeling.

  1. Physical properties and microstructure of Ti(CN)-based cermets with different WC particle size

    International Nuclear Information System (INIS)

    Deng, Ying; Deng, Ling; Xiong, Xiang; Ye, J.W.; Li, P.P.

    2014-01-01

    Ti(CN)-based cermets with different WC particle sizes from 0.2 to 4 μm were prepared at 1450 °C with 2 MPa Air pressure. The microstructure of cermets was investigated by scanning electron microscope (SEM), X-ray diffraction (XRD), Transmission electron microscope (TEM). The results showed that all the cermets with different WC particle sizes have a typical “core–rim” structure. With the increase of WC powder sizes, the frequency and portion of Ti(C 0.7 N 0.3 ) cores and rim are somewhat decreased while the portion of white core is increased, due to the relative dissolution rate decreasing. In addition, the fracture mode of Ti(C,N) based cermets is a mixture of trans-granular (primary) and inter-granular (subordinate) fracture. The TRS (about 1850 MPa) of the cermets fluctuate slightly with the WC particle sizes from 0.2 to 1.0 μm, but decrease evidently with WC particle sizes up to 2 μm

  2. Effect of SiC whisker addition on the microstructures and mechanical properties of Ti(C, N)-based cermets

    International Nuclear Information System (INIS)

    Wu, Peng; Zheng, Yong; Zhao, Yongle; Yu, Haizhou

    2011-01-01

    Ti(C, N)-based cermets with addition of SiC whisker (SiC w ) were prepared by vacuum sintering. The microstructures of the prepared cermets were investigated by using X-ray diffractometry (XRD) and scanning electron microscopy (SEM). Mechanical properties such as transverse rupture strength (TRS), fracture toughness (K IC ) and hardness (HRA) were also measured. It was found that the grain size of the cermets was affected by the SiC whisker addition. The cermets with 1.0 wt.% SiC whisker addition exhibited the smallest grain size. The porosities of the cermets increased with increasing SiC whisker additions. The addition of the SiC whisker had no influence on the phase constituents of the cermets. Compared with the cermets with no whisker addition, the highest TRS and fracture toughness for cermets with 1.0 wt.% SiC whisker addition increased by about 24% and 29%, respectively. The strengthening mechanisms were attributed to finer grain size, homogeneous microstructure and moderate thickness of rim phase. The toughening mechanisms were characterized by crack deflection, whisker bridging and whisker pulling-out.

  3. The Influence of Sintering Temperature of Reactive Sintered (Ti, MoC-Ni Cermets

    Directory of Open Access Journals (Sweden)

    Marek Jõeleht

    2015-09-01

    Full Text Available Titanium-molybdenum carbide nickel cermets ((Ti, MoC-Ni were produced using high energy milling and reactive sintering process. Compared to conventional TiC-NiMo cermet sintering the parameters for reactive sintered cermets vary since additional processes are present such as carbide synthesis. Therefore, it is essential to acquire information about the suitable sintering regime for reactive sintered cermets. One of the key parameters is the final sintering temperature when the liquid binder Ni forms the final matrix and vacancies inside the material are removed. The influence of the final sintering temperature is analyzed by scanning electron microscopy. Mechanical properties of the material are characterized by transverse rupture strength, hardness and fracture toughness.DOI: http://dx.doi.org/10.5755/j01.ms.21.3.7179

  4. Complexes of groups 3,4, the lanthanides and the actinides containing neutral phophorus donor ligands

    International Nuclear Information System (INIS)

    Fryzuk, M.D.; Haddad, T.S.; Berg, D.J.

    1990-01-01

    Of relevance to this review are complexes of the early transition elements, in particular groups 3 and 4 and the lanthanides and actinides. In this review the authors have attempted to collect all the data up to the end of 1988 for complexed of groups 3 and 4, the lanthanides and the actinides that contain phosphorus donor ligands. The 1989s have seen a renaissance of the use of phosphine donors for the early d elements (groups 3 and 4) and the f elements. Neutral phosphorus donors are defined as primary (PH 2 R), secondary (PH 2 ) or tertiary phosphines (PR 3 ), including complexes of phosphine, PH 3 . Also reviewed are complexes of PF 3 and phosphites, P(OR) 3 . Specifically excluded are phosphido derivates, PR 2 . The ability of a neutral phosphorus donor to bind the metals of groups 3 and 4, the lanthanides and the actinides is now well established. While there are still no examples of lanthanum or actinium phosphine complexes, such derivatives should be accessible at least for lanthanum. series. However, there is no obvious chemical reason to suggest that such derivatives cannot be generated. The phosphine ligands that appear to generate the most stable phosphine-metal interaction are chelating phosphines such as dmpe, trmpe and trimpsi. In addition, the use of the chelate effect in conjunction with a hard ligand such as the amide in - N(SiMe 2 CH 2 PMe 2 ) 2 , or an alkoxide as found in - OC(BU t ) 2 CH 2 PMe 2 , also appears to be effective in anchoring the phosphine donor to the metal. The majority of low oxidation state derivatives of the group 4 elements are stabilized by phosphine donors in contrast with other parts of the transition series where one finds that classic π-acceptor-type ligands such as CO or RNC are utilized. 233 refs

  5. High pressure gas-filled cermet spark gaps

    International Nuclear Information System (INIS)

    Avilov, Eh.A.; Yur'ev, A.L.

    2000-01-01

    The results of modernization of the R-48 and R-49 spark gaps making it possible to improve their electrical characteristics are presented. The design is described and characteristics of gas-filled cermet spark gaps are presented. By the voltage rise time of 5-6 μs in the Marx generator scheme they provide for the pulse break-through voltage of 120 and 150 kV. By the voltage rise time of 0.5-1 μs the break-through voltage of these spark gaps may be increased up to 130 and 220 kV. The proper commutation time is equal to ≤ 0.5 ns. Practical recommendations relative to designing cermet spark gaps are given [ru

  6. Nuclear fuel activity with minor actinides after their useful life in a BWR; Actividad del combustible nuclear con actinidos menores despues de su vida util en un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.; Ramirez S, J. R.; Alonso V, G., E-mail: eduardo.martinez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    Nuclear fuel used in nuclear power reactors has a life cycle, in which it provides energy, at the end of this cycle is withdrawn from the reactor core. This used fuel is known as spent nuclear fuel, a strong problem with this fuel is that when the fuel was irradiated in a nuclear reactor it leaves with an activity of approximately 1.229 x 10{sup 15} Bq. The aim of the transmutation of actinides from spent nuclear fuel is to reduce the activity of high level waste that must be stored in geological repositories and the lifetime of high level waste; these two achievements would reduce the number of necessary repositories, as well as the duration of storage. The present work is aimed at evaluating the activity of a nuclear fuel in which radioactive actinides could be recycled to remove most of the radioactive material, first establishing a reference of actinides production in the standard nuclear fuel of uranium at end of its burning in a BWR, and a fuel rod design containing 6% of actinides in an uranium matrix from the enrichment tails is proposed, then 4 standard uranium fuel rods are replaced by 4 actinide bars to evaluate the production and transmutation of the same, finally the reduction of actinide activity in the fuel is evaluated. (Author)

  7. High-Temperature Tolerance in Multi-Scale Cermet Solar-Selective Absorbing Coatings Prepared by Laser Cladding.

    Science.gov (United States)

    Pang, Xuming; Wei, Qian; Zhou, Jianxin; Ma, Huiyang

    2018-06-19

    In order to achieve cermet-based solar absorber coatings with long-term thermal stability at high temperatures, a novel single-layer, multi-scale TiC-Ni/Mo cermet coating was first prepared using laser cladding technology in atmosphere. The results show that the optical properties of the cermet coatings using laser cladding were much better than the preplaced coating. In addition, the thermal stability of the optical properties for the laser cladding coating were excellent after annealing at 650 °C for 200 h. The solar absorptance and thermal emittance of multi-scale cermet coating were 85% and 4.7% at 650 °C. The results show that multi-scale cermet materials are more suitable for solar-selective absorbing coating. In addition, laser cladding is a new technology that can be used for the preparation of spectrally-selective coatings.

  8. High-Temperature Tolerance in Multi-Scale Cermet Solar-Selective Absorbing Coatings Prepared by Laser Cladding

    Directory of Open Access Journals (Sweden)

    Xuming Pang

    2018-06-01

    Full Text Available In order to achieve cermet-based solar absorber coatings with long-term thermal stability at high temperatures, a novel single-layer, multi-scale TiC-Ni/Mo cermet coating was first prepared using laser cladding technology in atmosphere. The results show that the optical properties of the cermet coatings using laser cladding were much better than the preplaced coating. In addition, the thermal stability of the optical properties for the laser cladding coating were excellent after annealing at 650 °C for 200 h. The solar absorptance and thermal emittance of multi-scale cermet coating were 85% and 4.7% at 650 °C. The results show that multi-scale cermet materials are more suitable for solar-selective absorbing coating. In addition, laser cladding is a new technology that can be used for the preparation of spectrally-selective coatings.

  9. Synthesis and structural characterization of actinide oxalate compounds

    International Nuclear Information System (INIS)

    Tamain, C.

    2011-01-01

    Oxalic acid is a well-known reagent to recover actinides thanks to the very low solubility of An(IV) and An(III) oxalate compounds in acidic solution. Therefore, considering mixed-oxide fuel or considering minor actinides incorporation in ceramic fuel materials for transmutation, oxalic co-conversion is convenient to synthesize mixed oxalate compounds, precursors of oxide solid solutions. As the existing oxalate single crystal syntheses are not adaptable to the actinide-oxalate chemistry or to their manipulation constrains in gloves box, several original crystal growth methods were developed. They were first validate and optimized on lanthanides and uranium before the application to transuranium elements. The advanced investigations allow to better understand the syntheses and to define optimized chemical conditions to promote crystal growth. These new crystal growth methods were then applied to a large number of mixed An1(IV)-An2(III) or An1(IV)-An2(IV) systems and lead to the formation of the first original mixed An1(IV)-An2(III) and An1(IV)-An2(IV) oxalate single crystals. Finally thanks to the first thorough structural characterizations of these compounds, single crystal X-ray diffraction, EXAFS or micro-RAMAN, the particularly weak oxalate-actinide compounds structural database is enriched, which is essential for future studied nuclear fuel cycles. (author) [fr

  10. System and safety studies of accelerator driven systems for transmutation. Annual report 2007

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, Vasily; Fokau, Andrei; Persson, Calle; Runevall, Odd; Sandberg, Nils; Tesinsky, Milan; Wallenius, Janne; Youpeng Zhang (Div. of Reactor Physics, Royal Institute of Technology, Stockholm (Sweden))

    2008-05-15

    Within the project 'System and safety studies of accelerator driven systems for transmutation', research on design and safety of sub-critical reactors for recycling of minor actinides is performed. During 2007, the reactor physics division at KTH has calculated safety parameters for EFIT-400 with cermet fuel, permitting to start the transient safety analysis. The accuracy of different reactivity meters applied to the YALINA facility was assessed and neutron detection studies were performed. A model to address deviations from point kinetic behaviour was developed. Studies of basic radiation damage physics included calculations of vacancy formation and activation enthalpies in bcc niobium. In order to predict the oxygen potential of inert matrix fuels, a thermo-chemical model for mixed actinide oxides was implemented in a phase equilibrium code

  11. Etude structurale et propriétés des verres peralumineux de conditionnement des produits de fission et actinides mineurs"

    OpenAIRE

    Gasnier , Estelle

    2013-01-01

    In this work, peraluminous glasses (lack of alkaline and alkaline earth ions regarding aluminum) are under study to assess the potentiality of these matrices to confine fission products and minor actinides (FPA) at higher rate than current R7T7 glass (18,5 wt % FPA). The first part of this work aims at studying the physical and chemical properties of complex peraluminous glasses containing increasing FPA rate (18.5 to 32 wt %) to compare them with the specifications. The very low crystallizat...

  12. Actinide separation by electrorefining

    International Nuclear Information System (INIS)

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

    1995-01-01

    TRUMP-S is a pyrochemical process being developed for the recovery of actinides from PUREX wastes. This paper describes development of the electrochemical partitioning step for recovery of actinides in the TRUMP-S process. The objectives are to remove 99 % of each actinide from PUREX wastes, with a product that is > 90 % actinides. Laboratory tests indicate that > 99 % of actinides can be removed in the electrochemical partitioning step. A dynamic (not equilibrium) process model predicts that 90 wt % product actinide content can be achieved through 99 % actinide removal. Accuracy of model simulation results were confirmed in tests with rare earths. (authors)

  13. Working of Mo-TiC cermets for 'future nuclear systems'

    International Nuclear Information System (INIS)

    Allemand, Alexandre; Le Flem, Marion; Rousselet, Jerome

    2006-01-01

    The nuclear reactor cores (generation IV) will form an extremely severe environment (high temperature, severe and long irradiation...). These drastic criteria and the preoccupation to ensure a higher and higher safety level lead, beyond the preoccupations due to the feasibility of such reactors, to harsh choices in materials able to be used. Innovating materials such as Mo-TiC cermet are the subject of intense researches in the CEA. This study presents and compares two modes of Mo-TiC cermet working: the hot isostatic compression and the extrusion. Different compositions of Mo-TiC cermets are prepared by hot isostatic compression and extrusion, and then characterized in term of microstructural properties. At last, this study concludes to a very satisfying working by hot isostatic compression, nevertheless the extrusion has still to be improved. (O.M.)

  14. An Advanced TALSPEAK Concept for Separating Minor Actinides. Part 1. Process Optimization and Flowsheet Development

    Energy Technology Data Exchange (ETDEWEB)

    Lumetta, Gregg J. [Pacific Northwest National Laboratory, Nuclear Science and Engineering Group, Richland, WA, USA; Levitskaia, Tatiana G. [Pacific Northwest National Laboratory, Nuclear Science and Engineering Group, Richland, WA, USA; Wilden, Andreas [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Nukleare Entsorgung und Reaktorsicherheit (IEK-6), Jülich, Germany; Casella, Amanda J. [Pacific Northwest National Laboratory, Nuclear Science and Engineering Group, Richland, WA, USA; Hall, Gabriel B. [Pacific Northwest National Laboratory, Nuclear Science and Engineering Group, Richland, WA, USA; Lin, Leigh [Pacific Northwest National Laboratory, Nuclear Science and Engineering Group, Richland, WA, USA; Sinkov, Sergey I. [Pacific Northwest National Laboratory, Nuclear Science and Engineering Group, Richland, WA, USA; Law, Jack D. [Idaho National Laboratory, Aqueous Separations and Radiochemistry Department, Idaho Falls, ID, USA; Modolo, Giuseppe [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Nukleare Entsorgung und Reaktorsicherheit (IEK-6), Jülich, Germany

    2017-08-18

    A system is being developed to separate trivalent actinides from lanthanide fission product elements that uses 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester to extract the lanthanide ions into an organic phase, while the actinide ions are held in the citrate-buffered aqueous phase by complexation to N-(2-hydroxyethyl)ethylenediamine-N,N',N'-triacetic acid (HEDTA). Earlier investigations of this system using a 2-cm centrifugal contactor revealed that the relatively slow extraction of Sm3+, Eu3+, and Gd3+ resulted in low separation factors from Am3+. In the work reported here, adjustments to the aqueous phase chemistry were made to improve the extraction rates. The results suggest that increasing the concentration of the citric acid buffer from 0.2 to 0.6 mol/L, and lowering the pH from 3.1 to 2.6, significantly improved lanthanide extraction rates resulting in an actinide/lanthanide separation system suitable for deployment in centrifugal contactors. Experiments performed to evaluate whether the lanthanide extraction rates can be improved by replacing aqueous HEDTA with nitrilotriacetic acid (NTA) exhibited promising results. However, NTA exhibited an unsatisfactorily high distribution value for Am3+ under the extraction conditions examined.

  15. Structure and friction properties of cermet and oxide explosion coatings

    International Nuclear Information System (INIS)

    RGrigorov, A.I.; Semenov, A.P.; Fed'ko, Yu.P.; Shtejn, L.M.

    1977-01-01

    Conditions have been specified for spraying explosion coatings of cermets over Kh18N10T stainless steel and an aluminum alloy. A mixture of WC and CO powders served as a material for spraying. The method of micro-X-ray spectrum analysis has been used to study the structure of the transition zone between the coating and the substrate and to establish the mechanism responsible for the formation of a cermet layer

  16. Effect of heating rate on the mechanical properties and microstructure of Ti(C,N)-based cermets

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Qingzhong; Ai, Xing, E-mail: aixingsdu@163.com; Zhao, Jun; Zhang, Hongshan; Qin, Wenzhen; Gong, Feng

    2015-03-25

    An appropriate heating rate in the sintering process is crucial to obtain the Ti(C,N)-based cermets with superior properties. In this paper, Ti(C,N)-based cermets were sintered to investigate the influence of heating rate on the mechanical properties and microstructure of the cermet materials. The transverse rupture strength (TRS), Vickers hardness (HV) and fracture toughness (K{sub IC}) were tested. The microstructure, indention crack, fracture morphology and phase composition of the cermets were also studied by scanning electron microscope (SEM) with energy dispersive spectroscopy (EDS) and X-ray diffraction (XRD). The results reveal that the heating rate has a great influence on the mechanical properties and microstructure of Ti(C,N)-based cermets. The cermets sintered at the heating rate of 3 °C/min between 1300 °C and 1430 °C have the optimum comprehensive mechanical properties with a transverse rupture strength of 1605±107 MPa, a hardness of 12.02±0.25 GPa and a fracture toughness of 10.73±0.40 MPa m{sup 1/2}. The heating rate can affect the reaction among the constituents of Ti(C,N)-based cermets and then influence the elements distribution in the core–rim microstructures and the lattice parameter of Ti(C,N) phase. When the heating rate is between 2 °C/min and 5 °C/min, the lower the heating rate is, the coarser the Ti(C,N) grains become. A higher heating rate is detrimental to the formation of core–rim microstructures, and a lower heating rate can result in grain coarsening and inhomogeneous microstructure. The observation of indention cracks and fracture surfaces show that the intergranular cracks and intergranular fractures mainly occur in the cermets with larger binder mean free path and medium grains. While the cleavage fractures appear more in the cermets with grain coarsening, and the transgranular fractures exist more in the cermets with non-fully developed fine grains.

  17. Behaviour of contact layer material between cermet fuel element core and can

    International Nuclear Information System (INIS)

    Gavrilin, S.S.; Permyakov, L.N.; Simakov, G.A.; Chernikov, A.S.

    1996-01-01

    The structural state of the contact layer between the shell of the Zr1Nb alloy and cermet fuel element core containing up to 70% of uranium dioxides is experimental studied. The silumin alloy was used as contact material. The results of studies on interaction zones, formed on the Zr1Nb - silumin boundary after fuel elements manufacture and also under temperature conditions, modeling the maximum design and hypothetical accidents accompanied by the contact material melting, are presented [ru

  18. High performance W-AlN cermet solar coatings designed by modelling calculations and deposited by DC magnetron sputtering

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Qi-Chu [School of Physics, The University of Sydney, Sydney, NSW 2006 (Australia); Shen, Y.G. [Department of Manufacturing Engineering and Engineering Management, City University of Hong Kong (Hong Kong)

    2004-01-25

    High solar performance W-AlN cermet solar coatings were designed using a numerical computer model and deposited experimentally. In the numerical calculations aluminium oxynitride (AlON) was used as ceramic component. The dielectric function and then complex refractive index of W-AlON cermet materials were calculated using the Sheng's approximation. The layer thickness and W metal volume fraction were optimised to achieve maximum photo-thermal conversion efficiency for W-AlON cermet solar coatings on an Al reflector with a surface AlON ceramic anti-reflection layer. Optimisation calculations show that the W-AlON cermet solar coatings with two and three cermet layers have nearly identical solar absorptance, emittance and photo-thermal conversion efficiency that are much better than those for films with one cermet layer. The optimised calculated AlON/W-AlON/Al solar coating film with two cermet layers has a high solar absorptance of 0.953 and a low hemispherical emittance of 0.051 at 80C for a concentration factor of 2. The AlN/W-AlN/Al solar selective coatings with two cermet layers were deposited using two metal target direct current magnetron sputtering technology. During the deposition of W-AlN cermet layer, both Al and W targets were run simultaneously in a gas mixture of argon and nitrogen. By substrate rotation a multi-sub-layer system consisting of alternating AlN ceramic and W metallic sub-layers was deposited that can be considered as a macro-homogeneous W-AlN cermet layer. A solar absorptance of 0.955 and nearly normal emittance of 0.056 at 80C have been achieved for deposited W-AlN cermet solar coatings.

  19. Cermet crucible for metallurgical processing

    Science.gov (United States)

    Boring, Christopher P.

    1995-01-01

    A cermet crucible for metallurgically processing metals having high melting points comprising a body consisting essentially of a mixture of calcium oxide and erbium metal, the mixture comprising calcium oxide in a range between about 50 and 90% by weight and erbium metal in a range between about 10 and 50% by weight.

  20. Cermet Ni-ZrO2 by mechanical alloying

    International Nuclear Information System (INIS)

    Leite, Douglas Will

    2010-01-01

    The ZrO 2 and metallic Ni Cermet obtained by Mechanical Alloying - MA is studied in the present work with the objective to prepare solid oxide fuel cells anodes (SOFC). Metallic Ni is added under three different concentrations: 30, 40 and 50% volume. The millings were conducted in SPEX vibratory mill where the influence of milling time, process control additives efficiency, type and geometry of milling vessels were studied. The study of the influence of these variables was made under particle size analysis, surface area determination and resulting material morphology. The use of teflon vessel causes contamination by carbon. On the other side, steel vessel increases the contamination by metallic impurities. The several geometries projected and analyzed for the vessels showed that vessels with larger bottom radius (R.15) showed the best results. After conformation and sintering at 1300 degree C in argon atmosphere the samples reached densities between 60 and 80% of the theoretical density. Microstructures observed by scanning electron microscopy reveal good homogeneity in the Cermet phases distribution. The mechanical alloying technique was considered a good option to obtain Ni- ZrO 2 Cermet. (author)

  1. A new look at actinide recycle

    International Nuclear Information System (INIS)

    Burch, W.D.; Croff, A.G.; Rawlins, J.A.; Schulz, W.W.

    1991-01-01

    This paper will address the justification for reexamination of the value of recovering the minor actinides and certain fission products from spent light-water reactor fuels and describe some of the technical progress that has been made since the major studies of a decade ago. During this time, the US Environmental Protection Agency (EPA) and the Nuclear Regulatory Commission have begun establishing detailed criteria and regulations for geologic repositories. An examination of the hazards of waste disposal relative to the EPA release standards reveals that removal of 99.9% of the actinides (Pu, Am, and Np) reduces these hazards quite close to the EPA standards after 300 years' decay of the strontium and cesium. It may be also useful to remove and separately manage and dispose of certain of the long-lived fission products, such as 99 Tc and 129 I. Much additional work is required to fully assess the appropriate target recoveries as the hazards and risks are more closely examined and as the standards are reworked and refined. The two decades before the projected start of the US repository may present a window of opportunity to introduce several better management practices that act to simplify the repository safety issues. From a technical standpoint, significant progress has been made on recovery of the actinides from aqueous wastes though use of the TRUEX process. Additional work is required to demonstrate the application of the process to spent LWR fuels, but it appears straightforward. In addition, work at the Argonne National Laboratory on the liquid-metal reactor metal fuel cycle shows the relative simplicity of recycle of the actinides in that fast reactor cycle. Much work remains to fully demonstrate that actinides from all secondary waste streams can be removed to the target levels from both the aqueous reprocessing of LWR fuel and the pyro processes for the metal-fueled fast reactor. 9 refs., 2 figs

  2. Preparation of refractory cermet structures for lithium compatibility testing

    Science.gov (United States)

    Heestand, R. L.; Jones, R. A.; Wright, T. R.; Kizer, D. E.

    1973-01-01

    High-purity nitride and carbide cermets were synthesized for compatability testing in liquid lithium. A process was developed for the preparation of high-purity hafnium nitride powder, which was subsequently blended with tungsten powder or tantalum nitride and tungsten powders and fabricated into 3 in diameter billets by uniaxial hot pressing. Specimens were then cut from the billets for compatability testing. Similar processing techniques were applied to produce hafnium carbide and zirconium carbide cermets for use in the testing program. All billets produced were characterized with respect to chemistry, structure, density, and strength properties.

  3. Computer simulation of the optical properties of high-temperature cermet solar selective coatings

    Energy Technology Data Exchange (ETDEWEB)

    Nejati, M. Reza [K.N. Toosi Univ. of Technology, Dept. of Mechanical Engineering, Tehran (Iran); Fathollahi, V.; Asadi, M. Khalaji [AEOI, Center for Renewable Energy Research and Applications (CRERA), Tehran (Iran)

    2005-02-01

    A computer simulation is developed to calculate the solar absorptance and thermal emittance of various configurations of cermet solar selective coatings. Special attention has been paid to those material combinations, which are commonly used in high-temperature solar thermal applications. Moreover, other material combinations such as two-, three- and four-cermet-layer structures as solar selective coatings have been theoretically analyzed by computer simulation using three distinct physical models of Ping Sheng, Maxwell-Garnett and Bruggeman. The novel case of two-cermet-layer structure with different cermet components has also been investigated. The results were optimized by allowing the program to manipulate the metal volume fraction and thickness of each layer and the results compared to choose the best possible configuration. The calculated results are within the range of 0.91-0.97 for solar absorptance and 0.02-0.07 for thermal emittance at room temperature. (Author)

  4. The preparation of titanium-vanadium carbide/nickel cermets. Technical report

    International Nuclear Information System (INIS)

    Precht, W.; Sprissler, B.

    1976-01-01

    Titanium/vanadium alloy carbide rods were prepared by a zone melting procedure. Wetting studies were carried out using sections of the fused rods and candidate matrix material. It was established that nickel exhibits excellent wetting of (Ti, V) C, and accordingly cermet blends were prepared and liquid phase sintered. Processing parameters are discussed as well as their effect on the final microstructure. Alternate methods for cermet preparation are offered which use as received titanium carbide and vanadium carbide powders

  5. Literature review of thermal and radiation performance parameters for high-temperature, uranium dioxide fueled cermet materials

    International Nuclear Information System (INIS)

    Haertling, C.; Hanrahan, R.J.

    2007-01-01

    High-temperature fissile-fueled cermet literature was reviewed. Data are presented primarily for the W-UO 2 as this was the system most frequently studied; other reviewed systems include cermets with Mo, Re, or alloys as a matrix. Failure mechanisms for the cermets are typically degradation of mechanical integrity and loss of fuel. Mechanical failure can occur through stresses produced from dissimilar expansion coefficients, voids created from diffusion of dissimilar materials or formation of metal hydride and subsequent volume expansion. Fuel loss failure can occur by high temperature surface vaporization or by vaporization after loss of mechanical integrity. Techniques found to aid in retaining fuel include the use of coatings around UO 2 fuel particles, use of oxide stabilizers in the UO 2 , minimizing grain sizes in the metal matrix, minimizing impurities, controlling the cermet sintering atmosphere, and cladding around the cermet

  6. Investigation on microstructure and mechanical properties of Mo2FeB2 based cermets with and without PVA

    Science.gov (United States)

    Shen, Yupeng; Huang, Zhifu; Jian, Yongxin; Yang, Ming; Li, Kemin

    2018-03-01

    Mo2FeB2 based cermets with and without PVA have been investigated by x-ray diffractometry (XRD), x-ray photoelectron spectroscope (XPS) and scanning electron microscopy (SEM). The density and transverse rupture strength (TRS) of green compact, relative density, hardness (HRA), fracture toughness (KIC) and TRS of Mo2FeB2 based cermets were also measured. The results indicate that, compared with the Mo2FeB2 based cermets without PVA, the density of green compact with PVA can be improved slightly at the same pressure. However, the much higher TRS is obtained for the green compact without PVA. Meanwhile, Mo2FeB2 particles exhibit the finer and less congruity feature for Mo2FeB2 based cermets without PVA. In addition, the higher relative density, hardness, fracture toughness and TRS can be acquired for the cermets without PVA. Obviously, considering the mechanical properties and preparation period of Mo2FeB2 based cermets, no adding PVA is the optimized process of powder molding in the manufacture of Mo2FeB2 based cermets.

  7. Osteoblastic cell response to spark plasma-sintered zirconia/titanium cermets.

    Science.gov (United States)

    Fernandez-Garcia, Elisa; Guillem-Marti, Jordi; Gutierrez-Gonzalez, Carlos F; Fernandez, Adolfo; Ginebra, Maria-Pau; Lopez-Esteban, Sonia

    2015-01-01

    Ceramic/metal composites, cermets, arise from the idea to combine the dissimilar properties in the pure materials. This work aims to study the biocompatibility of new micro-nanostructured 3 Y-TZP/Ti materials with 25, 50 and 75 vol.% Ti, which have been successfully obtained by spark slasma sintering technology, as well as to correlate their surface properties (roughness, wettability and chemical composition) with the osteoblastic cell response. All samples had isotropic and slightly waved microstructure, with sub-micrometric average roughness. Composites with 75 vol.% Ti had the highest surface hydrophilicity. Surface chemical composition of the cermets correlated well with the relative amounts used for their fabrication. A cell viability rate over 80% dismissed any cytotoxicity risk due to manufacturing. Cell adhesion and early differentiation were significantly enhanced on materials containing the nanostructured 3 Y-TZP phase. Proliferation and differentiation of SaOS-2 were significantly improved in their late-stage on the composite with 75 vol.% Ti that, from the osseointegration standpoint, is presented as an excellent biomaterial for bone replacement. Thus, spark plasma sintering is consolidated as a suitable technology for manufacturing nanostructured biomaterials with enhanced bioactivity. © The Author(s) 2014 Reprints and permissions: sagepub.co.uk/journalsPermissions.nav.

  8. Japanese Fast Reactor Program for Homogeneous Actinide Recycling

    International Nuclear Information System (INIS)

    Ishikawa, Makoto; Nagata, Takashi; Kondo, Satoru

    2008-01-01

    In the present report, the homogeneous actinide recycling scenario of Fast Reactor (FR) Cycle Technology Development Project (FaCT) is summarized. First, the scenario of nuclear energy policy in Japan are briefly reviewed. Second, the basic plan of Japan to manage all minor actinide (MA) by recycling is summarized objectives of which are the efficiency increase of uranium resources, the environmental burden reduction, and the increase of nuclear non-proliferation potential. Third, recent results of reactor physics study related to MA-loaded FR cores are briefly described. Fourth, typical nuclear design of MA-loaded FR cores in the FaCT project and their main features are demonstrated with the feasibility to recycle all MA in the future FR equilibrium society. Finally, the research and development program to realize the MA recycling in Japan is introduced, including international cooperation projects. (authors)

  9. Sealing ability of cermet ionomer cement as a retrograde filling material.

    Science.gov (United States)

    Aktener, B O; Pehlivan, Y

    1993-03-01

    An in vitro dye leakage study was performed to compare the sealing ability of high copper amalgam with cavity varnish and cermet ionomer cement with and without varnish when used as retrofilling materials. The root canals of 54 maxillary anterior teeth were instrumented and obturated with gutta-percha and sealer. The apical 3 mm of the roots were resected and apical class I cavity preparations were made. The roots were then randomly divided into three groups and retrofilled with one of the experimental materials. After 72 h of immersion in India ink, the roots were cleared and evaluated for leakage with a stereomicroscope. Statistical analysis indicated that the cermet ionomer cement with varnish group had significantly less leakage than the amalgam group (P cermet ionomer cement without varnish group (P 0.05).

  10. The effectiveness of the ELSY concept with respect to minor actinide transmutation capability

    International Nuclear Information System (INIS)

    Grasso, Giacomo; Rocchi, Federico; Sumini, Marco; Artioli, Carlo; Monti, Stefano

    2010-01-01

    The task of partitioning and transmutation (PT) aims at the sustainability of new global nuclear scenarios for energy production, required by a continuously growing demand. The nuclear renaissance boosted by the breaking need of a reduction in CO 2 emissions, together with increasing safety and security requirements, is creating a clear interest in the Generation-IV philosophy. In particular, an effective management of minor actinides (MA) and their multi-recycling in innovative fast spectrum systems can lead to a minimisation of high-level wastes (HLW) to be disposed of in geological repositories. This study presents a PT application based on the European Lead-cooled System (ELSY), the 600 MWe Gen-IV lead-cooled fast reactor (LFR) under investigation in Europe within the 6. EURATOM Framework Programme. An 'adiabatic' core configuration is investigated here, for a system which can maintain a constant amount of both MA and plutonium during the whole fuel cycle, even without either axial or radial blankets. It is shown that an equilibrium concentration of MA exists, for which its production rate is exactly compensated by its transmutation rate. Any other concentration may enhance either their production or removal in such a way as to allow the system to evolve almost exponentially towards the equilibrium state. The practical feasibility of such an equilibrium core is then analysed: acceptable operative conditions might lead to a 'sustainable' nuclear system, the overall net outcome of which is the production of energy by burning a feed stream of uranium from one side, and the discharge of fission products (FP) only from the other side. (authors)

  11. Assessment of polyphase sintered iron-cobalt-iron boride cermets

    International Nuclear Information System (INIS)

    Nowacki, J.; Pieczonka, T.

    2004-01-01

    Sintering of iron, cobalt and boron powders has been analysed. As a result iron-iron boride, Fe-Fe 2 B and iron/cobalt boride with a slight admixture of molybdenum, Fe - Co - (FeMoCo) 2 B cermets have been produced. Iron was introduced to the mixture as the Astalloy Mo Hoeganaes grade powder. Elemental amorphous boron powder was used, and formation of borides occurred both during heating and isothermal sintering periods causing dimensional changes of the sintered body. Dilatometry was chosen to control basic phenomena taking place during multiphase sintering of investigated systems. The microstructure and phase constituents of sintered compacts were controlled as well. The cermets produced were substituted to: metallographic tests, X-ray analysis, measurements of hardness and of microhardness, and of wear in the process of sliding dry friction. Cermets are made up of two phases; hard grains of iron - cobalt boride, (FeCo) 2 B (1800 HV) constituting the reinforcement and a relatively soft and plastic eutectic mixture Fe 2 B - Co (400-500 HV) constituting the matrix. (author)

  12. Treatment of root fracture with accompanying resorption using cermet cement.

    Science.gov (United States)

    Lui, J L

    1992-02-01

    A method of treating an apical root fracture with accompanying resorption at the junction of the fracture fragments using glass-cermet cement is described. Endodontically, the material had previously been used for repair of lateral resorptive root defects and retrograde root fillings. Complete bone regeneration was observed three years post-operatively following treatment of the root fracture in the conventional manner. The various advantages of glass-cermet cement as a root filling material used in the technique described are discussed.

  13. Experimental Evaluation of Cermet Turbine Stator Blades for Use at Elevated Gas Temperatures

    Science.gov (United States)

    Chiarito, Patrick T.; Johnston, James R.

    1959-01-01

    The suitability of cermets for turbine stator blades of a modified turbojet engine was determined at an average turbine-inlet-gas temperature of 2000 F. Such an increase in temperature would yield a premium in thrust from a service engine. Because the cermet blades require no cooling, all the available compressor bleed air could be used to cool a turbine made from conventional ductile alloys. Cermet blades were first run in 100-hour endurance tests at normal gas temperatures in order to evaluate two methods for mounting them. The elevated gas-temperature test was then run using the method of support considered best for high-temperature operation. After 52 hours at 2000 F, one of the group of four cermet blades fractured probably because of end loads resulting from thermal distortion of the spacer band of the nozzle diaphragm. Improved design of a service engine would preclude this cause of premature failure.

  14. Effects of metal binder on the microstructure and mechanical properties of Ti(C,N)-based cermets

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Qingzhong; Ai, Xing, E-mail: aixingsdu@163.com; Zhao, Jun; Gong, Feng; Pang, Jiming; Wang, Yintao

    2015-09-25

    Highlights: • Ni–Co binder improves the solid solution reaction and the wetting of hard phases. • Cermets with 25 wt.% binder have evenly distributed grains with moderate rims. • Co/(Ni + Co) ratios influence the grain sizes and microstructure features of cermets. • The cermets with pure Co as binder exhibit optimal mechanical properties. - Abstract: To optimize the mechanical properties of Ti(C,N)-based cermets used as tool materials, the cermets with different Ni–Co binder contents and Co/(Ni + Co) weight ratios were prepared. The effects of metal binder content and Co/(Ni + Co) ratio on the microstructure and mechanical properties of Ti(C,N)-based cermets were investigated by scanning electron microscope (SEM) with energy dispersive spectroscopy (EDS), X-ray diffraction (XRD), and measuring the transverse rupture strength (TRS), Vickers hardness (HV) and fracture toughness (K{sub IC}). The experimental results reveal that increasing Ni–Co binder content can increase the thickness of rim phases by improving the solid solution reaction and the wetting of hard phases. The cermets with 25 wt.% binder addition present good comprehensive mechanical properties, which is attributed to the moderate rim phases and uniformly distributed Ti(C,N) grains. The Co/(Ni + Co) weight ratios in binder have a great influence on the grain sizes and microstructure features of Ti(C,N)-based cermets, in virtue of the synergic effects between the wettability of Co and the solubilizing capacity of Ni on hard phases. The cermets with pure Co as binder exhibit optimal mechanical properties with a TRS of 1767 ± 81 MPa, a hardness of 12.26 ± 0.10 GPa and a K{sub IC} of 8.40 ± 0.47 MPa m{sup 1/2}, which meet the requirements for tool materials. And the cermets with a Co/(Ni + Co) ratio of 0.2 have the second best mechanical properties with a TRS of 1848 ± 201 MPa, a hardness of 11.12 ± 0.40 GPa and a K{sub IC} of 9.43 ± 0.54 MPa m{sup 1/2}, in which the lower hardness can

  15. Comparison of Ti(C,N)-based cermets processed by hot-pressing sintering and conventional pressureless sintering

    International Nuclear Information System (INIS)

    Xu, Qingzhong; Ai, Xing; Zhao, Jun; Qin, Weizhen; Wang, Yintao; Gong, Feng

    2015-01-01

    Highlights: • The HP sintered Ti(C,N)-based cermets exhibit high hardness with fine grain size. • The PLS sintered cermets possess high mechanical properties with low porosity. • The applied pressure can rearrange particles and contribute to grain refinement. • The heating rate can greatly affect the solid and liquid phase sintering of cermets. - Abstract: A suitable sintering method is important to obtain the Ti(C,N)-based cermets with superior properties. In this paper, Ti(C,N)-based cermets were fabricated by hot-pressing sintering (HP) and conventional pressureless sintering (PLS) technology, respectively, to investigate the influence of different sintering methods on the microstructure and mechanical properties of cermets materials. The microstructure, fracture morphology, indention cracks and phase composition were observed and detected using scanning electron microscope (SEM), energy dispersive spectroscopy (EDS) and X-ray diffraction (XRD). The transverse rupture strength (TRS), Vickers hardness (HV) and fracture toughness (K IC ) were also measured. The results reveal that all of the Ti(C,N)-based cermets exhibit core–rim microstructures with black cores, white cores and grey rims embedded into metal binder phases. The grain size of the samples fabricated by HP is much finer and the structure is more compact than those fabricated by PLS, while there exist pores in the HP sintered samples. The sintering process has no influence on the phase composition of cermets, but affects the phase content and crystallinity. The samples fabricated by PLS present higher transverse rupture strength, fracture toughness and density than samples fabricated by HP. However, the HP sintered samples possess a higher hardness

  16. Nuclear data of the major actinide fuel materials

    Energy Technology Data Exchange (ETDEWEB)

    Poenitz, W.P.; Saussure, G. De

    1984-01-01

    The effect of nuclear data of the major actinide fuel materials on the design accuracy, economics and safety of nuclear power systems is discussed. Since most of the data are measured relative to measurement standards, in particular the fission cross-section of /sup 235/U, data must be examined to ensure that absolute measurements and relative measurements are correctly handled. Nuclear data of fissile materials, fertile materials and minor plutonium isotopes are discussed.

  17. The behaviour under irradiation of molybdenum matrix for inert matrix fuel containing americium oxide (CerMet concept)

    Energy Technology Data Exchange (ETDEWEB)

    D' Agata, E., E-mail: elio.dagata@ec.europa.eu [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, 1755 ZG Petten (Netherlands); Knol, S.; Fedorov, A.V. [Nuclear Research and Consultancy Group, P.O. Box 25, 1755 ZG Petten (Netherlands); Fernandez, A.; Somers, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Klaassen, F. [Nuclear Research and Consultancy Group, P.O. Box 25, 1755 ZG Petten (Netherlands)

    2015-10-15

    Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors or Accelerator Driven System (ADS, subcritical reactors dedicated to transmutation) of long-lived nuclides like {sup 241}Am is therefore an option for the reduction of radiotoxicity of waste packages to be stored in a repository. In order to safely burn americium in a fast reactor or ADS, it must be incorporated in a matrix that could be metallic (CerMet target) or ceramic (CerCer target). One of the most promising matrix to incorporate Am is molybdenum. In order to address the issues (swelling, stability under irradiation, gas retention and release) of using Mo as matrix to transmute Am, two irradiation experiments have been conducted recently at the High Flux Reactor (HFR) in Petten (The Netherland) namely HELIOS and BODEX. The BODEX experiment is a separate effect test, where the molybdenum behaviour is studied without the presence of fission products using {sup 10}B to “produce” helium, the HELIOS experiment included a more representative fuel target with the presence of Am and fission product. This paper covers the results of Post Irradiation Examination (PIE) of the two irradiation experiments mentioned above where molybdenum behaviour has been deeply investigated as possible matrix to transmute americium (CerMet fuel target). The behaviour of molybdenum looks satisfying at operating temperature but at high temperature (above 1000 °C) more investigation should be performed.

  18. Actinides reduction by recycling in a thermal reactor; Reduccion de actinidos por reciclado en un reactor termico

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Martinez C, E.; Balboa L, H., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    This work is directed towards the evaluation of an advanced nuclear fuel cycle in which radioactive actinides could be recycled to remove most of the radioactive material; firstly a production reference of actinides in standard nuclear fuel of uranium at the end of its burning in a BWR reactor is established, after a fuel containing plutonium is modeled to also calculate the actinides production in MOX fuel type. Also it proposes a design of fuel rod containing 6% of actinides in a matrix of uranium from the tails of enrichment, then four standard uranium fuel rods are replaced by actinides rods to evaluate the production and transmutation thereof, the same procedure was performed in the fuel type MOX and the end actinide reduction in the fuel was evaluated. (Author)

  19. Study of Some Innovant Reactors without on- Site Refueling with Triso and Cermet Fuel

    OpenAIRE

    A.Chetaine; A. Benchrif; H. Amsil; V. Kuznetsov; Y. Shimazu

    2012-01-01

    The evaluation of unit cell neutronic parameters and lifetime for some innovant reactors without on sit-refuling will be held in this work. the behavior of some small and medium reactors without on site refueling with triso and cermet fuel. For the FBNR long life except we propose to change the enrichment of the Cermet MFE to 9%. For the AFPR reactor we can see that the use of the Cermet MFE can extend the life of this reactor but to maintain the same life period for AFPR...

  20. Soudage par explosion thermique sous charge de cermets poreux à base de TiC-Ni sur substrat en acier-comportement tribologique Welding of porous TiC–Ni based cermets on substrate steel by thermal explosion under load-tribological behaviour

    Directory of Open Access Journals (Sweden)

    Lemboub Samia

    2013-11-01

    Full Text Available Dans ce travail, nous nous intéressons à l'élaboration de cermets à base de TiC-Ni par dispersion de particules de carbures, oxydes ou borures dans une matrice de nickel, grâce à la technique de l'explosion thermique sous une charge de 20 MPa. La combustion de mélanges actifs (Ti-C-Ni-An où An = Al2O3, MgO, SiC, TiB2, WC, basée sur la réaction de synthèse de TiC (ΔHf298K = −184 kJ/mole, génère des cermets complexes. Un court maintien sous charge du cermet à 1373 K, après l'explosion thermique, permet son soudage sur un substrat en acier XC55. Les cermets obtenus dans ces conditions demeurent poreux et conservent une porosité de l'ordre de 25–35 %. La densité relative du cermet, sa dureté et son comportement tribologique, dépendront de la nature de l'addition dans les mélanges de départ. Porous TiC-Ni based cermets were obtained by dispersion of carbides, oxides or borides particles in a nickel matrix thanks to the thermal explosion technique realized under a load of 20 MPa. The combustion of active mixtures (Ti-C-Ni-An where An = Al2O3, MgO, SiC, TiB2 or WC based on the titanium carbide reaction synthesis (ΔHf = −184 kJ/mol, generates porous complex cermets. After the thermal explosion, a short maintenance under load at 1373 K of the combustion product, allows at the same time the cermets welding on a carbon steel substrate. The obtained cermets under these conditions preserve a porosity of about 25–35%. The relative density, hardness and tribological behaviour of the complex cermets depend on the additions nature (An in the starting mixtures.

  1. Irradiation behaviour of UO2/Mo porous cermets for thermionic converters

    International Nuclear Information System (INIS)

    Stora, J.P.; Kauffmann, Y.

    1975-01-01

    Two types of UO 2 Mo porous cernets have been fabricated and irradiated in a Cythere irradiation device. The first cermet is constituted by little bits of dense fuel in which the two constituants are finely dispersed. The whole open porosity is located between the granules. This type of cermet is called breche (33.4vol%UO 2 , 51vol%Mo, 14.8vol%porosity). At the end of the irradiation the burn up was 19000MWd/t(U) and neither swelling of the cermet nor deformation of the can were noted. On the contrary, a shrinkage of the emitter was observed attributed to a fuel densification under irradiation. The second type of cermet is called macrogranule (36vol%UO 2 , 49vol%Mo 15vol%porosity). UO 2 granules of 0.07cm mean diameter are dispersed in the molybdenum matrix. The porosity is regularly distributed all around the UO 2 kernels. The post irradiation metrology shows that the emitter is fairly stable. Only a slight ovalisation of about 0.5% was noted, but the granules of UO 2 were redistributed inside the molybdenum matrix, overlapping the metallic cavity by a condensation-evaporation process. The matrix has crept into the central void and consequently the volume has grown and the whole porosity has increased from about 15% to about 23%. This creeping is due to the fission gas pressure in the molybdenum cavities after 3000 hours of irradiation. In conclusion two types of cermets have shown good behaviour under irradiation and should allow lifetimes of several thousand hours of operation for thermionic fuel elements [fr

  2. Strength and fracture mechanism of iron reinforced tricalcium phosphate cermet fabricated by spark plasma sintering.

    Science.gov (United States)

    Tkachenko, Serhii; Horynová, Miroslava; Casas-Luna, Mariano; Diaz-de-la-Torre, Sebastian; Dvořák, Karel; Celko, Ladislav; Kaiser, Jozef; Montufar, Edgar B

    2018-05-01

    The present work studies the microstructure and mechanical performance of tricalcium phosphate (TCP) based cermet toughened by iron particles. A novelty arises by the employment of spark plasma sintering for fabrication of the cermet. Results showed partial transformation of initial alpha TCP matrix to beta phase and the absence of oxidation of iron particles, as well as a lack of chemical reaction between TCP and iron components during sintering. The values of compressive and tensile strength of TCP/Fe cermet were 3.2 and 2.5 times, respectively, greater than those of monolithic TCP. Fracture analysis revealed the simultaneous action of crack-bridging and crack-deflection microstructural toughening mechanisms under compression. In contrast, under tension the reinforcing mechanism was only crack-bridging, being the reason for smaller increment of strength. Elastic properties of the cermet better matched values reported for human cortical bone. Thereby the new TCP/Fe cermet has potential for eventual use as a material for bone fractures fixation under load-bearing conditions. Copyright © 2018 Elsevier Ltd. All rights reserved.

  3. Actinide oxide photodiode and nuclear battery

    Energy Technology Data Exchange (ETDEWEB)

    Sykora, Milan; Usov, Igor

    2017-12-05

    Photodiodes and nuclear batteries may utilize actinide oxides, such a uranium oxide. An actinide oxide photodiode may include a first actinide oxide layer and a second actinide oxide layer deposited on the first actinide oxide layer. The first actinide oxide layer may be n-doped or p-doped. The second actinide oxide layer may be p-doped when the first actinide oxide layer is n-doped, and the second actinide oxide layer may be n-doped when the first actinide oxide layer is p-doped. The first actinide oxide layer and the second actinide oxide layer may form a p/n junction therebetween. Photodiodes including actinide oxides are better light absorbers, can be used in thinner films, and are more thermally stable than silicon, germanium, and gallium arsenide.

  4. Actinide-handling experience for training and education of future expert under J-ACTINET

    International Nuclear Information System (INIS)

    Osaka, Masahiko; Sato, Isamu; Miwa, Shuhei; Konashi, Kenji; Li, Dexin; Homma, Yoshiya; Yamamura, Tomoo; Hayashi, Hirokazu; Minato, Kazuo; Sekimoto, Syun; Kubota, Takumi; Fukutani, Satoshi; Hori, Junichi; Okumura, Ryo; Uehara, Akihiro; Fujii, Toshiyuki; Yamana, Hajimu; Kurosaki, Ken; Muta, Hiroaki; Ohishi, Yuji; Yamanaka, Shinsuke; Uno, Masayoshi; Yaita, Tsuyoshi

    2011-01-01

    Summer schools for future experts have successfully been completed under Japan Actinide Network (J-ACTINET) for the purpose of development of human resources who are expected to be engaged in every areas of actinide-research/engineering. The first summer school was held in Ibaraki-area in August 2009, followed by the second one in Kansai-area in August 2010. Two summer schools have focused on actual experiences of actinides in actinide-research fields for university students and young researchers/engineers as an introductory course of actinide-researches. Many efforts were made to awaken interests into actinide-researches inside the participants during short periods of schools, 3 to 4 days. As actinides must be handled inside special apparatuses such as an air-tight globe-box with well-trained and qualified technicians, programs were optimized for effective experiences of actinides-handling. Several quasi actinide-handling experiences at the actinide-research fields have attracted attentions of participants at the first school in Ibaraki-area. The actual experiments using actinides-containing solutions have been carried out at the second school in Kansai-area. Future summer schools will be held every year for the sustainable human resource development in various actinide-research fields, together with other training and education programs conducted by the J-ACTINET. (author)

  5. Thermodynamic Properties of Actinides and Actinide Compounds

    Science.gov (United States)

    Konings, Rudy J. M.; Morss, Lester R.; Fuger, Jean

    The necessity of obtaining accurate thermodynamic quantities for the actinide elements and their compounds was recognized at the outset of the Manhattan Project, when a dedicated team of scientists and engineers initiated the program to exploit nuclear energy for military purposes. Since the end of World War II, both fundamental and applied objectives have motivated a great deal of further study of actinide thermodynamics. This chapter brings together many research papers and critical reviews on this subject. It also seeks to assess, to systematize, and to predict important properties of the actinide elements, ions, and compounds, especially for species in which there is significant interest and for which there is an experimental basis for the prediction.

  6. Study of behavior of cermet fuel elements on IGR reactor under RIA type accident condition

    International Nuclear Information System (INIS)

    Vasil'ev, Yu.S.; Vurim, A.D.; Koltyshev, S.M.; Pakhnits, V.A.; Tukhvatulin, Sh.T.; Popov, V.V.; Ryzhkov, A.N.

    1996-01-01

    In 1993 December in IGR reactor of Inst. of Atomic Energy of National Nuclear Center of Republic of Kazakstan the second batch of in-pile testing of perspective cermet fuel elements under the condition, simulating RIA type accident was conducted. In the second batch of testing during eight start-ups 10 cermet fuel elements were examined. Among which 8 of monolith type and 2 fuel elements with false jacket beside cladding (FJF), as well as, 6 standard fuel elements of WWER-1000 type reactor with dioxide fuel were tested. 2 fuel elements - cermet and standard were placed into capsule filled with water. To measure energy release for the each start-up two fission monitor and inside core control gauge were placed. In all the start-ups operation mode of IGR was neutron pulse. Power of fuel element kept changing from 151 to 336 k W; energy release was 38-93 kJ/gr m 235 U; maximum temperature of cermet fuel was 1943-2173 K, of dioxide fuel - 1923-2843 K. The testing has demonstrated that operability of cermet fuel elements under reactivity accident condition with pulse width of 0,2 s is, at least, not less that operability of dioxide fuel elements, through advantages of cermet fuel under these conditions are revealed to the least extent

  7. Effect of composition on the degree of anisotropy of thermal expansion and electric resistance of cermet specimens of GeTe

    International Nuclear Information System (INIS)

    Barbakadze, K.G.; Vekua, T.S.; Ioseliani, M.I.; Kvitsiniya, K.M.

    1988-01-01

    A study was made on α temperature coefficient of thermal expansion and ρ specific electric resistance of cermet germanium telluride for alloys close to stoichiometric composition. It is shown that anisotropy of thermal expansion of cermet germanium telluride depends sufficiently on its composition. This dependence is clearly pronounced if tellurium content in alloys equals 50.4-51.2 at.%. The maximal anisotropy is observed in the alloy containing 50.8 at.% of tellurium. The temperature of extreme value of temperature coefficient of linear expansion decreases from 440 down to 373 deg.C for alloys with 49-50.8 at.% of tellurium, and grows from 373 up to 405 deg.C if tellurium content equals 50.8-52 at.%

  8. High temperature oxidation resistant cermet compositions

    Science.gov (United States)

    Phillips, W. M. (Inventor)

    1976-01-01

    Cermet compositions are designed to provide high temperature resistant refractory coatings on stainless steel or molybdenum substrates. A ceramic mixture of chromium oxide and aluminum oxide form a coating of chromium oxide as an oxidation barrier around the metal particles, to provide oxidation resistance for the metal particles.

  9. Use of cermet thin film resistors with nitride passivated metal insulator field effect transistor

    Science.gov (United States)

    Brown, G. A.; Harrap, V.

    1971-01-01

    Film deposition of cermet resistors on same chip with metal nitride oxide silicon field effect transistors permits protection of contamination sensitive active devices from contaminants produced in cermet deposition and definition processes. Additional advantages include lower cost, greater reliability, and space savings.

  10. Technical requirements for the actinide source-term waste test program

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, M.L.F.; Molecke, M.A.

    1993-10-01

    This document defines the technical requirements for a test program designed to measure time-dependent concentrations of actinide elements from contact-handled transuranic (CH TRU) waste immersed in brines similar to those found in the underground workings of the Waste Isolation Pilot Plant (WIPP). This test program wig determine the influences of TRU waste constituents on the concentrations of dissolved and suspended actinides relevant to the performance of the WIPP. These influences (which include pH, Eh, complexing agents, sorbent phases, and colloidal particles) can affect solubilities and colloidal mobilization of actinides. The test concept involves fully inundating several TRU waste types with simulated WIPP brines in sealed containers and monitoring the concentrations of actinide species in the leachate as a function of time. The results from this program will be used to test numeric models of actinide concentrations derived from laboratory studies. The model is required for WIPP performance assessment with respect to the Environmental Protection Agency`s 40 CFR Part 191B.

  11. Technical requirements for the actinide source-term waste test program

    International Nuclear Information System (INIS)

    Phillips, M.L.F.; Molecke, M.A.

    1993-10-01

    This document defines the technical requirements for a test program designed to measure time-dependent concentrations of actinide elements from contact-handled transuranic (CH TRU) waste immersed in brines similar to those found in the underground workings of the Waste Isolation Pilot Plant (WIPP). This test program wig determine the influences of TRU waste constituents on the concentrations of dissolved and suspended actinides relevant to the performance of the WIPP. These influences (which include pH, Eh, complexing agents, sorbent phases, and colloidal particles) can affect solubilities and colloidal mobilization of actinides. The test concept involves fully inundating several TRU waste types with simulated WIPP brines in sealed containers and monitoring the concentrations of actinide species in the leachate as a function of time. The results from this program will be used to test numeric models of actinide concentrations derived from laboratory studies. The model is required for WIPP performance assessment with respect to the Environmental Protection Agency's 40 CFR Part 191B

  12. Comparative studies of CERCER and CERMET fuels for EFIT from the viewpoint of core performance and safety

    International Nuclear Information System (INIS)

    Chen, X.N.; Rineiski, A.; Maschek, W.; Liu, P.; Boccaccini, C.M.; Sobolev, V.; Delage, F.; Rimpault, G.

    2011-01-01

    The European Facility for Industrial Transmutation (EFIT) has been developed within the 6. EU Framework by the EUROTRANS Program, aiming at a generic conceptual design of an accelerator driven transmuter. This paper deals with assessments of EFIT cores with CERCER and CERMET fuels from the viewpoint of core performance and safety. The conclusive remarks can be drawn as follows. Because of its much better thermal conductivity, the CERMET core can be designed by using thicker pins, so that it has the same or even better transmutation performance compared to the CERCER core. Both CERCER and CERMET fuels fulfill safety requirements. Moreover the CERMET fuel has higher fuel safety margins than the CERCER one. Preliminary analyses show that the CERMET total core power can be further increased by 50% at least without exceeding fuel and clad temperature limits. (authors)

  13. Metal-Matrix Hardmetal/Cermet Reinforced Composite Powders for Thermal Spray

    Directory of Open Access Journals (Sweden)

    Dmitri GOLJANDIN

    2012-03-01

    Full Text Available Recycling of materials is becoming increasingly important as industry response to public demands, that resources must be preserved and environment protected. To produce materials competitive in cost with primary product, secondary producers have to pursue new technologies and other innovations. For these purposes different recycling technologies for composite materials (oxidation, milling, remelting etc are widely used. The current paper studies hardmetal/cermet powders produced by mechanical milling technology. The following composite materials were studied: Cr3C2-Ni cermets and WC-Co hardmetal. Different disintegrator milling systems for production of powders with determined size and shape were used. Chemical composition of produced powders was analysed.  To estimate the properties of recycled hardmetal/cermet powders, sieving analysis, laser granulometry and angularity study were conducted. To describe the angularity of milled powders, spike parameter–quadric fit (SPQ was used and experiments for determination of SPQ sensitivity and precision to characterize particles angularity were performed. Images used for calculating SPQ were taken by SEM processed with Omnimet Image Analyser 22. The graphs of grindability and angularity were composed. Composite powders based on Fe- and Ni-self-fluxing alloys for thermal spray (plasma and HVOF were produced. Technological properties of powders and properties of thermal sprayed coatings from studied powders were investigated. The properties of spray powders reinforced with recycled hardmetal and cermet particles as alternatives for cost-sensitive applications were demonstrated.DOI: http://dx.doi.org/10.5755/j01.ms.18.1.1348

  14. Characterization of Nanometric-Sized Carbides Formed During Tempering of Carbide-Steel Cermets

    Directory of Open Access Journals (Sweden)

    Matus K.

    2016-06-01

    Full Text Available The aim of this article of this paper is to present issues related to characterization of nanometric-sized carbides, nitrides and/or carbonitrides formed during tempering of carbide-steel cermets. Closer examination of those materials is important because of hardness growth of carbide-steel cermet after tempering. The results obtained during research show that the upswing of hardness is significantly higher than for high-speed steels. Another interesting fact is the displacement of secondary hardness effect observed for this material to a higher tempering temperature range. Determined influence of the atmosphere in the sintering process on precipitations formed during tempering of carbide-steel cermets. So far examination of carbidesteel cermet produced by powder injection moulding was carried out mainly in the scanning electron microscope. A proper description of nanosized particles is both important and difficult as achievements of nanoscience and nanotechnology confirm the significant influence of nanocrystalline particles on material properties even if its mass fraction is undetectable by standard methods. The following research studies have been carried out using transmission electron microscopy, mainly selected area electron diffraction and energy dispersive spectroscopy. The obtained results and computer simulations comparison were made.

  15. Investigation of a Cermet Gas-turbine-blade Material of Titanium Carbide Infiltrated with Hastalloy C

    Science.gov (United States)

    Hoffman, Charles A

    1955-01-01

    A cermet composition was investigated as a potential material for gas-turbine blades. Blades of HS-21 alloy were also operated in the engine simultaneously to provide a basis of comparison. The cermet blades survived as long as approximately 312-1/2 hours at about 1500 degrees F with an average midspan centrifugal stress of approximately 11,500 psi. The alloy blade midspan stress was about 15,300 psi. Because of extensive damage to both types of blade due to external causes, a reliable comparison of operating lives could not be made. The cermet blades tended to fail in the airfoil rather than in the base, although the base was the usual location of failure in a prior study of cold-pressed and sintered cermets of other compositions with the same blade shape.

  16. Actinide recycling in reactors; Aktiniden-Rezyklierung in Reaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Kuesters, H.; Wiese, H.W.; Krieg, B.

    1995-08-01

    The objective is an assessment of the transmutation of long-lived actinides and fission products and the incineration of plutonium for reducing the risk potential of radioactive waste from reactors in comparison to direct waste disposal. The contribution gives an interim account on homogeneous and heterogeneous recycling of `risk nuclides` in thermal and fast reactors. Important results: - A homogeneous 5 percent admixture of minor actinides (MA) from N4-PWRs to EFR fuel would allow a transmutation not only of the EFR MA, but in addition of the MA from 5 or 6 PWRs of equal power. However, the incineration is restricted by safety considerations. - LWR have only a very low MA incineration potential, due to their disadvantageous neutron capture/fission ratio. - In order to keep the Cm inventory at a low level, it is advantageous to concentrate the Am heterogeneously in particular fuel elements or rods. (orig./HP)

  17. Some physics aspects of cermet and ceramic fast systems; Quelques aspects de la physique des reacteurs a neutrons rapides utilisant des cermets et des ceramiques comme combustibles; Nekotorye fizicheskie aspekty kermetnykh i keramicheskikh sistem na bystrykh nejtronakh; Algunos aspectos fisicos de los sistemas rapidos a base de combustibles cermet y ceramicos

    Energy Technology Data Exchange (ETDEWEB)

    Codd, J; James, M F; Mann, J E [United Kingdom Atomic Energy Authority, Reactor Group (United Kingdom)

    1962-03-15

    The characteristics of a system using an iron-based oxide cermet as fuel material are discussed. A transport theory investigation to develop methods of predicting the effect of core heterogeneity on reactivity and flux distribution is described. Some preliminary calculations are also given of resonance self-shielding and Doppler temperature effects in a cermet system. (author) [French] Les auteurs etudient les caracteristique s d'un reacteur utilisant comme combustible un cermet d'oxydes a armature de fer. Ils exposent une application de la theorie du transport a la mise au point des methodes permettant de prevoir l'effet de l'heterogeneite du coeur sur la reactivite et sur la distribution du flux. Ils donnent egalement quelques calculs preliminaires d'effets d'autoprotection due a la resonance et d'effet Doppler du a la chaleur dans un reacteur utilisant un cermet. (author) [Spanish] La memoria discute las caracteristicas de un sistema que emplea como combustible un oxido tipo cermet a base de hierro. Describe una investigacion de la teoria de transporte con miras a desarrollar metodos para evaluar el efecto de la heterogeneidad del cuerpo sobre la reactividad y la distribucion de flujo. Tambien da algunos calculos preliminares de los efectos del autoblindaje por resonancia y de la temperatura de Doppler en un sistema de tipo cermet. (author) [Russian] Obsuzhdayutsya kharakteristiki sistemy, ispol'zuyushchej v kachestve toplivnogo materiala oksidnye kermety, razrabotannye na osnove zheleza. Opisyvaetsya issledovanie teorii perenosa, chtoby razvit' metody predskazaniya vliyaniya geterogennosti aktivnoj zony na reaktivnost' i raspredelenie potoka. Dayutsya takzhe nekotorye predvaritel'nye raschety ehffektov rezonansnoj samozashchity i temperaturnogo ehffekta Dopplera v kermetnoj sisteme. (author)

  18. Feasibility studies of actinide recycle in LMFBRs as a waste management alternative

    International Nuclear Information System (INIS)

    Beaman, S.L.; Aitken, E.A.

    1976-01-01

    A strategy of actinide burnup in LMFBRs is being investigated as a waste management alternative to long term storage of high level nuclear waste. This strategy is being evaluated because many of the actinides in the waste from spent-fuel reprocessing have half-lives of thousands of years and an alternative to geological storage may be desired. From a radiological viewpoint, the actinides and their daughters dominate the waste hazard for decay times beyond about 400 years. Actinide burnup in LMFBRs may be an attractive alternative to geological storage because the actinides can be effectively transmuted to fission products which have significantly shorter half-lives. Actinide burnup in LMFBRs rather than LWRs is preferred because the ratio of fission reaction rate to capture reaction rate for the actinides is higher in an LMFBR, and an LMFBR is not so sensitive to the addition of the actinide isotopes. An actinide target assembly recycle scheme is evaluated to determine the effects of the actinides on the LMFBR performance, including local power peaking, breeding ratio, and fissile material requirements. Several schemes are evaluated to identify any major problems associated with reprocessing and fabrication of recycle actinide-containing assemblies. The overall efficiency of actinide burnout in LMFBRs is evaluated, and equilibrium cycle conditions are determined. It is concluded that actinide recycle in LMFBRs offers an attractive alternative to long term storage of the actinides, and does not significantly affect the performance of the host LMFBR. Assuming a 0.1 percent or less actinide loss during reprocessing, a 0.1 percent loss of less during fabrication, and proper recycle schemes, virtually all of the actinides produced by a fission reactor economy could be transmuted in fast reactors

  19. J-ACTINET activities of training and education for actinide science research

    International Nuclear Information System (INIS)

    Miato, Kazuo; Konashi, Kenji; Yamana, Hajimu; Yamanaka, Shinsuke; Nagasaki, Shinya; Ikeda, Yasuhisa; Sato, Seichi; Arita, Yuji; Idemitsu, Kazuya; Koyama, Tadafumi

    2011-01-01

    Actinide science research is indispensable to maintain sustainable development of innovative nuclear technology, especially advanced fuels, partitioning/reprocessing, and waste management. For actinide science research, special facilities with containment and radiation shields are needed to handle actinide materials since actinide elements are γ-, α- and neutron-emitters. The number of facilities for actinide science research has been decreased, especially in universities, due to the high maintenance cost. J-ACTINET was established in 2008 to promote and facilitate actinide science research in close cooperation with the facilities and to foster many of young scientists and engineers to be actively engaged in the fields of actinide science. The research program was carried out, through which young researchers were expected to learn how to make experiments with advanced experimental tools and to broaden their horizons. The summer schools and computational science school were held to provide students, graduate students, and young researchers with the opportunities to come into contact with actinide science research. In these schools, not only the lectures, but also the practical exercises were made as essential part. The overseas dispatch program was also carried out, where graduate students and young researchers were sent to the international summer schools and conferences. (author)

  20. Fractographic peculiarities of cermet tungsten fracture

    International Nuclear Information System (INIS)

    Stepanenko, V.A.; Babak, A.V.; Uskov, E.I.

    1982-01-01

    Effect of test temperature on fracture peculiarities of cermets tungsten with initial cellular structure of deformation is shown. Tungsten crack resistance increases at temperatures to Tsub(x) (ductile-brittle transition temperature) and decreases at temperatures above Tsub(x). The degree of ceramics tungsten plasticity realization depends on its crack resistance

  1. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Somers, J.; Van Den Durpel, L.

    2013-01-01

    The PUMA project - the acronym stands for “Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors” - was a Specific Targeted Research Project (STREP) within the Euratom 6th Framework (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO2-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR designs and their ability to accept a variety

  2. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Somers, J.; Van Den Durpel, L.; Chauvet, V.; Cerullo, N.; Cetnar, J.; Abram, T.; Bakker, K.; Bomboni, E.; Bernnat, W.; Domanska, J.G.; Girardi, E.; De Haas, J.B.M.; Hesketh, K.; Hiernaut, J.P.; Hossain, K.; Jonnet, J.; Kim, Y.; Kloosterman, J.L.; Kopec, M.; Murgatroyd, J.; Millington, D.; Lecarpentier, D.; Lomonaco, G.; McEachern, D.; Meier, A.; Mignanelli, M.; Nabielek, H.; Oppe, J.; Petrov, B.Y.; Pohl, C.; Ruetten, H.J.; Schihab, S.; Toury, G.; Trakas, C.; Venneri, F.; Verfondern, K.; Werner, H.; Wiss, T.; Zakova, J.

    2010-11-01

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO 2 -free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR designs and their ability to accept a

  3. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C., E-mail: kuijper@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Somers, J; Van Den Durpel, L; Chauvet, V; Cerullo, N; Cetnar, J; Abram, T; Bakker, K; Bomboni, E; Bernnat, W; Domanska, J G; Girardi, E; De Haas, J B.M.; Hesketh, K; Hiernaut, J P; Hossain, K; Jonnet, J; Kim, Y; Kloosterman, J L; Kopec, M; Murgatroyd, J; Millington, D; Lecarpentier, D; Lomonaco, G; McEachern, D; Meier, A; Mignanelli, M; Nabielek, H; Oppe, J; Petrov, B Y; Pohl, C; Ruetten, H J; Schihab, S; Toury, G; Trakas, C; Venneri, F; Verfondern, K; Werner, H; Wiss, T; Zakova, J

    2010-11-15

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO{sub 2}-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR designs and their ability to accept a

  4. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C., E-mail: kuijper@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Somers, J.; Van Den Durpel, L.; Chauvet, V.; Cerullo, N.; Cetnar, J.; Abram, T.; Bakker, K.; Bomboni, E.; Bernnat, W.; Domanska, J.G.; Girardi, E.; De Haas, J.B.M.; Hesketh, K.; Hiernaut, J.P.; Hossain, K.; Jonnet, J.; Kim, Y.; Kloosterman, J.L.; Kopec, M.; Murgatroyd, J.; Millington, D.; Lecarpentier, D.; Lomonaco, G.; McEachern, D.; Meier, A.; Mignanelli, M.; Nabielek, H.; Oppe, J.; Petrov, B.Y.; Pohl, C.; Ruetten, H.J.; Schihab, S.; Toury, G.; Trakas, C.; Venneri, F.; Verfondern, K.; Werner, H.; Wiss, T.; Zakova, J.

    2010-11-15

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO{sub 2}-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR

  5. Durability of cermet ionomer cement conditioned in different media.

    Science.gov (United States)

    el-Din, I M

    1992-01-01

    The glass ionomer cement has exhibited significant adhesion to hard tooth structures, and good cariostatic properties. The sintering of the silver alloy powder and glass ionomer cement "cermet cement" has provided additional improvement in the physical properties of the restorative material. These were flexural resistance, wear resistance, increased radio-opacity, hardness and porosity. The improvement in the physical properties of the cermet glass cements has provided an extension in their clinical use as core build up, lining for inlays, amalgam and composite restoratives, fissure filling, restoration of primary teeth, class II tunnel preparation, treatment of root caries and repair of defective metal margins in crown and inlays.

  6. Actinides-1981

    International Nuclear Information System (INIS)

    1981-09-01

    Abstracts of 134 papers which were presented at the Actinides-1981 conference are presented. Approximately half of these papers deal with electronic structure of the actinides. Others deal with solid state chemistry, nuclear physic, thermodynamic properties, solution chemistry, and applied chemistry

  7. Actinides-1981

    Energy Technology Data Exchange (ETDEWEB)

    1981-09-01

    Abstracts of 134 papers which were presented at the Actinides-1981 conference are presented. Approximately half of these papers deal with electronic structure of the actinides. Others deal with solid state chemistry, nuclear physic, thermodynamic properties, solution chemistry, and applied chemistry.

  8. Effect of TaC addition on the microstructures and mechanical properties of Ti(C, N)-based cermets

    International Nuclear Information System (INIS)

    Wu, Peng; Zheng, Yong; Zhao, Yongle; Yu, Haizhou

    2010-01-01

    The microstructures of the prepared Ti(C, N)-based cermets with various TaC additions were studied using X-ray diffractometry (XRD) and scanning electron microscopy (SEM). Mechanical properties such as transverse rupture strength (TRS), fracture toughness (K 1C ) and hardness (HRA) were also measured. The results showed that the grain size of the cermets decreased with increasing TaC addition, but too high TaC addition resulted in agglomeration of the grains. An increasing TaC addition increased the dissolution of tungsten, titanium, molybdenum and tantalum in the binder phase. The hardness of the cermets decreased slightly with increasing TaC addition. The transverse rupture strength was the highest for the cermets with 5 wt.% TaC addition, which was characterized by fine grains, homogeneous microstructure and the moderate thickness of rim phase in the binder. The fracture toughness of the cermets with TaC addition from 0 to 5 wt.% decreased obviously, which resulting from decreased grain size. The further decreasing of fracture toughness for the cermets with 7 wt.% TaC addition was due to increased porosity and interfacial tensile stress.

  9. Safe management of actinides in the nuclear fuel cycle: Role of mineralogy

    International Nuclear Information System (INIS)

    Ewing, R.C.

    2011-01-01

    During the past 60 years, more than 1800 metric tonnes of Pu, and substantial quantities of the 'minor' actinides, such as Np, Am and Cm, have been generated in nuclear reactors. Some of these transuranium elements can be a source of energy in fission reactions (e.g., 239 Pu), a source of fissile material for nuclear weapons (e.g., 239 Pu and 237 Np), and of environmental concern because of their long-half lives and radiotoxicity (e.g., 239 Pu and 237 Np). There are two basic strategies for the disposition of these heavy elements: (1) to 'burn' or transmute the actinides using nuclear reactors or accelerators; (2) to 'sequester' the actinides in chemically durable, radiation-resistant materials that are suitable for geologic disposal. There has been substantial interest in the use of actinide-bearing minerals, especially isometric pyrochlore, A 2 B 2 O 7 (A rare earths; B = Ti, Zr, Sn, Hf), for the immobilization of actinides, particularly plutonium, both as inert matrix fuels and nuclear waste forms. Systematic studies of rare-earth pyrochlores have led to the discovery that certain compositions (B = Zr, Hf) are stable to very high doses of alpha-decay event damage. Recent developments in our understanding of the properties of heavy element solids have opened up new possibilities for the design of advanced nuclear fuels and waste forms. (author)

  10. Comparative Study of f-Element Electronic Structure across a Series of Multimetallic Actinide, Lanthanide-Actinide and Lanthanum-Actinide Complexes Possessing Redox-Active Bridging Ligands

    Energy Technology Data Exchange (ETDEWEB)

    Schelter, Eric J.; Wu, Ruilian; Veauthier, Jacqueline M.; Bauer, Eric D.; Booth, Corwin H.; Thomson, Robert K.; Graves, Christopher R.; John, Kevin D.; Scott, Brian L.; Thompson, Joe D.; Morris, David E.; Kiplinger, Jaqueline L.

    2010-02-24

    A comparative examination of the electronic interactions across a series of trimetallic actinide and mixed lanthanide-actinide and lanthanum-actinide complexes is presented. Using reduced, radical terpyridyl ligands as conduits in a bridging framework to promote intramolecular metal-metal communication, studies containing structural, electrochemical, and X-ray absorption spectroscopy are presented for (C{sub 5}Me{sub 5}){sub 2}An[-N=C(Bn)(tpy-M{l_brace}C{sub 5}Me4R{r_brace}{sub 2})]{sub 2} (where An = Th{sup IV}, U{sup IV}; Bn = CH{sub 2}C{sub 6}H{sub 5}; M = La{sup III}, Sm{sup III}, Yb{sup III}, U{sup III}; R = H, Me, Et) to reveal effects dependent on the identities of the metal ions and R-groups. The electrochemical results show differences in redox energetics at the peripheral 'M' site between complexes and significant wave splitting of the metal- and ligand-based processes indicating substantial electronic interactions between multiple redox sites across the actinide-containing bridge. Most striking is the appearance of strong electronic coupling for the trimetallic Yb{sup III}-U{sup IV}-Yb{sup III}, Sm{sup III}-U{sup IV}-Sm{sup III}, and La{sup III}-U{sup IV}-La{sup III} complexes, [8]{sup -}, [9b]{sup -} and [10b]{sup -}, respectively, whose calculated comproportionation constant K{sub c} is slightly larger than that reported for the benchmark Creutz-Taube ion. X-ray absorption studies for monometallic metallocene complexes of U{sup III}, U{sup IV}, and U{sup V} reveal small but detectable energy differences in the 'white-line' feature of the uranium L{sub III}-edges consistent with these variations in nominal oxidation state. The sum of this data provides evidence of 5f/6d-orbital participation in bonding and electronic delocalization in these multimetallic f-element complexes. An improved, high-yielding synthesis of 4{prime}-cyano-2,2{prime}:6{prime},2{double_prime}-terpyridine is also reported.

  11. Influence of WC addition on the microstructure and mechanical properties of NbC-Co cermets

    International Nuclear Information System (INIS)

    Huang, S.G.; Li, L.; Van der Biest, O.; Vleugels, J.

    2007-01-01

    NbC-24.5 wt.% Co cermets with up to 30 wt.% WC were obtained by solid state hot pressing at 1300 o C under a pressure of 45 MPa for 10 min and pressureless liquid phase sintering at 1360 o C for 60 min. The effect of WC addition on the microstructure and mechanical properties of NbC-Co based cermets was investigated. The hot pressed cermets exhibited interconnected and irregular niobium carbide (NbC) or (Nb,W)C grains, whereas the shape of the NbC grains changed from faceted with rounded corners to spherical, as the WC content increased in the pressureless sintered cermets. The undissolved WC increased with increasing WC addition. A clear core/rim structure was observed in the hot pressed cermets with 10-30 wt.% WC additions, whereas this structure was gradually eliminated when pressureless sintering. The hardness remains nearly constant whereas the fracture toughness slightly increases with increasing WC addition. The dissolution of WC in the Co binder and NbC grains, as well as the formation of a solid solution (Nb,W)C phase were supported by thermodynamic calculations

  12. Working of Mo-TiC cermets for 'future nuclear systems'; Mise en forme de cermets Mo-TiC pour les 'Systemes Nucleaires du futur'

    Energy Technology Data Exchange (ETDEWEB)

    Allemand, Alexandre [CEA-Saclay, DRT/LITEN/LTMEx, 91191 Gif-sur-Yvette (France); Le Flem, Marion [CEA-Saclay, DEN/DMN/SRMA, 91191 Gif-sur-Yvette (France); Rousselet, Jerome [UTT Troyes, 10010 Troyes (France)

    2006-07-01

    The nuclear reactor cores (generation IV) will form an extremely severe environment (high temperature, severe and long irradiation...). These drastic criteria and the preoccupation to ensure a higher and higher safety level lead, beyond the preoccupations due to the feasibility of such reactors, to harsh choices in materials able to be used. Innovating materials such as Mo-TiC cermet are the subject of intense researches in the CEA. This study presents and compares two modes of Mo-TiC cermet working: the hot isostatic compression and the extrusion. Different compositions of Mo-TiC cermets are prepared by hot isostatic compression and extrusion, and then characterized in term of microstructural properties. At last, this study concludes to a very satisfying working by hot isostatic compression, nevertheless the extrusion has still to be improved. (O.M.)

  13. Zr-ZrO sub 2 cermet solar coatings designed by modelling calculations and deposited by dc magnetron sputtering

    CERN Document Server

    Zhang Qi Chu; Lee, K D; Shen, Y G

    2003-01-01

    High solar performance Zr-ZrO sub 2 cermet solar coatings were designed using a numerical computer model and deposited experimentally. The layer thickness and Zr metal volume fraction for the Zr-ZrO sub 2 cermet solar selective coatings on a Zr or Al reflector with a surface ZrO sub 2 or Al sub 2 O sub 3 anti-reflection layer were optimized to achieve maximum photo-thermal conversion efficiency at 80 deg. C under concentration factors of 1-20 using the downhill simplex method in multi-dimensions in the numerical calculation. The dielectric function and the complex refractive index of Zr-ZrO sub 2 cermet materials were calculated using Sheng's approximation. Optimization calculations show that Al sub 2 O sub 3 /Zr-ZrO sub 2 /Al solar coatings with two cermet layers and three cermet layers have nearly identical solar absorptance, emittance and photo-thermal conversion efficiency that are much better than those for films with one cermet layer. The optimized Al sub 2 O sub 3 /Zr-ZrO sub 2 /Al solar coating film w...

  14. Safe management of actinides in the nuclear fuel cycle: Role of mineralogy; La gestion des actinides dans le cycle du combustible nucleaire: le role de la mineralogie

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, R.C. [Department of Nuclear Engineering and Radiological Sciences, Department of Geological Sciences, Department of Materials Science and Engineering, University of Michigan, Ann Arbor, Michigan 48109-1005 (United States)

    2011-02-15

    During the past 60 years, more than 1800 metric tonnes of Pu, and substantial quantities of the 'minor' actinides, such as Np, Am and Cm, have been generated in nuclear reactors. Some of these transuranium elements can be a source of energy in fission reactions (e.g., {sup 239}Pu), a source of fissile material for nuclear weapons (e.g., {sup 239}Pu and {sup 237}Np), and of environmental concern because of their long-half lives and radiotoxicity (e.g., {sup 239}Pu and {sup 237}Np). There are two basic strategies for the disposition of these heavy elements: (1) to 'burn' or transmute the actinides using nuclear reactors or accelerators; (2) to 'sequester' the actinides in chemically durable, radiation-resistant materials that are suitable for geologic disposal. There has been substantial interest in the use of actinide-bearing minerals, especially isometric pyrochlore, A{sub 2}B{sub 2}O{sub 7} (A rare earths; B = Ti, Zr, Sn, Hf), for the immobilization of actinides, particularly plutonium, both as inert matrix fuels and nuclear waste forms. Systematic studies of rare-earth pyrochlores have led to the discovery that certain compositions (B = Zr, Hf) are stable to very high doses of alpha-decay event damage. Recent developments in our understanding of the properties of heavy element solids have opened up new possibilities for the design of advanced nuclear fuels and waste forms. (author)

  15. The anomalous behaviour of Ag-Al2O3 Cermet electroformed devices

    International Nuclear Information System (INIS)

    Khan, M.S.R.

    2003-06-01

    Cermet coating consisting of silver particles in an aluminium oxide matrix were prepared on glass substrates by vacuum deposition. Variation of the circulating current with potential difference was obtained in evaporated Al/Ag-Al 2 O 3 /Cu sandwich structures, 100 to 200 nm thick containing 10 wt % Ag. It was observed that the investigated sandwich structures exhibit anomalous behaviour such as electroforming with Voltage-Controlled-Negative Resistance (VCNR) in vacuo of ∼ 4 x 10 -6 torr. The formed characteristics were explained on the basis of filamentary model. (author)

  16. Electrochemical separation of actinides and fission products in molten salt electrolyte

    Energy Technology Data Exchange (ETDEWEB)

    Gay, R.L.; Grantham, L.F.; Fusselman, S.P. [Rockwell International/Rocketdyne Division, Canoga Park, CA (United States)] [and others

    1995-10-01

    Molten salt electrochemical separation may be applied to accelerator-based conversion (ABC) and transmutation systems by dissolving the fluoride transport salt in LiCl-KCl eutectic solvent. The resulting fluoride-chloride mixture will contain small concentrations of fission product rare earths (La, Nd, Gd, Pr, Ce, Eu, Sm, and Y) and actinides (U, Np, Pu, Am, and Cm). The Gibbs free energies of formation of the metal chlorides are grouped advantageously such that the actinides can be deposited on a solid cathode with the majority of the rare earths remaining in the electrolyte. Thus, the actinides are recycled for further transmutation. Rockwell and its partners have measured the thermodynamic properties of the metal chlorides of interest (rare earths and actinides) and demonstrated separation of actinides from rare earths in laboratory studies. A model is being developed to predict the performance of a commercial electrochemical cell for separations starting with PUREX compositions. This model predicts excellent separation of plutonium and other actinides from the rare earths in metal-salt systems.

  17. Interaction on boundary of current-conducting and glass-forming phases in cermet films under annealing

    International Nuclear Information System (INIS)

    Shulishova, O.I.; Zyrin, A.V.; Ismalgaliev, R.K.; Izmajlov, Sh.Z.; Kovylyaev, V.V.; Shevchuk, N.V.; Shcherbak, I.A.

    1990-01-01

    The electron-probe microanalysis permits investigating the interaction on the boundary of current-conducting and glass-binding phases in cermet films without noble metals on the base of ruthenium oxide. The performed studies along with experiments on model microsections subject to annealing in different media have shown the differences in the process of formation of structure and properties of cermet resistive elements as well as a significance of the oxidation process of current-conducting phase in formation of high working characteristics of cermet resistors on the base of hexaborides of the rare-earth elements

  18. Subsurface interactions of actinide species and microorganisms : implications for the bioremediation of actinide-organic mixtures

    International Nuclear Information System (INIS)

    Banaszak, J.E.; Reed, D.T.; Rittmann, B.E.

    1999-01-01

    By reviewing how microorganisms interact with actinides in subsurface environments, we assess how bioremediation controls the fate of actinides. Actinides often are co-contaminants with strong organic chelators, chlorinated solvents, and fuel hydrocarbons. Bioremediation can immobilize the actinides, biodegrade the co-contaminants, or both. Actinides at the IV oxidation state are the least soluble, and microorganisms accelerate precipitation by altering the actinide's oxidation state or its speciation. We describe how microorganisms directly oxidize or reduce actinides and how microbiological reactions that biodegrade strong organic chelators, alter the pH, and consume or produce precipitating anions strongly affect actinide speciation and, therefore, mobility. We explain why inhibition caused by chemical or radiolytic toxicities uniquely affects microbial reactions. Due to the complex interactions of the microbiological and chemical phenomena, mathematical modeling is an essential tool for research on and application of bioremediation involving co-contamination with actinides. We describe the development of mathematical models that link microbiological and geochemical reactions. Throughout, we identify the key research needs

  19. Subsurface interactions of actinide species and microorganisms : implications for the bioremediation of actinide-organic mixtures.

    Energy Technology Data Exchange (ETDEWEB)

    Banaszak, J.E.; Reed, D.T.; Rittmann, B.E.

    1999-02-12

    By reviewing how microorganisms interact with actinides in subsurface environments, we assess how bioremediation controls the fate of actinides. Actinides often are co-contaminants with strong organic chelators, chlorinated solvents, and fuel hydrocarbons. Bioremediation can immobilize the actinides, biodegrade the co-contaminants, or both. Actinides at the IV oxidation state are the least soluble, and microorganisms accelerate precipitation by altering the actinide's oxidation state or its speciation. We describe how microorganisms directly oxidize or reduce actinides and how microbiological reactions that biodegrade strong organic chelators, alter the pH, and consume or produce precipitating anions strongly affect actinide speciation and, therefore, mobility. We explain why inhibition caused by chemical or radiolytic toxicities uniquely affects microbial reactions. Due to the complex interactions of the microbiological and chemical phenomena, mathematical modeling is an essential tool for research on and application of bioremediation involving co-contamination with actinides. We describe the development of mathematical models that link microbiological and geochemical reactions. Throughout, we identify the key research needs.

  20. The INE-Beamline for Actinide Research at ANKA

    Science.gov (United States)

    Brendebach, Boris; Denecke, Melissa A.; Rothe, Jörg; Dardenne, Kathy; Römer, Jürgen

    2007-02-01

    The INE-Beamline for actinide research at the synchrotron source ANKA is now fully operational. This beamline was designed, built, and commissioned by the Institut für Nukleare Entsorgung (INE) at the Forschungszentrum Karlsruhe (FZK), Germany. It is dedicated to actinide speciation investigations related to nuclear waste disposal as well as applied and basic actinide research. Experiments on radioactive samples with activities up to 106 times the limit of exemption inside a safe and flexible double containment concept are possible. The close proximity of the beamline to INE's active laboratories is unique in Europe. Currently, experiments can be performed in an X-ray energy range from around 2.15 keV (P K edge) to 24.35 keV (Pd K edge). The INE-Beamline design is optimized for spectroscopy with emphasis on surface sensitive techniques. A microfocus option is presently being installed at the INE-Beamline. Access to the INE-Beamline is possible through cooperation with INE, through the ANKA proposal system and via the European Network of Excellence for Actinide Sciences (ACTINET).

  1. Cermet cathodes for strontium and magnesium-doped LaGaO3-based solid oxide fuel cells

    International Nuclear Information System (INIS)

    Datta, Pradyot; Bronin, D.I.; Majewski, P.; Aldinger, F.

    2009-01-01

    To check the suitability of La 0.9 Sr 0.1 Ga 0.85 Mg 0.15 O 3-δ -Ag cermets as cathode material for solid oxide fuel cell (SOFC) with Sr- and Mg-doped LaGaO 3 electrolyte a series of cermets with different Ag contents were prepared by conventional sintering process. The chemical compatibility between La 0.9 Sr 0.1 Ga 0.85 Mg 0.15 O 3-δ (LSGM) and Ag was investigated by X-ray diffraction, scanning electron microscopy, and X-ray photoelectron spectroscopy. Thermal expansion coefficient of the cermets was measured as a function of Ag content and was found to increase with increasing metallic content. Oxygen adsorption at the surface of the cermets could be detected but no reaction or solid solubility between LSGM and Ag was found. It was noticed that a minimum of 30 wt.% Ag is needed to form a cermet with percolating network. From impedance spectroscopy measurement activation energy for the polarization conductance was found to be around 110 kJ mol -1

  2. Sensitivity of DF-ICP-MS, PERALS and alpha-spectrometry for the determination of actinides. A comparison

    International Nuclear Information System (INIS)

    Ayranov, M.; Kraehenbuehl, U.

    2009-01-01

    We applied three techniques (DF-ICP-MS, PERALS and alpha-spectrometry) for the determination of minor actinides at environmental levels. For each method the limit of detection and the resolution were estimated in order to study the content and isotopic composition of the actinides. Two international reference materials, IAEA-135 (Irish Sea Sediment) and IAEA-300 (Baltic Sea sediment) were analyzed for activity concentrations of 238 Pu, 239 Pu, 240 Pu, 241 Pu and 241 Am. The sensitivities of the three determination techniques were compared. (author)

  3. Lanthanides and actinides extraction by calixarenes containing CMPO groups

    International Nuclear Information System (INIS)

    Garcia Carrera, A.

    2001-01-01

    In the framework of the French program SPIN concerning the radioactive waste management, researches are performed to develop processes allowing the separation of long-lived radioisotopes in order to their transmutation or their specific conditioning. These studies deal with the extraction and the separation of trivalent lanthanides and actinides in acid solution. Many systems ''calixarene-diluent-aqueous phase'' are examined by extraction liquid-liquid and membrane transport. The extraction efficiency and the selectivity of the synthesized calixarene-CMPO and of the CMPO are compared with these cations, as the nitric acid extraction by these molecules. (A.L.B.)

  4. Effect of Carbon Content on the Microstructure and Mechanical Properties of NbC-Ni Based Cermets

    Directory of Open Access Journals (Sweden)

    Shuigen Huang

    2018-03-01

    Full Text Available The aim of this work was to correlate the overall carbon content in NbC-Ni, NbC-Ni-VC and NbC-Ni-Mo starting powders with the resulting microstructure, hardness, and fracture toughness of Ni-bonded NbC cermets. A series of NbC-Ni, NbC-Ni-VC and NbC-Ni-Mo cermets with different carbon content were prepared by conventional liquid phase sintering for 1 h at 1420 °C in vacuum. Microstructural analysis of the fully densified cermets was performed by electron probe microanalysis (EPMA to assess the effect of carbon and VC or Mo additions on the NbC grain growth and morphology. A decreased carbon content in the starting powder mixtures resulted in increased dissolution of Nb, V, and Mo in the Ni binder and a decreased C/Nb ratio in the NbC based carbide phase. The Vickers hardness (HV30 and Palmqvist indentation toughness were found to decrease significantly with an increasing carbon content in the Mo-free cermets, whereas an antagonistic correlation between hardness and toughness was obtained as a function of the Mo-content in Mo-modified NbC cermets. To obtain optimized mechanical properties, methods to control the total carbon content of NbC-Ni mixtures were proposed and the prepared cermets were investigated in detail.

  5. Sigma Team for Advanced Actinide Recycle FY2015 Accomplishments and Directions

    Energy Technology Data Exchange (ETDEWEB)

    Moyer, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-30

    The Sigma Team for Minor Actinide Recycle (STAAR) has made notable progress in FY 2015 toward the overarching goal to develop more efficient separation methods for actinides in support of the United States Department of Energy (USDOE) objective of sustainable fuel cycles. Research in STAAR has been emphasizing the separation of americium and other minor actinides (MAs) to enable closed nuclear fuel recycle options, mainly within the paradigm of aqueous reprocessing of used oxide nuclear fuel dissolved in nitric acid. Its major scientific challenge concerns achieving selectivity for trivalent actinides vs lanthanides. Not only is this challenge yielding to research advances, but technology concepts such as ALSEP (Actinide Lanthanide Separation) are maturing toward demonstration readiness. Efforts are organized in five task areas: 1) combining bifunctional neutral extractants with an acidic extractant to form a single process solvent, developing a process flowsheet, and demonstrating it at bench scale; 2) oxidation of Am(III) to Am(VI) and subsequent separation with other multivalent actinides; 3) developing an effective soft-donor solvent system for An(III) selective extraction using mixed N,O-donor or all-N donor extractants such as triazinyl pyridine compounds; 4) testing of inorganic and hybrid-type ion exchange materials for MA separations; and 5) computer-aided molecular design to identify altogether new extractants and complexants and theory-based experimental data interpretation. Within these tasks, two strategies are employed, one involving oxidation of americium to its pentavalent or hexavalent state and one that seeks to selectively complex trivalent americium either in the aqueous phase or the solvent phase. Solvent extraction represents the primary separation method employed, though ion exchange and crystallization play an important role. Highlights of accomplishments include: Confirmation of the first-ever electrolytic oxidation of Am(III) in a

  6. The effect of explosive compacting on the properties of the 60% LaCr03-40% Cr cermet

    International Nuclear Information System (INIS)

    Atroshenko, E.S.; Barykin, B.M.; Ivanov, V.S.; Krasulin, Yu.L.; Spirodonov, E.G.

    1976-01-01

    A technique of explosive compacting of composite cermet materials (M + M0) has been used for producing large blanks with a density close to theoretical. A study has been made of the properties of an explosively compacted 60% LaCr0 3 -40% Cr cermet over a wide temperature range. Cermets compacted explosively are shown to have a number of advantages over ones prepared by conventional powder metallurgical techniques. (author)

  7. Conversion of actinide solutions for the production of MA bearing fuels for Gen IV fast reactor systems

    International Nuclear Information System (INIS)

    Fernandez, A.; McGinley, J.; Somers, J.

    2008-01-01

    The conversion of the solution to solid for fuels containing minor actinides for accelerator driven systems or Gen IV fast reactors cannot be made by conventional ammonia or oxalate precipitation as is the case in today's reprocessing plant. The small particle size and concomitant dust that is produced in subsequent processing steps will not permit use of these processes on industrial scale. Innovation is needed to avoid dust generating powders, and indeed to simplify the processes themselves. Two such processing routes have been developed at the JRC-ITU. The sol gel route has been used to produce fuel containing Am and Np for the SUPERFACT, TRABANT and other irradiation experiments. The infiltration process has also been established and fuels have been produced for the FUTURIX and HELIOS experiments. (authors)

  8. Conversion of actinide solutions for the production of MA bearing fuels for Gen IV fast reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, A.; McGinley, J.; Somers, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements P.O.Box 2340, Karlsruhe, D-76125 (Germany)

    2008-07-01

    The conversion of the solution to solid for fuels containing minor actinides for accelerator driven systems or Gen IV fast reactors cannot be made by conventional ammonia or oxalate precipitation as is the case in today's reprocessing plant. The small particle size and concomitant dust that is produced in subsequent processing steps will not permit use of these processes on industrial scale. Innovation is needed to avoid dust generating powders, and indeed to simplify the processes themselves. Two such processing routes have been developed at the JRC-ITU. The sol gel route has been used to produce fuel containing Am and Np for the SUPERFACT, TRABANT and other irradiation experiments. The infiltration process has also been established and fuels have been produced for the FUTURIX and HELIOS experiments. (authors)

  9. A core performance study on an actinide recycling 'zero-sodium-void worth' core

    International Nuclear Information System (INIS)

    Kawashima, M.; Nakagawa, M.; Yamaoka, M.; Kasahara, F.

    1994-01-01

    A core performance study was made for an absorber-type parfait core (A-APC) as one of 'Zero-sodium-void-worth' core concepts. This evaluation study pursued different two aspects; one for transuranic (TRU) management strategy, and another for a loss-of-coolant anticipated transient behavior considering the unique core configuration. The results indicated that this core has a large flexibility for actinide recycling in terms of self-sufficiency and minor actinide burning. The result also showed that this core has kept a large mitigation potential for ULOF events as well as a simple flat core concept, reflecting detailed three dimensional core bowing behavior for the A-APC configuration. (author)

  10. Processing microstructure property correlation of porous Ni-YSZ cermets anode for SOFC application

    International Nuclear Information System (INIS)

    Pratihar, Swadesh K.; Dassharma, A.; Maiti, H.S.

    2005-01-01

    The present paper investigates microstructural properties and electrical conductivity of cermets prepared by a solid-state technique, a liquid-dispersion technique and a novel electroless coating technique. The Ni-YSZ processed through different techniques shows varying temperature-conductivity behaviour. The cermets synthesised by electroless coating were found to be electronically conducting with 20 vol% nickel, which is substantially lower than that normally reported. The conductivity of Ni-YSZ cermets was found highest for the samples prepared by an electroless coating technique and lowest for the samples prepared by a solid-state technique, the samples prepared from liquid-dispersion show an intermediate value for a constant nickel content. The variation in electrical conductivity has been well explained from the microstructure of the samples

  11. Composite coating containing WC/12Co cermet and Fe-based metallic glass deposited by high-velocity oxygen fuel spraying

    International Nuclear Information System (INIS)

    Terajima, Takeshi; Takeuchi, Fumiya; Nakata, Kazuhiro; Adachi, Shinichiro; Nakashima, Koji; Igarashi, Takanori

    2010-01-01

    A composite coating containing WC/12Co cermet and Fe 43 Cr 16 Mo 16 C 15 B 10 metallic glass was successfully deposited onto type 304 stainless steel by high-velocity oxygen fuel (HVOF) spraying, and the microstructure and tribological properties were investigated. The microstructure of the coating was characterized by scanning electron microscopy/electron probe micro-analysis (SEM/EPMA) and X-ray diffractometry (XRD). The hardness, adhesion strength and tribological properties of the coating were tested with a Vickers hardness tester, tensile tester and reciprocating wear tester, respectively. The composite coating, in which flattened WC/12Co was embedded in amorphous Fe 43 Cr 16 Mo 16 C 15 B 10 layers, exhibited high hardness, good wear resistance and a low friction coefficient compared to the monolithic coating. The addition of 8% WC/12Co to the Fe 43 Cr 16 Mo 16 C 15 B 10 matrix increased the cross-sectional hardness from 660 to 870 HV and reduced the friction coefficient from 0.65 to 0.5. WC/12Co reinforcement plays an important role in improving the tribological properties of the Fe 43 Cr 16 Mo 16 C 15 B 10 coating.

  12. Corrosion characteristics of several thermal spray cermet-coating/alloy systems

    International Nuclear Information System (INIS)

    Ashary, A.A.; Tucker, R.C. Jr.

    1991-01-01

    The corrosion characteristics of a thermal spray multiphase cermet coating can be quite complex. Factors such as porosity and galvanic effects between different phases in the coating and the substrate, as well as the inherent general and localized corrosion resistance of each phase, can play an important role. The present paper describes the corrosion of several cermet-coating/alloy systems as studied by a potentiodynamic cyclic polarization technique. The corrosion of these coating systems was found to be most often dominated by corrosion of the metallic phases in the coating or of the substrate alloy. (orig.)

  13. Transmutation studies of minor actinides in high intensity neutron fluxes

    International Nuclear Information System (INIS)

    Fioni, G.; Bolognese, T.; Cribier, M.; Marie, F.; Roettger, S.; Faust, H.; Leconte, Ph.

    1999-01-01

    Integral measurements of nuclear data and of the transmutation potential in specific neutron fluxes, constitute the fastest and essential way to overcome to the large uncertainties present in the nuclear data libraries. In the frame of the activities of the Directorate for Science of Matter (DSM) of the French Atomic Energy Authority (CEA), a new project is proposed so as to carry out integral measurements relevant for nuclear waste transmutation systems. A new beam tube will be installed to irradiate actinides and fission fragment samples at different distances from the fuel element of the ILL reactor. Variable neutron energy spectra could then be obtained by choosing the distance between the sample and the fuel element, opening the way to the determination of the ideal physical conditions to incinerate nuclear waste in hybrid transmutation systems. (author)

  14. Organophosphorus reagents in actinide separations: Unique tools for production, cleanup and disposal

    International Nuclear Information System (INIS)

    Nash, K. L.

    2000-01-01

    Interactions of actinide ions with phosphate and organophosphorus reagents have figured prominently in nuclear science and technology, particularly in the hydrometallurgical processing of irradiated nuclear fuel. Actinide interactions with phosphorus-containing species impact all aspects from the stability of naturally occurring actinides in phosphate mineral phases through the application of the bismuth phosphate and PUREX processes for large-scale production of transuranic elements to the development of analytical separation and environment restoration processes based on new organophosphorus reagents. In this report, an overview of the unique role of organophosphorus compounds in actinide production, disposal, and environment restoration is presented. The broad utility of these reagents and their unique chemical properties is emphasized

  15. Subsurface interactions of actinide species and microorganisms. Implications for the bioremediation of actinide-organic mixtures

    International Nuclear Information System (INIS)

    Banaszak, J.E.; Rittmann, B.E.; Reed, D.T.

    1999-01-01

    By reviewing how microorganisms interact with actinides in subsurface environments, the way how bioremediation controls the fate of actinides is assessed. Actinides often are co-contaminants with strong organic chelators, chlorinated solvents, and fuel hydrocarbons. Bioremediation can immobilize the actinides, biodegrade the co-contaminants, or both. Actinides at the IV oxidation state are the least soluble, and microorganisms accelerate precipitation by altering the actinide's oxidation state or its speciation. The way how microorganisms directly oxidize or reduce actinides and how microbiological reactions that biodegrade strong organic chelators, alter the pH, and consume or produce precipitating anions strongly affect actinide speciation and, therefore, mobility is described. Why inhibition caused by chemical or radiolytic toxicities uniquely affects microbial reactions is explained. Due to the complex interactions of the microbiological and chemical phenomena, mathematical modeling is an essential tool for research on and application of bioremediation involving co-contamination with actinides. Development of mathematical models that link microbiological and geochemical reactions is described. Throughout, the key research needs are identified. (author)

  16. Advanced propulsion engine assessment based on a cermet reactor

    Science.gov (United States)

    Parsley, Randy C.

    1993-01-01

    A preferred Pratt & Whitney conceptual Nuclear Thermal Rocket Engine (NTRE) has been designed based on the fundamental NASA priorities of safety, reliability, cost, and performance. The basic philosophy underlying the design of the XNR2000 is the utilization of the most reliable form of ultrahigh temperature nuclear fuel and development of a core configuration which is optimized for uniform power distribution, operational flexibility, power maneuverability, weight, and robustness. The P&W NTRE system employs a fast spectrum, cermet fueled reactor configured in an expander cycle to ensure maximum operational safety. The cermet fuel form provides retention of fuel and fission products as well as high strength. A high level of confidence is provided by benchmark analysis and independent evaluations.

  17. Cermet reinforcement of a weakened endodontically treated root: a case report.

    Science.gov (United States)

    Lui, J L

    1992-08-01

    Many clinical applications have been recommended for glass-cermet cement because of its improved properties compared to the original glass-ionomer cements. It has also been accepted as a dentinal substitute that can strengthen teeth. In this paper, an additional clinical application for glass-cermet cement, the reinforcement of weakened endodontically treated roots, is suggested. This technique is in keeping with the trends of tooth conservation and the use of an adhesive restorative material in the restoration of severely damaged teeth by a conservative approach.

  18. Glass-ceramic nuclear waste forms obtained by crystallization of SiO 2-Al 2O 3-CaO-ZrO 2-TiO 2 glasses containing lanthanides (Ce, Nd, Eu, Gd, Yb) and actinides (Th): Study of the crystallization from the surface

    Science.gov (United States)

    Loiseau, P.; Caurant, D.

    2010-07-01

    Glass-ceramic materials containing zirconolite (nominally CaZrTi 2O 7) crystals in their bulk can be envisaged as potential waste forms for minor actinides (Np, Am, Cm) and Pu immobilization. In this study such matrices are synthesized by crystallization of SiO 2-Al 2O 3-CaO-ZrO 2-TiO 2 glasses containing lanthanides (Ce, Nd, Eu, Gd, Yb) and actinides (Th) as surrogates. A thin partially crystallized layer containing titanite and anorthite (nominally CaTiSiO 5 and CaAl 2Si 2O 8, respectively) growing from glass surface is also observed. The effect of the nature and concentration of surrogates on the structure, the microstructure and the composition of the crystals formed in the surface layer is presented in this paper. Titanite is the only crystalline phase able to significantly incorporate trivalent lanthanides whereas ThO 2 precipitates in the layer. The crystal growth thermal treatment duration (2-300 h) at high temperature (1050-1200 °C) is shown to strongly affect glass-ceramics microstructure. For the system studied in this paper, it appears that zirconolite is not thermodynamically stable in comparison with titanite growing form glass surface. Nevertheless, for kinetic reasons, such transformation (i.e. zirconolite disappearance to the benefit of titanite) is not expected to occur during interim storage and disposal of the glass-ceramic waste forms because their temperature will never exceed a few hundred degrees.

  19. Evaluation of actinide partitioning and transmutation in light-water reactors

    International Nuclear Information System (INIS)

    Collins, Emory D.; Renier, John-Paul

    2004-01-01

    Advanced Fuel Cycle Initiative (AFCI) studies were made to evaluate the feasibility of multicycle transmutation of plutonium and the minor actinides (MAs) in light-water reactors (LWRs). Results showed that significant repository benefits, cost reductions, proliferation resistance, and effective use of facilities can be obtained. Key advantages are shown to be made possible by processing 30-year-decayed spent fuel rather than the more traditional 5-year-decayed fuel. (authors)

  20. Towards an interpretation of the mechanism of the actinides(III)/lanthanides(III) separation by synergistic solvent extraction with nitrogen-containing polydendate ligands

    International Nuclear Information System (INIS)

    Francois, N.

    2000-01-01

    In the field of the separation of long-lived radionuclides from the wastes produced by nuclear fuel reprocessing, aromatic nitrogen-containing polydendate ligands are potential candidates for the selective extraction, alone or in synergistic mixture with acidic extractants, of trivalent actinides from trivalent lanthanides. The first part of this work deals with the complexation of trivalent f cations with various nitrogen-containing ligands (poly-pyridine analogues). Time-resolved laser-induced fluorimetry (TRLIF) and UV-visible spectrophotometry were used to determine the nature and evaluate the stability of each complex. Among the ligands studied, the least basic Me-Btp proved to be highly selective towards americium(III) in acidic solution. In the second part, two synergistic systems (nitrogen-containing polydendate ligand and lipophilic carboxylic acid) are studied and compared in regard to the extraction and separation of lanthanides(III) and actinides(III). TRLIF and gamma spectrometry allowed the nature of the extracted complexes and the optimal conditions of efficiency of both systems to be determined. Comparison between these different studies showed that the selectivity of complexation of trivalent f cations by a given nitrogen-containing polydendate ligand could not always be linked to the Am(III)Eu(III) selectivity reached in synergistic extraction. The latter depends on the 'balance' between the acid-basic properties on the one hand, and on the hard-soft characteristics on the other hand, of both components of synergistic system. (author)