Heavy water moderated gas-cooled reactors

International Nuclear Information System (INIS)

Bailly du Bois, B.; Bernard, J.L.; Naudet, R.; Roche, R.

1964-01-01

France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [fr

Submersion-Subcritical Safe Space (S4) reactor

International Nuclear Information System (INIS)

King, Jeffrey C.; El-Genk, Mohamed S.

2006-01-01

The Submersion-Subcritical Safe Space (S 4 ) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S 4 reactor has a Mo-14% Re solid core, loaded with uranium nitride fuel, cooled by He-30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S 4 reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S 4 reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S 4 reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

Fabrication of the 4. set of fuel elements for the experimental pile EL2; Fabrication du 4. jeu de barreaux de la pile d'essai EL2

Energy Technology Data Exchange (ETDEWEB)

Ringot, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The reactor EL2 is the second atomic reactor built in France. It is a laboratory reactor using heavy water and natural uranium. Its cooling circuit operates with compressed CO{sub 2} gas at 8 kg/cm{sup 2} pressure. The subject of this lecture is the manufacturing of the fourth set of rods. The principle of uranium-can connection is exposed: that is the principle of a pre-pressed bound can. The EL2 reactor has been a prototype with respect to this aspect of the question, and a prototype which has been quite satisfactory. The main steps of the fabrication are exposed: the {gamma} phase extension of uranium, the machining, the three canning (die canning, hydraulic canning, compressed air treatment), the automatic argon arc welding of cups and the different manufacturing controls. (author) [French] Le reacteur EL2 est le deuxieme reacteur construit en France. C'est un reacteur de recherches qui utilise de l'eau lourde et de l'uranium naturel. Il est refroidi par du gaz carbonique sous 8 kg/cm{sup 2} de pression. On etudie dans cet expose la fabrication du quatrieme jeu d'elements combustibles. Le principe de la liaison uranium-gaine est expose: c'est celui d'une gaine precontrainte. La pile EL2 a constitue un prototype a ce point de vue, prototype qui a donne entiere satisfaction. Les principales etapes de la fabrication sont ensuite expliquees: le filage {gamma} de l'uranium, l'usinage des barreaux, les trois operations de gainages (gainage par filiere, gainage hydraulique, gainage a chaud), la soudure automatique des bouchons a l'argon-arc et les differents controles de fabrication. (auteur)

Fabrication of the 4. set of fuel elements for the experimental pile EL2

International Nuclear Information System (INIS)

Ringot, C.

1958-01-01

The reactor EL2 is the second atomic reactor built in France. It is a laboratory reactor using heavy water and natural uranium. Its cooling circuit operates with compressed CO 2 gas at 8 kg/cm 2 pressure. The subject of this lecture is the manufacturing of the fourth set of rods. The principle of uranium-can connection is exposed: that is the principle of a pre-pressed bound can. The EL2 reactor has been a prototype with respect to this aspect of the question, and a prototype which has been quite satisfactory. The main steps of the fabrication are exposed: the γ phase extension of uranium, the machining, the three canning (die canning, hydraulic canning, compressed air treatment), the automatic argon arc welding of cups and the different manufacturing controls. (author) [fr

Characteristics and construction problems of EL 4; Caracteristiques et problemes de construction d'EL4

Energy Technology Data Exchange (ETDEWEB)

Carle, R; Schulhof, P [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires; Sevin, Ph [Electricite de France (EDF), 75 - Paris (France); Buttin, J [Societe INDATOM (France)

1964-07-01

EL 4 is the prototype of a new series of reactors moderated with heavy water and cooled by carbon dioxyde. It was studied with a double purpose: - to realize a sufficiently large and complete reactor for testing the problems of construction and operation of this type of reactors; - to keep in this installation the possibility of adapting the improvements (new materials, improved fuel elements) studied in the R. and D. program. The first objective could not be considered without a considerable volume of preliminary work. This work included particularly the construction and testing from 1962 to 1964 of several prototype channels, out of pile, but under the actual temperature and pressure conditions. These tests showed the proper behaviour of the materials under the severe mechanical and chemical conditions of the reactor. These installations will, furthermore, be available for testing further modifications before adapting them to the reactor. Important tests concerning the safety of the reactor in the event of an explosion of the CO{sub 2} circuit have also been carried out. Construction itself began in July 1962 under the double direction of the Commissariat a l'Energie Atomique and Electricite de France. The civil engineering work will be finished in 1964. The airtight containment (in which it was considered preferable to house the reactor because of its prototype character) was built in pre-stressed concrete, a method which appeared to be particularly rapid and easy. The main piece of the reactor is the heavy water tank. This tank is composed of two double end plates equipped with 216 tubes and joined by a steel cylinder. Some difficult welding problems according to the specifications were solved during the construction of these bottoms. A rigorous series of controls by radiographies and ultrasonic methods was adapted to a complicated geometry. The assembly of the reactor unit, and particularly of the array of tubes supplying the 216 channels, was studied in every

G4-STORK: A Geant4-based Monte Carlo reactor kinetics simulation code

International Nuclear Information System (INIS)

Russell, Liam; Buijs, Adriaan; Jonkmans, Guy

2014-01-01

Highlights: • G4-STORK is a new, time-dependent, Monte Carlo code for reactor physics applications. • G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. • G4-STORK was designed to simulate short-term fluctuations in reactor cores. • G4-STORK is well suited for simulating sub- and supercritical assemblies. • G4-STORK was verified through comparisons with DRAGON and MCNP. - Abstract: In this paper we introduce G4-STORK (Geant4 STOchastic Reactor Kinetics), a new, time-dependent, Monte Carlo particle tracking code for reactor physics applications. G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. The toolkit provides the fundamental physics models and particle tracking algorithms that track each particle in space and time. It is a framework for further development (e.g. for projects such as G4-STORK). G4-STORK derives reactor physics parameters (e.g. k eff ) from the continuous evolution of a population of neutrons in space and time in the given simulation geometry. In this paper we detail the major additions to the Geant4 toolkit that were necessary to create G4-STORK. These include a renormalization process that maintains a manageable number of neutrons in the simulation even in very sub- or supercritical systems, scoring processes (e.g. recording fission locations, total neutrons produced and lost, etc.) that allow G4-STORK to calculate the reactor physics parameters, and dynamic simulation geometries that can change over the course of simulation to illicit reactor kinetics responses (e.g. fuel temperature reactivity feedback). The additions are verified through simple simulations and code-to-code comparisons with established reactor physics codes such as DRAGON and MCNP. Additionally, G4-STORK was developed to run a single simulation in parallel over many processors using MPI (Message Passing Interface) pipes

Shutdown channels and fitted interlocks in atomic reactors

International Nuclear Information System (INIS)

Furet, J.; Landauer, C.

1968-01-01

This catalogue consists of tables (one per reactor) giving the following information: number and type of detectors, range of the shutdown channels, nature of the associated electronics, thresholds setting off the alarms, fitted interlocks. These cards have been drawn up with a view to an examination of the reactors safety by the 'Reactor Safety Sub-Commission', they take into account the latest decisions. The reactors involved in this review are: Azur, Cabri, Castor-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, and Ulysse. (authors) [fr

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

Corrosión de componentes de aluminio en el reactor RP-10

OpenAIRE

Morales Larrea, Soledad; Tenorio de la Cruz, Favio

1983-01-01

Se analiza la velocidad de corrosión, por pérdida de peso de placas de aluminio a diversos valores de pH y temperatura en solución acuosa. El estudio se hace simulando las condiciones de trabajo (pH y temperatura) a las que pueden verse sometidas las vainas de aluminio de los elementos combustibles del reactor RP-10.

Influencia de la velocidad de carga orgánica sobre el proceso de digestión anaerobia de aguas de lavado de aceitunas de almazara en reactores de lecho fluidizado

Directory of Open Access Journals (Sweden)

Borja, Rafael

1998-02-01

Full Text Available A study of the anaerobic digestion process of wastewaters from the washing of olives prior to the oil production process was carried out in a fluidized bed reactor with sepiolite immobilised biomass at 35°G. The bioreactor worked satisfactorily using Influent COD concentrations of 4.5,3.5,2.5 and 1.5 g COD/I for a hydraulic retention time (HRT range of 4.5 to 1,25 days. COD removal efficiencies in the range 50-90% were achieved in the reactor, when evaluated at organic loading rates of between 0.46 and 2.25 g COD/I day using the highest influent substrate concentration (S_o = 4.5 g COD/I. COD and volatile fatty acid (VFA concentration were Increased In parallel with the increase of organic loading rate in the reactor, VFA concentration showing a maximum value of 1.55 g acetic acid/I at the most unfavourable case studied (S_o = 4,5 g COD/I and HRT = 1.25 days. The increase in effluent VFA concentrations was always counteracted by the high alkalinity values (3440-4670 mg CaCO₃/l which brought about the high stability of the process and values of the alkalinity/VFA ratio lower than 0.3-0.5, except for the above-mentioned case, limit value over which the anaerobic process is destabilized. The yield coefficient of methane production was 0.281 methane STP/g COD removed.

Se ha efectuado un estudio del proceso de digestión anaerobia en régimen continuo de aguas de lavado de aceitunas de almazara en un reactor de lecho fluidizado utilizando sepiolita como soporte de adhesión bacteriana a la temperatura de 35°C. El biorreactor operó de modo satisfactorio utilizando concentraciones de alimento de 4,5, 3,5, 2,5, y 1,5 g DQO/I en un rango de 4,5 a 1,25 días de tiempo de retención hidráulico (TRH. Se han obtenido porcentajes de eliminación de DQO entre el 50 y el 90% para velocidades de carga orgánica variables entre 0,46 y 2,25 g DQO/I día utilizando la concentración de alimento más elevada (S

Session 4: Test of a reactor for water-gas-shift reaction on a 3 kW{sub el.} scale at direct combination with auto-thermal reforming

Energy Technology Data Exchange (ETDEWEB)

Pasel, J.; Cremer, P.; Peters, R.; Stolten, D. [Forschungszentrum Julich GmbH, Institute for Materials and Processes in Energy Systems (IWV 3), Julich (Germany)

2004-07-01

The goal of the work described in this paper was to test a reactor for WGS reaction on a larger scale of approx. 3 kW{sub el.} and to demonstrate a successful direct combination of two important components of fuel processing, i.e. a combination of ATR with WGS reaction. The value for the electric power of 3 kW{sub el.} fulfils quite well the demands of a technical application of a fuel cell system if e.g. a so-called Auxiliary Power Unit (APU) is considered. An APU can be used in passenger cars, heavy duty vehicles, ships and air planes. (authors)

Programmed elimination of neutronic poisons in nuclear reactors

International Nuclear Information System (INIS)

Perriere, G. de la

1967-11-01

This work deals with the use of salts of elements having a large neutron capture cross-section, so-called 'soluble poisons' which are dissolved in the moderating water to control the reactivity of heavy-water reactors, and more particularly to compensate the xenon effect in the reactor EL 4. The report describes the controlled elimination of these poisons by fixation on ion-exchange resins. The poisons considered are lithium-6, cadmium and gadolinium in the sulphate form, and boron as boric acid. The thermodynamic and kinetic constants of the ion-exchange reactions were first determined and a study was then made of the fixation of these compounds in beds of small-calibre resins placed in columns. Lithium-6 is the poison which is most easily applicable to compensate the xenon effect in the reactor EL 4. It can be eliminated rapidly and completely from heavy water, and its use does not lead to supplementary problems of protection against the gamma radiation of the reactor circuits. (author) [fr

JENDL-4.0 benchmarking for fission reactor applications

International Nuclear Information System (INIS)

Chiba, Go; Okumura, Keisuke; Sugino, Kazuteru; Nagaya, Yasunobu; Yokoyama, Kenji; Kugo, Teruhiko; Ishikawa, Makoto; Okajima, Shigeaki

2011-01-01

Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0. (author)

Efecto de dos metales pesados, cadmio y níquel, sobre la eficiencia de remoción de carga orgánica de un reactor UASB a escala de laboratorio

Directory of Open Access Journals (Sweden)

Luis Eduardo Forero

2004-01-01

Full Text Available Se realizaron ensayos en tres reactores UASB de tres litros cada uno, a un tiempo de retención hidráulico (TRH de cuatro horas y carga orgánica volumétrica de 4,8 g/L/d. Después de la fase inicial de arranque, tiempo de 4.000 horas para los tres reactores, se procedió a afectarlos de la siguiente forma: el primer reactor fue alimentado con 5 mg/L de cloruro de cadmio en forma continua, el segundo reactor fue alimentado con 10 mg/L de cloruro de níquel en forma continua también, mientras que el tercer reactor no se afectó con sustancia alguna y sirvió como control. La eficiencia de remoción de demanda química de oxígeno (DQO del primer reactor cambió del 60% de la fase de arranque (fase 1 al 18% en la fase afectada con cadmio (fase dos; la eficiencia de remoción de DQO en el reactor dos pasó del 60 al 24% y a su vez para el reactor tres control no hubo cambio significativo en dicha eficiencia. A su vez el reactor uno acumuló el cadmio en el lodo, mientras que el reactor dos no hizo lo propio con el níquel.

Estudio de criticidad del reactor MSBR con SCALE

OpenAIRE

Criado Martín, Alejandro Fernando

2011-01-01

El presente proyecto final de carrera se enmarca en el convenio de colaboración entre el Consejo de Seguridad Nuclear (CSN) y la Universitat Politècnica de Catalunya (UPC) para la realización de proyectos en el ámbito de la seguridad nuclear y la protección radiológica. El proyecto estudia la criticidad del reactor Molten Salt Breeder Reactor (MSBR) mediante el código de simulación SCALE. El MSBR es un reactor de sales fundidas concebido y diseñado por ORNL, con una composic...

Burnup dependent core neutronic calculations for research and training reactors via SCALE4.4

International Nuclear Information System (INIS)

Tombakoglu, M.; Cecen, Y.

2001-01-01

In this work, the full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 code system. KENOV.a module of SCALE4.4 code system is utilized for full core neutronic analysis. The ORIGEN-S module is coupled with the KENOV.a module to perform burnup dependent neutronic analyses. Results of neutronic calculations for 1 st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. These results are extended to burnup dependent core calculations of TRIGA Mark-II research reactors. The code system developed here is similar to the code system that couples MCNP and ORIGEN2.(author)

The use and evolution of the CEA research reactors

International Nuclear Information System (INIS)

Rossillon, F.; Chauvez, C.

1964-01-01

The authors successively examine the different research reactors in use in the French C.E.A. Nuclear Centres. They trace briefly their histories, describing how they have been used up to the present, and how they have been adapted to changes in programme by means of certain modifications. They also describe the reasons which have led to the elaboration of the project for the new reactor Osiris. Zoe, the oldest reactor in the CEA, has been in service in the Centre de Fontenay-aux-Roses since 1948. It is used mainly for measurements of absorption cross-sections in graphite, and for various short irradiations which do not require high fluxes. The reactor EL 2, in service since 1952, was used for the first studies on gas cooling. It has also been widely used for the production of radioisotopes and for a large number of experiments in the fields of physics, metallurgy and physical chemistry. The ageing of certain elements of the reactor has led to the decision to close it down in the near future The reactor EL 3 has been widely used for experiments in physics and in the investigation of fuels. The possibilities of the reactor in fast neutron irradiations will be considerably improved by the adoption of a new type of core (the 'snow crystal' structure). Triton-I, a 2 MW swimming-pool reactor, is used for the most part for fast neutron and gamma irradiations. The modifications being carried out on it at present should result in an increase in the power of the reactor up to 4 or 5 MW. In a neighbouring compartment is housed Triton-II which is of the same general structure, as Triton-I, but whose maximum power is 100 kW. Triton-II is used solely for studies on shielding. Melusine, a 2 MW swimming-pool reactor, has been in use in the Centre d'Etudes Nucleaires de Grenoble since 1959. It has supported a very high programme concerned mainly with solid state physics, fundamental research into refractory fissile materials and special graphites, and the study of the behaviour of

Chernobyl: recovery operations and the entombment of Reactor 4

International Nuclear Information System (INIS)

Dalziel, S.P.C.

1988-01-01

The immediate actions taken following the accident at the Chernobyl-number 4 reactor in April 1986 are described. These included actions to put out the fires, initial medical aid and the dropping of sand, lead, dolomite and boron onto the reactor from helicopters. Following this the chamber below the damaged reactor core was filled with concrete to prevent any further explosions or meltdown. The reactor was subsequently entombed in steel and concrete. The evacuation of the surrounding area is also mentioned. (U.K.)

Operating Experience with the BR-5 Reactor; Experience acquise aupres du reacteur BR-5; Opyt ehkspluatatsii reaktora BR-5; Experiencia practica con el reactor BR-5

Energy Technology Data Exchange (ETDEWEB)

Lejpunskij, A. I.; Kazachkovskij, O. D.; Pinkhasik, M. S.; Aristarkhov, N. N.; Karpov, A. V.; Larin, E. P.; Efimov, I. A.

1963-10-15

The paper discusses the carrying-out of repair and maintenance work on the radioactive liquid-metal circuit of the BR-5 fast neutron reactor. Attention is also given to problems of reactor operation after achievement of the planned 2% fuel burn-up with some disturbance of leak-tightness in individual fuel elements. An account is given of experience in discharging the active section, examining the condition and leak-tightness of the fuel elements, and decontaminating the equipment and piping of the first radioactive circuit after reaching 5% fuel burn-up. (author) [French] Dans ce memoire les auteurs decrivent l'execution des reparations et des travaux d'entretien dans le circuit radioactif liquide-metal du reacteur a neutrons rapides BR-5. Ils etudient egalement les problemes lies au fonctionnement du reacteur au taux de combustion de 2% prevu avec quelques defauts d'etancheite dans des elements combustibles particuliers. Ils decrivent le dechargementen zone active et examinent les conditions d'etancheite des elements combustibles. Ainsi que la decontamination de l'appareillage et des tuyauteries du premier circuit radioactif apres avoir atteint un taux de combustion de 5%. (author) [Spanish] En la memoria se examinan los problemas planteados por el mantenimiento del circuito radiactivo de metal liquido del reactor de neutrones rapidos BR-5. Se tratan cuestiones relacionadas con la explotacion del reactor una vez alcanzado el grado de combustion de 2%, previsto en el proyecto y luego de producirse ciertas alteraciones de la densidad de determinados elementos combustibles. Se describen la experiencia adquirida durante la descarga del cuerpo del reactor, las investigaciones del estado general y de la hermeticidad de los elementos combustibles y las operaciones de descontaminacion de la instalacion y de las tuberias del circuito radiactivo primario despues de alcanzado un grado de combustion de 5%. (author) [Russian] V doklade rassmatrivayutsya voprosy proizvodstva

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

B7-2 expressed on EL4 lymphoma suppresses antitumor immunity by an interleukin 4-dependent mechanism.

Science.gov (United States)

Stremmel, C; Greenfield, E A; Howard, E; Freeman, G J; Kuchroo, V K

1999-03-15

For T cells to become functionally activated they require at least two signals. The B7 costimulatory molecules B7-1 and B7-2 provide the "second signal" pivotal for T cell activation. In this report, we studied the relative roles of B7-1 and B7-2 molecules in the induction of antitumor immunity to the T cell thymoma, EL4. We generated EL4 tumor cells that expressed B7-1, B7-2, and B7-1+B7-2 by transfecting murine cDNAs. Our results demonstrate that EL4-B7-1 cells are completely rejected in syngeneic mice. Unlike EL4-B7-1 cells, we find that EL4-B7-2 cells are not rejected but progressively grow in the mice. A B7-1- and B7-2-EL4 double transfectant was generated by introducing B7-2 cDNA into the EL4-B7-1 tumor line that regressed in vivo. The EL4-B7-1+B7-2 double transfectant was not rejected when implanted into syngeneic mice but progressively grew to produce tumors. The double transfectant EL4 cells could costimulate T cell proliferation that could be blocked by anti-B7-1 antibodies, anti-B7-2 antibodies, or hCTLA4 immunoglobulin, showing that the B7-1 and B7-2 molecules expressed on the EL4 cells were functional. In vivo, treatment of mice implanted with double-transfected EL4 cells with anti-B7-2 monoclonal antibody resulted in tumor rejection. Furthermore, the EL4-B7-2 and EL4-B7-1+B7-2 cells, but not the wild-type EL4 cells, were rejected in interleukin 4 (IL-4) knockout mice. Our data suggests that B7-2 expressed on some T cell tumors inhibits development of antitumor immunity, and IL-4 appears to play a critical role in abrogation of the antitumor immune response.

Fuel slugs considered for use in the high flux reactor EL3; Elements combustibles envisages pour la pile a haut flux EL 3

Energy Technology Data Exchange (ETDEWEB)

Stohr, J A; Caillat, R; Gauthron, M; Montagne, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

EL3 was designed essentially for the study, under irradiation conditions, of materials used in the construction of atomic reactors. The study schedule allocates considerable time and effort to new types of fuel slugs. The present report described the various types of slug being tested or scheduled for tests. After laboratory study, each slug is tested in an experimental cell in the pile. The best are retained and used to charge the reactor (the present charge is purely provisional to permit first criticality and power rise tests)ren. [French] La pile EL3 est essentiellement destinee a l'etude sous irradiation des materiaux utilises dans la construction des reacteurs atomiques. Dans ce programme, une tres large part est reservee a l'etude de nouveaux elements combustibles. Le present rapport decrit les differentes solutions de cartouches dont l'essai est envisage ou en cours. Apres etude en laboratoire, chacune de ces solutions est testee dans une cellule experimentale en pile. Les meilleures seront retenues pour constituer le chargement normal de la pile (le chargement actuel etant essentiellement une solution provisoire qui a permis la divergence de la pile et les premiers essais de montee en puissance). (auteur)

An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

International Nuclear Information System (INIS)

Santos, Adimir dos; Nascimento, Jamil Alves do

2002-01-01

An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 %Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pinch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

International Nuclear Information System (INIS)

Santos, A. dos; Nascimento, J.A. do

2002-01-01

An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 % Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pitch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

Metodología para resolver la ecuación del transporte con el código de ordenadas discretas TORT en el reactor IPEN/MB-01

OpenAIRE

Bernal, A.; Abarca Giménez, Agustín; Barrachina Celda, Teresa María; Miró Herrero, Rafael; Verdú Martín, Gumersindo Jesús

2013-01-01

La resolución de la Ecuación del Transporte Neutrónico en estado estacionario en reactores nucleares de tipo piscina, se consigue normalmente por medio de 2 métodos numéricos diferentes: Monte Carlo (estocástico) y Ordenadas Discretas (determinista). El método de las Ordenadas Discretas resuelve la Ecuación del Transporte Neutrónico para un conjunto de determinadas direcciones, obteniendo un conjunto de ecuaciones y soluciones para cada dirección, donde la solución para cada dirección es el f...

Estudio hidrodinámico de reactores empacados de flujo ascendente(REFA)

OpenAIRE

Díaz Marrero, Miguel Ángel; Dueñas Moreno, Jaime; Cabrera Díaz, Ania

2014-01-01

En el presente trabajo se emplean las técnicas de estímulo respuesta para estudiar los modelos de flujos de dos reactores tipo REFA, con volúmenes de 3,4 y 6 litros respectivamente, usando tiempos de retención hidráulicos y trazadores diferentes en ambos. Se determinaron las curvas de concentración contra tiempo para ambos reactores y se realizó el análisis comparativo de un grupo de relaciones entre los diferentes tiempos que se obtuvieron en los gráficos. Se aplica con los mismos experiment...

Nuclear energy. The innovations of the N4 reactor

International Nuclear Information System (INIS)

Anon.

1998-01-01

The coupling to the electric network of the two first units of N4 type reactors, on the site of Chooz in the Ardennes, marks the third great step of the French nuclear programme of PWR type reactors, after the realization of 34 units of 900 MWe and 20 units of 1300 M We. The nuclear boiler N4, realizes a new evolution in power, in performances and in reliability. (N.C.)

Depleted Reactor Analysis With MCNP-4B

International Nuclear Information System (INIS)

Caner, M.; Silverman, L.; Bettan, M.

2004-01-01

Monte Carlo neutronics calculations are mostly done for fresh reactor cores. There is today an ongoing activity in the development of Monte Carlo plus burnup code systems made possible by the fast gains in computer processor speeds. In this work we investigate the use of MCNP-4B for the calculation of a depleted core of the Soreq reactor (IRR-1). The number densities as function of burnup were taken from the WIMS-D/4 cell code calculations. This particular code coupling has been implemented before. The Monte Carlo code MCNP-4B calculates the coupled transport of neutrons and photons for complicated geometries. We have done neutronics calculations of the IRR-1 core with the WIMS and CITATION codes in the past Also, we have developed an MCNP model of the IRR-1 standard fuel for a criticality safety calculation of a spent fuel storage pool

The fuel element of the first charge for EL 4; presentation, main problems arising in the research, production problems; L'element combustible du 1. jeu de EL 4; presentation, problemes essentiels poses par l'etude, problemes de fabrication

Energy Technology Data Exchange (ETDEWEB)

Ringot, C; Bailly, H; Bujas, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The fuel element making up the first charge for EL-4 is made of slightly enriched uranium oxide canned in stainless steel. This fuel element makes it possible to operate the reactor in the safest conditions awaiting the development of the fuel which will be finally adopted; this will have a low absorption can: beryllium, or a zirconium copper alloy. The 500 mm assembly is made up of 19 small rods placed on 3 rings, inside a graphite jacket. The solution adopted was a solution using completely independent small rods. This report deals with possible problems resulting from their study and production. (authors) [French] L'element combustible du 1er jeu EL-4 est un element combustible a oxyde d'uranium legerement enrichi gaine d'acier inoxydable. C'est un element combustible permettant de faire fonctionner le reacteur EL 4 dans des conditions aussi sures que possible avant de mettre au point le combustible definitif qui sera a gaine peu absorbante: beryllium, ou alliage zirconium-cuivre. L'assemblage de longueur 500 mm est constitue de 19 crayons places sur 3 couronnes, a l'interieur d'une chemise de graphite. La solution adoptee a ete une solution a crayons independants les uns des autres. Ce rapport traite des problemes eventuels poses par leur etude et leur fabrication. (auteurs)

Validation of SCALE4.4a for Calculation of Xe-Sm Transients After a Scram of the BR2 Reactor

International Nuclear Information System (INIS)

Kalcheva, S.; Ponsard, B.; Koonen, E.

2007-01-01

The aim of this report is to validate the computational modules system SCALE4.4a for evaluation of reactivity changes, macroscopic absorption cross sections and calculations of the positions of the Control Rods during their motion in Xe-Sm transient after a scram of the BR-2 reactor. The rapid shutting down of the reactor by inserting of negative reactivity by the Control Rods is known as a reactor scram. Following reactor scram, a large xenon and samarium buildup occur in the reactor, which may appreciably affect the multiplication factor of the core due to enormous neutron absorption. The validation of the calculations of Xe-Sm transients by SCALE4.4a has been performed on the measurements of the positions of the Control Rods during their motion in Xe-Sm transients of the BR-2 reactor and on comparison with the calculations by the standard procedure XESM, developed at the BR-2 reactor. A final conclusion is made that the SCALE4.4a modules system can be used for evaluation of Xe-Sm transients of the BR-2 reactor. The utilization of the code is simple, the computational time takes from few seconds.

Heavy water moderated gas-cooled reactors; Filiere eau lourde - gaz

Energy Technology Data Exchange (ETDEWEB)

Bailly du Bois, B; Bernard, J L; Naudet, R; Roche, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [French] La France, qui a base son effort principal pour la production d'energie nucleaire sur la filiere des reacteurs a uranium naturel et graphite refroidis par gaz, et qui a un programme a plus

B7-2 Expressed on EL4 Lymphoma Suppresses Antitumor Immunity by an Interleukin 4–dependent Mechanism

Science.gov (United States)

Stremmel, C.; Greenfield, E.A.; Howard, E.; Freeman, G.J.; Kuchroo, V.K.

1999-01-01

For T cells to become functionally activated they require at least two signals. The B7 costimulatory molecules B7-1 and B7-2 provide the “second signal” pivotal for T cell activation. In this report, we studied the relative roles of B7-1 and B7-2 molecules in the induction of antitumor immunity to the T cell thymoma, EL4. We generated EL4 tumor cells that expressed B7-1, B7-2, and B7-1+B7-2 by transfecting murine cDNAs. Our results demonstrate that EL4–B7-1 cells are completely rejected in syngeneic mice. Unlike EL4–B7-1 cells, we find that EL4–B7-2 cells are not rejected but progressively grow in the mice. A B7-1– and B7-2–EL4 double transfectant was generated by introducing B7-2 cDNA into the EL4–B7-1 tumor line that regressed in vivo. The EL4–B7-1+B7-2 double transfectant was not rejected when implanted into syngeneic mice but progressively grew to produce tumors. The double transfectant EL4 cells could costimulate T cell proliferation that could be blocked by anti–B7-1 antibodies, anti–B7-2 antibodies, or hCTLA4 immunoglobulin, showing that the B7-1 and B7-2 molecules expressed on the EL4 cells were functional. In vivo, treatment of mice implanted with double-transfected EL4 cells with anti–B7-2 monoclonal antibody resulted in tumor rejection. Furthermore, the EL4–B7-2 and EL4–B7-1+B7-2 cells, but not the wild-type EL4 cells, were rejected in interleukin 4 (IL-4) knockout mice. Our data suggests that B7-2 expressed on some T cell tumors inhibits development of antitumor immunity, and IL-4 appears to play a critical role in abrogation of the antitumor immune response. PMID:10075975

Research reactor fuel bundle design review by means of hydrodynamic testing; Ensayos hidrodinamicos para verificacion de diseno de un elemento combustible para reactores de investigacion

Energy Technology Data Exchange (ETDEWEB)

Pastorini, A; Belinco, C [Comision Nacional de Energia Atomica, San Martin (Argentina). Centro Atomico Constituyentes

1998-12-31

During the design steps of a fuel bundle for a nuclear reactor, some vibration tests are usually necessary to verify the prototype dynamical response characteristics and the structural integrity. To perform these tests, the known hydrodynamic loop facilities are used to evaluate the vibrational response of the bundle under the different flow conditions that may appear in the reactor. This paper describes the tests performed on a 19 plate fuel bundle prototype designed for a low power research reactor. The tests were done in order to know the dynamical characteristics of the plates and also of the whole bundle under different flow rate conditions. The paper includes a description of the test facilities and the results obtained during the dynamical characterization tests and some preliminary comments about the tests under flowing water are also presented. (author) 4 refs., 12 figs., 4 tabs. [Espanol] Durante el diseno de un elemento combustible para un reactor nuclear se requiere de la realizacion de ensayos con el objeto de verificar el comportamiento de ese diseno y permitir, de ser necesario, la introduccion de modificaciones al mismo. Para verificar las caracteristicas de respuesta dinamica e integridad estructural, se realizan ensayos de vibraciones que incluyen someter al prototipo a condiciones de circulacion del fluido similares a las que soportara durante la operacion del reactor. Estos ensayos se realizan en facilidades de ensayos conocidas como circuitos hidrodinamicos, que permiten no solo someter el prototipo al flujo de fluido, sino tambien obtener una adecuada caracterizacion de la respuesta del mismo a traves del luso de sensores de distinto tipo. En este trabajo se describen los ensayos realizados sobre un prototipo de elemento combustible de 19 placas destinado a un reactor de investigacion multiproposito de baja potencia. Los ensayos tuvieron como objetivo conocer la respuesta dinamica de las placas individuales y del elemento combustible en su

Aumento de la actividad metanogenica en lodos granulares, precipitando calcio en el nejayote mediante el burbujeo de CO2

OpenAIRE

Ferreira-Rolón, A.; Ramírez-Romero, G.; Ramírez-Vives, F.

2014-01-01

Más del 60% del agua potable utilizada en el proceso de nixtamalización se desecha al ambiente como agua residual (nejayote). Esta contiene: una alta concentración de materia orgánica (20-30 gDQO L-1), cantidades considerables de Ca2+, PO4-2 y otros nutrientes. En este trabajo se evalúan por un lado, la producción de metano en un reactor de lecho de lodos de flujo ascendente (UASB), diluyendo el nejayote con agua residual municipal a diferentes cargas orgánicas; y por el otro, el efecto del c...

INFLUENCIA DE LA CARGA ORGÁNICA SOBRE LA EFICIENCIA DE REACTORES RBC DE TRES ETAPAS EN EL TRATAMIENTO DE UN EFLUENTE INDUSTRIAL SINTÉTICO

Directory of Open Access Journals (Sweden)

Elisabeth Behling de Calmón

2012-01-01

Full Text Available El presente artículo incluye el estudio de la influencia de la carga orgánica (CO sobre la eficiencia de reactores biológicos rotativos de contacto (RBC, aerobios de tres etapas, al tratar un efluente industrial sintético, con la finalidad de establecer la adecuación del efluente final con respecto a los límites de descarga establecidos en la normativa de Venezuela. Durante la experimentación, se evaluaron pH, alcalinidad total, oxígeno disuelto, demanda biológica de oxígeno (DBO, demanda química de oxígeno (DQO, sólidos suspendidos totales (SST, nitrógeno total Kjeldahl (NTK, NH 4 + , NO 2 - y NO 3 - , de acuerdo con los métodos estándares. La variación de la CO aplicada se obtuvo mediante modificación del tiempo de retención hidráulico (TRH, (24, 12 y 6 h, y se mantuvo constante la DQO de entrada (influente sintético de sacarosa+ urea. La mayor eficiencia de remoción de DQO se obtuvo para un CO global de 11,68 gDQO/m 2 .d (96,25%; TRH=12 h. Para los TRH 24 y 12 h, la eficiencia de remoción global de N-total fue de 66,92 y 62,95%, respectivamente. La mayor remoción de C y de N se obtuvo en la primera etapa de los reactores y se logró cumplir con el límite venezolano permisible de descarga para DQO (<350 mg/L. La posible ocurrencia del proceso de nitrificación repercutió sobre el aumento de las concentraciones de nitrógeno inorgánico en el efluente final.

CD4 expression on EL4 cells as an epiphenomenon of retroviral transduction and selection.

Science.gov (United States)

Logan, Grant J; Spinoulas, Afroditi; Alexander, Stephen I; Smythe, Jason A; Alexander, Ian E

2004-04-01

The EL4 murine tumour cell line, isolated from a chemically induced lymphoma over 50 years ago, has been extensively exploited in immunological research. The conclusions drawn from many of these studies have been based on the presumption that EL4 cells maintain a stable phenotype during experimental manipulation. To the contrary, we have observed 100-fold greater expression of cell surface CD4 (CD4(high)) on a subpopulation of EL4 cells following retroviral transduction and G418 selection when compared with unmodified populations. Although the mechanism responsible for this effect remains to be elucidated, the unexpected expression of CD4, a molecule that functions as both a coreceptor with the T-cell receptor and ligand for the pro-inflammatory cytokine IL-16, has the potential to influence experimental outcomes. Upregulation of CD4 should be excluded when EL4 cells are utilized in experiments requiring a consistent immuno-phenotype.

4 puentes sobre el Rin Alemania Federal

Directory of Open Access Journals (Sweden)

Idelberger, Klaus

1978-05-01

Full Text Available This article describes two different ways of carrying out expansion of the number of lanes, from 4 to 6, on a bridge, without having to interrupt traffic: — The bridge with lanes for metropolitan train-trolley and automobiles between the city of Cologne-Deutz (Deutz and Mülheim are very large outskirts of Cologne, to the east of the Rhine, will be expanded between 1977 and 1979 from 4 to 6 lanes by means of an adjacent bridge of similar characteristics. — Another bridge —the suspension bridge for both metropolitan train-trolley and vehicles— existing between Cologne-Riehl and Cologne-Mülheim was expanded between 1974 and 1977 from 4 to 6 lanes of traffic, distributing in a more favorable manner the transversal section with the existing support structure. At the same time the upper part of the bridge was newly paved and reinforced to adapt to the strictest construction standards, in order to avoid future bulging problems.

Se describen en este artículo dos maneras diferentes de realizar la ampliación del número de carriles —de 4 a 6— de los puentes, sin necesidad de interrumpir el tráfico. — El puente de vigas mixtas para tranvía-ferrocarril metropolitano y automóviles entre la ciudad de Colonia-Deutz (Deutz y Mülheim son grandes barrios de Colonia, al este del Rin se ampliará entre los años 1977 y 1979 de 4 a 6 carriles de circulación por medio de uno adyacente de características similares. — Otro puente —el puente colgante para tranvía-ferrocarril metropolitano y vehículos— existente entre Colonia-Riehl y Colonia-Mülheim se amplió entre los años 1974 y 1977 de 4 a 6 carriles de circulación, distribuyendo de modo más favorable la sección transversal con la estructura portante existente. Al mismo tiempo se asfaltó nuevamente la parte superior del puente y se reforzó adaptándose a las normas de construcción más estrictas, con el fin de evitar en el futuro problemas de abolladura.

Biohydrogen production from diary processing wastewater by anaerobic biofilm reactors

Energy Technology Data Exchange (ETDEWEB)

Rios-Gonzalez, L.J.; Moreno-Davila, I.M.; Rodriguez-Martinez, J.; Garza-Garcia, Y. [Universidad Autonoma de Coahuila, Saltillo, Coahuila (Mexico)]. E-mail: leopoldo.rios@mail.uadec.mx

2009-09-15

to be employed for hydrogen production. [Spanish] Este articulo describe la produccion biologica de hidrogeno a partir de agua residual diaria via fermentacion anaerobica utilizando choque termico pretratado (100 grados centigrados, 30 min.) y procedimientos de tratamiento acido para enriquecer selectivamente el hidrogeno produciendo consorcios mezclados antes de la inoculacion de reactores por lote. El biorreactor empleado para el consorcio de inmovilizacion se opero a temperatura mesofilica (ambiente) (20{+-}3 grados centigrados), bajo condiciones acidofilicas (pH 4.0-4.5), HRT (2h), y un soporte natural para generar hidrogeno produciendo biopelicula de consorcios mezclados: Opuntia imbricata. El reactor se opero inicialmente con sorbitol (5g/L) durante 60 dias de operacion. Las pruebas de lote se llevaron a cabo empleando 20{+-}0.02g de soporte natural con biopelicula. Los experimentos de lote se realizaron para investigar el efecto de la DQO ((2.9-21.1 g-DQO/L), a pH inicial de 7.0, 32{+-}1 grados centigrados. La produccion maxima de hidrogeno se obtuvo a 21.1 g-COD/L. Se efectuaron experimentos del efecto del pH empleando una concentracion de sustrato optima (21.2 g-COD/L), a pH de 4 a 7 y 11.32 (pH de agua residual diaria) y 32{+-}1 grados centigrados. Los resultados de los experimentos indican que el cultivo inicial optimo fue de pH 4.0, pero podemos considerar tambien una produccion estable de hidrogeno a pH 11.32 (pH de agua residual diaria), por lo que se pudo evitar ajustar el pH, y usar agua residual diaria como queda en el proceso de produccion de queso. El pH operacional de 4.0 esta 1.5 unidades por debajo del reportado antes correspondiente al hidrogeno que producen los organismos. La influencia del efecto de la temperatura se realizo usando la concentracion de sustrato optima (21.2 g-COD/L), dos niveles de pH: 4.0 y 11.32, y cuatro diferentes temperaturas: 16{+-}3 grados centigrados (temperatura ambiente), 32{+-}1 grados centigrados, 45{+-}1 grados

A basic design of SR4 instrumentation and control system for research reactor

International Nuclear Information System (INIS)

Syahrudin Yusuf; M Subhan; Ikhsan Shobari; Sutomo Budihardjo

2010-01-01

An SR4 instrumentation and control systems of research reactor is the equipment of nuclear research reactors as power protection devices and control systems. The equipment is to monitor safety parameters and process parameters in the state of reactor shut down, start-up, and in operation at fixed power. In the engineering of Instrumentation and control systems SR4 research reactor, its basic design consists of technical specifications of the reactor protection system devices, technical specifications of the reactor power control system devices, technical specifications information system devices, and systems process termination cabling as a support system. This basic design is used as the basis for the preparation of detailed design and subsequent engineering development of instrumentation systems and control system integrated. (author)

Reactor power cutback system test experience at YGN 4

International Nuclear Information System (INIS)

Chi, Sung Goo; Kim, Se Chang; Seo, Jong Tae; Eom, Young Meen; Wook, Jeong Dae; Choi, Young Boo

1995-01-01

YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor POwer Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/ Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems

Thermal and fast reactor benchmark testing of ENDF/B-6.4

International Nuclear Information System (INIS)

Liu Guisheng

1999-01-01

The benchmark testing for B-6.4 was done with the same benchmark experiments and calculating method as for B-6.2. The effective multiplication factors k eff , central reaction rate ratios of fast assemblies and lattice cell reaction rate ratios of thermal lattice cell assemblies were calculated and compared with testing results of B-6.2 and CENDL-2. It is obvious that 238 U data files are most important for the calculations of large fast reactors and lattice thermal reactors. However, 238 U data in the new version of ENDF/B-6 have not been renewed. Only data of 235 U, 27 Al, 14 N and 2 D have been renewed in ENDF/B-6.4. Therefor, it will be shown that the thermal reactor benchmark testing results are remarkably improved and the fast reactor benchmark testing results are not improved

Decree no. 96-978 from October 31, 1996 giving permission to the French atomic energy Commission (CEA) to create a basic nuclear installation intended to maintain under supervision and in an intermediate dismantling state the old basic nuclear installation no. 28, named Monts d'Arree-EL 4 nuclear power plant (a decommissioned reactor), in the Monts d'Arree site of the Loqueffret town (Finistere, Brittany)

International Nuclear Information System (INIS)

Borotra, F.; Lepage, C.

1996-01-01

This decree from the French ministry of industry and postal services gives permission to the CEA to create a new basic nuclear installation, named EL 4D, which is devoted to the storage of materials from the partially dismantled Monts d'Arree EL 4 reactor. Thus, the CEA is allowed to carry out confining works on the reactor building with the closure of all apertures with the exception of the personnel entry sieve, on the circuits and equipments of the reactor vessel with the plugging of fuel channels, heavy water, helium and demineralized water pipes and of the heads of control rod drive mechanisms and other channels, and on the primary coolant circuit outside the reactor vessel and the steam generators with the installation of welded hatches. The irradiated fuels building, the solid wastes repository, the ventilation building, the heavy water and helium circuits, the fuel handling systems and the effluents treatment plant will be completely dismantled. The other buildings will be pulled down. The rest of the decree enumerates the general technical and safety prescriptions which have to be followed in order to ensure the protection of the personnel against ionizing radiations and of the environment against radioactive pollution. (J.S.)

Application of WIMSD-4 for ''MARIA'' reactor lattice calculations

International Nuclear Information System (INIS)

Andrzejewski, K.; Kulikowska, T.

1993-12-01

A general description of the WIMSD-4 lattice code is given with the emphasis on available geometrical models. The difficulties encountered while modelling reactor lattices with the tubular type fuel elements are explained. Then the analysis of code options allowing to overcome these difficulties is carried out. Eventually, recommendations of options and input parameters for calculations of MARIA reactor lattice with satisfactory accuracy are given. During the work a set of modifications had to be introduced leading to a new code version called WIMS-S. Another version, under the name WIMS-T has been developed to allow for burnup calculations of the MARIA reactor lattice with improved resonance approach. (author). 14 refs, 6 figs, 10 tabs

REACTORES DISCONTINUOS SECUENCIALES: UNA TECNOLOGÍA VERSÁTIL EN EL TRATAMIENTO DE AGUAS RESIDUALES

Directory of Open Access Journals (Sweden)

Juan Fernando Muñoz Paredes

2014-01-01

Full Text Available El interés en eliminar contaminantes de las aguas residuales se ha incrementado en los últimos años. Existe una constante búsqueda de nuevos y mejores diseños que permitan la implementación de sistemas de tratamiento confiables, de bajo costo y que ofrezcan mejores resultados. Esta tarea ha sido realmente un reto, debido a la generación de múltiples tipos de vertimientos líquidos, con diferentes características y composición y, en particular, al cumplimiento de estrictas y diferentes regulaciones que en los distintos países se establecen en materia de control ambiental. La investigación en el área de la ingeniería y las ciencias ambientales ha permitido el desarrollo de la tecnología de reactores discontinuos secuenciales (SBR, por sus siglas en inglés, como una variación del proceso de lodos activados convencional para la eliminación de materia orgánica y de nutrientes de las aguas residuales. El presente artículo de revisión describe la importancia de este tipo de sistemas, teniendo en cuenta las generalidades del tratamiento, la descripción técnica del sistema, los parámetros de operación más importantes, el mecanismo biológico del proceso de eliminación y las diferentes modificaciones al diseño original. Finalmente, se encontró que este tipo de sistema es de gran utilidad y ofrece muchas ventajas en la eliminación de nutrientes y materia orgánica, en comparación con los sistemas convencionales de tratamiento, además de que se puede aplicar para el tratamiento de gran cantidad y diversidad de aguas residuales tanto domésticas como industriales.

In-reactor testing of self-powered neutron detectors and miniature fission chambers

International Nuclear Information System (INIS)

Duchene, J.; LeMeur, R.; Verdant, R.

1975-01-01

The CEA has tested a variety of ''slow'' self-powered neutron detectors with rhodium, silver and vanadium emitters. Currently there are 120 vanadium detectors in the EL4 heavy water reactor. In addition, ''fast'' detectors with cobalt emitters have been tested at Saclay and 50 of these are in reactor. Other studies are concerned with 6 mm diameter miniature fission chambers. Two fast response chambers with argon-nitrogen filling gas became slow during irradiation, but operated to 600 deg C. An argon filled chamber of 4.7 mm diameter, for traversing in core system in pressurized water reactor, has shown satisfactory test results. (author)

On disruption of reactor core of the Chernobylsk-4 reactor (retrospective analysis of experiments and facts)

International Nuclear Information System (INIS)

Platonov, P.A.

2007-01-01

Fragments of graphite blocks from the damaged Chernobyl NPP, unit 4 are studied, the results are analyzed. The temperature of the graphite blocks at the moment of accident release from the reactor is evaluated. Results of studying the fragments of fuel channel and fuel dispersion are considered. The fuel heat content at the moment of the explosion is evaluated and some conclusions are made about the character of the reactor core destruction [ru

Power Reactor Design at Zero Power; Etudes de Reacteurs de Puissance, au Moyen de Machines de Puissance Zero; Konstruktsiya ehnergeticheskogo reaktora nulevoj moshchnosti; Diseno de Reactores Generadores con Ayuda de Reactores de Potencia Nula

Energy Technology Data Exchange (ETDEWEB)

Redman, W. C.; Plumlee, K. E.; Baird, Q. L. [Argonne National Laboratory, Argonne, IL (United States)

1964-02-15

de los elementos combustibles de oxido de uranio-oxido de torio en agua pesada, destacando principalmente los datos necesarios para el diseAo de un segundo cuerpo para el reactor experimental de agua hirviente de Argonne; 2. Preparacien de una maqueta de reactor de investigacion de flujo elevado que permitira veriflcar los calculos efectuados durante el estudio, determinar la geometrra optima y evaluar el efecto de la combustion; 3. Determinacion de las distribuciones energeticas y del efecto de inmersion de los elementos combustibles sobre la reactividad en el caso de un reactor de agua hirviente con sobrecalentador incorporado; 4. Diseflo de un cuerpo de reactor reproductor plutonfgeno de neutrones ripidos, refrigerado por sodio y alimentado con {sup 235}U, que constituiri la carga inicial del segundo reactor reproductor experimental (EBR-II) de Argonne; 5. Estudio de las caracterfcticas de un reactor de dos zonas (termica y rapida) que sufren interaccidn. Al discutir estos programas, los autores explican tambien en que factores se basd la eleccion de los experimentos en conjuntos exponenciales y criticos sin envenenamiento en maquetas de potencia nula, asi como de los experimentos in situ, que sirvieron para obtener los datos necesarios. Tambien describen la importancia de los trabajos analiticos complementarios. La memoria presenta ejemplos especfficos para demostrar en que medida se pueden obtener datos sobre el diseno del reactor antes de explotarlo en regimen normal. Entre estos datos se cuenta el margen de paro, el exceso de reactividad necesario para el funcionamiento, los coeficientes de temperatura, la eficacia de las barras de control y de seguridad, la cinetica del reactor, los esquemas de produccion de energia, los requisitos que ha de cumplir la fuente neutronica de puesta en marcha, y la sensibilidad de los instrumentos, los blindajes y la economfa neutronica. El estudio de los experimentos realizados recientemente con reactores de potencia nula

Biodegradation of Jet Fuel-4 (JP-4) in Sequencing Batch Reactors

Science.gov (United States)

1992-06-01

nalw~eo %CUMENTATION PAGE__ _ _ _ _ _ _ _ _O 74S Ab -A258 020 L AW POi~W6 DATI .~ TYP AIMqm ,-& 0 U. glbs A~ I ma"&LFUN Mu BIODEGRADATION OF JET FUEL...Specific Objectives of This Proposal Are: 1. To assess the ability of C. resinae , P. chrysosporium and selected bacterial consortia to degrade individual...chemical components of JP-4. 2. To develop a sequencing batch reactor that utilizes C. resinae to degrade chemical components of JP-4 in contaminated

Shutdown channels and fitted interlocks in atomic reactors; Chaines de securite et verrouillages installes sur les piles atomiques

Energy Technology Data Exchange (ETDEWEB)

Furet, J; Landauer, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1968-07-01

This catalogue consists of tables (one per reactor) giving the following information: number and type of detectors, range of the shutdown channels, nature of the associated electronics, thresholds setting off the alarms, fitted interlocks. These cards have been drawn up with a view to an examination of the reactors safety by the 'Reactor Safety Sub-Commission', they take into account the latest decisions. The reactors involved in this review are: Azur, Cabri, Castor-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, and Ulysse. (authors) [French] Ce catalogue est compose d'un ensemble de tableaux (a raison de un tableau par pile) donnant les renseignements suivants: nombre et nature des detecteurs, dynamique des chaines, nature de l'electronique associee, seuils provoquant des actions de securite, verrouillages installes. Ces fiches ont ete etablies en vue de l'examen de la securite des piles par la 'Sous-Commission de Surete des Piles', et tiennent compte des decisions de celle-ci. Les reacteurs concernes sont: Azur, Cabri, Cator-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, et Ulysse. (auteurs)

Experiment operations plan for the MT-4 experiment in the NRU reactor

International Nuclear Information System (INIS)

Russcher, G.E.; Wilson, C.L.; Parchen, L.J.; Marshall, R.K.; Hesson, G.M.; Webb, B.J.; Freshley, M.D.

1983-06-01

A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the formal operations plan for MT-4 - the fourth materials deformation experiment conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. A major objective of MT-4 was to simulate a pressurized water reactor LOCA that could induce fuel rod cladding deformation and rupture due to a short-term adiabatic transient and a peak fuel cladding temperature of 1200K (1700 0 F)

El ferrocarril colombiano: 4 temas recurrentes en la literatura

Directory of Open Access Journals (Sweden)

Luis Márquez

2017-04-01

Full Text Available El objetivo de este trabajo es identificar qué se ha investigado sobre el ferrocarril colombiano y qué temas deberían ser objeto de estudio para comprender mejor el futuro del transporte de carga por ferrocarril en Colombia. El trabajo está basado en una revisión descriptiva de la literatura que permitió identificar 4 temas recurrentes sobre el ferrocarril colombiano: los problemas administrativos del Estado, la competencia con el modo carretero, el transporte del carbón y algunos elementos técnicos como el trazado, la capacidad ferroviaria, la eficiencia, la integración modal y la modelación del transporte. Además de la revisión, que es de gran valor para los tomadores de decisiones sobre comercio internacional, política económica y economía regional, son identificadas temáticas de investigación futura.

LEW 88180, LEW 87119, and ALH 85119: New EH6, EL7, and EL4 Enstatite Chondrites

Science.gov (United States)

Zhang, Y.; Benoit, P. H.; Sears, D. W. G.

1993-07-01

The EH and EL chondrites formed in a uniquely reducing environment, containing low-Fe pyroxene, abundant metal, and a number of unusual sulphides and other minerals [1]. An important aspect of their history is that while the EL chondrites consist predominantly of metamorphosed meteorites, the EH consist primarily of little-metamorphosed meteorites (e.g., [2]), and yet EL chondrites have lower equilibrium temperatures than EH chondrite [3,4]. To help understand this observation and its implication for the history of the classes, we have been searching for new enstatite chondrites, looking especially for meteorites of previously unknown chemical-petrologic class. Using our normal INAA methods [5] and sample splits of 100-200 mg, the bulk composition of nine Antarctic enstatite chondrites and one fall were determined. The data were used to assign the meteorites to chemical classes, the Ni/Ir vs. Al/V plot (Fig. 1) being especially useful since it uses the refractory element difference between EH and EL chondrites and is insensitive to metal-silicate heterogeneity. The well-analyzed Qingzhen was included to check our method. ALH84170, ALH84206, and EET87746, which Mason described as E3, E4, and E4 were all found to be EH chondrites [6]. Our data for the three paired EL3 chondrites were discussed earlier (MAC88136, 88180, and 88184) [7,8]. LEW88180, LEW87119, and ALH85119, which Mason described as type E6, E6, and E4 respectively [6], are EH, EL, and EL; thus LEW88180 and ALH85119 appear to be the first EH6 and EL4 chondrites. The compositions of kamacite, phosphide, and niningerite-alabandite (Fig. 2) for ALH84170, ALH84206, EET87746, LEW88180, and ALH85119 are consistent with Mason's petrologic type assignments [6]. The mineral composition of LEW88180 (2.7% Si and 9.4% Ni in the kamacite, 7.8% Ni in the phosphide, and 60% FeS in the niningerite) confirms our classification of this meteorite as EH6. ALH85119 contains kamacite with 0.5% Si and 7% Ni, phosphide with 46

Fast breeder reactors--lecture 4

International Nuclear Information System (INIS)

Marshall, W.; Davies, L.M.

1986-01-01

This paper discusses the economics of fast breeder reactors. An algebraic background is presented which represents the various views expressed by different nations regarding the cost of fast breeder reactors and their associated fuel cycle services, the timescale by which they might be available, and the simultaneous variations in the price of uranium. Actual presentations made by individual countries in recent discussions serve to verify the general nature of this present discussion. It is assumed that if nuclear power is to make a long term contribution to the needs of the world, the introduction of fast breeder reactors is both essential and necessary

In-reactor oxidation of zircaloy-4 under low water vapor pressures

Science.gov (United States)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2015-01-01

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 °C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

In-reactor oxidation of zircaloy-4 under low water vapor pressures

International Nuclear Information System (INIS)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2015-01-01

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 ℃). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

Operating reactors licensing actions summary. Vol.4, No. 4

International Nuclear Information System (INIS)

1984-06-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

Evaluación y estandarización del análisis por activación neutrónica según el método del k-sub cero en el reactor nuclear RP-10: Estudio preliminar empleando irradiaciones cortas

OpenAIRE

Montoya Rossi, Eduardo Haroldo

1995-01-01

Se ha estandarizado una posición de irradiación del reactor nuclear RP-10 para el uso del análisis por activación neutrónica según el método del k sub cero, empleando la convención de Högdahl y se ha evaluado el comportamiento de dicho método respecto a la exactitud y precisión de los resultados obtenidos en el análisis multielemental cuantitativo de diversos materiales certificados de referencia. Para comprobar que el método analítico se encuentra totalmente bajo control estadístico, se ha e...

Differential Expression of Ccn4 and Other Genes Between Metastatic and Non-metastatic EL4 Mouse Lymphoma Cells

OpenAIRE

S. CHAHAL, MANPREET; TERESA KU, H.; ZHANG, ZHIHONG; M. LEGASPI, CHRISTIAN; LUO, ANGELA; M. HOPKINS, MANDI; E. MEIER, KATHRYN

2016-01-01

Background: Previous work characterized variants of the EL4 murine lymphoma cell line. Some are non-metastatic, and others metastatic, in syngenic mice. In addition, metastatic EL4 cells were stably transfected with phospholipase D2 (PLD2), which further enhanced metastasis. Materials and Methods: Microarray analyses of mRNA expression was performed for non-metastatic, metastatic, and PLD2-expressing metastatic EL4 cells. Results: Many differences were observed between non-metastatic and meta...

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

2004-01-01

Full text: The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

2004-01-01

The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance in the

The 4th surveillance testing for Kori unit 3 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-10-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

SCALE-4 analysis of pressurized water reactor critical configurations. Volume 1: Summary

International Nuclear Information System (INIS)

DeHart, M.D.

1995-03-01

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original fresh composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power's Surry Unit 1 reactor for the Cycle 2 core; Volume 4 documents the calculations performed based on GPU Nuclear Corporation's Three Mile Island Unit 1 Cycle 5 core; and, lastly, Volume 5 describes the analysis of Virginia Power's North Anna Unit 1 Cycle 5 core. Each of the reactor-specific volumes provides the details of calculations performed to determine the effective multiplication factor for each reactor core for one or more critical configurations using the SCALE-4 system; these results are summarized in this volume. Differences between the core designs and their possible impact on the criticality calculations are also discussed. Finally, results are presented for additional analyses performed to verify that solutions were sufficiently converged

Questions about the reactor accident with Chernobyl-4

International Nuclear Information System (INIS)

Heijboer, R.J.

1986-01-01

The author presents an inventory of existing information about the Chernobyl-4 accident. Several possible scenarios are described and a comparison is drawn with the Three Mile Island-2 accident. The author concludes that the event is connected to an inherent instability of the RBMK-1000 reactor type. (G.J.P.)

Application of the neutron noise technique for measurement of reactivity for subcritical reactor RA-4

International Nuclear Information System (INIS)

Orso, J; Marenzana, A

2012-01-01

Reactor core RA-4 is divided into two parts that come together to start reactor. The reactor with core separate has the largest subcritical condition, this condition is more secure and therefore the reactor shutdown. In this paper measurements are made of the decay constant of the neutron prompt ' P ', using the α-Rossi and α-Feynman methods to calculate the reactivity of the reactor core for different positions. Both techniques are compared and reactivity is obtained for several position of the reactor core using the α-Rossi technical which is obtained a function that gives the reactivity depending on the separation of the core length. Both techniques are verified using a no multiplicative system. Reactivity values for different position of the core obtained by α-Rossi technique are: $[0 cm] = (-11+/-1) dollar, $[3 cm] = (-7+/-1) dollar, $[3.5 cm] (-5.5+/-0.8) dollar, $[4.2 cm] = (-3.8+/-0.3) dollar y $[4.5] = (-3.0+/-0.1) dollar (author)

Desempeño de un reactor biológico secuencial (RBS en el tratamiento de aguas residuales domésticas

Directory of Open Access Journals (Sweden)

Carmen Cárdenas

2012-07-01

Full Text Available Título en ingles: performance of a sequencing Batch Reactor (SBR in the treatment of domestic sewage Resumen: Se estudió la remoción biológica de materia orgánica y nutrientes de un agua residual doméstica empleando un Reactor Biológico Secuencial (RBS a escala piloto. El estudio fue dividido en cuatro fases en las que se modifico la carga orgánica y la duración de las etapas anaerobia, aerobia y anoxica que conforman cada ciclo de tratamiento, considerando edades de lodo de 10 y 7,5 días. Durante las Fases I y II se operó el sistema con bajos valores de carga másica: 0,364 y 0,220 kg.DQO/Kg.SSV.dia, mientras que durante las Fases III y IV se emplearon cargas mayores: 0,665 y 0,737 kg.DQO/Kg.SSV.dia respectivamente. Los resultados obtenidos muestran que las mayores eficiencias de remoción de materia orgánica en términos de DBO se alcanzaron durante la Fases III (91% y IV (82%, con remoción de fósforo superior a 40%. En cuanto al proceso de nitrificación durante las Fases I y II se registraron tasas de 0,032 y 0,024 kg.N-NH3/kg.SSV.dia, esto debido al menor contenido de materia orgánica y a la baja relación DBO/NKT, mientras que durante las Fases III y IV estas fueron menores: 0,015 kg.N-NH3/kg.SSV.dia durante la Fase III y 0,020 kg.N-NH3/kg.SSV.dia en la Fase IV, sin embargo, fue en estas fases donde se alcanzaron los mayores niveles de desnitrificación durante la etapa anóxica, favorecido por una relación C/N adecuada, próxima de 4 kg.DBO/kg.N-NO3- y la presencia de un substrato de fácil biodegradación. Los resultados obtenidos muestran los RBS como una alternativa eficiente y viable en el tratamiento de aguas residuales domésticas Palabras clave: reactor biológico secuencial; nitrificación/desnitrificación; remoción biológica; edad de lodo. Abstract: It was studied the biological removal of organic matter and nutrients from domestic wastewater using a Sequential Biological Reactor (SBR at pilot scale. The study

EVALUACIÓN DE LA CALIDAD NUTRICIONAL Y MORFOLOGÍA DEL GRANO DE VARIEDADES AMARGAS DE QUINUA BENEFICIADAS EN SECO, MEDIANTE EL NOVEDOSO EMPLEO DE UN REACTOR DE LECHO FLUIDIZADO DE TIPO SURTIDOR

Directory of Open Access Journals (Sweden)

Carla Quiroga Ledezma

2011-01-01

Por tanto, se puede concluir que la quinua beneficiada en el reactor de lecho fluidizado de tipo surtidor tiene una calidad igual o mejor a la quinua que ha sido escarificada, lavada y secada durante el beneficiado.

Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

International Nuclear Information System (INIS)

1992-07-01

Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility

OSCAR-4 Code System Application to the SAFARI-1 Reactor

International Nuclear Information System (INIS)

Stander, Gerhardt; Prinsloo, Rian H.; Tomasevic, Djordje I.; Mueller, Erwin

2008-01-01

The OSCAR reactor calculation code system consists of a two-dimensional lattice code, the three-dimensional nodal core simulator code MGRAC and related service codes. The major difference between the new version of the OSCAR system, OSCAR-4, and its predecessor, OSCAR-3, is the new version of MGRAC which contains many new features and model enhancements. In this work some of the major improvements in the nodal diffusion solution method, history tracking, nuclide transmutation and cross section models are described. As part of the validation process of the OSCAR-4 code system (specifically the new MGRAC version), some of the new models are tested by comparing computational results to SAFARI-1 reactor plant data for a number of operational cycles and for varying applications. A specific application of the new features allows correct modeling of, amongst others, the movement of fuel-follower type control rods and dynamic in-core irradiation schedules. It is found that the effect of the improved control rod model, applied over multiple cycles of the SAFARI-1 reactor operation history, has a significant effect on in-cycle reactivity prediction and fuel depletion. (authors)

Determination of the neutron spectrum at different locations in the Argentine RA-1 Reactor; Determinacion del espectro neutronico en distintas posiciones del reactor RA-1

Energy Technology Data Exchange (ETDEWEB)

Lerner, A M; Madariaga, M R [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

1999-12-31

Full text: It is well known that the RA-1 reactor is used to irradiate different types of materials with neutrons. The Radio dosimetry Group (which belongs to the Nuclear Regulatory Authority) uses its fast column for the design, calibration and set up of criticality dosimeters as well as for a quick assessment of the dose to workers in case of an accident. With such purpose, Au(1), Au under Cd and In(2) foils were irradiated to estimate absolute thermal, epithermal and fast neutron fluxes at the irradiation location. The accuracy of this estimation is higher when the response to the present neutron spectrum of the different materials constituting the detectors is better known. This, in turn, requires the previous knowledge of such spectrum (a detailed energy dependence of neutron flux) at the analysed location. In this work a neutronic calculation is presented at the fast irradiation location. The whole calculation was carried out following two different methodologies, and considering a power of 40 kW. The reactor and its surroundings were represented by a simplified one-dimensional model, as a concentric cylindrical set of regions. Figures are drawn representing fast and thermal fluxes (with the cut at 0.4 eV) as a function of the distance to the core centre. The neutron flux (in n/cm{sup 2}sec.eV) as a function of energy is also shown at the fast irradiation location. Values of flux (in n/cm{sup 2}.sec.eV) are also provided as a function of energy in other typical locations, as well as the equivalent integrated flux values (in n/cm{sup 2}.sec). ((1) According to the reaction Au{sup 197}(n,{gamma})Au{sup 198}, having a cross section of {sigma}{sub 0}=98.8b for thermal neutrons. (2) According to the reaction In{sup 115}(n,n`)In{sup 115m}, with a cross section of some 70 mb for neutrons with energies above 1.2MeV). (author) [Espanol] Texto completo: Como se sabe, el reactor RA1 se utiliza para irradiar con neutrones distintos tipos de materiales. El grupo de

[Establishment and identification of mouse lymphoma cell line EL4 expressing red fluorescent protein].

Science.gov (United States)

Li, Yan-Jie; Cao, Jiang; Chen, Chong; Wang, Dong-Yang; Zeng, Ling-Yu; Pan, Xiu-Ying; Xu, Kai-Lin

2010-02-01

This study was purposed to construct a lentiviral vector encoding red fluorescent protein (DsRed) and transfect DsRed into EL4 cells for establishing mouse leukemia/lymphoma model expressing DsRed. The bicistronic SIN lentiviral transfer plasmid containing the genes encoding neo and internal ribosomal entry site-red fluorescent protein (IRES-DsRed) was constructed. Human embryonic kidney 293FT cells were co-transfected with the three plasmids by liposome method. The viral particles were collected and used to transfect EL4 cells, then the cells were selected by G418. The results showed that the plasmid pXZ208-neo-IRES-DsRed was constructed successfully, and the viral titer reached to 10(6) U/ml. EL4 cells were transfected by the viral solution efficiently. The transfected EL4 cells expressing DsRed survived in the final concentration 600 microg/ml of G418. The expression of DsRed in the transfected EL4 cells was demonstrated by fluorescence microscopy and flow cytometry. In conclusion, the EL4/DsRed cell line was established successfully.

SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configurations

International Nuclear Information System (INIS)

Gauld, I.C.

2000-01-01

Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the US, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k eff values within about 1% Δk/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models

Summary of the 4th workshop on the reduced-moderation water reactor

International Nuclear Information System (INIS)

Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi

2001-09-01

The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

B7-2 Expressed on EL4 Lymphoma Suppresses Antitumor Immunity by an Interleukin 4–dependent Mechanism

OpenAIRE

Stremmel, C.; Greenfield, E.A.; Howard, E.; Freeman, G.J.; Kuchroo, V.K.

1999-01-01

For T cells to become functionally activated they require at least two signals. The B7 costimulatory molecules B7-1 and B7-2 provide the “second signal” pivotal for T cell activation. In this report, we studied the relative roles of B7-1 and B7-2 molecules in the induction of antitumor immunity to the T cell thymoma, EL4. We generated EL4 tumor cells that expressed B7-1, B7-2, and B7-1+B7-2 by transfecting murine cDNAs. Our results demonstrate that EL4–B7-1 cells are completely rejected in sy...

Differential downstream functions of protein kinase Ceta and -theta in EL4 mouse thymoma cells.

Science.gov (United States)

Resnick, M S; Kang, B S; Luu, D; Wickham, J T; Sando, J J; Hahn, C S

1998-10-16

Sensitive EL4 mouse thymoma cells (s-EL4) respond to phorbol esters with growth inhibition, adherence to substrate, and production of cytokines including interleukin 2. Since these cells express several of the phorbol ester-sensitive protein kinase C (PKC) isozymes, the function of each isozyme remains unclear. Previous studies demonstrated that s-EL4 cells expressed substantially more PKCeta and PKCtheta than did EL4 cells resistant to phorbol esters (r-EL4). To examine potential roles for PKCeta and PKCtheta in EL4 cells, wild type and constitutively active versions of the isozymes were transiently expressed using a Sindbis virus system. Expression of constitutively active PKCeta, but not PKCtheta, in s- and r-EL4 cells altered cell morphology and cytoskeletal structure in a manner similar to that of phorbol ester treatment, suggesting a role for PKCeta in cytoskeletal organization. Prolonged treatment of s-EL4 cells with phorbol esters results in inhibition of cell cycling along with a decreased expression of most of the PKC isozymes, including PKCtheta. Introduction of virally expressed PKCtheta, but not PKCeta, overcame the inhibitory effects of the prolonged phorbol ester treatment on cell cycle progression, suggesting a possible involvement of PKCtheta in cell cycle regulation. These results support differential functions for PKCeta and PKCtheta in T cell activation.

Repair of EL4 leaks

International Nuclear Information System (INIS)

1985-03-01

The reactor shutdown was decided on the 15th of November 1984, because the evolution of the carbon dioxide quantity in the helium blanket of the heavy water. Leaks have been localized on three different channels. Repairs have been made in hard conditions taking into account the reactor state (materials strongly irradiated). The restart has been authorized on the 24th of January 1985 [fr

Summary of the 4th workshop on the reduced-moderation water reactor

Energy Technology Data Exchange (ETDEWEB)

Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-09-01

The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

CFD for Nuclear Reactor Safety Applications (CFD4NRS-4) - Workshop Proceedings

International Nuclear Information System (INIS)

2014-01-01

Following the CFD4NRS workshops held in Garching, Germany (Sept. 2006), Grenoble, France (Sep. 2008) and Washington D.C., USA (Sept. 2010), this Workshop is intended to extend the forum created for numerical analysts and experimentalists to exchange information in the application of CFD and CMFD to NRS issues and in guiding nuclear reactor design thinking. The workshop includes single-phase and multi-phase CFD applications, and offers the opportunity to present new experimental data for CFD validation. More emphasis has been given to the experiments, especially on two-phase flow, for advanced CMFD modelling for which sophisticated measurement techniques are required. Understanding of the physics has been depen before starting numerical analysis. Single-phase and multi-phase CFD simulations with a focus on validation were performed in areas such as: single-phase heat transfer, boiling flows, free-surface flows, direct contact condensation and turbulent mixing. These relate to NRS-relevant issues, such as pressurised thermal shock, critical heat flux, pool heat exchangers, boron dilution, hydrogen distribution in containments, thermal striping, etc. The use of systematic error quantification and the application of BPGs were strongly encouraged. Experiments providing data suitable for CFD or CMFD validation were also presented. These included local measurements using multi-sensor probes, laser-based techniques (LDV, PIV or LIF), hot-film/wire anemometry, imaging, or other advanced measuring techniques. There were over 150 registered participants at the CFD4NRS-4 workshop. The programme consisted of 48 technical papers. Of these, 44 were presented orally and 4 as posters. An additional 8 posters related to the OECD/NEA-KAERI sponsored CFD benchmark exercise on turbulent mixing in a rod bundle with spacers (MATiS-H) were presented and a special session was allocated for 6 video presentations. In addition, five keynote lectures were given by distinguished experts. The

9 CFR 75.4 - Interstate movement of equine infectious anemia reactors and approval of laboratories, diagnostic...

Science.gov (United States)

2010-01-01

... infectious anemia reactors and approval of laboratories, diagnostic facilities, and research facilities. 75.4... IN HORSES, ASSES, PONIES, MULES, AND ZEBRAS Equine Infectious Anemia (swamp Fever) § 75.4 Interstate movement of equine infectious anemia reactors and approval of laboratories, diagnostic facilities, and...

Deficient tyrosine phosphorylation of c-Cbl and associated proteins in phorbol ester-resistant EL4 mouse thymoma cells.

Science.gov (United States)

Luo, X; Sando, J J

1997-05-02

Two tyrosine phosphoproteins in phorbol ester-sensitive EL4 (S-EL4) mouse thymoma cells have been identified as the p120 c-Cbl protooncogene product and the p85 subunit of phosphatidylinositol 3-kinase. Tyrosine phosphorylation of p120 and p85 increased rapidly after phorbol ester stimulation. Phorbol ester-resistant EL4 (R-EL4) cells expressed comparable amounts of c-Cbl and phosphatidylinositol 3-kinase protein but greatly diminished tyrosine phosphorylation. Co-immunoprecipitation experiments revealed complexes of c-Cbl with p85, and of p85 with the tyrosine kinase Lck in phorbol ester-stimulated S-EL4 but not in unstimulated S-EL4 or in R-EL4 cells. In vitro binding of c-Cbl with Lck SH2 or SH3 domains was detected in both S-EL4 and R-EL4 cells, suggesting that c-Cbl, p85, and Lck may form a ternary complex. In vitro kinase assays revealed phosphorylation of p85 by Lck only in phorbol ester-stimulated S-EL4 cells. Collectively, these results suggest that Cbl-p85 and Lck-p85 complexes may form in unstimulated S-EL4 and R-EL4 cells but were not detected due to absence of tyrosine phosphorylation of p85. Greatly decreased tyrosine phosphorylation of c-Cbl and p85 in the complexes may contribute to the failure of R-EL4 cells to respond to phorbol ester.

English Language for Teachers (EL4T): a course for EFL teachers

OpenAIRE

Shrestha, Prithvi

2013-01-01

This paper reports on the design and implementation of EL4T in a large-scale project. EL4T is a self-study mobile technology-based ESP course designed to enhance Bangladeshi school English language teachers’ English language skills and pedagogical practices. Key implications of developing this course for ESP in EFL contexts will be presented.

Influencia del solvente en el espectro ultravioleta del 4-nitrobifenilo

Directory of Open Access Journals (Sweden)

Jaime de la Zerda L.

2009-05-01

Full Text Available Se estudiaron los efectos de solvente en la banda del 4-nitrobifenilo en los alcoholes metanol, etanol, I propanol, 2 propanol, 1-butanol, 2-metilpropanol, que fueron purificados cuidadosamente, determinándose luego el grado de pureza que resultó satisfactorio en casi todos los casos. Por ser el agua la impureza persistente en todos ellos, fue necesario estudiar su efecto con cierto detalle.

Measurement of the in-pile core temperature of an EL-4 pencil element, first charge (can of type-347 stainless steel, 0.4 mm thick, UO{sub 2} fuel, 11 mm diameter). Determination of the apparent thermal conductivity integral of in-pile UO{sub 2}; Mesure de la temperature a coeur en pile d'un crayon EL-4 1er jeu (gaine acier inoxydable, nuance 347 - epaisseur 0,4 mm - combustible UO{sub 2} - diametre 11 mm). Determination de l'integrale de conductibilite thermique apparente de l'UO{sub 2} en pile

Energy Technology Data Exchange (ETDEWEB)

Lavaud, B; Ringot, C; Vignesoult, N [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

1966-11-01

The core temperature of a pencil fuel element depends on the thermal conductivity of the UO{sub 2}, and on the UO{sub 2}-can contact. This temperature may be known accurately only if in-pile tests using the actual geometry are carried out. The test described concerns the measurement of the core- temperature of an EL-4 fuel element, first charge, having a stainless steel can. This temperature is measured at the center of the in-pile pencil element using a high-temperature thermocouple (W-Re with Ta sheath). The element is subjected to operating conditions similar to those of EL-4, both for the specific power and the can temperature and for the pressure acting on the can. The specific power is obtained in the EL-3 reactor using a slightly higher enrichment for the UO{sub 2} than that planned for EL-4. The required can temperature and pressure are obtained using a Zircaloy-2 irradiation container filled with NaK, adapted for use in the EL-3 reactor. The core temperatures of the UO{sub 2}, and that of the can surface are measured. The power is calculated from the heat exchanges in the container calibrated in the laboratory. The temperature drop at the UO{sub 2}-can interface is deduced from laboratory measurements carried out under comparable heat flux conditions, and in a gas atmosphere corresponding to the beginning of the life-time of the fuel element. It is possible to draw an integral conductivity curve. It is also possible to check the temperature distribution in the oxide, as deduced from the thermal conductivity integral, by micro-graphic examination of the oxide structure. (authors) [French] La temperature a coeur d'un crayon combustible est fonction de la conductibilite thermique de l'UO{sub 2}, mais aussi du contact UO{sub 2}-gaine. Les essais de mesure en geometrie reelle en pile sont les seuls qui permettent d'avoir une connaissance exacte de cette valeur. L'essai dont il est question dans ce rapport a trait a la mesure de la temperature a coeur d

Calcium signals and caspase-12 participated in paraoxon-induced apoptosis in EL4 cells.

Science.gov (United States)

Li, Lan; Cao, Zhiheng; Jia, Pengfei; Wang, Ziren

2010-04-01

In order to investigate whether calcium signals participate in paraoxon (POX)-induced apoptosis in EL4 cells, real-time laser scanning confocal microscopy (LSCM) was used to detect Ca(2+) changes during the POX application. Apoptotic rates of EL4 cells and caspase-12 expression were also evaluated. POX (1-10nM) increased intracellular calcium concentration ([Ca(2+)]i) in EL4 cells in a dose-dependent manner at early stage (0-2h) of POX application, and apoptotic rates of EL4 cells after treatment with POX for 16h were also increased in a dose-dependent manner. Pre-treatment with EGTA, heparin or procaine attenuated POX-induced [Ca(2+)]i elevation and apoptosis. Additionally, POX up-regulated caspase-12 expression in a dose-dependent manner, and pre-treatment with EGTA, heparin or procaine significantly inhibited POX-induced increase of caspase-12 expression. Our results suggested that POX induced [Ca(2+)]i elevation in EL4 cells at the early stage of POX-induced apoptosis, which might involve Ca(2+) efflux from the endoplasmic reticulum (ER) and Ca(2+) influx from extracellular medium. Calcium signals and caspase-12 were important upstream messengers in POX-induced apoptosis in EL4 cells. The ER-associated pathway possibly operated in this apoptosis. Copyright (c) 2010 Elsevier Ltd. All rights reserved.

Present status of fusion reactor materials, 4

International Nuclear Information System (INIS)

Nagasaki, Ryukichi; Shiraishi, Kensuke; Watanabe, Hitoshi; Murakami, Yoshio; Takamura, Saburo

1982-01-01

Recently, the design of fusion reactors such as Intor has been carried out, and various properties that fusion reactor materials should have been clarified. In the Japan Atomic Energy Research Institute, the research and development of materials aiming at a tokamak type experimental fusion reactor are in progress. In this paper, the problems, the present status of research and development and the future plan about the surface materials and structural materials for the first wall, blanket materials and magnet materials are explained. The construction of the critical plasma testing facility JT-60 developed by JAERI has progressed smoothly, and the operation is expected in 1985. The research changes from that of plasma physics to that of reactor technology. In tokamak type fusion reactors, high temperature D-T plasma is contained with strong magnetic field in vacuum vessels, and the neutrons produced by nuclear reaction, charged particles diffusing from plasma and neutral particles by charge exchange strike the first wall. The PCA by improving 316 stainless steel is used as the structural material, and TiC coating techniques are developed. As the blanket material, Li 2 O is studied, and superconducting magnets are developed. (Koko, I.)

Sedimentación, solubilización, y prefermentación de aguas residuales en un reactor biopelícula

OpenAIRE

Cuevas-Rodríguez, Germán; Tejero Monzón, Iñaki

2003-01-01

Esta investigación fue realizada con el objetivo de desarrollar un nuevo reactor prefermentador de aguas residuales para incrementar los porcentajes de sedimentación, hidrólisis y prefermentación de la materia orgánica contenida en el agua residual bruta, empleando una sola unidad de pretratamiento y, de esta manera, poder remplazar el decantador primario por este nuevo reactor. El estudio fue realizado en un reactor biopelícula de lecho sumergido fijo, empacado con un medio de soporte BLASF‚...

Integral physics data for fast-reactor design; Donnees de physique integrale intervenant dans les etudes de reacteur a neutrons rapides; Integral'nye fizicheskie dannye dlya raschetov reaktorov na bystrykh nejtronakh; Datos fisicos integrales para el diseno de reactores rapidos

Energy Technology Data Exchange (ETDEWEB)

Loewenstein, W B; Meneghetti, D [Argonne National Laboratory, Argonne, IL (United States)

1962-03-15

examinent ces donnees et decrivent leurs domaines d'application. Ils montrent que dans certaines analyses de spectre et d'etat critique, les resultats experimentaux et analytiques sont limites. Ils font des suggestions sur l'orientation des recherches experimentales et analytiques a venir. Elles combleraient le fosse entre la theorie et l'experience qui existe dans les systemes 'connus'. Ces propositions comprennent egalement des suggestions en vue de 'consolider' la physique de modeles theoriques de grands reacteurs surgenerateurs a neutrons rapides. (author) [Spanish] La preparacion del capitulo dedicado a la fisica de los reactores rapidos, en la segunda edicion de la publicacion 'Reactor Physics Constants' que aparecera en breve, exigio la recopilacion de los datos disponibles sobre experimentos integrales. La eleccion de los datos integrales de fisica de los reactores rapidos que se ha de incluir en esa seccion se baso en los dos criterios siguientes: a) que los datos provengan de sistemas relativamente simples que se presten para un analisis teorico sencillo; y b) que se trate de sistemas complejos que representan prototipos o maquetas que ofrecen interes general para el estudio de los reactores de potencia rapidos. Se fijo el primer criterio con la intencion de registrar los datos integrales de aquellos sistemas que tienen una utilidad mas general en la verificacion de los parametros y los procedimientos de calculo de las secciones eficaces. El segundo criterio se basa en la presentacion de los datos corrientes sobre sistemas reales de reactores de potencia reproductores rapidos. Estos son demasiado complicados para permitir un analisis teorico sencillo. Constituyen una demostracion de la complejidad del reactor real si se compara con la instalacion critica de experimentacio n mas esquematica y mas facil de analizar. Los datos fisicos integrales que intrevienen en el diseno de reactores constituyen el resultado de mediciones efectuadas en conjuntos criticos o

[Antitumor effects of matrix protein of vesicular stomatic virus on EL4 lymphoma mice].

Science.gov (United States)

Lin, Shi-jia; Yu, Qin-mei; Meng, Wen-tong; Wen, Yan-jun; Chen, Li-juan; Niu, Ting

2011-03-01

To explore antitumor effects of plasmid pcDNA3. 1-MP encoding matrix protein of vesicular stomatitis virus (VSV) complexed with cationic liposome (DOTAP:CHOL) in mice with EL4 lymphoma. C57BL/6 mouse model with EL4 lymphoma was established. Sixty mice bearing EL4 lymphoma were divided randomly into five groups including Lip-MP, Lip-pVAX, Lip, ADM and NS groups, which were intravenously injected with liposome-pcDNA 3. 1-MP complex, liposome-pVAX complex, empty liposome, Adriamycin and normal saline respectively every three days. Tumor volumes and survival time were monitored. Microvessel density and tumor proliferative index in tumor tissues were determined by CD31, Ki-67 immunohistochemistry staining, meanwhile the tumor apoptosis index was measured by TUNEL method. From 6 days after treatments on, the tumor volume in Lip-MP group was much smaller than that in Lip-pVAX, Lip and NS group (P EL4 tumor cells in vivo (P EL4 lymphoma, which may be related to the induction of tumor cell apoptosis, inhibition of tumor angiogenesis, and suppression of tumor cell proliferation.

Los niveles de anticuerpos anti factor plaquetario 4-heparina y el índice 4T para trombocitopenia inducida por heparina

Directory of Open Access Journals (Sweden)

Marta E. Martinuzzo

2012-02-01

Full Text Available La trombocitopenia inducida por heparina (HIT es un efecto adverso del tratamiento con heparina, mediada por anticuerpos anti complejo factor plaquetario 4 (PF4-heparina (HPIA. La HIT es frecuentemente moderada pero pueden desarrollarse complicaciones trombóticas. El diagnóstico precoz es importante. La detección de HPIA por ELISA tiene alta sensibilidad pero baja especificidad (títulos bajos sin significación clínica. El índice de las 4T (índice 4T puede detectar pacientes con alto riesgo de HIT. El propósito del estudio fue correlacionar los niveles de HPIA y el índice 4T de un grupo de pacientes derivados a nuestro centro. Evaluamos 84 pacientes, 34 de ellos desarrollaron trombosis. Cada médico completó un cuestionario clínico que fue remitido con la muestra a nuestro centro. Los cuestionarios fueron analizados por un investigador externo y el índice 4T se calculó previamente al ensayo. Los HPIA se determinaron por un ELISA (Asserachrom HPIA que detecta los 3 isotipos, IgG, IgM e IgA, único reactivo disponible en Argentina. Los resultados se expresaron como porcentaje de absorbancia (%ABS. La correlación del índice 4T con los HPIA fue 0.472 (rho spearman, p < 0.001. Los pacientes con índice 4T ≥ 6 presentaban %ABS mayores que los ≤ 5 (67 vs. 39, p < 0.001. Aquéllos con trombosis presentaron títulos mayores que los que no la desarrollaron (%ABS 59 vs. 39, p = 0.017. En conclusión: Los títulos altos de HPIA medidos por ELISA, que detecta los 3 isotipos, correlacionaron claramente con el índice 4T ≥ 6 y fueron más frecuentes en los pacientes con trombosis, coincidiendo con lo ya descripto para ensayos de ELISA específicos para isotipo IgG.

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

Science.gov (United States)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo codes for transient reactor analysis

International Nuclear Information System (INIS)

Hoogenboom, J.E.

2013-01-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branch-less collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires the coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3*3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3*3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail. (authors)

Análisis de la sensibilidad paramétrica en reactores de lecho fijo

Directory of Open Access Journals (Sweden)

Hermes A. Rangel Jara

1992-05-01

Full Text Available En la búsqueda de los reactores de lecho fijo que ofrezcan una seguridad y permitanmaximizar la conversión -para una determinada longitud del reactor- se analizan los tres arreglos más comunes (paralelo, contracorriente y temperatura constante, con respecto al medio de enfriamiento. Como casos de aplicación se estudiaron la oxidación parcial de O-xileno para producir anhidrido ftálico como producto único en el primer caso y teniendo en cuenta reacciones paralelas y consecutivas para el segundo caso. El sistema de ecuaciones variacionales originado a partir del sistema de ecuaciones diferenciales del modelo del reactor sirve para solucionar el problema de valores de frontera y adicionalmente la sensibilidad paramétrica de las diferentes variables. Mediante un análisis de la sensibilidad paramétrica y de otras ventajas resultantes el arreglo en paralelo puede considerarse como la alternativa más atractiva.

Regulation of Id2 expression in EL4 T lymphoma cells overexpressing growth hormone.

Science.gov (United States)

Weigent, Douglas A

2009-01-01

In previous studies, we have shown that overexpression of growth hormone (GH) in cells of the immune system upregulates proteins involved in cell growth and protects from apoptosis. Here, we report that overexpression of GH in EL4 T lymphoma cells (GHo) also significantly increased levels of the inhibitor of differentiation-2 (Id2). The increase in Id2 was suggested in both Id2 promoter luciferase assays and by Western analysis for Id2 protein. To identify the regulatory elements that mediate transcriptional activation by GH in the Id2 promoter, promoter deletion analysis was performed. Deletion analysis revealed that transactivation involved a 301-132bp region upstream to the Id2 transcriptional start site. The pattern in the human GHo Jurkat T lymphoma cell line paralleled that found in the mouse GHo EL4 T lymphoma cell line. Significantly less Id2 was detected in the nucleus of GHo EL4 T lymphoma cells compared to vector alone controls. Although serum increased the levels of Id2 in control vector alone cells, no difference was found in the total levels of Id2 in GHo EL4 T lymphoma cells treated with or without serum. The increase in Id2 expression in GHo EL4 T lymphoma cells measured by Id2 promoter luciferase expression and Western blot analysis was blocked by the overexpression of a dominant-negative mutant of STAT5. The results suggest that in EL4 T lymphoma cells overexpressing GH, there is an upregulation of Id2 protein that appears to involve STAT protein activity.

Estudio de ecotoxicidad y biodegradabilidad de ibuprofeno en un reactor aerobio de lodos activos de mezcla completa

OpenAIRE

Zambrano Flores, Johanna Vanessa

2013-01-01

Es importante conocer qué efectos de toxicidad aguda y crónica presenta el Ibuprofeno, así como los posibles efectos tóxicos que a largo plazo puedan producirse sobre la biomasa activa presente en las plantas depuradoras de aguas residuales contaminadas con este compuesto. Para ello, se realizó el montaje de un reactor aerobio de fangos activos de mezcla completa. Primero, se alimentó al reactor únicamente con agua residual sintética para el arranque y operación estacionaria del reactor. Desp...

Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

International Nuclear Information System (INIS)

Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

2008-01-01

The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

Calculations of fuel burn-up and radionuclide inventory in the syrian miniature neutron source reactor using the WIMSD4 code

International Nuclear Information System (INIS)

Khattab, K.

2005-01-01

Calculations of the fuel burn up and radionuclide inventory in the Miniature Neutron Source Reactor after 10 years (the reactor core expected life) of the reactor operating time are presented in this paper. The WIMSD4 code is used to generate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burnt up and plutonium produced in the reactor core, the concentrations and radioactivities of the most important fission product and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core are calculated using the WIMSD4 code as well

Characteristics of Al Alloy as a Material for Hydrolysis Reactor of NaBH4

International Nuclear Information System (INIS)

Jung, Hyeon-Seong; Oh, Sung-June; Jeong, Jae-Jin; Na, Il-Chai; Chu, Cheun-Ho; Park, Kwon-Pil; Chu, Cheun-Ho

2015-01-01

Aluminum alloy was examined as a material of low weight reactor for hydrolysis of NaBH 4 . Aluminum is dissolved with alkali, but there is NaOH as a stabilizer in NaBH 4 solution. To decrease corrosion rate of aluminum, decrease NaOH concentration and this result in loss of NaBH 4 during storage of NaBH 4 solution. Therefore stability of NaBH 4 and corrosion of aluminum should be considered in determining the optimum NaOH concentration. NaBH 4 stability and corrosion rate of aluminum were measured by hydrogen evolution rate. NaBH 4 stability was tested at 20-50 .deg. C and aluminum corrosion was measured at 60-90 .deg. C. The optimum concentration of NaOH was 0.3 wt%, considering both NaBH 4 stability and aluminun corrosion. NaBH 4 hydrolysis reaction continued 200min in aluminum No 6061 alloy reactor with 0.3 wt% NaOH at 80-90 .deg. C.

Heat transfer tests conducted on full-scale model, to investigate cooling conditions of EL.3 experimental reactor

International Nuclear Information System (INIS)

Raievski, R.; Bousquet, M.; Braudeau, M.; Milliat, M.

1958-01-01

For such high heat flux density as is released in the channels of EL3 reactor (2.10 6 kcal/m 2 h on the hottest point) cooling conditions have proved to be satisfactory, that is free from nucleate boiling. The arrangements provided for these tests and the technique used for measurements (of temperature particularly) are specified. Two fields have been investigated: in the former (forced convection without nucleate boiling) a good agreement is found with Colburn's formula. The influence of the ratio L/D is pointed out. The latter field is of forced convection with beginning of nucleate boiling; there the observed raise of the transfer coefficient has been shown occurring with some delay. (author) [fr

Selective up-regulation of protein kinase C eta in phorbol ester-sensitive versus -resistant EL4 mouse thymoma cells.

Science.gov (United States)

Resnick, M S; Luo, X; Vinton, E G; Sando, J J

1997-06-01

Stimulation of sensitive EL4 mouse thymoma cells (s-EL4) with phorbol esters results in production of interleukin 2 (IL-2), adherence to a plastic substrate, and growth inhibition, whereas a phorbol ester-resistant variant (r-EL4) fails to respond. Previous studies revealed substantially decreased expression of protein kinase C (PKC) epsilon in the r-EL4 versus s-EL4 cells. This work has been extended to examine the more recently described PKC isozymes. Western and Northern analyses revealed a marked decrease in PKC eta and theta in r-EL4 as compared to s-EL4 cells. Treatment of these lines with phorbol ester for 24 h resulted in down-regulation of all PKC isozymes examined except PKC eta, which was up-regulated in the s-EL4 cells at the time of maximal IL-2 production. Two newly isolated EL4 clones, resistant to phorbol ester-induced growth inhibition but still exhibiting the phorbol ester-induced adherence and IL-2 production, both expressed PKC eta and theta. Collectively, these observations suggest a dissociation of growth inhibition from adherence and IL-2 production pathways and a potential role for PKC eta in the latter.

Removal of FePO4 and Fe3(PO4)2 crystals on the surface of passive fillers in Fe0/GAC reactor using the acclimated bacteria

International Nuclear Information System (INIS)

Lai, Bo; Zhou, Yuexi; Yang, Ping; Wang, Juling; Yang, Jinghui; Li, Huiqiang

2012-01-01

Highlights: ► Fe 3 (PO 4 ) 2 and FePO 4 crystals would weaken treatment efficiency of Fe 0 /GAC reactor. ► Fe 3 (PO 4 ) 2 and FePO 4 crystals could be removed by the acclimated bacteria. ► FeS and sulfur in the passive film would be removed by the sulfur-oxidizing bacteria. ► Develop a cost-effective bio-regeneration technology for the passive fillers. - Abstract: As past studies presented, there is obvious defect that the fillers in the Fe 0 /GAC reactor begin to be passive after about 60 d continuous running, although the complicated, toxic and refractory ABS resin wastewater can be pretreated efficiently by the Fe 0 /GAC reactor. During the process, the Fe 3 (PO 4 ) 2 and FePO 4 crystals with high density in the passive film are formed by the reaction between PO 4 3− and Fe 2+ /Fe 3+ . Meanwhile, they obstruct the formation of macroscopic galvanic cells between Fe 0 and GAC, which will lower the wastewater treatment efficiency of Fe 0 /GAC reactor. In this study, in order to remove the Fe 3 (PO 4 ) 2 and FePO 4 crystals on the surface of the passive fillers, the bacteria were acclimated in the passive Fe 0 /GAC reactor. According to the results, it can be concluded that the Fe 3 (PO 4 ) 2 and FePO 4 crystals with high density in the passive film could be decomposed or removed by the joint action between the typical propionic acid type fermentation bacteria and sulfate reducing bacteria (SRB), whereas the PO 4 3− ions from the decomposition of the Fe 3 (PO 4 ) 2 and FePO 4 crystals were released into aqueous solution which would be discharged from the passive Fe 0 /GAC reactor. Furthermore, the remained FeS and sulfur (S) in the passive film also can be decomposed or removed easily by the oxidation of the sulfur-oxidizing bacteria. This study provides some theoretical references for the further study of a cost-effective bio-regeneration technology to solve the passive problems of the fillers in the zero-valent iron (ZVI) or Fe 0 /GAC reactor.

The different generation of nuclear reactors from Generation-1 to Generation-4

International Nuclear Information System (INIS)

Cognet, G.

2010-01-01

In this work author deals with the history of the development of nuclear reactors from Generation-1 to Generation-4. The fuel cycle and radioactive waste management as well as major accidents are presented, too.

Tests of the RBMK-1500 reactor fuel assemblies in the Leningrad reactor

International Nuclear Information System (INIS)

Aden, V.C.; Varovin, I.A.; Vorontsov, B.A.

1981-01-01

Test of fuel assemblies of the RBMK-1500 reactor is conducted in the reactor of the Leningrad NPP unit 2 for proving the calculational values of critical power of the RBMK-1500 reactor fuel assemblies adopted in design. The experiment presupposes the maximal approximation of the fuel assembly operation parameters to the calculational critical parameters without bringing into the mode of heat transfer crisis. The experiments are carried out at 500, 850 and 900 MW(el) of the reactor. The maximal channel power made up 472 kW at 20.5 t/h coolant flow rate and 49% mass steam content at the outlet of the channel. It was concluded that there was supply up to the heat transfer crisis in all the investigated modes. Data of temperature measurings of the fuel element cans, readings of the devices of the failure control system of the fuel element cans and external inspection of the assemblies after the tests testify to it [ru

Improvement of macrophage dysfunction by administration of anti-transforming growth factor-beta antibody in EL4-bearing hosts.

Science.gov (United States)

Maeda, H; Tsuru, S; Shiraishi, A

1994-11-01

An experimental therapy for improvement of macrophage dysfunction caused by transforming growth factor-beta (TGF-beta) was tried in EL4 tumor-bearing mice. TGF-beta was detected in cell-free ascitic fluid from EL4-bearers, but not in that from normal mice, by western blot analysis. The ascites also showed growth-suppressive activity against Mv1Lu cells, and the suppressive activity was potentiated by transient acidification. To investigate whether the functions of peritoneal macrophages were suppressed in EL4-bearers, the abilities to produce nitric oxide and tumor necrosis factor-alpha (TNF-alpha) upon lipopolysaccharide (LPS) stimulation were measured. Both abilities of macrophages in EL4-bearing mice were suppressed remarkably on day 9, and decreased further by day 14, compared with non-tumor-bearing controls. TGF-beta activity was abrogated by administration of anti-TGF-beta antibody to EL4-bearing mice. While a large amount of TGF-beta was detected in ascitic fluid from control EL4-bearers, little TGF-beta was detectable in ascites from EL4-bearers given anti-TGF-beta antibody. Furthermore, while control macrophages exhibited little or no production of nitric oxide and TNF-alpha on LPS stimulation in vitro, macrophages from EL4-bearers administered with anti-TGF-beta antibody showed the same ability as normal macrophages. These results clearly indicate that TGF-beta contributes to macrophage dysfunction and that the administration of specific antibody for TGF-beta reverses macrophage dysfunction in EL4-bearing hosts.

Tratamiento de aguas industriales mediante reactor biológico de membranas

OpenAIRE

Aznar Jiménez, Antonio

2008-01-01

El Laboratorio de Ingeniería para el Tratamiento de Aguas de la Universidad Carlos III de Madrid, de investigación y servicios en el tratamiento de aguas residuales, optimiza el diseño y puesta a punto de reactores biológicos de membranas (MBR), indicados para obtener agua depurada de alta calidad y/o aumentar la capacidad de tratamiento.

REVISIÓN DE LAS EXPERIENCIAS EN EL TRATAMIENTO DE AGUAS RESIDUALES DOMÉSTICAS MEDIANTE REACTORES UASB EN COCHABAMBA-BOLIVIA COMPARADAS CON LAS DE LATINOAMÉRICA, INDIA Y EUROPA

Directory of Open Access Journals (Sweden)

Vanessa Gandarillas R.

2017-06-01

Full Text Available El proceso de tratamiento anaeróbico es cada vez más reconocido como opción viable para la protección del medio ambiente y la conservación de los recursos y representa, combinado con otros métodos adecuados, un sistema de tratamiento de aguas residuales sostenible y apropiado para los países en vías de desarrollo. El tratamiento anaeróbico de aguas residuales se está utilizando con éxito en los países tropicales desde la década de los 80, y con resultados alentadores en las regiones subtropicales y templadas. En este estudio se revisan las principales características del tratamiento de aguas residuales domésticas, con énfasis en el Reactor Anaeróbico de Mantos de Lodos de Flujo Ascendente (RALF o Upflow Anaerobic Sludge Blanket (UASB de su sigla en inglés. Se revisaron las aplicaciones del proceso UASB en Europa, Asia y las Américas. En Latino América y en particular en Bolivia el uso de esta tecnología ha incrementado considerablemente debido a la pequeña demanda de terreno requerida para su implementación y a sus atractivos costos de inversión y operación. La revisión mostró, en base a la experiencia de operación de distintas plantas de tratamiento ubicadas en los valles de Cochabamba, que los reactores anaerobios de mantos de lodos de flujo ascendente apropiadamente diseñados, son adecuados para el tratamiento de aguas residuales domésticas en las regiones de los valles y llanos de Bolivia debido a que presentan condiciones ambientales que hacen que el uso de esta biotecnología anaerobia sea favorable bajo la perspectiva del desarrollo sostenible.

Effect of RNAi p21 gene on uncoupling of EL-4 cells induced by X-irradiation

International Nuclear Information System (INIS)

Ju Guizhi; Yan Fengqin; Fu Shibo; Shen Bo; Sun Shilong; Yang Ying; Li Pengwu

2008-01-01

Objective: To investigate the effect of RNAi p21 gene on uncoupling of EL-4 cells induced by X-irradiation. Methods: Construction of RNAi p21 plasmid of pSileneer3.1-H1 neo-p21 was performed. Lipofectamine transfection assay was used to transfer the p21siBNA into EL-4 cells. Fluorescent staining and flow cytometry (FCM) analysis were employed for measurement of protein expression. Fluorescent staining of propidium iodide (PI) and FCM were used for measurement of potyploid cells. Results: In dose-effect experiment it was found that the expression of P21 protein of EL-4 cells increased significantly 24 h after X- irradiation with different doses compared with sham-inadiated control. In time course experiment it was found that the expression of P21 protein of EL-4 cells increased significantly at 8 h to 72 h after 4.0 Gy X-irradiation compared with sham-irradiated control. The results showed that the number of polyploid cells in EL-4 cells was not changed markedly after X-irradiation with doses of 0.5-6.0 Gy. After RNA interference with p21 gene, the expression of P21 protein of EL-4 cells decreased significantly 24 h and 48 h after 4.0 Gy X-irradiation in transfection of plasmid of pSilencer3.1-H1 neo-p21 compared with transfection of plasmid of pSilencer3.1-H1 nco control. And at the same time, the number of polyploid cells in EL-4 cells was increased significantly in transfection of plasmid of pSilencer3.1-H1 neo-p21 compared with transfection of plasmid of pSilencer3.1-H1 nco control. Conclusions: Uncoupling could be induced by X-irradiation in EL-4 cells following BNAi p21 gene, suggesting that P21 protein may play an important role in uncoupling induced by X-rays. (authors)

Report on the Survey of the Design Review of New Reactor Applications. Volume 4: Reactor Coolant and Associated Systems

International Nuclear Information System (INIS)

Downey, Steven; Monninger, John; Nevalainen, Janne; Joyer, Philippe; Koley, Jaharlal; Kawamura, Tomonori; Chung, Yeon-Ki; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Monninger, John; Choi, Young-Joon; )

2017-01-01

At the tenth meeting of the Committee on Nuclear Regulatory Activities (CNRA) Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the Working Group agreed to present the responses to the Second Phase, or Design Phase, of the licensing process survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This report provides a discussion of the survey responses related to the Reactor Coolant and Associated Systems category. The Reactor Coolant and Associated Systems category includes the following technical topics: overpressure protection, reactor coolant pressure boundary, reactor vessel, and design of the reactor coolant system. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the level of effort needed to perform the review. Based on a comparison of the information provided by the member countries in response to the survey, the following observations were made: - Although the description of the information provided by the applicant differs in scope and level of detail among the member countries that provided responses, there are similarities in the information that is required. - All of the technical topics covered in the survey are reviewed in some manner by all of the regulatory authorities that provided responses. - It is common to consider operating experience and lessons learnt from the current fleet during the review process. - The most commonly and consistently identified technical expertise needed to perform design reviews related to this category are mechanical engineering and materials engineering. The complete survey

A pulsed fast reactor; Un reacteur pulse a neutrons rapides; Impul'snyj reaktor na bystrykh nejtronakh; Reactor rapido pulsado

Energy Technology Data Exchange (ETDEWEB)

Blokhin, G. E.; Blokhintsev, D. I.; Blyumkina, Yu. A.; Bondarenko, I. I.; Deryagin, B. N.; Zajmovskij, A. S.; Zinov' ev, V. P.; Kazachkovskij, O. D.; Krasnoyarov, N. V.; Lejpunskij, A. I.; Malykh, V. A.; Nazarov, P. M.; Nikolaev, S. K.; Stavisskij, Yu. Ya.; Ukraintsev, F. I.; Frank, I. M.; Shapiro, F. Ji.; Yazvitskij, Yu. S. [Akademiya Nauk, Moscow, SSSR (Russian Federation)

1962-03-15

temperature de la reactivite. (author) [Spanish] Desde diciembre de 1960, el reactor de impulsos de neutrones rapidos IBR viene funcionando a su potencia nominal en el Instituto Central de Investigaciones Nucleares. Dicho reactor se utiliza como fuente pulsante de neutrones para realizar experimentos de fisica por el metodo del tiempo de vuelo. Se llevan a cabo determinacione s de las secciones eficaces totales, de las secciones eficaces de captura de neutrones intermedios, estudios de las interacciones de los neutrones lentos con los solidos y los liquidos y mediciones de los espectros neutronicos en distintos medios. Los autores describen las principales caracteristicas constructivas del reactor y los resultados de los estudios realizados mediante el mismo. Este reactor trabaja con arreglo a un regimen de impulsos periodicos. Los impulsos de potencia se originan cuando la parte movil del cuerpo, fijada a un disco giratorio, atraviesa la parte estacionaria con una velocidad del orden de los 230 m/s. Gracias a la presencia de una zona movible auxiliar, es posible variar la frecuencia de los impulsos de potencia entre 2,3 y 88 impulsos por segundo. La potencia media del reactor es de 1 kW y la duracion media de los impulsos, de 36 {mu}s. El reactor esta provisto de un sistema de mando y de seguridad que vela por el mantenimiento automatico de la potencia del reactor en su valor medio, asi como por su rapida detencion en caso de perturbacion del funcionamiento. Tambien posee el reactor conductores neutronicos de vacio, que se utilizan en los experimentos de tiempo de vuelo. El conducto principal tiene 1000 m de longitud. En el proceso de puesta en marcha y durante las investigaciones fisicas realizadas con el reactor, se estudio el efecto del desplazamiento de los organos de regulacion y de las partes moviles del cuerpo sobre la reactividad, se determino la duracion de los impulsos a distintos regimenes de trabajo del reactor y se estudiaron las fluctuaciones de la amplitud de

Cytotoxic T lymphocyte recognition of HLA-A/B antigens introduced into EL4 cells by cell-liposome fusion.

Science.gov (United States)

Engelhard, V H; Powers, G A; Moore, L C; Holterman, M J; Correa-Freire, M C

1984-01-01

HLA-A2 and -B7 antigens were introduced into EL4 (H-2b) cells by cell-liposome fusion and were used as targets or stimulators for cytotoxic T lymphocytes (CTL) generated in C57B1/6 (H-2b) mice. It was found that such EL4-HLA cells were not recognized by CTL that had been raised against either a human cell line bearing these HLA antigens or the purified HLA-A2 and -B7 antigens reconstituted into liposomes. In addition, EL4-HLA cells were not capable of inducing CTL that could recognize a human cell line bearing HLA-A2 and -B7 antigens. Instead, EL4-HLA cells induced CTL that specifically lysed EL4-HLA cells and not human cells expressing HLA-A2 and -B7. CTL recognition required the presence of HLA antigens on the EL4 cell surface and was inhibited by antibodies against either H-2b or HLA-A/B. Monoclonal antibody binding studies showed that the expected polymorphic determinants of the HLA-A2 and -B7 antigens were still present on EL4-HLA cells. However, the specificity of CTL or their precursors that are capable of recognizing HLA-A2 or -B7 was altered after these antigens became associated with the EL4 surface. Possible explanations for these results are discussed.

Control of E.L.4. dismantling products

International Nuclear Information System (INIS)

Rotelle, O.; Dionisio Gomes, A.; Pinaux, M.

1998-01-01

Te Brennilis power plant ( a 70 MW heavy water moderated reactor, cooled with Co2) was shut down in July 1985, after 18 years of service. The fuel elements and the heavy water were transferred to facilities for treatment and temporary storage. The second phase of the Brennilis power plant de-construction began in September 1997. Except for the reactor building, the concrete structures will be drained and demolished after inspection and declassification. In the reactor building, only the heavy water circuits will be disassembled. Once this operation has been carried out, new work will be started to provide reactor containment and monitoring. Two types of wastes are produced by the de-construction operations (concrete, steel, nonferrous metals, electrical wiring, insulation, filters, technological wastes): low activity:medium activity wastes, i;e; stored on the surface after being treated and packaged; for very low activity wastes, two possibilities are under consideration: creation of a dedicated surface storage facility and recycling through approved channels. (N.C.)

The stimulation of EL-4 cells to produce interleukin-2 and its potential use in immunocytotoxicity testing

International Nuclear Information System (INIS)

Lasek, W.; Steer, S.; Clothier, R.; Balls, M.

1989-01-01

The ability of EL-4 thymoma cells to produce interleukin-2 (IL-2) following exposure to phorbol-12-myristate 13-acetate (PMA) and Concanavalin A (Con A) has been studied in vitro using medium containing either 10% or 1% fetal calf serum (FCS). The potent stimulatory effect of PMA on IL-2 production by EL-4 cells has been confirmed by measuring 3H-thymidine incorporation by the IL-2-dependent T cell line, CTLL-2, in the presence of conditioned medium (CM) from stimulated cultures. EL-4 cells produced several times more IL-2 when cultured in medium containing 10% FCS than when only 1% FCS was present. Added together, PMA and Con A acted synergistically in some EL-4 cell cultures. The ability of E:-4 cells to produce IL-2 was maintained after further incubation without stimulants. CM with IL-2 activity from stimulated EL-4 cells could prove useful in immunotoxicity testing

Connected analysis nuclear-thermo-hydraulic of parallel channels of a BWR reactor using distributed computation; Analisis acoplado nuclear-termohidraulico de canales paralelos de un reactor BWR empleando computacion distribuida

Energy Technology Data Exchange (ETDEWEB)

Campos Gonzalez, Rina Margarita

2007-07-15

developed work only concentrates in the reactor core, but taking advantage of the modularity that PVM offers, it is possible to add component such as separators and steam dryers, lines of steam and feed water to obtain a model of a complete closed circuit. The applications concentrate mainly in the training of personnel in the phenomenology of the BWR, and as an investigation tool in the study of the dynamics of BWR reactors. The oscillations out of phase study presents challenges at the moment as are the explanation of the variation of the neutral line with time, non azimuthal but axial oscillations out of phase, etc. So far a first model oriented in this direction is at hand. [Spanish] Este trabajo consiste en la integracion de tres modelos desarrollados previamente los cuales se encuentran ampliamente descritos en la literatura: modelo del canal termohidraulico, modelo de la neutronica modal y el modelo de los lazos de recirculacion. La herramienta utilizada para este acoplamiento de modelos es el sistema PVM, Parallel Virtual Machine, que permitio paralelizar el modelo mediante el concepto de computacion distribuida. La finalidad de hacer este acoplamiento de modelos es la de obtener una herramienta mas completa que represente mejor la configuracion real y la fenomenologia del nucleo de un reactor BWR, obteniendo asi mejores resultados. Ademas mantener la flexibilidad de mejorar el modelo resultante en cualquier momento, ya que los modelos muy complejos o sofisticados resultan dificiles de mejorar siendo imposible modificar las ecuaciones que utilizan y pueden incluir variables que no son de importancia primaria en el problema tratado o que enmascaren relaciones entre variables debido al exceso de resultados. Tambien el mantener la flexibilidad de agregar modelos de componentes o sistemas del reactor BWR, todo esto dependiendo de las necesidades del modelado. Se eligio a la planta sueca Ringhals para caracterizar el modelo acoplado resultante por contar con un Benchmark

J.O. no. 10 of the 13 january 2004, page 991, text no. 14. Decrees, orders. General texts. Decree no. 2004-47 of the 12 january 2004 modifying the decree no. 96-978 of the 31 october 1996 relative to the creation of the nuclear installation no. 162 named EL4-D, installation of storage for the Monts d'Arree power plants materials

International Nuclear Information System (INIS)

2004-06-01

The reactor EL4, implemented in december 1966, has definitely shutdown on the 31 july 1985. This reactor was an industrial prototype, built and exploited by Cea and EDF. In the framework of the partial dismantling of the installation, the decree 96-978 of the 31 october 1996 agreed the Cea to modify the installation to become a storage installation. (A.L.B.)

Algoritmo para el 4ap haciendo uso de la metaheuristica sistema hormiga

Directory of Open Access Journals (Sweden)

Manuel Vicente Centeno Romero

2008-07-01

Full Text Available La metaheurística sistema hormiga consiste en la analogía entre el procedimiento que utilizan las hormigas reales para la búsqueda de alimentos, encontrando la ruta más corta, y los problemas de optimización combinatoria para encontrar la mejor solución. Entre estos problemas se encuentra el de asignación multidimensional (mAP, el cual es un problema NP-difícil para m > 2. En la actualidad no se ha desarrollado trabajo alguno sobre el 4AP (mAP con m=4, tampoco existe publicación sobre la aplicación del sistema hormiga para el mAP. En este trabajo se desarrolló un software que hace uso de la metaheurística sistema hormiga para encontrar la mejor solución o aproximada al 4AP, con problemas generados aleatoriamente. Se implementó una metodología híbrida entre la metodología de Investigación de Operaciones descrita por Taha (1991 y la ingeniería de software (1990. El número de asignaciones tomadas en cuenta para verificar qué tan buenas son las soluciones arrojadas por el software, varía desde n=2 hasta n=25, obteniéndose soluciones exactas para 2 £ n £ 6, ya que al compararse dichas soluciones con las dadas por XPRESS (software de programación lineal que se usa para resolver modelos matemáticos se observa la igualdad en los resultados. Para n _ 7, XPRESS (la versión utilizada no ofrece respuesta alguna, pero el software desarrollado arroja buenos resultados en un tiempo computacional razonable, considerando el número de asignaciones que se procesan en cada n. The metaheuristic ant system consists on the analogy among the procedure used by the real ants for searching food, for finding the shortest route; and the problems of combinatoria optimization to find the best solution. Among these problems we can find the multidimensional assignment problem (mAP, which is a NP-difficult problem for m > 2. At the present time, there isn‘t any work that has been developed, on this topic about the 4AP (mAP with m=4, neither

De "átomos para la paz" a los reactores de potencia: Tecnología y política nuclear en la Argentina (1955-1976)

OpenAIRE

Hurtado de Mendoza, Diego

2005-01-01

Durante el período 1955-76, el programa nuclear argentino se integró a la arena internacional; su Comisión Nacional de Energía Atómica construyó cuatro reactores de investigación, adquirió a una empresa alemana y puso en marcha el primer reactor de potencia Atucha I, y compró a una empresa canadiense un segundo reactor de potencia. En este artículo se examinan estos desarrollos en relación con el contexto político local y con el panorama nuclear internacional. En particular, se analizan la po...

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2008

International Nuclear Information System (INIS)

2014-02-01

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment and practice in the nuclear reactor training, Irradiation for activation analyses, RI productions, fission tracks etc. In the fiscal year 2008, the research reactor JRR-3 was operated for 7 cycles (cycle operation : 26days/cycle) for utilization sharing of facility. The research reactor JRR-4 was not operated in 2008. Because a crack was found on the weld of the aluminum cladding of a graphite reflector element. JRR-4 has remained shutdown until the reflector elements were replaced. The volume contains 250activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, and others submitted by the users in JAEA and other Organizations. (author)

Estudio de un reactor catalítico para la obtención de gas de síntesis

OpenAIRE

Romero Sayago, Sara Isabel

2016-01-01

Este trabajo se centra en el estudio del proceso de reformado de gas natural con vapor de agua para producir gas de síntesis. Un compuesto, que como su nombre indica, es de gran importancia en la síntesis de muchos productos. En concreto, se estudia el reactor heterogéneo catalítico donde tiene lugar la reacción de reformado. Mediante un programa de simulación de procesos químicos, se optimiza el proceso de reformado para obtener un rendimiento elevado en el reactor con el mínimo consumo e...

J.O. no. 10 of the 13 january 2004, page 991, text no. 14. Decrees, orders. General texts. Decree no. 2004-47 of the 12 january 2004 modifying the decree no. 96-978 of the 31 october 1996 relative to the creation of the nuclear installation no. 162 named EL4-D, installation of storage for the Monts d'Arree power plants materials; J.O. no. 10 du 13 janvier 2004, page 991, texte no. 14. Decrets, arretes, circulaires. Textes generaux. Decret no. 2004-47 du 12 janvier 2004 modifiant le decret no. 96-978 du 31 octobre 1996 relatif a la creation de l'installation nucleaire de base no. 162 denommee EL4-D, installation d'entreposage de materiels de la centrale nucleaire des monts d'Arree

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-06-01

The reactor EL4, implemented in december 1966, has definitely shutdown on the 31 july 1985. This reactor was an industrial prototype, built and exploited by Cea and EDF. In the framework of the partial dismantling of the installation, the decree 96-978 of the 31 october 1996 agreed the Cea to modify the installation to become a storage installation. (A.L.B.)

Differential Expression of Ccn4 and Other Genes Between Metastatic and Non-metastatic EL4 Mouse Lymphoma Cells.

Science.gov (United States)

Chahal, Manpreet S; Ku, H Teresa; Zhang, Zhihong; Legaspi, Christian M; Luo, Angela; Hopkins, Mandi M; Meier, Kathryn E

Previous work characterized variants of the EL4 murine lymphoma cell line. Some are non-metastatic, and others metastatic, in syngenic mice. In addition, metastatic EL4 cells were stably transfected with phospholipase D2 (PLD2), which further enhanced metastasis. Microarray analyses of mRNA expression was performed for non-metastatic, metastatic, and PLD2-expressing metastatic EL4 cells. Many differences were observed between non-metastatic and metastatic cell lines. One of the most striking new findings was up-regulation of mRNA for the matricellular protein WNT1-inducible signaling pathway protein 1 (CCN4) in metastatic cells; increased protein expression was verified by immunoblotting and immunocytochemistry. Other differentially expressed genes included those for reproductive homeobox 5 (Rhox5; increased in metastatic) and cystatin 7 (Cst7; decreased in metastatic). Differences between PLD2-expressing and parental cell lines were limited but included the signaling proteins Ras guanyl releasing protein 1 (RGS18; increased with PLD2) and suppressor of cytokine signaling 2 (SOCS2; decreased with PLD2). The results provide insights into signaling pathways potentially involved in conferring metastatic ability on lymphoma cells. Copyright© 2016, International Institute of Anticancer Research (Dr. John G. Delinasios), All rights reserved.

Core reactor simulation of the Central Laguna Verde (CLV) reactor in stationary state and an example of the application in the recharge options analysis of cycle 3; Simulacion del nucleo del reactor de la Central Laguna Verde (CLV) en estado estacionario y ejemplo de aplicacion en el analisis de alternativas de recarga del ciclo 3

Energy Technology Data Exchange (ETDEWEB)

Ocampo Mansilla, Hector; Francois Lacouture, Juan Luis; Blanco Lara, Jesus; Cortes Campos, Carlos Cristobal; Esquivias Montoya, Jesus; Esquivel Torres, Jose Luis; Martin del Campo Marquez, Cecilia [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Montes Tadeo, Jose Luis [Instituto Nacional de Investigaciones Nucleares (ININ), Salazar (Mexico); Sanchez Herrera, Luciano; Torres Alvarez, Carlos [Comision Federal de Electricidad (CFE), Mexico, D. F. (Mexico)

1991-12-31

The results are presented of a study requested by Comision Federal de Electricidad (CFE) for the analysis of Cycle 3, of Unit No. 1 of the Laguna Verde Nuclear Power Station (CLNV) and determine the burning effect impact, carried out with the starting tests and the operation of Cycles 1 and 2 on base of the cycle extension known as coastdown. The calculations were realized with the Code Package FMS for fuel managing, using the Code PRESTO-B that analyzes the reactor in detailed form in three dimensions an in stationary state. In the study the schemes of fraction of recharge proposed by General Electric (GE) were analyzed with the effect of cycle extension. The initial design value of 100 assemblies for Cycle 3, GE proposes to increase such fraction from 112 to 120 assemblies. This impacts the cost of the second recharge and the purpose of this investigation is to analyze options with higher fuel enrichment in U-235 to minimize the number of assemblies in this recharge. The analyses effected show that the designs proposed by GE do not fulfill the required energy proposed for the cycle, even using in the recharge only fuel with 3.03% of enrichment. It is proposed, likewise, the fuel enrichment up to 3.25% to satisfy the energy demand with a minimum of assemblies. [Espanol] Se presentan los resultados de un estudio solicitado por la Comision Federal de Electricidad (CFE) para analizar el ciclo 3, de la unidad 1 de la Central Laguna Verde (CLV), y determinar el impacto del efecto de quemado llevado a cabo con las pruebas de arranque y por la operacion de los ciclos 1 y 2 con base en la tecnica de alargamiento del ciclo conocida como coastdown1. Los calculos se realizaron con el paquete de codigos FMS para la administracion de combustible, usando el codigo PRESTO-B que analiza el reactor en forma detallada en tres dimensiones y en estado estacionario. Se analizaron en el estudio los esquemas de fraccion de recarga propuesta por la General Electric (GE) con el efecto de

Core reactor simulation of the Central Laguna Verde (CLV) reactor in stationary state and an example of the application in the recharge options analysis of cycle 3; Simulacion del nucleo del reactor de la Central Laguna Verde (CLV) en estado estacionario y ejemplo de aplicacion en el analisis de alternativas de recarga del ciclo 3

Energy Technology Data Exchange (ETDEWEB)

Ocampo Mansilla, Hector; Francois Lacouture, Juan Luis; Blanco Lara, Jesus; Cortes Campos, Carlos Cristobal; Esquivias Montoya, Jesus; Esquivel Torres, Jose Luis; Martin del Campo Marquez, Cecilia [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Montes Tadeo, Jose Luis [Instituto Nacional de Investigaciones Nucleares (ININ), Salazar (Mexico); Sanchez Herrera, Luciano; Torres Alvarez, Carlos [Comision Federal de Electricidad (CFE), Mexico, D. F. (Mexico)

1992-12-31

The results are presented of a study requested by Comision Federal de Electricidad (CFE) for the analysis of Cycle 3, of Unit No. 1 of the Laguna Verde Nuclear Power Station (CLNV) and determine the burning effect impact, carried out with the starting tests and the operation of Cycles 1 and 2 on base of the cycle extension known as coastdown. The calculations were realized with the Code Package FMS for fuel managing, using the Code PRESTO-B that analyzes the reactor in detailed form in three dimensions an in stationary state. In the study the schemes of fraction of recharge proposed by General Electric (GE) were analyzed with the effect of cycle extension. The initial design value of 100 assemblies for Cycle 3, GE proposes to increase such fraction from 112 to 120 assemblies. This impacts the cost of the second recharge and the purpose of this investigation is to analyze options with higher fuel enrichment in U-235 to minimize the number of assemblies in this recharge. The analyses effected show that the designs proposed by GE do not fulfill the required energy proposed for the cycle, even using in the recharge only fuel with 3.03% of enrichment. It is proposed, likewise, the fuel enrichment up to 3.25% to satisfy the energy demand with a minimum of assemblies. [Espanol] Se presentan los resultados de un estudio solicitado por la Comision Federal de Electricidad (CFE) para analizar el ciclo 3, de la unidad 1 de la Central Laguna Verde (CLV), y determinar el impacto del efecto de quemado llevado a cabo con las pruebas de arranque y por la operacion de los ciclos 1 y 2 con base en la tecnica de alargamiento del ciclo conocida como coastdown1. Los calculos se realizaron con el paquete de codigos FMS para la administracion de combustible, usando el codigo PRESTO-B que analiza el reactor en forma detallada en tres dimensiones y en estado estacionario. Se analizaron en el estudio los esquemas de fraccion de recarga propuesta por la General Electric (GE) con el efecto de

HYDROGEN KINETICS LIMITATION OF AN AUTOTROPHIC SULPHATE REDUCTION REACTOR

Directory of Open Access Journals (Sweden)

CÉSAR SÁEZ-NAVARRETE

2012-01-01

Full Text Available El uso de sustratos inorgánicos podría reducir los costos y simplificar la operación de sistemas de tratamiento de aguas que utilizan bacterias reductoras de sulfato. Sin embargo, el uso de H2 como sustrato energético y la bioproducción de H2S podrían provocar limitaciones cinéticas. El objetivo de este estudio fue evaluar las condiciones en las que la capacidad de transferencia de masa de un bioreactor de reducción de sulfato, limita su cinética de reducción. La cinética del reactor fue obtenida monitoreando la presión del sistema en condiciones de no limitación por sulfato. Se concluyó que el diseño del bioreactor debería basarse en sus propiedades de transferencia. La tasa de consumo de H2 alcanzó un máximo de 10-4 M/min, para una tasa de reducción de sulfato de 3.4 g·L-1·d-1. Para evitar limitación por H2 se requirió un kLa de 1.48 min-1 a 1.2·109 cells/L (1.23·10-9 L·min-1·cell-1, valor relevante para propósitos de escalamiento.

DINÁMICA DE UN REACTOR DE BIOPELÍCULA ANAEROBIA TIPO INTERCAMBIADOR DE CALOR (RBAIC

Directory of Open Access Journals (Sweden)

Carlos Ramiro Escalera Vásquez

2005-01-01

Full Text Available Las características dinámicas de un reactor de biopelícula anaerobio tipo intercambiador de calor (RBAIC, usado para el tratamiento de aguas residuales de melazas, fueron estudiadas experimentalmente. Se realizaron experimentos para estudiar la respuesta del reactor a las sobrecargas orgánicas. También se estudiaron los efectos de los cambios de temperatura de las paredes calientes y las temperaturas ambientales, sobre la eficiencia del reactor, bajo condiciones de estado estacionario. Se demostró que el RBAIC es estable ante la ocurrencia de sobrecarga orgánica. Se concluyó que existe una separación de fases microbianas dentro del reactor, en condiciones normales de operación. Es decir, las bacterias acidogénicas predominan en la masa líquida recirculante y las heteroacetogénicas y metanogénicas lo hacen en la biopelícula adherida a las paredes calientes de transferencia de calor, lo cual implica que los cambios de la temperatura de la pared afectan de mayor manera a la eficiencia de remoción, que los cambios de temperatura del entorno. El RBAIC es una configuración  novedosa, con características energéticas favorables para el tratamiento de aguas residuales de la industria alimenticia.

Design of the fuel element 'snow-flake' in uranium oxide, canned with aluminium, for the experimental reactor EL 3 (1960); Etude d'un element combustible en oxyde d'uranium gaine d'aluminium, type ''cristal de neige'' pour la pile EL 3 (1960)

Energy Technology Data Exchange (ETDEWEB)

Gauthron, M; Guibert, B [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

This report sums up the main studies have been carried out on the fuel element 'Snowflake' (uranium oxide, canned with aluminium), designed to replace the present element of the experimental reactor EL3 in order to increase the reactivity without modifying the neutron flux/thermal power ratio. (author) [French] Ce rapport resume les principales etudes qui ont ete faites sur l'element combustible 'Cristal de Neige' (a oxyde d'uranium, gaine d'aluminium) destine a remnlacer l'element actuel du reacteur experimental EL3, afin d'en augmenter la reactivite sans modifier le rapport flux neutronique-puissance thermique. (auteur)

Software for the nuclear reactor dynamics study using time series processing; Software para el estudio de la dinamica de reactores nucleares mediante el procesamiento de series temporales

Energy Technology Data Exchange (ETDEWEB)

Valero, Esbel T.; Montesino, Maria E. [Instituto Superior de Ciencia y Tecnologia Nuclear (ISCTN), La Habana (Cuba)

1997-12-01

The parametric monitoring in Nuclear Power Plant (NPP) permits the operational surveillance of nuclear reactor. The methods employed in order to process this information such as FFT, autoregressive models and other, have some limitations when those regimens in which appear strongly non-linear behaviors are analyzed. In last years the chaos theory has offered new ways in order to explain complex dynamic behaviors. This paper describes a software (ECASET) that allow, by time series processing from NPP`s acquisition system, to characterize the nuclear reactor dynamic as a complex dynamical system. Here we show using ECASET`s results the possibility of classifying the different regimens appearing in nuclear reactors. The results of several temporal series processing from real systems are introduced. This type of analysis complements the results obtained with traditional methods and can constitute a new tool for monitoring nuclear reactors. (author). 13 refs., 3 figs.

Safety of research reactors. Topical issues paper no. 4

International Nuclear Information System (INIS)

Alcala-Ruiz, F.; Ferraz-Bastos, J.L.; Kim, S.C.; Voth, M.; Boeck, H.; Dimeglio, F.; Litai, D.

2001-01-01

Assessment of Research Reactors (INSARR) missions. The prime objective of these missions has been to conduct a comprehensive operational safety review of the research reactor facility and to verify compliance with the IAEA Safety Standards. The methods used during an INSARR mission have been collected and analysed. Some of the important issues identified are the following: general ageing of the facility; uncertain status of many research reactors (in extended shutdown); indefinite deferral of return to operation or decommissioning; inadequate regulatory supervision; insufficient systematic (periodic) reassessment of safety; lack of quality assurance (QA) programmes; lack of an international safety convention or arrangement; lack of financial support for safety measures (e.g. safety reassessment, safety upgrading, decommissioning) and utilization; lack of clear utilization programmes; inadequate emergency preparedness; inadequate safety documentation (e.g. safety analysis report, operating rules and procedures, emergency plan); inadequate funding of shutdown reactors; weak safety culture; loss of expertise and corporate memory; loss of information concerning radioactive materials contained in retired experimental devices stored in the facility indefinitely; obsolescence of equipment and lack of spare parts; inadequate training and qualifications of regulators and operators; safety implications of new fuel types. These issues have been addressed by the IAEA Secretariat and the chairman of the International Nuclear Safety Advisory Group (INSAG). INSAG has identified three major safety issues that are: the increasing age of research reactors, the number of research reactors that are not operating anymore but have not been decommissioned, and the number of research reactors in countries that do not have appropriate regulatory authorities. This issue paper discusses the concerns generated by an analysis of the results of INSARR missions and those expressed by INSAG. The

Study of cold neutron sources: Implementation and validation of a complete computation scheme for research reactor using Monte Carlo codes TRIPOLI-4.4 and McStas

International Nuclear Information System (INIS)

Campioni, Guillaume; Mounier, Claude

2006-01-01

The main goal of the thesis about studies of cold neutrons sources (CNS) in research reactors was to create a complete set of tools to design efficiently CNS. The work raises the problem to run accurate simulations of experimental devices inside reactor reflector valid for parametric studies. On one hand, deterministic codes have reasonable computation times but introduce problems for geometrical description. On the other hand, Monte Carlo codes give the possibility to compute on precise geometry, but need computation times so important that parametric studies are impossible. To decrease this computation time, several developments were made in the Monte Carlo code TRIPOLI-4.4. An uncoupling technique is used to isolate a study zone in the complete reactor geometry. By recording boundary conditions (incoming flux), further simulations can be launched for parametric studies with a computation time reduced by a factor 60 (case of the cold neutron source of the Orphee reactor). The short response time allows to lead parametric studies using Monte Carlo code. Moreover, using biasing methods, the flux can be recorded on the surface of neutrons guides entries (low solid angle) with a further gain of running time. Finally, the implementation of a coupling module between TRIPOLI- 4.4 and the Monte Carlo code McStas for research in condensed matter field gives the possibility to obtain fluxes after transmission through neutrons guides, thus to have the neutron flux received by samples studied by scientists of condensed matter. This set of developments, involving TRIPOLI-4.4 and McStas, represent a complete computation scheme for research reactors: from nuclear core, where neutrons are created, to the exit of neutrons guides, on samples of matter. This complete calculation scheme is tested against ILL4 measurements of flux in cold neutron guides. (authors)

Producción de hidrógeno a partir del tratamiento anaerobio de vinazas en un reactor UASB

Directory of Open Access Journals (Sweden)

César González-Ugalde

2014-09-01

Bajo condiciones mesofílicas (37 °C, un pH de operación de aproximadamente 5,50, una concentración del sustrato de 20 000 mg DQO/L y un tiempo de retención hidráulica (TRH de seis horas, la producción promedio de hidrógeno obtenida en el reactor UASB fue de 1,68 mL H2/h/L, con una tasa máxima de 13,4 mL H2/h/L. El porcentaje de remoción de DQO en el proceso de fermentación alcanzó valores máximos del 43%, con un promedio cercano al 20%. Tanto la producción de hidrógeno como la remoción de DQO presentaron una dependencia inversamente proporcional al TRH. Los resultados obtenidos en este estudio demuestran que la fermentación anaerobia en un reactor UASB abre la posibilidad de utilizar las vinazas para producir hidrógeno molecular de forma sostenible.

Simulación CFD de la transferencia de calor en un reactor de hidrotratamiento de aceites vegetales de segunda generación

OpenAIRE

Mendoza Sépulveda, César Camilo

2013-01-01

Resumen: Se desarrolló un modelo CFD que permite representar la transferencia de calor en un reactor de hidrotratamiento de aceites vegetales. Este modelo permitió evaluar la transferencia de calor para distintas configuraciones del reactor. En el proceso de hidrotratamiento de aceites vegetales se transforma el aceite en un líquido con cero contenido de azufre y excelentes propiedades como combustible diesel. El proceso se basa en la adición de hidrógeno a alta presión en un reactor de lecho...

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

International Nuclear Information System (INIS)

2014-02-01

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2005

International Nuclear Information System (INIS)

2007-03-01

In the fiscal year 2005, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation : 137 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography. Irradiation for activation analyses, radioisotope (RI) productions, fission tracks. Irradiation test of reactor materials etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT). Prompt gamma-ray analyses. Sensitivity measurement of radiation detectors. Experiment in the nuclear reactor training. Practice of Reactor operation. Irradiation for activation analyses, RI productions, fission tracks etc. The volume contains 100 activity reports, which are categorized into the fields of neutron scattering (9 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2006

International Nuclear Information System (INIS)

2009-01-01

In the fiscal year 2006, the research reactor JRR-3 was operated 7 cycles (cycle operation: 26 days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation: 151 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. The volume contains 294 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-02-15

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

Characteristics of Al Alloy as a Material for Hydrolysis Reactor of NaBH{sub 4}

Energy Technology Data Exchange (ETDEWEB)

Jung, Hyeon-Seong; Oh, Sung-June; Jeong, Jae-Jin; Na, Il-Chai; Chu, Cheun-Ho; Park, Kwon-Pil [Sunchon National University, Suncheon (Korea, Republic of); Chu, Cheun-Ho [ETIS Co, Gimpo (Korea, Republic of)

2015-12-15

Aluminum alloy was examined as a material of low weight reactor for hydrolysis of NaBH{sub 4}. Aluminum is dissolved with alkali, but there is NaOH as a stabilizer in NaBH{sub 4} solution. To decrease corrosion rate of aluminum, decrease NaOH concentration and this result in loss of NaBH{sub 4} during storage of NaBH{sub 4} solution. Therefore stability of NaBH{sub 4} and corrosion of aluminum should be considered in determining the optimum NaOH concentration. NaBH{sub 4} stability and corrosion rate of aluminum were measured by hydrogen evolution rate. NaBH{sub 4} stability was tested at 20-50 .deg. C and aluminum corrosion was measured at 60-90 .deg. C. The optimum concentration of NaOH was 0.3 wt%, considering both NaBH{sub 4} stability and aluminun corrosion. NaBH{sub 4} hydrolysis reaction continued 200min in aluminum No 6061 alloy reactor with 0.3 wt% NaOH at 80-90 .deg. C.

Interactions of RuO4(g) with different surfaces in nuclear reactor containments

International Nuclear Information System (INIS)

Holm, J.; Glaenneskog, H.; Ekberg, C.

2008-07-01

During a severe nuclear reactor accident with air ingress, ruthenium in the form of RuO4 can be released from the nuclear fuel. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. This work has investigated the distribution of RuO4 between an aqueous and gaseous phase in the temperature interval of 20-50 deg. C by on-line measurements with an experimental set-up made of glass. The experiments showed that RuO4 is almost immediately distributed in the aqueous phase after its introduction in the set-up in the entire temperature interval. However, the deposition of ruthenium on the glass surfaces in the system was significant. The speciation of the ruthenium on the glass surfaces was studied by SEM-EDX and ESCA and was determined to be the expected RuO2. Experiments of interactions between gaseous ruthenium tetroxide and the metals aluminium, copper and zinc have been investigated. The metals were treated by RuO4 (g) at room temperature and analyzed with ESCA, SEM and XRD. The analyses show that the black ruthenium deposits on the metal surfaces were RuO2, i.e. the RuO4 (g) has been transformed on the metal surfaces to RuO2(s). The analyses showed also that there was a significant deposition of ruthenium tetroxide especially on the copper and zinc samples. Aluminium has a lower ability to deposit gaseous ruthenium tetroxide than the other metals. The conclusion that can be made from the results is that surfaces in nuclear reactor containments will likely reduce the source term in the case of a severe accident in a nuclear power plant. (au)

Differential expression of FAK and Pyk2 in metastatic and non-metastatic EL4 lymphoma cell lines.

Science.gov (United States)

Zhang, Zhihong; Knoepp, Stewart M; Ku, Hsun; Sansbury, Heather M; Xie, Yuhuan; Chahal, Manpreet S; Tomlinson, Stephen; Meier, Kathryn E

2011-08-01

The murine EL4 lymphoma cell line exists in variants that are either sensitive or resistant to phorbol 12-myristate 13-acetate (PMA). In sensitive cells, PMA causes Erk MAPK activation and Erk-mediated growth arrest. In resistant cells, PMA induces a low level of Erk activation, without growth arrest. A relatively unexplored aspect of the phenotypes is that resistant cells are more adherent to culture substrate than are sensitive cells. In this study, the roles of the protein tyrosine kinases FAK and Pyk2 in EL4 phenotype were examined, with a particular emphasis on the role of these proteins in metastasis. FAK is expressed only in PMA-resistant (or intermediate phenotype) EL4 cells, correlating with enhanced cell-substrate adherence, while Pyk2 is more highly expressed in non-adherent PMA-sensitive cells. PMA treatment causes modulation of mRNA for FAK (up-regulation) and Pyk2 (down-regulation) in PMA-sensitive but not PMA-resistant EL4 cells. The increase in Pyk2 mRNA is correlated with an increase in Pyk2 protein expression. The roles of FAK in cell phenotype were further explored using transfection and knockdown experiments. The results showed that FAK does not play a major role in modulating PMA-induced Erk activation in EL4 cells. However, the knockdown studies demonstrated that FAK expression is required for proliferation and migration of PMA-resistant cells. In an experimental metastasis model using syngeneic mice, only FAK-expressing (PMA-resistant) EL4 cells form liver tumors. Taken together, these studies suggest that FAK expression promotes metastasis of EL4 lymphoma cells.

Evaluation of Tehran research reactor (TRR) control rod worth using MCNP4C computer code

International Nuclear Information System (INIS)

Hosseini, Mohammad; Vosoughi, Naser; Hosseini, Seyed Abolfazl

2010-01-01

The main objective of reactor control system is to provide a safe reactor starting up, operation and shutting down. Calculation or measurement of precise values of control rod worth is of great importance in Tehran Research Reactor (TRR), considering the fact that they are the only controlling tools in the reactor. In present paper, simulation of TRR in First Operation Cycle (FOC) and in cold and clean core for the calculation of total and integral worth of control nods is reported. MCNP4C computer code has been used for all simulation process. Two method have been used for control rods worth calculation in this paper, namely the direct approach and perturbation method. It is shown that while the direct approach is appropriate for worth calculation of both the shim and the regulating control rods, the perturbation method is just suitable for tiny reactivity changes, i.e. for small initial part of regulating rods. Results of simulation are compared with the reported data in Safety Analysis Report (SAR) of Tehran research reactor and showed satisfactory agreement. (author)

Dispositivo de posicionamiento de muestras biológicas para su irradiación en un canal radial de un reactor nuclear // Biological samples positioning device for irradiations on a radial channel at the nuclear research reactor

Directory of Open Access Journals (Sweden)

Maritza Rodríguez - Gual

2010-05-01

Full Text Available ResumenPor la demanda de un dispositivo experimental para el posicionamiento de las muestras biológicaspara su irradiación en un canal radial de un reactor nuclear de investigaciones en funcionamiento, seconstruyó y se puso en marcha un dispositivo para la colocación y retirada de las muestras en laposición de irradiación de dicho canal. Se efectuaron las valoraciones económicas comparando conotro tipo de dispositivo con las mismas funciones. Este trabajo formó parte de un proyectointernacional entre Cuba y Brasil que abarcó el estudio de los daños inducidos por diferentes tipos deradiación ionizante en moléculas de ADN. La solución propuesta es comprobada experimentalmente,lo que demuestra la validez práctica del dispositivo. Como resultado del trabajo, el dispositivoexperimental para la irradiación de las muestras biológicas se encuentra instalado y funcionando yapor 5 años en el canal radial # 3(BH#3 Palabras claves: reactor nuclear de investigaciones, dispositivo para posicionamiento de muestras,___________________________________________________________________________AbstractFor the demand of an experimental device for biological samples positioning system for irradiationson a radial channel at the nuclear research reactor in operation was constructed and started up adevice for the place and remove of the biological samples from the irradiation channels withoutinterrupting the operation of the reactor. The economical valuations are effected comparing withanother type of device with the same functions. This work formed part of an international projectbetween Cuba and Brazil that undertook the study of the induced damages by various types ofionizing radiation in DNA molecules. Was experimentally tested the proposed solution, whichdemonstrates the practical validity of the device. As a result of the work, the experimental device forbiological samples irradiations are installed and operating in the radial beam hole #3(BH#3

Multipurpose research reactors

International Nuclear Information System (INIS)

1988-01-01

The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

Tratamiento biológico del lixiviado generado en el relleno sanitario "El Guayabal" de la ciudad San José de Cúcuta

OpenAIRE

Alexander Álvarez Contreras; John Hermógenes Suárez Gelvez

2006-01-01

En este trabajo se realizó un diagnóstico de calidad y cantidad del lixiviado generado en el relleno sanitario El Guayabal de la ciudad San José de Cúcuta, y se evaluaron dos sistemas de tratamiento biológico a escala laboratorio para este lixiviado. El lixiviado en el momento de la experiencia presentaba un rango de DQO de 7.650 a 28.250 mg/L. Los sistemas de tratamiento ensayados fueron: un reactor anaerobio del tipo UASB y un sistema de Biodiscos. La carga máxima asimilada p...

FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ``Alternative Teams,`` chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S&S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT`s analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option.

Actualización del sistema SCADA y de control para los reactores MQ5 y MQ6 de la planta de Pinturas Condor, Sherwin Williams Ecuador

Directory of Open Access Journals (Sweden)

Jonathan Reinoso

2013-12-01

Full Text Available El presente documento describe la actualización del sistema SCADA para los reactores MQ5 y MQ6 de la planta de Pinturas Condor mediante el software Intouch y la actualización del sistema de control del reactor MQ5 implementado en un controlador lógico programable (PLC de marca SCHNEIDER, además de la arquitectura de control realizada en el proyecto. El sistema SCADA y de control de los reactores permiten la visualización y control de los datos y variables más relevantes durante las diferentes fases de producción de resinas en los reactores MQ5 y MQ6.

Phorbol esters induce interleukin 2 mRNA in sensitive but not in resistant EL4 cells

International Nuclear Information System (INIS)

Harrison, J.R.; Lynch, K.R.; Sando, J.J.

1986-01-01

Phorbol ester (PE) sensitive EL4 cells are growth-inhibited and produce interleukin 2 (IL2) when treated with PE. Resistant EL4 cells lack both responses. To determine whether the defect in resistant cells occurs pre or post-transcriptionally, an assay for IL2 mRNA was developed using a synthetic oligonucleotide to mouse IL2 as a probe. Total RNA (15 μg) from cells +/- PE was electrophoresed, blotted onto a cationic nylon membrane, and probed with radiolabeled oligomer. This probe hybridized to a 1.1 kb band in RNA from PE-treated sensitive cells. This RNA was detectable within 3h of PE administration, was clearly visible by 6h, and peaked by 9 to 12h. No bands hybridizing with the IL2 probe were detected in RNA isolated from unstimulated cells or from resistant EL4 cells at any time following PE stimulation. Since levels of the protooncogene c-myc have been shown to decrease in a number of cell lines during differentiation and growth inhibition, total RNA from EL4 cells was probed with a nick-translated plasmid containing the protein coding region of the c-myc gene. In PE sensitive cells, levels of c-myc RNA are markedly reduced by 3h. In a pilot experiment with resistant cells, c-myc levels appeared to remain constant. These results demonstrate that PE induced IL2 mRNA in PE sensitive but not resistant EL4 cells. Sensitive and resistant EL4 cell lines provide a useful model for the investigation of the regulation of gene expression by PE

El ecobarrio: proyecto de sensibilización ambiental. El caso de Villa 4 Álamos de Maipú (Chile)

OpenAIRE

Ubeira, Felipe; Quiroga, Carolina

2001-01-01

En la presente investigación se analizaron las consecuencias que ha tenido el desarrollo del primer proyecto de eco barrio en los habitantes de la villa 4 Álamos de Maipú, en la ciudad de Santiago de Chile. El proyecto es impulsado por la organización comunitaria “Ceibo”, integrada por habitantes de la villa, quienes tras enfrentarse a varios conflictos medioambientales y problemas sociales de distinto tipo, han ido en busca de nuevas maneras de habitar la ciudad, debiendo lidiar con factores...

Research reactor utilization. Summary reports of three study group meetings: Irradiation techniques at research reactors, held in Istanbul 15-19 November 1965; Research reactor operation and maintenance problems, held in Caracas 6-10 December 1965; and Research reactor utilization in the Far East, held in Lucas Heights 28 February - 4 March 1966

International Nuclear Information System (INIS)

1967-01-01

The three sections of this book, which are summary reports of three Study Group meetings of the IAEA: Irradiation techniques at research reactors, Istanbul, 15-19 November 1965; Research reactor operation and maintenance problems, Caracas, 6-10 December 1965; and Research reactor utilization in the Far East, Lucas Heights, Australia, 28 February - 4 March 1966. These meetings were the latest in a series designed to promote efficient utilization of research reactors, to disseminate information on advances in techniques, to discuss common problems in reactor operations, and to outline some advanced areas of reactor-based research. (author)

[Experimental study of interleukin-12 gene vaccines in the treatment of low-load malignant lymphoma (EL4)].

Science.gov (United States)

Jiang, Q; Da, W; Ou, Y

2001-11-01

Two kinds of murine interleukin-12 (mIL-12) fusion gene vaccines were used to treat the murine low-load malignant T cell lymphoma EL4 as minimal residual disease (MRD) model. C57BL/6 synergistical mice were subcutaneously inoculated with 1 x 10(6) wild-type (wt) EL4 tumor cells as low-load lymphoma model treated with two mIL-12 gene vaccines. Package cell line PA317/12 producing mIL-12 retrovirus (RV) was used as in vivo vaccine and EL4 tumor cells transferred with mIL-12 gene as ex vivo vaccine. In both mIL-12 gene vaccine-treated groups, there was no tumor growth in 50% mice 60 days after inoculation. Nine of these no tumor growth mice were re-challenged with 5 x 10(5) wt EL4 cells, and 5 of them survived without tumors in another 60 days. All control mice died with tumors within one month after inoculation. Among those developed tumors in both vaccine-treated groups, the development of tumors was delayed, the survival period prolonged (P EL4 MRD in C57BL/6 mice.

The concept of the sodium cooled small fast reactor 4S and the analyses of the loss of flow events

International Nuclear Information System (INIS)

Nishi, Yoshihisa; Ueda, Nobuyuki; Koga, Tomonari; Matsumiya, Hisato

2007-01-01

CRIEPI has been developing the 4S reactor (Super Safe, Small and Simple reactor) for application in dispersed energy supply and multipurpose use, in conjunction with Toshiba Corporation. The 4S is sodium cooled fast reactor and their electrical output has two options of 10MWe and 50MWe. In this paper, 10MWe 4S (4S-10M) was proposed. 4S-10M has some unique features. It employs a burn-up control system with annular reflector in place of the control rod that requires the frequent maintenance service. The core life time of the 4S-10M is 30 years and the fuel transport is not required during core life time. All temperature feedback coefficients are negative during core life time. In the latest design for 4S-10M, a pool and tall type reactor design was selected to reduce the construction cost. Two types of decay heat removal system (Reactor Vessel Auxiliary Cooling System; RVACS, Intermediate Reactor Auxiliary Cooling System; IRACS) using natural convection power were adopted. It is necessary to confirm that these two heat removal system can operate appropriately. The transition analyses were executed by the CERES code to evaluate the design feasibility and the thermal hydraulic characteristics of the 4S-10M. CERES is a multi-dimensional plant dynamics simulation code for liquid metal reactors developed by the CRIEPI. CERES can perform simulations ranging from forced circulation (full/partial power operation) to natural circulation. Components (pumps, IHXs, SGs, pipings, etc.) of the reactor are modeled as one-dimensional. Multi-dimensional plena are connected to such components. Two loss-of-flow accident sequences are considered. In the first case, it is assumed that the primary and the secondary pump were stopped by the total station black out. The reactor shut down system was assumed to be success. This sequence is referred to as the protected loss-of-flow accident (PLOF). In the second case, it is assumed that the reactor shut down systems fail to operate and the

Study of the origin of elements of the uranium-235 family observed in excess in the vicinity of the experimental nuclear EL4 reactor under dismantling. Lessons got at this day and conclusions

International Nuclear Information System (INIS)

2007-01-01

This study resumes the discovery of an excess of actinium 227 found around by EL4 nuclear reactor actually in dismantling. The search for the origin of this excess revealed a real inquiry of investigation during three years. Because a nuclear reactor existed in this area a particular attention will have concerned this region. The doubt became the line of conduct to find the answer to the human or natural origin of this excess. Finally and against any evidence, it appears that the origin of this phenomenon was natural, consequence of the particular local geology. The detail of the different investigations is given: search of a possible correlation with the composition of elevations constituent of lanes, search (and underlining) of new sites in the surroundings of the Rusquec pond and the Plouenez station, study of the atmospheric deposits under winds of the nuclear power plant and in the east direction, search of a possible relationship with the gaseous effluents of the nuclear power plant in the past, historical study of radioactive effluents releases in the fifty last years by the analysis of the sedimentary deposits in the Saint-Herbiot reservoir, search of a possible correlation between the excess of actinium 227 and the nuclear power plant activity; search of a possible correlation with a human activity without any relationship with the nuclear activities, search of a correlation with the underground waters, search of a correlation with the geological context, collect of information on the possible transfers in direction of the food chain, determination of the radiological composition of the underground waters ( not perturbed by human activity), search of the cause of an excess of actinium 227 in the old channel of liquid effluents release of the nuclear power plant. The results are given and discussed. And contrary to all expectations the origin of the excess of actinium 227 is completely natural. (N.C.)

TNF induction of EL4 hyposensitivity to lysis by recombinant (soluble) and membrane-associated TNFs: TNF binding, internalization, and degradation.

Science.gov (United States)

Fishman, M; Costlow, M

1994-04-01

EL4 mouse thymoma cells sensitive to TNF-mediated lysis only in the presence of cycloheximide (S-EL4) or in the presence or absence of cycloheximide (N-EL4) were used in these experiments. Murine tumor cell line (S-EL4) sensitivity to TNF cytotoxicity is augmented when cycloheximide is added together with TNF or when cycloheximide is added 1 hr before or after TNF. No enhanced sensitivity is observed when target cells are incubated with cycloheximide 2-4 hr before or after the addition of TNF. In the absence of cycloheximide, S-EL4 cells preexposed to murine TNF are less susceptible to lysis by TNF and TNF receptor-conjugated TNF but are lysed by integral membrane TNF. TNF-induced hyposensitivity is partially reversed by actinomycin D or by culturing the preexposed cells for 4 hr prior to TNF lytic assay. TNF preincubation of N- and S-EL4 cells results in an immediate decrease in 125I-TNF binding due to TNF receptor occupancy. Recovery of TNF-R occupancy and TNF internalization were subsequently noted.

The use and evolution of the CEA research reactors; Utilisation et evolution des reacteurs de recherche du C.E.A

Energy Technology Data Exchange (ETDEWEB)

Rossillon, F; Chauvez, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The authors successively examine the different research reactors in use in the French C.E.A. Nuclear Centres. They trace briefly their histories, describing how they have been used up to the present, and how they have been adapted to changes in programme by means of certain modifications. They also describe the reasons which have led to the elaboration of the project for the new reactor Osiris. Zoe, the oldest reactor in the CEA, has been in service in the Centre de Fontenay-aux-Roses since 1948. It is used mainly for measurements of absorption cross-sections in graphite, and for various short irradiations which do not require high fluxes. The reactor EL 2, in service since 1952, was used for the first studies on gas cooling. It has also been widely used for the production of radioisotopes and for a large number of experiments in the fields of physics, metallurgy and physical chemistry. The ageing of certain elements of the reactor has led to the decision to close it down in the near future The reactor EL 3 has been widely used for experiments in physics and in the investigation of fuels. The possibilities of the reactor in fast neutron irradiations will be considerably improved by the adoption of a new type of core (the 'snow crystal' structure). Triton-I, a 2 MW swimming-pool reactor, is used for the most part for fast neutron and gamma irradiations. The modifications being carried out on it at present should result in an increase in the power of the reactor up to 4 or 5 MW. In a neighbouring compartment is housed Triton-II which is of the same general structure, as Triton-I, but whose maximum power is 100 kW. Triton-II is used solely for studies on shielding. Melusine, a 2 MW swimming-pool reactor, has been in use in the Centre d'Etudes Nucleaires de Grenoble since 1959. It has supported a very high programme concerned mainly with solid state physics, fundamental research into refractory fissile materials and special graphites, and the study of the behaviour of

Resveratrol (trans-3,5,4'-trihydroxystilbene) suppresses EL4 tumor growth by induction of apoptosis involving reciprocal regulation of SIRT1 and NF-κB.

Science.gov (United States)

Singh, Narendra P; Singh, Udai P; Hegde, Venkatesh L; Guan, Hongbing; Hofseth, Lorne; Nagarkatti, Mitzi; Nagarkatti, Prakash S

2011-08-01

Understanding the molecular mechanisms through which natural products and dietary supplements exhibit anticancer properties is crucial and can lead to drug discovery and chemoprevention. The current study sheds new light on the mode of action of resveratrol (RES), a plant-derived polyphenolic compound, against EL-4 lymphoma growth. Immuno-compromised NOD/SCID mice injected with EL-4 tumor cells and treated with RES (100 mg/kg body weight) showed delayed development and progression of tumor growth and increased mean survival time. RES caused apoptosis in EL4 cells through activation of aryl hydrocarbon receptor (AhR) and upregulation of Fas and FasL expression in vitro. Blocking of RES-induced apoptosis in EL4 cells by FasL mAb, cleavage of caspases and PARP, and release of cytochorme c, demonstrated the participation of both extrinsic and intrinsic pathways of apoptosis. RES also induced upregulation of silent mating type information regulation 2 homolog, 1 (SIRT1) and downregulation of nuclear factor kappa B (NF-κB) in EL4 cells. siRNA-mediated downregulation of SIRT1 in EL4 cells increased the activation of NF-κB but decreased RES-mediated apoptosis, indicating the critical role of SIRT1 in apoptosis via blocking activation of NF-κB. These data suggest that RES-induced SIRT1 upregulation promotes tumor cell apoptosis through negative regulation of NF-κB, leading to suppression of tumor growth. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Evaluación del comportamiento hidráulico en un reactor anaerobio de doble cámara (RADCA

Directory of Open Access Journals (Sweden)

Nancy Rincón

2011-01-01

tales como cortos circuitos, zonas muertas y recirculación interna afectan su desempeño. En esta investigación se evaluó el comportamiento hidráulico de un reactor anaerobio de doble cámara (RADCA de 534,5 L (cámara 1=305 L y cámara 2= 229,5 L como innovación tecnológica de los reactores UASB. El RADCA fue alimentado con agua residual municipal (ARM de la ciudad de Maracaibo, Venezuela; cada una de las cámaras fueron inoculadas con lodo granular (20% v/v proveniente de una cervecería local. La evaluación hidráulica se realizó en la fase líquida y en operación utilizando Li+ (LiCl como trazador aplicado de forma instantánea en el afluente a tiempo de retención hidráulico teórico (TRHt de 6 horas; 3,4 h en la cámara 1 y 2,6 para la cámara 2. El RADCA describió un flujo pistón en ambas cámaras y una eficiencia hidráulica cercana a la unidad (1 indicando una presencia casi nula de zonas muertas. La eficiencia de remoción de la DQO total (DQOT del RADCA se mantuvo en el rango de 59,77% a 74,64% con un promedio de 68,26%. Para las cámaras 1 y 2 la eficiencia promedio fue 60,4 y 20,94% con una producción de biogás (L/h de 2,768 y 0,541 respectivamente.

Las movilizaciones del 4 de febrero y el 6 de marzo de 2008. Una lectura de las representaciones sociales en el discurso de la prensa nacional

Directory of Open Access Journals (Sweden)

Carolina Jaramillo Correa

2010-01-01

Full Text Available Las convocatorias y la respuesta a las marchas del 4 de febrero y el 6 de marzo de 2008 se han constituido como uno de los mayores impactos en términos de movilización ciudadana en Colombia. Los medios masivos de comunicación fueron actores decisivos para la divulgación de las marchas, y de persuasión para que los ciudadanos se sumaran o no a éstas. El equipo de investigación del proyecto Representaciones de las movilizaciones sociales por la paz en la prensa colombiana: 4 de febrero y 6 de marzo de 2008 recopiló las noticias sobre las marchas publicadas en tres diarios impresos de circulación nacional: el diario El Tiempo (el mayor periódico nacional, el diario El Colombiano (que representa la región del país con el índice más alto de movilizaciones sociales y el semanario Voz (que representa los sectores políticos de izquierda y pertenece al partido comunista. Este artículo presenta la clasificación y categorización de información desde la perspectiva del análisis del discurso (ad y una aproximación a los resultados de la categorización cualitativa de la información.

The Role of Exponential and PCTR Experiments at Hanford in the Design of Large Power Reactors; Roles Respectifs des Experiences Exponentielles et du Reacteur d'Etude des Constantes Physiques de Hanford dans les Etudes de Grands Reacteurs de Puissance; Znachenie ehksponentsial'nykh opytov i opytov na reaktore PCTR pri proektirovanii bol'shikh ehnergeticheskikh reaktorov v khehnforde; Papel de los Experimentos Exponenciales y del Reactor PCTR de Hanford en el Proyecto de Grandes Reactores de Potencia

Energy Technology Data Exchange (ETDEWEB)

Heineman, R. E. [General Electric Company, Richland, WA (United States)

1964-02-15

que ces installations permettent d'utiliser, en vue de faire face aux besoins de donnees experimentales de plus en plus diverses. Il faut avoir tous ces renseignements presents a l'esprit si l'on veut prevoir comment evolueront les besoins et les tendances dans l'emploi de ces installations pour les etudes de reacteurs de puissance. Le memoire decrit brievement le Reacteur d'etude des reseaux a haute temperature et indique comment on se propose de l'utiliser dans le cadre de cette evolution. (author) [Spanish] Desde hace casi 15 anos se vienen realizando en los laboratorios de Hanford mediciones exponenciales en reticulados de grafito* uranio. Aunque los resultados de dichos experimentos se emplearon para establecer los laplacianos de reactores de produccion, contribuyeron tambien a ampliar los conocimientos sobre la fisica de estos sistemas. Muy pronto se reconocio que la utilidad del experimento exponencial quedaba limitada por sus grandes dimensiones y por su escasa sensibilidad a pequenas perturbaciones localizadas del sistema. Por ello se comenzo a idear un experimento integral en un reactor que reduciria al minimo la cantidad de materiales necesarios para obtener datos significativos. A tal efecto, se construyo una instalacion critica perfeccionada de varias regiones, que se denomino PCTR (reactor para estudio de constantes fisicas). Este reactor se ha empleado para determinar las constantes fisicas de varios reactores de potencia. Ademas, ha servido como instalacion de uso general para medir secciones eficaces y para determinar los parametros diferenciales e integrales de fisica de los reactores correspondientes a diversos tipos de medios multiplicadores. Los reactores exponenciales se emplearon despues de construir el PCTR, a pesar de que este cumplio ampliamente sus promesas. El autor proporciona diversos datos tipicos obtenidos con estas dos instalaciones y compara sus papeles respectivos para el estudio de nuevos reactores de potencia, para justificar la

[Regulatory T cells inhibit proliferation of mouse lymphoma cell line EL4 in vitro].

Science.gov (United States)

Zhang, Chen; Kong, Yan; Guo, Jun; Ying, Zhi-Tao; Yuan, Zhi-Hong; Zhang, Yun-Tao; Zheng, Wen; Song, Yu-Qin; Li, Ping-Ping; Zhu, Jun

2010-10-01

This study was aimed to investigate the effect of regulatory T (Treg) cells on the T cell lymphoma EL4 cells and its mechanism in vitro. C57BL/6 mouse Treg cells were isolated by magnetic cell sorting (MACS). The purity of Treg cells and their expression of Foxp3 were identified by flow cytometry (FCM) and PT-PCR respectively. The suppression of Treg cells on EL4 cells was detected by 3H-TdR method. At the same time, enzyme-linked immunosorbent assay (ELISA) was used to detect the secretion of cytokine TGF-β1 and IL-10. The results showed that CD4+CD25+ T cells could be successfully isolated by MACS with the purity reaching 94.52% and the expression of Foxp3 reaching 84.72%. After sorting, the expression of Foxp3 mRNA could be detected by RT-PCR. 3H-TdR assay confirmed that regulatory T cells could suppress the proliferation of EL4 cells with or without antigen presenting cells (APC) or dendritic cells (DC), APC or DC might effectively enhance the suppression. In addition, DC alone also suppressed the proliferation. TGF-β1 and IL-10 could be detected in the supernatant by ELISA. It is concluded that the Treg cells can obviously suppress the proliferation of T cell lymphoma cells in vitro, APC or DC can enhance this suppressive effect, while the DC alone also can suppress the proliferation of EL4 cells, the TGF-β1 and IL-10 cytokine pathway may be one of the mechanisms of suppression.

FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

International Nuclear Information System (INIS)

1996-09-01

Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES ampersand H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ''Alternative Teams,'' chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S ampersand S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT's analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option

Lysosomal processing of sialoglycoconjugates in a wheat germ agglutinin resistant variant of EL4 murine leukemia cells

International Nuclear Information System (INIS)

Devino, N.L.

1989-01-01

Metabolic studies were undertaken in EL4 murine leukemia in WB6, a wheat germ agglutinin-resistant variant of EL4, in order to identify any differences in lysosomal processing of sialoglyco-conjugates. Five lysosomal acid hydrolases, acetylesterase, acid phosphatase, β-galactosidase, α-mannosidase, and neuraminidase, were studied using fluorescent 4-methylumbelliferyl substrates. No significant differences were found in the total activity of any of these enzymes in EL4 and WB6. Cells were incubated in the presence of N-acetylmannosamine, the metabolic precursor of sialic acid (N-acetylneuraminic acid). Free sialic acid accumulated in the lysosomes of WB6 but not of EL4. The accumulation of lysosomal free sialic acid in WB6 showed a dependence on the concentration of N-acetylmannosamine in the growth medium. Metabolic labelling with [6- 3 H]-N-acetylmannosamine showed that WB6 accumulated lysosomal free sialic acid even at very low concentrations of N-acetylmannosamine. The two cell lines differed in their distribution of radiolabelled neutral sugars, free sialic acid, and sialoglycoproteins. The velocity of 3 H-sialic acid release was 3.7-fold lower in WB6 than in EL4, suggesting that WB6 has a defect in lysosomal sialic acid transport. The metabolic consequences of this defect are examined, in light of other biochemical and immunological data on these cells

Análisis de estabilidad del reactor PFTR para una reacción con cinética de primer orden utilizando la funcional de Lyapunov

Directory of Open Access Journals (Sweden)

Héctor Armando Durán Peralta

2007-01-01

Full Text Available Abunda la literatura referente al análisis de estabilidad de reactores con parámetros globalizados de concentración y temperatura (por ejemplo el CSTR, en cambio es escasa la literatura sobre la estabilidad de reactores con parámetros distribuidos donde existe distribución espacial de concentración y temperatura, como es el caso del reactor tubular PFTR. Este documento analiza la estabilidad del reactor PFTR isotérmico y no isotérmico para una reacción con cinética de primer orden utilizando la funcional de Lyapunov. Se trabaja con una cinética de primer orden pues un objetivo de este artículo es mostrar cómo se aplica la funcional de Lyapunov al análisis de un reactor de parámetros distribuidos, dado que es casi inexistente la literatura sobre el método de la funcional de Lyapunov aplicada a la estabilidad de reactores (técnica usada en el análisis de estabilidad de sistemas en ingeniería eléctrica. El análisis de estabilidad dio como resultado perfiles de temperatura y concentración asintóticamente estables para los casos PFTR isotérmico, no isotérmico con constante cinética independiente de la temperatura y PFTR no isotérmico adiabático. Para el PFTR con retiro de calor el análisis condujo a una región de estabilidad asintótica y a una región incierta donde puede o no haber oscilaciones.

RasGRP1 confers the phorbol ester-sensitive phenotype to EL4 lymphoma cells.

Science.gov (United States)

Han, Shujie; Knoepp, Stewart M; Hallman, Mark A; Meier, Kathryn E

2007-01-01

The murine EL4 lymphoma cell line exists in variants that are either sensitive or resistant to the tumor promoter phorbol 12-myristate 13-acetate (PMA). In sensitive EL4 cells, PMA causes robust Erk mitogen-activated protein kinase activation that results in growth arrest. In resistant cells, PMA induces minimal Erk activation, without growth arrest. PMA stimulates IL-2 production in sensitive, but not resistant, cells. The role of RasGRP1, a PMA-activated guanine nucleotide exchange factor for Ras, in EL4 phenotype was examined. Endogenous RasGRP1 protein is expressed at much higher levels in sensitive than in resistant cells. PMA-induced Ras activation is observed in sensitive cells but not in resistant cells lacking Ras-GRP1. PMA induces down-regulation of RasGRP1 protein in sensitive cells but increases RasGRP1 in resistant cells. Transfection of RasGRP1 into resistant cells enhances PMA-induced Erk activation. In the reverse experiment, introduction of small interfering RNA (siRNA) for RasGRP1 suppresses PMA-induced Ras and Erk activations in sensitive cells. Sensitive cells incubated with siRNA for RasGRP1 exhibit the PMA-resistant phenotype, in that they are able to proliferate in the presence of PMA and do not secrete IL-2 when stimulated with PMA. These studies indicate that the PMA-sensitive phenotype, as previously defined for the EL4 cell line, is conferred by endogenous expression of RasGRP1 protein.

Construction of Research Reactors for Gen 3 and Gen 4 Reactors Development

International Nuclear Information System (INIS)

Behar, Christophe

2014-01-01

Christophe Behar, Director of the Nuclear Energy Division at CEA, detailed the different kind of research reactors and the issues in term of investment, use, side application such as the medical isotopes production

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2007

International Nuclear Information System (INIS)

2012-03-01

In the fiscal year 2007, the research reactor JRR-3 was operated for 7 cycles (cycle operation : 26days/cycle) and the JRR-4 was operated for 92 days. JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment and practice in the nuclear reactor training, Irradiation for activation analyses, RI productions, fission tracks, etc. The volume contains 262 activity reports, which are categorized into the fields of neutron scattering (10 subcategories), neutron radiography, neutron activation analyses, prompt gamma-ray analyses, and others submitted by the users in JAEA and other Organizations. (author)

Canadian supercritical water reactor modeling using G4STORK

International Nuclear Information System (INIS)

Ford, W.; Buijs, A.

2015-01-01

The Canadian Supercritical Water Reactor design was simulated using G4STORK. The results showed the expected trends but the determined Keff of 1.253±0.001 with a Coolant Void Reactivity (CVR) of -25mk differed greatly from the results achieved using MCNP of Keff=1.2914 and a CVR of -14mk. This discrepancy is partly due to the different data libraries used and the mixing of different temperature libraries in MCNP, but is also likely due to a difference in the physics methodology. Work is ongoing to further clarify reasons for discrepancies and improve the efficiency of the simulation. (author)

Canadian supercritical water reactor modeling using G4STORK

Energy Technology Data Exchange (ETDEWEB)

Ford, W.; Buijs, A. [McMaster University, Hamilton, ON (Canada)

2015-07-01

The Canadian Supercritical Water Reactor design was simulated using G4STORK. The results showed the expected trends but the determined Keff of 1.253±0.001 with a Coolant Void Reactivity (CVR) of -25mk differed greatly from the results achieved using MCNP of Keff=1.2914 and a CVR of -14mk. This discrepancy is partly due to the different data libraries used and the mixing of different temperature libraries in MCNP, but is also likely due to a difference in the physics methodology. Work is ongoing to further clarify reasons for discrepancies and improve the efficiency of the simulation. (author)

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

International Nuclear Information System (INIS)

Vaibhaw, Kumar; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

2008-01-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (∼300 deg. C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation (F n ) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process

NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

International Nuclear Information System (INIS)

Van Nieuwenhove, R.

2012-01-01

A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

Science.gov (United States)

Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

2008-12-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

Energy Technology Data Exchange (ETDEWEB)

Van Nieuwenhove, R. (Institutt for Energiteknikk, OECD Halden Reactor Project (Norway))

2012-01-15

A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

Operating reactors licensing actions summary. Volume 4, No. 9

International Nuclear Information System (INIS)

1984-11-01

This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the division of licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Operating reactors licensing actions summary. Vol. 4, No. 2

International Nuclear Information System (INIS)

1984-04-01

This summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Neutronic and thermo-hydraulic design of LEU core for Japan Research Reactor 4

International Nuclear Information System (INIS)

Arigane, Kenji; Watanabe, Shukichi; Tsuruta, Harumichi

1988-04-01

As a part of the Reduced Enrichment Research and Test Reactor (RERTR) program in JAERI, the enrichment reduction for Japan Research Reactor 4 (JRR-4) is in progress. A fuel element using a 19.75 % enriched UAlx-Al dispersion type with a uranium density of 2.2 g/cm 3 was designed as the LEU fuel and the neutronic and thermo-hydraulic performances of the LEU core were compared with those of the current HEU core. The results of the neutronic design are as follows: (1) the excess reactivity of the LEU core becomes about 1 % Δk/k less, (2) the thermal neutron flux in the fuel region decreases about 25 % on the average, (3) the thermal neutron fluxes in the irradiation pipes are almost the same and (4) the core burnup lifetime becomes about 20 % longer. The thermo-hydraulic design also shows that: (1) the fuel plate surface temperature decreases about 10 deg C due to the increase of the number of fuel plates and (2) the temperature margin with respect to the ONB temperature increases. Therefore, it is confirmed that the same utilization performance as the HEU core is attainable with the LEU core. (author)

The Chernobyl-4 Reactor and the possible causes of the accident

International Nuclear Information System (INIS)

Motte, F.

1986-01-01

A description and information about the Chernobyl nuclear reactor is given. Some comparison elements between the RBMK reactor type and GCR, CANDU, SGHWR and Hanford N reactor types are presented. A scenario of the possible causes of the accident is discussed. (A.F.)

Pretreatment of low dose radiation reduces radiation-induced apoptosis in mouse lymphoma (EL4) cells.

Science.gov (United States)

Kim, J H; Hyun, S J; Yoon, M Y; Ji, Y H; Cho, C K; Yoo, S Y

1997-06-01

Induction of an adaptive response to ionizing radiation in mouse lymphoma (EL4) cells was studied by using cell survival fraction and apoptotic nucleosomal DNA fragmentation as biological end points. Cells in early log phase were pre-exposed to low dose of gamma-rays (0.01 Gy) 4 or 20 hrs prior to high dose gamma-ray (4, 8 and 12 Gy for cell survival fraction analysis; 8 Gy for DNA fragmentation analysis) irradiation. Then cell survival fractions and the extent of DNA fragmentation were measured. Significant adaptive response, increase in cell survival fraction and decrease in the extent of DNA fragmentation were induced when low and high dose gamma-ray irradiation time interval was 4 hr. Addition of protein or RNA synthesis inhibitor, cycloheximide or 5,6-dichloro-1-beta-d-ribofuranosylbenzimidazole (DRFB), respectively during adaptation period, the period from low dose gamma-ray irradiation to high dose gamma-ray irradiation, was able to inhibit the induction of adaptive response, which is the reduction of the extent DNA fragmentation in irradiated EL4 cells. These data suggest that the induction of adaptive response to ionizing radiation in EL4 cells required both protein and RNA synthesis.

Specific anti-EL4-lymphoma immunity in mice cured 2 years earlier with doxorubicin and interleukin-2.

Science.gov (United States)

Ehrke, M J; Verstovsek, S; Zaleskis, G; Ho, R L; Ujházy, P; Maccubbin, D L; Mihich, E

1996-05-01

This laboratory has reported the conditions for an effective, non-toxic, chemoimmunotherapy utilizing doxorubicin in combination with prolonged administration of interleukin-2 and the identification of the critical role of activated CD8+ T cells in the therapeutic effect. Mice (C57BL/6) cured in those studies have been followed for the remainder of their life spans. These mice, approximately 2 months of age when initially inoculated with syngeneic EL4 lymphoma, survived for more than 2 years, the normal life span of C57BL/6 mice. Mice 4 months old reinoculated with the EL4 cells all survived. At about 1 year of age mice were sacrificed and the ability of their thymocytes and splenocytes to develop specific CD8+ anti-EL4 activity was as high as it had been at the time of tumor rejection. At about 2 years of age EL4 was reimplanted into mice; all of them survived. These surviving mice, at 2 years 2 months of age, as well as a group of 2-year-old mice not rechallenged, were killed and functional antitumor activity and phenotype characteristics of various lymphocyte populations were determined in comparison to those of young and age-matched control mice. The phenotyping of the lymphocytes from the cured mice indicated very notable differences in subset distribution and increased CD44 expression. Functionally they developed high levels of anti-EL4 activity, which was ablated by combined treatment with monoclonal antibodies against CD8 and CD44, indicating the role of memory cells. Consistent with cells from aged mice, these same cell populations had a very reduced allogeneic responsiveness. It appears that cured mice have developed an immune memory specific for EL4.

Present status of decommissioning in the Musashi Reactor Facility (4)

International Nuclear Information System (INIS)

Uchiyama, Takafumi; Tanzawa, Tomio; Mitsuhashi, Ishi; Morishima, Kayoko; Matsumoto, Tetsuo

2012-01-01

The decommissioning of the Musashi reactor was decided in 2003. Permanent shutdown of the reactor and stopping the operational functions were conducted in 2004. Transportation of the spent fuels was finished in 2006. After 2007, the system and equipment stopping the functions were stored as installed in the reactor facility as radioactive wastes. After separating nonradioactive wastes such as concretes from radioactive wastes with a contamination test, stopping the functions of liquid waste management facility was performed with newly installed drainage facility for radioisotope use in 2010. Solid waste management facility was also dismantled and removed in the same way as liquid waste management facility in 2011. Radioactive wastes packed in containers were moved and stored in the reactor facility. (T. Tanaka)

Calculations of fuel burn up and radionuclide inventories in the Syrian miniature neutron source reactor using the WIMSD4 and CITATION codes

International Nuclear Information System (INIS)

Khattab, K.

2005-01-01

The WIMSD4 code is used to generate the fuel group constants and the infinite multiplication factor as a function of the reactor operating time for 10, 20, and 30 k W operating power levels. The uranium burn up rate and burn up percentage, the amounts of the plutonium isotopes, the concentrations and radioactivities of the fission products and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core are calculated using the WIMSD4 code as well. The CITATION code is used to calculate the changes in the effective multiplication factor of the reactor.(author)

IAEA safety standards for research reactors

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

A simulation of a pebble bed reactor core by the MCNP-4C computer code

Directory of Open Access Journals (Sweden)

Bakhshayesh Moshkbar Khalil

2009-01-01

Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

Nuclear reactor technology progress report, vol. 4

International Nuclear Information System (INIS)

1981-01-01

The works of the Engineering Section, Fast Experimental Reactor Division, are roughly classified into the technologies concerning the reactor core, abnormality monitoring, the plant, purity control and operation planning. In this paper, the activities of the Engineering Section, the operational results of Joyo and the foreign informations on FBRs in this quarter are reported. The second regular inspection carried out successively from the previous quarter was completed, and the fourth cycle operation of Joyo at 75 MW was started. The measurement of CP around the primary system pipings and equipments, the preliminary test of a core flow meter for Monju, and the various characteristic tests were carried out during this period. 2 N reports, 1 SA report and 63 memos were drawn up in this quarter. The test plan to be carried out during the period of the fourth to sixth cycle operations in this last year using the MK-1 core was formed and decided. Various meetings within and outside the division are reported. The data obtained in the operational characteristic test and special test are shown. As the results concerning the reactor technologies, the development of dosimetry techniques, the measurement and analysis of the core characteristics, the measurement of the temperature and flow velocity of coolant at the fuel assembly exit, the system pressure loss in the primary cooling system and others are reported. (Kako, I.)

Proceedings of the 4. CSNI workshop on the chemistry of iodine in reactor safety

Energy Technology Data Exchange (ETDEWEB)

Guentay, S [ed.; Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1996-12-01

The 4. OECD workshop on the chemistry of iodine in reactor safety was held in Wuerenlingen, Switzerland from June 10th to 12th, 1996. It was organised in collaboration with the Laboratory for Safety and Accident Research of the Paul Scherrer Institute. About seventy experts from fourteen OECD member countries attended the meeting, as well as experts from Latvia and the Commission of the European Communities. Thirty-four papers were presented in five sessions on various aspects of national and international programmes, integral and intermediate-scale experiments, experimental homogeneous phase chemistry, surface processes, thermodynamic and kinetic studies and safety applications. Throughout the meeting, emphasis was placed on detailed and open discussions. The purpose of the workshop was to exchange information on the iodine chemistry and other important fission products relevant to reactor safety, to discuss the status of the open issues identified during the previous workshop held in 1991, to define reactor safety issues and to discuss developments and future plans. (author) figs., tabs., refs.

Proceedings of the 4. CSNI workshop on the chemistry of iodine in reactor safety

International Nuclear Information System (INIS)

Guentay, S.

1996-12-01

The 4. OECD workshop on the chemistry of iodine in reactor safety was held in Wuerenlingen, Switzerland from June 10th to 12th, 1996. It was organised in collaboration with the Laboratory for Safety and Accident Research of the Paul Scherrer Institute. About seventy experts from fourteen OECD member countries attended the meeting, as well as experts from Latvia and the Commission of the European Communities. Thirty-four papers were presented in five sessions on various aspects of national and international programmes, integral and intermediate-scale experiments, experimental homogeneous phase chemistry, surface processes, thermodynamic and kinetic studies and safety applications. Throughout the meeting, emphasis was placed on detailed and open discussions. The purpose of the workshop was to exchange information on the iodine chemistry and other important fission products relevant to reactor safety, to discuss the status of the open issues identified during the previous workshop held in 1991, to define reactor safety issues and to discuss developments and future plans. (author) figs., tabs., refs

TRATAMIENTO DE AGUAS RESIDUALES DE UNA INDUSTRIA PROCESADORA DE PESCADO EN REACTORES ANAERÓBICOS DISCONTINUOS

Directory of Open Access Journals (Sweden)

Julio César Marín Leal

2015-01-01

Full Text Available En el presente trabajo se evaluó el tratamiento de las aguas residuales de una industria procesadorade pescado de la ciudad de Manta (Ecuador, en reactores anaeróbicos discontinuos, y se estableciósu adecuación a las normas ambientales vigentes en materia de vertido. Para ello, se realizaronensayos de laboratorio en reactores discontinuos de 1 L, con un tiempo de contacto de 24 h yprovistos de un lodo anaerobio procedente de una planta de tratamiento de aguas residualesdomésticas. Dicho efluente fue diluido con agua destilada en proporciones de 33%, 66% y 100%,correspondientes a las etapas I, II y III, respectivamente. Durante cada etapa se monitorearon los siguientes parámetros, de acuerdo con los métodos estándares: pH, alcalinidad total, DBO5.20,DQO, nitrito, amonio, nitrógeno total Kjeldahl (NTK, ortofostato, sulfato, sólidos suspendidostotales (SST y sólidos suspendidos volátiles (SSV. Los resultados muestran porcentajes deremoción de materia orgánica, expresados como DBO5.20(37.9±4.1%; 41.8±7.6% y 46.2±3.2% yDQO (34.7±9.7%; 36.9±9.2% y 43.8±4.1%, para las etapas I, II y III, respectivamente, relativamentebajos como resultado del origen del inóculo usado, así como del contenido relativo de sales en elefluente industrial. Las remociones de amonio, NTK y ortofosfato estuvieron entre 60-95%, 25-37% y 6-25%, respectivamente. Bajo las condiciones de los ensayos realizados, el efluente tratadorequiriere de la aplicación de un postratamiento para reducir el contenido de materia orgánica ynutrientes a los límites permisibles de descarga establecidos en la República de Ecuador.

The accident of Chernobylsk-4 reactor and its consequences

International Nuclear Information System (INIS)

1986-01-01

This report deals with the particulars of the accident as communicated by the Soviet delegation at an IAEA meeting by the and of August 1986. It was stated that the consequences emanated from the inherent instability of the design of the reactor, the deviation from the safety rules by the operators and the lack of a sight reactor containment. (G.B.)

Reaction Cross Section Calculations in Neutron Induced Reactions and GEANT4 Simulation of Hadronic Interactions for the Reactor Moderator Material BeO

Directory of Open Access Journals (Sweden)

Veli ÇAPALI

2016-05-01

Full Text Available BeO is one of the most common moderator material for neutron moderation; due to its high density, neutron capture cross section and physical-chemical properties that provides usage at elevated temperatures. As it’s known, for various applications in the field of reactor design and neutron capture, reaction cross–section data are required. The cross–sections of (n,α, (n,2n, (n,t, (n,EL and (n,TOT reactions for 9Be and 16O nuclei have been calculated by using TALYS 1.6 Two Component Exciton model and EMPIRE 3.2 Exciton model in this study. Hadronic interactions of low energetic neutrons and generated isotopes–particles have been investigated for a situation in which BeO was used as a neutron moderator by using GEANT4, which is a powerful simulation software. In addition, energy deposition along BeO material has been obtained. Results from performed calculations were compared with the experimental nuclear reaction data exist in EXFOR.

Magnetic susceptibility as a direct measure of oxidation state in LiFePO4 batteries and cyclic water gas shift reactors.

Science.gov (United States)

Kadyk, Thomas; Eikerling, Michael

2015-08-14

The possibility of correlating the magnetic susceptibility to the oxidation state of the porous active mass in a chemical or electrochemical reactor was analyzed. The magnetic permeability was calculated using a hierarchical model of the reactor. This model was applied to two practical examples: LiFePO4 batteries, in which the oxidation state corresponds with the state-of-charge, and cyclic water gas shift reactors, in which the oxidation state corresponds to the depletion of the catalyst. In LiFePO4 batteries phase separation of the lithiated and delithiated phases in the LiFePO4 particles in the positive electrode gives rise to a hysteresis effect, i.e. the magnetic permeability depends on the history of the electrode. During fast charge or discharge, non-uniform lithium distributionin the electrode decreases the hysteresis effect. However, the overall sensitivity of the magnetic response to the state-of-charge lies in the range of 0.03%, which makes practical measurement challenging. In cyclic water gas shift reactors, the sensitivity is 4 orders of magnitude higher and without phase separation, no hysteresis occurs. This shows that the method is suitable for such reactors, in which large changes of the magnetic permeability of the active material occurs.

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2007. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and tandem accelerator

International Nuclear Information System (INIS)

Miyazaki, Osamu; Awa, Yasuaki; Isaka, Koji; Kutsukake, Kenichi; Komeda, Masao; Shibata, Ko; Hiyama, Kazuhisa; Suzuki, Mayu; Sone, Takuya; Ohuchi, Tomoaki; Terakado, Yuichi; Sataka, Masao

2009-06-01

The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor-3), JRR-4(Japan Research Reactor-4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2007 and March 31, 2008. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator. (2) Utilization of research reactors and tandem accelerator. (3) Upgrading of utilization techniques of research reactors and tandem accelerator. (4) Safety administration for research reactors and tandem accelerator. (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, commendation, plans and outcomes in service and technical developments and so on. (author)

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2010. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and Tandem Accelerator

International Nuclear Information System (INIS)

Ishii, Tetsuro; Nakamura, Kiyoshi; Kawamata, Satoshi; Yamada, Yusuke; Kawashima, Kazuhiro; Asozu, Takuhiro; Nakamura, Takemi; Arai, Masaji; Yoshinari, Shuji; Sataka, Masao

2012-03-01

The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2010 and March 31, 2011. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator, (2) Utilization of research reactors and tandem accelerator, (3) Upgrading of utilization techniques of research reactors and tandem accelerator, (4) Safety administration for research reactors and tandem accelerator, (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, commendation, outcomes in service and technical developments and so on. (author)

Generalities about nuclear reactors

International Nuclear Information System (INIS)

Jaouen, C.; Beroux, P.

2012-01-01

From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

REMOVAL OF ORGANIC MATTER AND TOXICITY IN AN UPFLOW IMMOBILIZED BIOMASS ANAEROBIC REACTOR TREATING HOSPITAL WASTEWATER: PRELIMINARY EVALUATION

Directory of Open Access Journals (Sweden)

MÓNICA PORRAS TORRES

2013-01-01

Full Text Available El objetivo de esta investigación consistió en evaluar el desempeño de un reactor anaerobio de flujo ascendente de biomasa inmovilizada (RAFABI tratando un efluente hospitalario real. Se estudió la remoción de materia orgánica y toxicidad, por medio de análisis como UV254, DQOfiltrada y determinación del porcentaje de inhibición en el crecimiento de la raíz de la cebolla. Los resultados mostraron que el proceso biológico estuvo estable durante los 287 días de operación continua, el valor medio de la relación AI/AP fue de 1.21±0.08, indicando que no hubo acumulación de ácidos en el sistema. Sin embargo, los valores de la eficiencia de remoción de DQOfiltrada, 56±15% y UV254, 21±36%, no fueron representativos. La toxicidad se redujo en 50%. Con base en lo anterior, es necesario utilizar el reactor anaerobio en combinación con otros procesos como por ejemplo los procesos de oxidación avanzada, para continuar reduciendo la materia orgánica recalcitrante al proceso anaerobio. Se comprobó la capacidad que tienen los reactores anaerobios de biomasa inmovilizada para remover la toxicidad.

Collective occupational dose for nuclear reactors of the 2., 3. and 4. generation

International Nuclear Information System (INIS)

Guidez, J.; Saturnin, A.

2016-01-01

In France during reactor operation the individual occupational doses are collected and recorded according to the law. When you sum up all the individual doses you get the yearly collective dose expressed in Man.Sv/year. This piece of information can be used to make comparisons between various types of reactors and between reactors of the same type. The results show a steady decrease of the collective dose for all types of reactors over the time except for CANDU reactors for which a slight increase of the dose has appeared since the years 1996-1998. The decrease is due to the continuous improvement of reactor operating and to changes in the reactor design. There is also a constant gap over time between the collective dose for a BWR reactor (1.12 Man.Sv/y) and a PWR reactor 0.60 Man.Sv/y), this gap is certainly due to N 16 nuclide that is created in the primary circuit and transported to turbines in the case of a BWR reactor. For sodium-cooled fast reactors (RNR-Na) the collective dose is below 0.40 Man.Sv/y except for the BN-600 reactor. (A.C.)

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

Phosphorylation of paxillin via the ERK mitogen-activated protein kinase cascade in EL4 thymoma cells.

Science.gov (United States)

Ku, H; Meier, K E

2000-04-14

Intracellular signals can regulate cell adhesion via several mechanisms in a process referred to as "inside-out" signaling. In phorbol ester-sensitive EL4 thymoma cells, phorbol-12-myristate 13-acetate (PMA) induces activation of extracellular signal-regulated kinase (ERK) mitogen-activated protein kinases and promotes cell adhesion. In this study, clonal EL4 cell lines with varying abilities to activate ERKs in response to PMA were used to examine signaling events occurring downstream of ERK activation. Paxillin, a multifunctional docking protein involved in cell adhesion, was phosphorylated on serine/threonine residues in response to PMA treatment. This response was correlated with the extent and time course of ERK activation. PMA-induced phosphorylation of paxillin was inhibited by compounds that block the ERK activation pathway in EL4 cells, primary murine thymocytes, and primary murine splenocytes. Paxillin was phosphorylated in vitro by purified active ERK2. Two-dimensional electrophoresis revealed that PMA treatment generated a complex pattern of phosphorylated paxillin species in intact cells, some of which were generated by ERK-mediated phosphorylation in vitro. An ERK pathway inhibitor interfered with PMA-induced adhesion of sensitive EL4 cells to substrate. These findings describe a novel inside-out signaling pathway by which the ERK cascade may regulate events involved in adhesion.

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2011. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and tandem accelerator

International Nuclear Information System (INIS)

Ishii, Tetsuro; Nakamura, Kiyoshi; Kawamata, Satoshi; Ishikuro, Yasuhiro; Kawashima, Kazuhito; Kabumoto, Hiroshi; Nakamura, Takemi; Tamura, Itaru; Kawasaki, Sayuri; Sataka, Masao

2013-03-01

The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2011 and March 31, 2012. The activities were categorized into six service/development fields: (1) Recovery from the Great East Japan Earthquake, (2) Operation and maintenance of research reactors and tandem accelerator, (3) Utilization of research reactors and tandem accelerator, (4) Upgrading of utilization techniques of research reactors and tandem accelerator, (5) Safety administration for research reactors and tandem accelerator, (6) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, number of staff members dispatched to Fukushima for the technical assistance, commendation, outcomes in service and technical developments and so on. (author)

State of development of gas cooled reactors in the Union of Soviet Socialist Republics

International Nuclear Information System (INIS)

Grebennik, V.N.; Mosevitskij, I.S.

1991-01-01

In the context of the programme for the development of gas-cooled reactors in the USSR it is reported that pilot plants with VGR-50 MW(el) and VG-400 MW(el) have been developed up to the stage of engineering design and that now the efforts are concentrated on the project of pilot-commercial reactor plant VGM (PCRP VGM) of a modular type with unit thermal power of 200-250 MW. The installation is designed to solve the main scientific and engineering problems of construction of high-temperature gas-cooled reactors, to test equipment components, and to show advantages of the given type of installations having the enhanced safety and capability to generate high-potential heat. The status of work on the PCRP VGM project is described. 3 refs, 1 fig., 1 tab

An initial assessment of the Chernobyl-4 reactor accident release source

International Nuclear Information System (INIS)

Macdonald, H.F.; ApSimon, H.M.; Wilson, J.J.N.

1986-07-01

The long-range atmospheric dispersion model MESOS has been used to provide a preliminary evaluation of the effects over Western Europe of radioactivity released during the accident which occurred at the Chernobyl-4 reactor in the USSR in April 1986. The results of this analysis have been compared with observations during the first week or so following the accident of airborne contamination levels at a range of locations across Europe in order to obtain an estimate of accident release source. The work presented here was performed during the 6-8 weeks following the accident and the results obtained will be subject to refinement as more detailed data become available. However, at this early stage they indicate a release source for the Chernobyl accident, expressed as a fraction of the estimated reactor core inventory, of approx. 15-20% of the iodine and caesium isotopes, approx. 1% of the ruthenium and lesser amounts of the other fission products and actinides, together with an implied major fraction of the krypton and xenon noble gases. (author)

Examination policy concerning the additional installation of No. 3 and No. 4 reactors in Takahama Nuclear Power Station and No. 3 and No. 4 reactors in Fukushima No. 2 Nuclear Power Station

International Nuclear Information System (INIS)

1980-01-01

The Nuclear Safety Commission decided the annual examination policy on the modification of reactor installation in Takahama Nuclear Power Station to construct No. 3 and No. 4 reactors inquired under date of November 26, 1979, by the Minister of International Trade and Industry, so that the examination results of the accident in Three Mile Island nuclear power station are reflected to the examination for the purpose of improving reactor safety. The examination results of the accident in Three Mile Island power station are being investigated by the Committee on Examination of Reactor Safety, based on the policy shown in ''On the second report of the special committee examining the accident in a nuclear power station in the U.S.'' determined by the Nuclear Safety Commission under date of September 13, 1979. Though the Committee will further clarify the past guideline about the items concerning the criteria, design and operation management, the Committee decided the tentative policy to reflect it to safety examination. Further, a table is attached, in which 52 items to be reflected to the security measures are classified from the viewpoint of necessity to reflect them to the final examination. This table includes 13 items of criteria and examination, 7 items related to design, 10 items related to operation management, 10 antidisaster items, and 12 items related to safety research. (Wakatsuki, Y.)

UCLA program in reactor studies: The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1991-01-01

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

Four energy group neutron flux distribution in the Syrian miniature neutron source reactor using the WIMSD4 and CITATION code

International Nuclear Information System (INIS)

Khattab, K.; Omar, H.; Ghazi, N.

2009-01-01

A 3-D (R, θ , Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the point wise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation site with relative difference less than 7% and 5% respectively. (author)

Estudio de la distribución de los tiempos de residencia en un reactor tubular para la hidrólisis de lecitina de soja con fosfolipasa A2 inmovilizada

Directory of Open Access Journals (Sweden)

Zaritzky, N.

2001-10-01

efectuada mediante el uso de enzima fosfolipasa A2 inmovilizada, liberando un ácido graso de la posición C-2 de los fosfolípidos para obtener un producto enriquecido en lisolecitinas. La reacción enzimática sigue una cinética de primer orden cuando las concentraciones de sustrato están dentro del rango: 6,34 10-3 y 19,0 10-3M. El valor de la constante de velocidad es: k= 9,88 10-2 min-1 cuando la enzima está inmovilizada sobre alúmina. Se construyó un reactor que permite la circulación del fluido a través del soporte. El soporte seleccionado fue alúmina en consideración a sus buenas propiedades mecánicas y a su bajo costo. Fue analizado el comportamiento del flujo en el reactor, y cuanto este se aparta del modelo ideal de flujo en pistón, inyectando una solución de 1 % NaCl (trazador en forma de inyección por impulso. La medición de la conductividad de la solución efluente resultó adecuada para la determinación de los tiempos de residencia. El sistema mostró comportamiento lineal. Se analizaron los tiempos de residencia en el reactor utilizando tres diferentes volúmenes de flujo para diferentes arreglos de soporte y material inerte. Se calcularon las fracciones no convertidas en el reactor y se observaron las diferencias a la salida en comparación a las de un reactor de flujo en pistón, precisamente porque se generan canalizaciones y cortocircuitos en la columna. La conversión máxima resultó para las más altas concentraciones de sustrato y para el menor flujo de alimentación. El módulo de dispersión resultó bastante mayor que el límite que introduce una curva gaussiana para el caso en el cual el grado de suposición de alta dispersión fue corregido. El reactor alcanzó un comportamiento similar al de un reactor de mezcla completa y se concluyó que son importantes el grado de retromezcla, la formación de remolinos y zonas de redistribución de material.

Análisis de estabilidad del reactor PFTR para una reacción con cinética de primer orden utilizando la funcional de Lyapunov

OpenAIRE

Héctor Armando Durán Peralta; Luis Fernando Córdoba C

2007-01-01

Abunda la literatura referente al análisis de estabilidad de reactores con parámetros globalizados de concentración y temperatura (por ejemplo el CSTR), en cambio es escasa la literatura sobre la estabilidad de reactores con parámetros distribuidos donde existe distribución espacial de concentración y temperatura, como es el caso del reactor tubular PFTR. Este documento analiza la estabilidad del reactor PFTR isotérmico y no isotérmico para una reacción con cinética de primer orden utilizando...

Análisis de estabilidad del reactor pftr para una reacción con cinética de primer orden utilizando la funcional de lyapunov

OpenAIRE

Durán Peralta, Héctor Armando; Córdoba C, Luis Fernando

2010-01-01

Abunda la literatura referente al análisis de estabilidad de reactores con parámetros globalizados de concentración y temperatura (por ejemplo el CSTR), en cambio es escasa la literatura sobre la estabilidad de reactores con parámetros distribuidos donde existe distribución espacial de concentración y temperatura, como es el caso del reactor tubular PFTR. Este documento analiza la estabilidad del reactor PFTR isotérmico y no isotérmico para una reacción con cinética de primer orden utilizand...

Homogeneity of Continuum Model of an Unsteady State Fixed Bed Reactor for Lean CH4 Oxidation

Directory of Open Access Journals (Sweden)

Subagjo

2014-07-01

Full Text Available In this study, the homogeneity of the continuum model of a fixed bed reactor operated in steady state and unsteady state systems for lean CH4 oxidation is investigated. The steady-state fixed bed reactor system was operated under once-through direction, while the unsteady-state fixed bed reactor system was operated under flow reversal. The governing equations consisting of mass and energy balances were solved using the FlexPDE software package, version 6. The model selection is indispensable for an effective calculation since the simulation of a reverse flow reactor is time-consuming. The homogeneous and heterogeneous models for steady state operation gave similar conversions and temperature profiles, with a deviation of 0.12 to 0.14%. For reverse flow operation, the deviations of the continuum models of thepseudo-homogeneous and heterogeneous models were in the range of 25-65%. It is suggested that pseudo-homogeneous models can be applied to steady state systems, whereas heterogeneous models have to be applied to unsteady state systems.

Special Nuclear Material Control by the Power Reactor Operator; Controle des Matieres Nucleaires Speciales par l'Exploitant d'une Centrale Nucleaire; Spetsial'nyj kontrol' nalichiya yadernykh materialov operatorom ehnergeticheskogo reaktora; Control de Materiales Nucleares Especiales por Parte de Quienes Operan el Reactor de Potencia

Energy Technology Data Exchange (ETDEWEB)

Cordin, R. A. [Yankee Atomic Electric Company, Boston, MA (United States)

1966-02-15

matieres nucleaires ne se limite pas S de simples travaux d'inventaire mais sert de base a beaucoup d'autres activites qui font partie integrante du programme d'operations de tout reacteur, par exemple les expeditions de combustible irradie, le traitement chimique du combustible epuise et la comptabilite du combustible recupere et des matieres produites au cours du fonctionnement du reacteur, et l'institution et l'application d'un regime d'assurance satisfaisant. (author) [Spanish] Combustible relativamente nuevo y sumamente valioso para la produccion de energia electrica, el uranio requiere un control muy minucioso desde el momento en que la direccion de una central asume la responsabilidad financiera inherente a su posesion hasta que como combustible parcialmente agotado se transfiere a otra instalacion en la que se recupera la parte que no se ha consumido. Antes de que se descubriera la posibilidad de emplear la energia nuclear para producir electricidad, la mayor parte de las empresas que actualmente explotan centrales nucleares explotaban centrales alimentadas con combustibles fosiles y hablan establecido sistemas de control relativamente completos y adecuados para los combustibles de ese tipo. Los responsables de las centrales nucleoelectricas deben disponer de sistemas no menos adecuados para controlar los materiales nucleares especiales que utilizan. La explotacion de los reactores de potencia no es una ciencia antigua, pero durante el tiempo relativamente corto que ha transcurrido desde que se inicio su empleo los ingenieros y hombres de ciencia han mejorado continuamente el diseflo del equipo y los metodos de trabajo con objeto de disminuir los costos de produccion y de lograr que las centrales nucleares puedan competir en el plano economico con las centrales clasicas. La administracion de los materiales nucleares debe efectuarse con metodos modernos y eficientes a fin de que los adelantos tecnologicos que han permitido reducir los costos no resulten inutiles

Evaluation of the OSCAR-4/MCNP calculation methodology for radioisotope production in the SAFARI-1 reactor

International Nuclear Information System (INIS)

Karriem, Z.; Zamonsky, O.M.

2014-01-01

The South African Nuclear Energy Corporation SOC Ltd (Necsa) is a state owned nuclear facility which owns and operates SAFARI-1, a 20 MW material testing reactor. SAFARI-1 is a multi-purpose reactor and is used for the production of radioisotopes through in-core sample irradiation. The Radiation and Reactor Theory (RRT) Section of Necsa supports SAFARI-1 operations with nuclear engineering analyses which include core-reload design, core-follow and radiation transport analyses. The primary computer codes that are used for the analyses are the OSCAR-4 nodal diffusion core simulator and the Monte Carlo transport code MCNP. RRT has developed a calculation methodology based on OSCAR-4 and MCNP to simulate the diverse in-core irradiation conditions in SAFARI-1, for the purpose of radioisotope production. In this paper we present the OSCAR-4/MCNP calculation methodology and the software tools that were developed for rapid and reliable construction of MCNP analysis models. The paper will present the application and accuracy of the methodology for the production of yttrium-90 ( 90 Y) and will include comparisons between calculation results and experimental measurements. The paper will also present sensitivity analyses that were performed to determine the effects of control rod bank position, representation of core depletion state and sample loading configuration, on the calculated 90 Y sample activity. (author)

Estudio de distribución de tiempos de residencia en un reactor biológico de lecho empacado cerámico

Directory of Open Access Journals (Sweden)

Tatiana Rodríguez Chaparro

2004-01-01

Full Text Available La distribución de tiempos de residencia de un reactor es una característica del mezclado que ocurre dentro de él [ ] 1 [ ] 2 ; su determinación es básica para el diseño de cualquier tipo de reactor en escala real. El objetivo de esta investigación consistió en determinar la distribución de tiempos de residencia en un reactor biológico de lecho empacado cerámico (anillos a partir de pruebas con trazadores. Los resultados obtenidos utilizando las técnicas de inyección por paso y pulso fueron 34.577 seg., y 17.745 seg., respectivamente, y la dispersión calculada infinita. Lo anterior permite concluir que en reactores de lecho empacado cerámico (anillos las moléculas del trazador se distribuyen uniformemente en todo el sistema. Los ensayos se realizaron en un modelo a escala laboratorio.

The Application of Non-Metallic Core Materials in a High-Temperature Reactor Experiment; Utilisation de materes non metalliques dans le coeur d'un reacteur experimental a haute temperature; Ispol'zovanie nemetallicheskikh materialov dlya aktivnoj zony vysokotemperaturnogo opytnogo reaktora; Empleo de materiales no metalicos en el nucleo de un reactor experimental de alta temperatura

Energy Technology Data Exchange (ETDEWEB)

Huddle, R. A.U.; Shepherd, L. R. [Organization for Economic Co-Operation and Development, Dragon Project, Atomic Energy Establishment, Winfrith, Dorset (United Kingdom)

1963-11-15

haute temperature refroidis par un gaz, qui soient rentables pour la production d'electricite. (author) [Spanish] El proyecto DRAGON de la O.C.F.E., relativo a un reactor de alta temperatura, se emprendio con el fin de perfeccionar la tecnologia de esa categoria de reactores refrigerados por gas y, como parte de esta labor, con objeto de construir y explotar un reactor experimental de 20 MW (t). El reactor, cuya construccion esta por terminarse, es del tipo refrigerado por helio, con una temperatura de salida del nucleo de 750{sup o}C; como combustible se utiliza {sup 235}U y torio como materiaj fertil. Una particularidad de este reactor es la ausencia de materiales metalicos en su nucleo. En razon de la elevada temperatura de funcionamiento, a saber, hasta 1050{sup o}C en la superficie de los elementos combustibles y superior a 1500{sup o}C en las regiones mas calientes de los mismos se utilizan meteriales refractarios no metalicos. Todos los materiales del nucleo se hallan incorporados en los elementos combustibles, lo que permite alcanzar una amplia superficie de transmision de calor en relacion con el volumen del nucleo y, por tanto, una potencia especifica media elevada que hace posible reducir las dimensiones del sistema. Cada elemento combustible esta formado-por un haz-de tubos de grafito que contienen los materiales fisionables y fertiles en forma de carburos incorporados en pastillas de grafito. Una corriente purgadora de helio refrigerante que atraviesa la zona-central de cada barra de combustible es extraida por la base, desde donde pasa a una instalacion de tratamiento para eliminar los productos de fision y otras impurezas antes de retornar el helio al reactor. Por este procedimiento se reduce el escape de productos de fision del combustible ceramico, cuya temperatura es muy elevada, al circuito primario de refrigeracion. Se exponen los problemas inherentes al estudio y produccion de combustibles ceramicos y de grafito para este reactor, asi como al

The short term effects of Low-dose-rate Radiation on EL4 Lymphoma Cell

International Nuclear Information System (INIS)

Bong, Jin Jong; Kang, Yu Mi; Shin, Suk Chull; Choi, Moo Hyun; Choi, Seung Jin; Kim, Hee Sun; Lee, Kyung Mi

2012-01-01

To determine the biological effects of low-dose-rate radiation ( 137 Cs, 2.95 mGy/h) on EL4 lymphoma cells during 24 h, we investigated the expression of genes related to apoptosis, cell cycle arrest, DNA repair, iron transport, and ribonucleotide reductase. EL4 cells were continuously exposed to low-dose-rate radiation (total dose: 70.8 mGy) for 24 h. We analyzed cell proliferation and apoptosis by trypan blue exclusion and flow cytometry, gene expression by real-time PCR, and protein levels with the apoptosis ELISA kit. Apoptosis increased in the Low-dose-rate irradiated cells, but cell number did not differ between non- (Non-IR) and Low-dose-rate irradiated (LDR-IR) cells. In concordance with apoptotic rate, the transcriptional activity of ATM, p53, p21, and Parp was upregulated in the LDR-IR cells. Similarly, Phospho-p53 (Ser15), cleaved caspase 3 (Asp175), and cleaved Parp (Asp214) expression was upregulated in the LDR-IR cells. No difference was observed in the mRNA expression of DNA repair-related genes (Msh2, Msh3, Wrn, Lig4, Neil3, ERCC8, and ERCC6) between Non-IR and LDR-IR cells. Interestingly, the mRNA of Trfc was upregulated in the LDR-IR cells. Therefore, we suggest that short-term Low-dose-rate radiation activates apoptosis in EL4 lymphoma cells.

The short term effects of Low-dose-rate Radiation on EL4 Lymphoma Cell

Energy Technology Data Exchange (ETDEWEB)

Bong, Jin Jong; Kang, Yu Mi; Shin, Suk Chull; Choi, Moo Hyun; Choi, Seung Jin; Kim, Hee Sun [Radiation Health Research Institute, Korea Hydro and Nuclear Power Co., Ltd, Seoul (Korea, Republic of); Lee, Kyung Mi [Global Research Lab, BAERI Institute, Dept. of Biochemistry and Molecular Biology, Korea University College of Medicine, Seoul (Korea, Republic of)

2012-06-15

To determine the biological effects of low-dose-rate radiation ({sup 137}Cs, 2.95 mGy/h) on EL4 lymphoma cells during 24 h, we investigated the expression of genes related to apoptosis, cell cycle arrest, DNA repair, iron transport, and ribonucleotide reductase. EL4 cells were continuously exposed to low-dose-rate radiation (total dose: 70.8 mGy) for 24 h. We analyzed cell proliferation and apoptosis by trypan blue exclusion and flow cytometry, gene expression by real-time PCR, and protein levels with the apoptosis ELISA kit. Apoptosis increased in the Low-dose-rate irradiated cells, but cell number did not differ between non- (Non-IR) and Low-dose-rate irradiated (LDR-IR) cells. In concordance with apoptotic rate, the transcriptional activity of ATM, p53, p21, and Parp was upregulated in the LDR-IR cells. Similarly, Phospho-p53 (Ser15), cleaved caspase 3 (Asp175), and cleaved Parp (Asp214) expression was upregulated in the LDR-IR cells. No difference was observed in the mRNA expression of DNA repair-related genes (Msh2, Msh3, Wrn, Lig4, Neil3, ERCC8, and ERCC6) between Non-IR and LDR-IR cells. Interestingly, the mRNA of Trfc was upregulated in the LDR-IR cells. Therefore, we suggest that short-term Low-dose-rate radiation activates apoptosis in EL4 lymphoma cells.

Effect of the plutonium isotopic composition on the performance of fast reactors; Effet de la composition isotopique du plutonium sur le rendement de reacteurs a neutrons rapides; Vliyanie izotopnogo sostava plutoniya na rabotu reaktorov na bystrykh nejtronakh; Efectos de la composicion isotopica del plutonio sobre el funcionamiento de los reactores rapidos

Energy Technology Data Exchange (ETDEWEB)

Yiftah, S [Israel Atomic Energy Commission (Israel)

1962-03-15

constate egalement que dans le plutonium utilise comme combustible sous forme de metal, d'oxyde ou de carbure, les variations de la reactivite, apres enlevement de 40% du sodium initialement present dans le coeur, sont plus negatives (ou moins positives) si le plutonium a une plus forte teneur en isotopes superieurs. (author) [Spanish] La composicion isotopica del plutonio empleado como combustible en los reactores de neutrones rapidos depende de su procedencia. En principio, existen tres fuentes posibles, a saber: a) los reactores generadores de plutonio; b) los reactores de potencia termicos (alimentados con uranio natural o enriquecido); c) las envolturas fertiles de los reactores de neutrones rapidos. La fuente a), y hasta cierto punto la fuente c), proporcionan plutonio-239 relativamente puro, mientras que el plutonio de la fuente b) es rico en {sup 240}Pu, {sup 241}Pu y {sup 242}Pu. La cantidad de impurezas depende, del tipo de reactor empleado, del grado de combustion y, en general, de las condiciones de irradiacion del combustible. Cabe preguntarse, entonces, si es posible emplear cualquier clase de plutonio como combustible en los reactores de neutrones rapidos. Con el proposito de estudiar los efectos que ejerce sobre el funcionamiento de los reactores la composicion isotopica del combustible de plutonio, en forma de metal, oxido o carburo, se realizo una serie de calculos de geometria segun una teoria de difusion de 16 grupos, utilizando el conjunto de secciones eficaces de 16 grupos recientemente establecido por Yiftah, Okrent y Moldauer; esos calculos se aplicaron a plutonio de tres composiciones isotopicas diferentes, empezando por {sup 239}Pu puro y aumentando la concentracion de los isotopos de numero masico mas elevado. Se han estudiado sistemas con cuerpos de 800, 1500 y 2500 litros de volumen, que son los tipicos de los reactores generadores rapidos de grandes dimensiones. Se llega a la conclusion de que, si solo se tienen en cuenta los isotopos

Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

Energy Technology Data Exchange (ETDEWEB)

El-Messeiry, A M [National Center for Nuclear Safety and Radiation Control, Atomic Energy Authority, Cairo (Egypt)

1996-03-01

In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs.

Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

International Nuclear Information System (INIS)

El-Messeiry, A.M.

1996-01-01

In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs

Calculation of the effective delayed neutron fraction by TRIPOLI-4 code for IPEN/MB-01 research reactor

International Nuclear Information System (INIS)

Lee, Y.K.; Hugot, F.X.

2011-01-01

The effective delayed neutron fraction βeff is an important reactor physics parameter. Its calculation within the multi-group deterministic transport code can be performed with the aid of adjoint flux weighted integrations. However, in continuous energy Monte Carlo transport code, the adjoint weighted βeff calculation becomes complicated due to the backward treatment of the anisotropy scattering. In TRIPOLI-4 continuous energy Monte Carlo code, the βeff calculation was performed by a two-run method, one run with delayed neutrons and second with only the contribution from prompt fission neutrons. To improve the uncertainty of the βeff two-run calculation for the experimental reactors, two simple and fast one-run methods to estimate the βeff in the continuous energy simulation have been implemented into the TRIPOLI-4 code. First approach is an improved one of the Bretscher's prompt method and second one based on the proposal of Nauchi and Kameyama. In these one-run methods, the prompt and the delayed neutrons are first tagged. Their tracking and statistics are separated performed. The new βeff calculations have been optimized in the power iteration cycles so as to estimate the production of prompt and delayed neutrons from the prompt and delayed neutrons of previous generation. To validate the new βeff calculation by TRIPOLI-4, several benchmarks including fast and thermal systems have been considered. In this paper the recent measurements of βeff in the research reactor IPEN/MB-01 have been benchmarked. The basic components of the βeff and the Keff have been also calculated so as to understand the influences of the cross sections and the delayed neutron yields on the reactor reactivity calculations. Three nuclear data libraries, ENDF/BVI.r4, ENDF/B-VII.0, and JEFF-3.1 were taken into account in this study. (author)

The effects of tributyltin oxide and deoxynivalenol on the transcriptome of the mouse thymoma cell line EL-4

NARCIS (Netherlands)

Schmeits, P.J.M.; Kol, S.; Loveren, van H.; Peijnenburg, A.A.C.M.; Hendriksen, P.J.M.

2014-01-01

The main goal of this study was to assess the potential of the mouse thymoma EL-4 cell line in screening for chemical induced immunotoxicity. Therefore, EL-4 cells were exposed to two well-known immunotoxicants, organotin compound tributyltin oxide (TBTO, 0.5 and 1 µM for 3 or 6 h) and the mycotoxin

El deseo del bien o del bien aparente en ética a nicómaco iii.4

OpenAIRE

Gualdrón, Miguel

2010-01-01

La noción de “fin”, y el deseo de este fin, son cruciales dentro de la teoría aristotélica de la acción, en tanto sustentan no sólo la posibilidad misma de actuar, sino también la evaluación ética de la acción. El deseo de un fin en la acción es identificado en varios lugares de la obra de Aristóteles con el deseo del bien. En el capítulo 4 del libro III de la Ética a Nicómaco, Aristóteles plantea un problema en lo que concierne a este deseo, según el cual el deseo puede ser del bien en sí o ...

Generation IV reactors: economics

International Nuclear Information System (INIS)

Dupraz, B.; Bertel, E.

2003-01-01

The operating nuclear reactors were built over a short period: no more than 10 years and today their average age rounds 18 years. EDF (French electricity company) plans to renew its reactor park over a far longer period : 30 years from 2020 to 2050. According to EDF this objective implies 3 constraints: 1) a service life of 50 to 60 years for a significant part of the present operating reactors, 2) to be ready to built a generation 3+ unit in 2020 which infers the third constraint: 3) to launch the construction of an EPR (European pressurized reactor) prototype as soon as possible in order to have it operating in 2010. In this scheme, generation 4 reactor will benefit the feedback experience of generation 3 and will take over in 2030. Economic analysis is an important tool that has been used by the generation 4 international forum to select the likely future reactor systems. This analysis is based on 4 independent criteria: the basic construction cost, the construction time, the operation and maintenance costs and the fuel cycle cost. This analysis leads to the evaluation of the global cost of electricity generation and of the total investment required for each of the reactor system. The former defines the economic competitiveness in a de-regulated energy market while the latter is linked to the financial risk taken by the investor. It appears, within the limits of the assumptions and models used, that generation 4 reactors will be characterized by a better competitiveness and an equivalent financial risk when compared with the previous generation. (A.C.)

Control Rods in high-Flux Swimming-Pool Reactors; Les Barres de Controle dans les Piles Piscines a Haut Flux; Reguliruyushchie sterzhni dlya reaktorov bassejnovogo tipa s vysokoj plotnost'yu nejtronnogo potoka; Las Barras de Control en los Reactores Tipo Piscina de Flujo Elevado

Energy Technology Data Exchange (ETDEWEB)

Ageroni, P.; Blum, P.; Denielou, G.; Denis, P.; Meunier, C. [Centre d' Etudes Nucleaires de Grenoble (France)

1964-06-15

etudes en cours sur les barres de controle des piles piscine a coeur ouvert fonctionnant dans la bande de 10 a 30 MW. (author) [Spanish] La memoria examina los problemas planteados por las barras de control en los reactores de investigacion de tipo piscina abierta, de alta potencia especifica y elevado flujo, basandose en calculos y experimentos efectuados durante la construccion del reactor SILOE. Expone asimismo la experiencia adquirida con las barras de control mientras el reactor funcionaba a 13 MW. Examina sucesivamente: a) Los balances de reactividad y los valores de esta para los diversos tipos de barras de control que se han ensayado (cadmio, B4C, tierras raras, y combinaciones de estas sustancias); b) los picos de flujo que la presencia de barras de control crea en el cuerpo del reactor, su influencia sobre la potencia especifica, los flujos rapidos que se pueden obtener y los medios para incrementarlos; c) los problemas tecnologicos planteados por la construccion de las barras; d) los problemas de refrigeracion, vibracion, deformacion y tiempo necesario para introducirlos en el reactor. Para terminar, describe someramente loe estudios que se estan realizando con las barras de control de reactores de tipo piscina de cuerpo abierto cuando funcionan en el intervalo de potencias comprendido entre 10 y 30 MW. (author) [Russian] 1. V svete raschetov i jeksperimentov, provedennyh pri postrojke reaktora STLOE , rassmatrivajutsja problemy, voznikajushhie v svjazi s regulirujushhimi sterzhnjami dlja issledovatel'skih reaktorov otkrytogo bassejnovogo tipa s bol'shoj udel'noj moshhnost'ju i s vysokoj plotnost'ju nejtronnogo potoka. Privodjatsja takzhe rezul'taty ispytanija jetogo reaktora pri .moshhnosti 13 mgvt v svjazi s razlichnymi regulirujushhimi sterzhnjami. 2. Posledov atel'no rassmatrivajutsja sledujushhie problemy: a) balans reaktivnosti i reaktivnaja sposobnost' podvergnutyh ispytanijam regulirujushhih sterzhnej razlichnyh tipov (kadmij, B{sub 4}C , redkie zemli

Las tecnologías de organización de las TICS para el desarrollo (ICT4D), más allá del éxito y el fracaso : el caso de Infodesarrollo en Ecuador

OpenAIRE

Jiménez Becerra, Javier Andrés

2012-01-01

La Red Ecuatoriana de Información y Comunicación para el Desarrollo (Infodesarrollo), desde su inicio, fue considerada como una de las experiencias más exitosas de América Latina en el campo de las tecnologías de información y comunicación (TICS)1 para el desarrollo, así como un actor fundamental de las políticas públicas, en este tema, en Ecuador por 4 años (2005 -2008). Sin embargo, al finalizar el 2009 había desparecido del escenario de las políticas públicas en Ecuador y su continuidad...

Molten salt reactor concept

International Nuclear Information System (INIS)

Sood, D.D.

1980-01-01

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

CONSTRUCCIÓN DE UN REACTOR DISCONTINUO PARA LA OBTENCIÓN DE BIODIESEL A PARTIR DEL ACEITE DE Ricinus communis

Directory of Open Access Journals (Sweden)

Yolimar Fernández

2014-01-01

Full Text Available Se construyó un reactor discontinuo para obtener biodiesel a partir de 5 litros de extracto obtenido de la semilla de Ricinus communis. El reactor es de acero inoxidable, con longitud de 29 cm; diámetro interno de 15,24 cm y fondo cónico de 20cm de largo, espesor de la pared de 0,2cm, resistencia tubular de 1000 W y motor de 110 volt. Se extrajo y se comparó con las normas respectivas las propiedades físicas y químicas del aceite crudo. Se realizaron pruebas preliminares de transesterificación del aceite catalizadas con NaOH para constatar la viabilidad de la reacción y definir las condiciones operacionales. El biodiesel obtenido fue caracterizado y comparado con referencias presentes en la literatura. Los resultaron mostraron que es posible obtener el biocombustible en el reactor discontinuo con un grado de conversión 88%; confirmando su aplicación en reacciones de transesterificación en medio básico.

Application of Box-Wilson experimental design method for 2,4-dinitrotoluene treatment in a sequential anaerobic migrating blanket reactor (AMBR)/aerobic completely stirred tank reactor (CSTR) system

International Nuclear Information System (INIS)

Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

2011-01-01

A sequential aerobic completely stirred tank reactor (CSTR) following the anaerobic migrating blanket reactor (AMBR) was used to treat a synthetic wastewater containing 2,4-dinitrotoluene (2,4-DNT). A Box-Wilson statistical experiment design was used to determine the effects of 2,4-DNT and the hydraulic retention times (HRTs) on 2,4-DNT and COD removal efficiencies in the AMBR reactor. The 2,4-DNT concentrations in the feed (0-280 mg/L) and the HRT (0.5-10 days) were considered as the independent variables while the 2,4-DNT and chemical oxygen demand (COD) removal efficiencies, total and methane gas productions, methane gas percentage, pH, total volatile fatty acid (TVFA) and total volatile fatty acid/bicarbonate alkalinity (TVFA/Bic.Alk.) ratio were considered as the objective functions in the Box-Wilson statistical experiment design in the AMBR. The predicted data for the parameters given above were determined from the response functions by regression analysis of the experimental data and exhibited excellent agreement with the experimental results. The optimum HRT which gave the maximum COD (97.00%) and 2,4-DNT removal (99.90%) efficiencies was between 5 and 10 days at influent 2,4-DNT concentrations 1-280 mg/L in the AMBR. The aerobic CSTR was used for removals of residual COD remaining from the AMBR, and for metabolites of 2,4-DNT. The maximum COD removal efficiency was 99% at an HRT of 1.89 days at a 2,4-DNT concentration of 239 mg/L in the aerobic CSTR. It was found that 280 mg/L 2,4-DNT transformed to 2,4-diaminotoluene (2,4-DAT) via 2-amino-4-nitrotoluene (2-A-4-NT) and 4-amino-2-nitrotoluene (4-A-2-NT) in the AMBR. The maximum 2,4-DAT removal was 82% at an HRT of 8.61 days in the aerobic CSTR. The maximum total COD and 2,4-DNT removal efficiencies were 99.00% and 99.99%, respectively, at an influent 2,4-DNT concentration of 239 mg/L and at 1.89 days of HRT in the sequential AMBR/CSTR.

Application of Box-Wilson experimental design method for 2,4-dinitrotoluene treatment in a sequential anaerobic migrating blanket reactor (AMBR)/aerobic completely stirred tank reactor (CSTR) system

Energy Technology Data Exchange (ETDEWEB)

Kuscu, Ozlem Selcuk, E-mail: oselcuk@mmf.sdu.edu.tr [Department of Environmental Engineering, Engineering and Architecture Faculty, Sueleyman Demirel University, Cuenuer Campus, 32260 Isparta (Turkey); Sponza, Delia Teresa [Dokuz Eyluel University, Engineering Faculty, Environmental Engineering Department, Buca Kaynaklar campus, Izmir (Turkey)

2011-03-15

A sequential aerobic completely stirred tank reactor (CSTR) following the anaerobic migrating blanket reactor (AMBR) was used to treat a synthetic wastewater containing 2,4-dinitrotoluene (2,4-DNT). A Box-Wilson statistical experiment design was used to determine the effects of 2,4-DNT and the hydraulic retention times (HRTs) on 2,4-DNT and COD removal efficiencies in the AMBR reactor. The 2,4-DNT concentrations in the feed (0-280 mg/L) and the HRT (0.5-10 days) were considered as the independent variables while the 2,4-DNT and chemical oxygen demand (COD) removal efficiencies, total and methane gas productions, methane gas percentage, pH, total volatile fatty acid (TVFA) and total volatile fatty acid/bicarbonate alkalinity (TVFA/Bic.Alk.) ratio were considered as the objective functions in the Box-Wilson statistical experiment design in the AMBR. The predicted data for the parameters given above were determined from the response functions by regression analysis of the experimental data and exhibited excellent agreement with the experimental results. The optimum HRT which gave the maximum COD (97.00%) and 2,4-DNT removal (99.90%) efficiencies was between 5 and 10 days at influent 2,4-DNT concentrations 1-280 mg/L in the AMBR. The aerobic CSTR was used for removals of residual COD remaining from the AMBR, and for metabolites of 2,4-DNT. The maximum COD removal efficiency was 99% at an HRT of 1.89 days at a 2,4-DNT concentration of 239 mg/L in the aerobic CSTR. It was found that 280 mg/L 2,4-DNT transformed to 2,4-diaminotoluene (2,4-DAT) via 2-amino-4-nitrotoluene (2-A-4-NT) and 4-amino-2-nitrotoluene (4-A-2-NT) in the AMBR. The maximum 2,4-DAT removal was 82% at an HRT of 8.61 days in the aerobic CSTR. The maximum total COD and 2,4-DNT removal efficiencies were 99.00% and 99.99%, respectively, at an influent 2,4-DNT concentration of 239 mg/L and at 1.89 days of HRT in the sequential AMBR/CSTR.

Application of Box-Wilson experimental design method for 2,4-dinitrotoluene treatment in a sequential anaerobic migrating blanket reactor (AMBR)/aerobic completely stirred tank reactor (CSTR) system.

Science.gov (United States)

Kuşçu, Özlem Selçuk; Sponza, Delia Teresa

2011-03-15

A sequential aerobic completely stirred tank reactor (CSTR) following the anaerobic migrating blanket reactor (AMBR) was used to treat a synthetic wastewater containing 2,4-dinitrotoluene (2,4-DNT). A Box-Wilson statistical experiment design was used to determine the effects of 2,4-DNT and the hydraulic retention times (HRTs) on 2,4-DNT and COD removal efficiencies in the AMBR reactor. The 2,4-DNT concentrations in the feed (0-280 mg/L) and the HRT (0.5-10 days) were considered as the independent variables while the 2,4-DNT and chemical oxygen demand (COD) removal efficiencies, total and methane gas productions, methane gas percentage, pH, total volatile fatty acid (TVFA) and total volatile fatty acid/bicarbonate alkalinity (TVFA/Bic.Alk.) ratio were considered as the objective functions in the Box-Wilson statistical experiment design in the AMBR. The predicted data for the parameters given above were determined from the response functions by regression analysis of the experimental data and exhibited excellent agreement with the experimental results. The optimum HRT which gave the maximum COD (97.00%) and 2,4-DNT removal (99.90%) efficiencies was between 5 and 10 days at influent 2,4-DNT concentrations 1-280 mg/L in the AMBR. The aerobic CSTR was used for removals of residual COD remaining from the AMBR, and for metabolites of 2,4-DNT. The maximum COD removal efficiency was 99% at an HRT of 1.89 days at a 2,4-DNT concentration of 239 mg/L in the aerobic CSTR. It was found that 280 mg/L 2,4-DNT transformed to 2,4-diaminotoluene (2,4-DAT) via 2-amino-4-nitrotoluene (2-A-4-NT) and 4-amino-2-nitrotoluene (4-A-2-NT) in the AMBR. The maximum 2,4-DAT removal was 82% at an HRT of 8.61 days in the aerobic CSTR. The maximum total COD and 2,4-DNT removal efficiencies were 99.00% and 99.99%, respectively, at an influent 2,4-DNT concentration of 239 mg/L and at 1.89 days of HRT in the sequential AMBR/CSTR. Copyright © 2011 Elsevier B.V. All rights reserved.

Study of a fuel assembly for the nuclear reactor of IV generation cooled with supercritical water; Estudio de un ensamble de combustible para el reactor nuclear de generacion IV enfriado con agua supercritica

Energy Technology Data Exchange (ETDEWEB)

Barragan M, A.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Espinosa P, G., E-mail: albrm29@yahoo.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (MX)

2011-11-15

In this work a neutron study is presented about a square assembly design of double line of fuel rods, with moderator box to the center of the arrangement, for a nuclear reactor cooled with supercritical water (SCWR). The SCWR reactor was chosen by the characteristics of its design, mainly because is based in light water reactors (PWR and BWR), and the operational experience that has of them allow to use models and similar programs to simulate the fuel and the nucleus of this type of reactors. To develop the necessary models and to carry out the design and analysis of the SCWR reactor, the neutron codes MCNPX and Helios were used. The reason of using both codes, is because the code MCNPX used thoroughly in the neutron simulation of these reactors, it has been our reference code to analyze the results obtained with the Helios code which results are more efficient because its calculation times are minors. In the nucleus design the same parameters for both codes were considered. The results show that the design with Helios is a viable option to simulate these reactors since their values of the neutrons multiplication factor are very similar to those obtained with MCNPX. On the other hand, it could be corroborated that the CASMO-4 code is inadequate to simulate the fuel to the temperature conditions and water pressure in the SCWR. (Author)

Technology assessment HTR. Part 4. Power upscaling of High Temperature Reactors

International Nuclear Information System (INIS)

Van Heek, A.I.

1996-06-01

Designs of nuclear reactors can be classified in evolutionary, revolutionary and innovative designs. An innovative design is the High Temperature Reactor (HTR). Introduction of innovative reactors has not been successful until now. Globally, three requirements for this reactors for successful market introduction can be identified: (1) Societal support for nuclear energy, or if separable, for this reactor type, should be repaired; (2) After market introduction the innovative plant must be able to operate economically competitive; and (3) The costs of market introduction of an innovative reactor design must be limited. Until now all reactor designs classified as innovative have not yet been realized. High temperature reactors exist in many different designs. Common features are: helium coolant, graphite moderator and coated particle fuel. The combination of these creates the potential to fulfill the first requirement (public support), and similarly a hurdle to the second requirement (economical operation). All three problems existing in the eyes of the public are addressed, while a high degree of transparency is reached, making the design understandable also by others than nuclear experts. A consequence of designing according to the social support requirement is a limitation of the unit power level. The usual method to make nuclear power plants economically competitive, i.e. just raising the power level (economy of scale) could not be applied anymore. Therefore other means of cost decreasing had to be used: modularization and simplification. These ideas are explained. Since all existing HTRs are currently out of operation, additional experience from two small HTRs under construction at this moment in the Far East will be essential. In the history of HTR designs, an evolutionary path can be identified. The early designs had a philosophy of safety and economics very similar to those of LWR. Modularization was introduced to attain economic viability and the design was

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

1999-04-01

The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

[Establishment of EL4 tumor-bearing mouse models and investigation on immunological mechanisms of anti-tumor effect of melphalan].

Science.gov (United States)

Li, Mo-lin; Li, Chuan-gang; Shu, Xiao-hong; Jia, Yu-jie; Qin, Zhi-hai

2006-03-01

To establish mouse lymphoma EL4 tumor-bearing mouse models in wild type C57BL/6 mice and nude C57BL/6 mice respectively, and to further investigate the immunological mechanisms of anti-tumor effect of melphalan. Mouse lymphoma EL4 cells were inoculated subcutaneously into wild type C57BL/6 mice (immune-competent mice). Twelve days later, melphalan of different doses were administered intraperitoneally to treat these wild type C57BL/6 tuomr-bearing mice. Tumor sizes were observed and recorded subsequently to find out the minimal dose of melphalan that could cure the tuomr-bearing mice. Then the same amount of EL4 tumor cells were inoculated subcutaneously into wild type C57BL/6 mice and nude C57BL/6 mice (T cell-deficient mice) simultaneously, which had the same genetic background of C57BL/6. Twelve days later, melphalan of the minimal dose was given intraperitoneally to treat both the wild type and nude C57BL/6 tuomr-bearing mice. Tumor sizes were observed and recorded in these two different types of mice subsequently. A single dose of melphalan (7.5 mg/kg) could cure EL4 tumor-bearing wild type C57BL/6 mice, but could not induce tumor regression in EL4 tumor-bearing nude C57BL/6 mice. A single dose of melphalan has obvious anti-tumor effect on mouse lymphoma EL4 tumor-bearing wild type C57BL/6mice, which requires the involvement of T lymphocytes in the host probably related to their killing functions.

Comportamiento eléctrico del compuesto Bi5FeTi3O15 y de sus soluciones sólidas con CaBi4Ti4O15

Directory of Open Access Journals (Sweden)

Durán, P.

1999-12-01

Full Text Available Bi5FeTi3O15 (BiFT compound has been prepared by solid state reaction between the corresponding oxides. Its crystalline structure has been established by X ray Diffraction, (XRD. Ceramic samples with apparent density > 95% Dth have been sintered. On these samples, electrical conductivity and Curie temperature have been measured. Solid solutions of Bi5FeTi3O15 (BiFT and CaBi4Ti4O15 (CBiT have been prepared. On poled samples of these solid solutions, piezoelectric parameters have been established. The BiFT compound shows electrical conductivity values very similar to those of the Bi4Ti3O12 (BiT compound. The electrical conductivity of solid solutions is a function of CBiT amount. A possible electrical conductivity mechanism which is different of that accepted for the BiT compound is discussed.Se ha preparado Bi5FeTi3O15 (BiFT por reacción en estado sólido de los óxidos correspondientes. Se ha determinado su estructura cristalina por Difracción de Rayos X (DRX. Se han preparado compactos sinterizados con densidades superiores al 95%. Se ha determinado su temperatura de Curie, y la conductividad eléctrica entre 150 y 850ºC. Se han preparado soluciones sólidas de Bi5FeTi3O15 con CaBi4Ti4O15, (CBiT y se han determinado los mismos parámetros de temperatura de Curie y de conductividad para ellas. En las soluciones sólidas se han determinado los parámetros Piezoeléctricos de muestras polarizadas Debe destacarse que el compuesto Bi5FeTi3O15 presenta unos valores de conducción eléctrica más próximos a los correspondientes al Bi4Ti3O12 (BiT que a los de los compuestos MeBi4Ti4O15. La conductividad eléctrica de las soluciones sólidas varía con el contenido de CBiT. Se discute la posible existencia de un modelo de conducción eléctrica que difiere del aceptado hasta el momento para el BiT, basado en los defectos localizados en las capas Bi2O2 2-.

Safeguarding research reactors

International Nuclear Information System (INIS)

Powers, J.A.

1983-03-01

The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

. El nivel de los Factores que afectan el calendario de vacunación en niños menores de 4 años del centro de salud Año Nuevo-2016

OpenAIRE

La Rosa Asencios, Maura America

2017-01-01

El presente trabajo de investigación tuvo como objetivo determinar el nivel de los factores que afectan el calendario de vacunación de los niños menores de 4 años en el centro de salud Año Nuevo en el año 2016 el método empleado es netamente descriptivo simple diseño no experimental es cuantitativo la muestra estuvo constituida por 54 madres que acuden al servicio de vacunación el instrumento utilizado es un formulario tipo cuestionario que consta de 30 preguntas las 10 prim...

Problems of creating fuel elements for fast gas-cooled reactors working on N2O4-dissociating coolant

International Nuclear Information System (INIS)

Nesterenko, V.B.; Zelensky, V.F.; Kolykhan, L.I.; Karpenko, G.V.; Krasnorutsky, V.S.; Isakov, V.P.; Ashikhmin, V.P.; Permyakov, L.N.

1985-01-01

A variant of fast gas-cooled reactors is one using dissociating N 2 O 4 nitrogen tetroxide as a coolant. This type of reactors is promising because of great thermal effects of dissociation reactions while heating and recombination while cooling; small latent heat of evaporation; high heat transfer coefficient owing to additional heat transfer in a chemical reaction; high N 2 O 4 density in a gas state at operation parameters. The mentioned advantages give possibility to create a small turbine, heat exchange apparatus and to get high heat production in the active zone. All this opens new ways to increase power plants effectiveness

Effects of phorbol ester on mitogen-activated protein kinase kinase activity in wild-type and phorbol ester-resistant EL4 thymoma cells.

Science.gov (United States)

Gause, K C; Homma, M K; Licciardi, K A; Seger, R; Ahn, N G; Peterson, M J; Krebs, E G; Meier, K E

1993-08-05

Phorbol ester-sensitive and -resistant EL4 thymoma cell lines differ in their ability to activate mitogen-activated protein kinase (MAPK) in response to phorbol ester. Treatment of wild-type EL4 cells with phorbol ester results in the rapid activations of MAPK and pp90rsk kinase, a substrate for MAPK, while neither kinase is activated in response to phorbol ester in variant EL4 cells. This study examines the activation of MAPK kinase (MAPKK), an activator of MAPK, in wild-type and variant EL4 cells. Phosphorylation of a 40-kDa substrate, identified as MAPK, was observed following in vitro phosphorylation reactions using cytosolic extracts or Mono Q column fractions prepared from phorbol ester-treated wild-type EL4 cells. MAPKK activity coeluted with a portion of the inactive MAPK upon Mono Q anion-exchange chromatography, permitting detection of the MAPKK activity in fractions containing both kinases. This MAPKK activity was present in phorbol ester-treated wild-type cells, but not in phorbol ester-treated variant cells or in untreated wild-type or variant cells. The MAPKK from wild-type cells was able to activate MAPK prepared from either wild-type or variant cells. MAPKK activity could be stimulated in both wildtype and variant EL4 cells in response to treatment of cells with okadaic acid. These results indicate that the failure of variant EL4 cells to activate MAP kinase in response to phorbol ester is due to a failure to activate MAPKK. Therefore, the step that confers phorbol ester resistance to variant EL4 cells lies between the activation of protein kinase C and the activation of MAPKK.

EURATOM's Programme of Participation in Power Reactor Construction; Le programme de participation d'Euratom aux reacteurs de puissance; Programma uchastiya v razrabotke ehnergeticheskikh reaktorov Evratoma; El programa de participacion de la Euratom en la construccion y explotacion de reactores de potencia

Energy Technology Data Exchange (ETDEWEB)

Ramadier, R. C.; Parker, E. [Communaute Europoenne de l' Energie Atomique, Bruxelles (Belgium)

1963-10-15

annees de leur utilisation. Cette participation de l'Euratom a eu pour effet d'encourager la construction de certaines de ces centrales nucleaires. Elle a permis en outre et permettra encore d'acquerir des informations extremement utiles au cours des annees qui viennent, annees au cours desquelles les problemes de fonctionnement se poseront d'une facon decisive pour le developpement de l'energie atomique. (author) [Spanish] Para fomentar e l desarrollo de una industria nuclear europea, la Comision de la EURATOM utiliza, entre otros medios, su programa de ''participacion mancomunada'', que le permite cooperar con aportaciones de 32 millones de unidades de cuenta AME como maximo, en la construccion y explotacion de reactores de potencia. Como contrapartida, obtiene informaciones sobre el diseno, la construccion, la puesta en marcha y el funcionamiento de esos reactores. Hasta hoy la Comision ha firmado contratos con tres empresas: a) la Societa Elettronucleare Nazionale (SENN), que hace construir en Italia una central de 150 MW(e) netos equipada con un reactor de agua hirviente de ciclo doble; b) la Societa Italiana Meridionale Energia Atomica (SIMEA), que ha emprendido en Italia la construccion de una central de 200 MW(e) netos equipada con un reactor de uranio natural-grafito-anhidrido carbonico; c) la Societe d'Energie Nucleaire Franco-Belge des Ardennes (SENA), que ha emprendido en la frontera franco-belga la construccion de una central equipada con un reactor de agua a presion que podra alcanzar - y probablemente superar - una potencia neta de 242 MW(e). La Comision ha recibido, ademas, una solicitud del grupo Rheinisch-Westfalisches Elektrizitetswerk- Bayernwerke (RWE-BW) y otra de la N.V. Samenwerkende Electriciteits-Productiebedrijven para participar en la construccion y explotacion de otros dos reactores de potencia: la primera se refiere a un reactor de 237 MW(e) de agua hirviente y de ciclo doble, y la segunda a un reactor de 50 MW(e) de agua hirviente, de ciclo

Estudio preliminar para el tratamiento de lixiviados en un reactor de biodiscos

OpenAIRE

Ordóñez Losada, Paola Jimena; Betancur Pérez, Alonso

2003-01-01

El presente trabajo hace parte de un proyecto de investigación de la Universidad Nacional de Colombia Sede Manizales y EMAS (Empresa Metropolitana de Aseo S.A. E.S.P) para encontrar la mejor alternativa para el tratamiento de los lixiviados del relleno sanitario “La Esmeralda” de la ciudad de Manizales, con el fin de cumplir la legislación ambiental vigente sobre vertimientos líquidos industriales a las aguas superficiales. Se analizó en forma preliminar la aplicación de la tecnología biodisc...

From 3D to 4D seismic tomography at El Hierro Island (Canary Islands, Spain)

Science.gov (United States)

Garcia-Yeguas, A.; Koulakov, I.; Jakovlev, A.; Ibáñez, J. M.

2012-04-01

In this work we are going to show the advantages of a dynamic tomography 4D, versus a static image 3D related with a volcanic reactivation and eruption at El Hierro island (Canary Islands, Spain). In this process a high number of earthquakes before and during the eruptive processes have been registered. We are going to show a 3D image as an average of the velocity structure and then the characteristics and physical properties on the medium, including the presence or not of magma. This image will be complemented with its evolution along the time, observing its volcanic dynamic and its influence over the medium properties, including its power as an important element on early warnings protocols. After more than forty years of quiet at Canary Islands, since 1971 with Teneguía eruption at La Palma Island, and more than 200 years on El Hierro Island (The last eruption known at El Hierro took place in 1793, volcán de Lomo Negro), on 19th July on 2011 the Spanish seismic national network, administered by IGN (Instituto Geográfico Nacional), detected an increase of local seismic activity below El Hierro island (Canary Islands, Spain). Since this moment an intense swarm took place, with more than 11000 events, until 11th December, with magnitudes (MLg) from 0.2 to 4.4. In this period two eruptive processes have been declared in front of the South coast of El Hierro island, and they have not finished yet. This seismic swarm has allowed carrying out a 3D seismic tomography, using P and S waves traveltimes. It has showed a low velocity from the North to the South. On the other hand, we have performed a 4D seismic tomography, taking the events occurred at different intervals of time. We can observe the evolution of the negative anomaly along the time, from the North to the South, where has taken place La Restinga submarine eruption. 4D seismic tomography is an innovative and powerful tool able to show the evolution in time of a volcanic process.

Medical aspects of the nuclear accident in the Chernobylsk-4 reactor

International Nuclear Information System (INIS)

Arndt, D.; Schmidt, W.

1989-01-01

The Kiev conference on the Chernobylsk reactor accident was concerned with the following items: (1) Medical consequences and organization of medical assistance as well as aftercare of radiation-exposed persons. (2) Analysis of the postirradiation situation and judgement of the consequences of the accident as to the USSR population. (3) Peculiarities of external and internal radiation exposure of the population in the area controlled. (4) Organization and efficiency of the epidemiological register of the USSR. (5) Organization and judgement of educational work and public relations concerning the sanitary conditions in populations exposed to an increased contamination

Expression of Caspase-3, P53 in EL-4 cells induced by ionizing radiation and its biological implications

International Nuclear Information System (INIS)

Ju Guizhi; Shen Bo; Sun Shilong; Yan Fengqin; Fu Shibo; Li Pengwu

2006-01-01

Objective: To investigate the effect of ionizing radiation on the expressions of Caspase-3 and P53 proteins in EL-4 cells and its implications in the induction of apoptosis and polyploid cells. Methods: EL- 4 cells were irradiated with 4.0 Gy X-rays (180 kV, 15 mA, 0.287 Gy/min). Fluorescent staining and flow cytometry analysis were used to measure protein expression, apoptosis and polyploid cells. Results: It was found that the expression of Caspase-3 protein was increased significantly at 8 h and 12 h after the irradiation compared with sham-irradiated control (P<0.05), and the expression of P53 protein was also increased significantly at 2,4,8,12 and 24 h after the irradiation compared with sham-irradiated control (P<0.05 or P<0.01). The results showed that apoptosis of EL-4 cells was increased significantly at 2,4,8,12,24,48, and 72 h after 4.0 Gy irradiation compared with sham-irradiated control (P<0.05 or P<0.01 or P<0.001). However, no significant change in the number of polyploidy cells was found during the period from 2 to 48 h after the irradiation with 4.0 Gy X-rays. Conclusions: It is indicated that the expressions of Caspase-3 and P53 protein in EL-4 cells can be induced by ionizing radiation, and play an important role in the induction of apoptosis; the molecular pathway for polyploid formation might be P53-independent. (authors)

Performance Characteristics of the Experimental Boiling Water Reactor from 0 to 100 MW(t); Performances de l'EBWR de 0 a 100 MW; Rabochaya kharakteristika ehksperimental'nogo kipyashchego reaktora EBWR pri moshchnosti 0 - 100 mgvt.; Rendimiento del reactor experimental de agua hirviente (EBWR) entre 0 y 100 MW

Energy Technology Data Exchange (ETDEWEB)

Iskenderian, A.; Lipinski, W. C.; Petrick, M.; Wimunc, E. A. [Argonne National Laboratory, Argonne, IL (United States)

1963-10-15

controle; emploi d'elements enrichis; fonctions de transfert; analyse des bruits; certaines mesures de flux; stabilite, etc. En outre, on a pu savoir comment se comportaient certaines pieces et parties constitutives du reacteur et si elles restaient intactes, notamment le systeme de controle de l'acide borique, le niveau des rayonnements, la distribution des produits de la corrosion, les vices de fonctionnement du materiel; les barreaux de combustible et les barres de commande, etc. Les performances de l'EBWR dependent presque exclusivement de l'entrainement de la vapeur dans le tube d'eau, de Tentraftiement du liquide par la vapeur qui se degage, et, indirectement, de remplacement de l'interface vraie dans le caisson. C'est l'entrafhement de la vapeur dans le tube d'eau qui domine pour les puissances inferieures. Au-dessus de 65 MW les performances du reacteur sont radicalement modifiees. La vitesse de degagement de la vapeur atteint 33 cm/s et la hauteur du dome de vapeur descend a 1m. Dans ces conditions, il se produit un entrainement de liquide par la vapeur qui augmente rapidement lorsqu'on augmente la puissance. Le reacteur cesse alors de se comporter comme un reacteur a eau bouillante a cycle direct; en un sens, il fonctionne comme un reacteur a deux cycles, en circulation naturelle. (author) [Spanish] El 25 de mayo de 1962, el Laboratorio Nacional de Argonne fue autorizado por la USAEC a poner en funcionamiento el EBWR con una potencia de 100 MW. En el marco de la administracion de su sistema de salvaguardias el Organismo Internacional de Energia Atomica dio su aprobacion el 11 de julio de 1962. El 15 de noviembre del mismo ano, el reactor alcanzo la potencia de 100 MW. El programa experimental ejecutado con el reactor EBWR de 100 MW quedo completado el 6 de diciembre de 1962. Uno de los principales propositos del mismo consistia en dotar al reactor de los instrumentos necesarios para obtener datos e informaciones sobre el rendimiento de este tipo de reactor

SIMULACION DE LA PUESTA EN MARCHA DE UN REACTOR DE BIOPELÍCULA ANAEROBIA TIPO INTERCAMBIADOR DE CALOR

Directory of Open Access Journals (Sweden)

Ramiro Escalera Vásquez

2009-01-01

Full Text Available Se ha desarrollado un  modelo de reactor que considera la separación de fases microbianas dentro de un reactor anaerobio tipo intercambiador de calor, donde las bacterias acidogénicas predominan en la masa líquida recirculante y las heteroacetogénicas y metanogénicas lo hacen en la biopelícula adherida a las paredes. El modelo considera también las resistencias difusionales a la transferencia de masa ocasionadas por la capa laminar y la biopelícula. También se consideran las reacciones paralelas y consecutivas propias de la degradación anaerobia de compuestos orgánicos fácilmente biodegradables, por ejemplo, residuos industriales de altas concentraciones de carbohidratos. El modelo de reactor y las ecuaciones pseudo-analíticas para la estimación de los factores de efectividad, desarrolladas para otro tipo de bioreactores anaerobios tales como lechos empacados y fluidizados, pueden utilizarse para estimar la eficacia y evaluar el funcionamiento de un Reactor de Biopelícula Anaerobia tipo Intercambiador de Calor (RBAIC . En este trabajo se ha verificado que los resultados del modelo concuerdan con los resultados experimentales de la eficacia y funcionamiento del RBAIC, dentro del periodo de puesta en marcha.

Plan de Auditoría informática para el Grupo El Comercio C.A. con aplicación de la metodología COBIT 4.1

OpenAIRE

Bastidas Fierro, Carmen Alezandra

2014-01-01

This paper titling called: Plan Audit Computing Group Trade CA, with application of the COBIT 4.1 methodology seeks an audit departments Drafting and Technology using COBIT El presente trabajo de titulación denominado: Plan de Auditoría Informática para Grupo El Comercio C.A., con aplicación de la metodología COBIT 4.1 tiene por objeto realizar una auditoría en los departamentos de Redacción y Tecnología usando COBIT

{gamma} activity and heating of rods in EL2 and EL3; Activitiy {gamma} et echauffement des barres de EL2 et EL3

Energy Technology Data Exchange (ETDEWEB)

Lalere, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1959-07-01

A method is described for calculating the {gamma} activity of uranium rods, given the mean flux in which they are irradiated, the time they remain in the pile and the duration of deactivation. This calculation leads to numerical formulae which may be applied to the rods of the two reactors. It allows the saturation activities to be foreseen both for EL2 and for EL3, taking into recount the minimum times necessary for extraction. Measurements have been carried out, and the results are in good agreement with those foreseen by calculation. In the last section this method is used to calculate the heating of the irradiated rods. (author) [French] Une methode est indiquee ici, qui permet de calculer l'activite {gamma} des barres d'uranium connaissant le flux moyen dans lequel elles ont ete irradiees, leur temps de sejour en pile et la duree de la desactivation. Ce calcul conduit a des formules numeriques que l'on peut appliquer aux barres des deux reacteurs. Il permet de prevoir les activites atteintes a saturation, tant a EL2 qu'a EL3, compte tenu des temps minima necessaires a l'extraction. Des mesures ont ete faites: les resultats sont en bon accord avec les previsions du calcul. Enfin, en derniere partie, cette methode est utilisee pour calculer l'echauffement des barres irradiees. (auteur)

MULTIVARIABLE ANALYSIS OF 2,4-D HERBICIDE PHOTOCATALYTIC DEGRADATION

Directory of Open Access Journals (Sweden)

ANDRÉS F. LÓPEZ-VÁSQUEZ

2011-01-01

Full Text Available La degradación del herbicida 2,4-D en suspensiones de TiO2 en agua real fue evaluada bajo condiciones de irradiación artificial. El análisis multivariable de metodología de superficie de respuesta (MSR, se aplicó para evaluar el efecto de variables como la concentración de catalizador y pesticida, el pH y el caudal volumétrico sobre la reacción fotocatalítica en dos fotorreactores catalíticos: placa plana y tubular. La variable de respuesta fue la mineralización del pesticida expresada como porcentaje de degradación de carbono orgánico total (COT después de cuatro horas de irradiación. Para el fotorreactor tubular, los cuatro factores tuvieron la misma significancia sobre la degradación, mientras que para el fotorreactor de placa plana inclinada, sólo la concentración de catalizador y el pH tuvieron significancia. La MSR fue una técnica adecuada para obtener parámetros de operación óptimos de un proceso fotocatalítico con un reactor específico y dentro de un rango de estudio determinado.

Effect of Utilization of Silicide Fuel with the Density 4.8 gU/cc on the Kinetic Parameters of RSG-GAS Reactor

International Nuclear Information System (INIS)

Setiyanto; Sembiring, Tagor M.; Pinem, Surian

2007-01-01

Presently, the RSG-GAS reactor using silicide fuel element of 2.96 gU/cc. For increasing reactor operation time, its planning to change to higher density fuel. The kinetic calculation of silicide core with density 4.8 gU/cc has been carried out, since it has an influence on the reactor operation safety. The calculated kinetic parameters are the effective delayed neutron fraction, the delayed neutron decay constant, prompt neutron lifetime and feedback reactivity coefficient very important for reactor operation safety. the calculation is performed in 2-dimensional neutron diffusion-perturbation method using modified Batan-2DIFF code. The calculation showed that the effective delayed neutron fraction is 7. 03256x10 -03 , total delay neutron time constant is 7.85820x10 -02 s -1 and the prompt neutron lifetime is 55.4900 μs. The result of prompt neutron lifetime smaller 10 % compare with silicide fuel of 4.8 gU/cc. The calculated results showed that all of the feedback reactivity coefficient silicide core 4.8 gU/cc is negative. Totally, the feedback reactivity coefficient of silicide fuel of 4.8 gU/cc is 10% less than that of silicide fuel of 2.96 gU/cc. The results shown that kinetic parameters result decrease compared with the silicide core with density 2.96 gU/cc, but no significant influence in the RSG-GAS reactor operation. (author)

El cuerpo, el gueto y el Estado penal

Directory of Open Access Journals (Sweden)

Loïc Wacquant

2010-02-01

Full Text Available

Este artículo analiza el enfoque del autor sobre la etnografía, la teoría social, y la

Elements on reactor control

International Nuclear Information System (INIS)

Bruna, G.B.

1998-01-01

In order to achieve the two-fold goal of maximizing the energy obtained from reactor fuel and ensuring the large flexibility of plant operation in respect to safety regulations and keeping the reactor integrity the control of PWRs is generally based on real time monitoring and analysing of independent neutronic parameters: thermal power release, axial power distribution in the core and temperatures of the primary loop. Two control chains more or less coupled according to the control chosen mode are in charge of the control of these parameters. With the brief history of control in French power reactors the advanced X control mode adopted by Framatome for N4 plants is described in detail. A summary of N4 reactor control and protection system is included

Design of the fuel element 'snow-flake' in uranium oxide, canned with aluminium, for the experimental reactor EL 3 (1960); Etude d'un element combustible en oxyde d'uranium gaine d'aluminium, type ''cristal de neige'' pour la pile EL 3 (1960)

Energy Technology Data Exchange (ETDEWEB)

Gauthron, M.; Guibert, B. [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

This report sums up the main studies have been carried out on the fuel element 'Snowflake' (uranium oxide, canned with aluminium), designed to replace the present element of the experimental reactor EL3 in order to increase the reactivity without modifying the neutron flux/thermal power ratio. (author) [French] Ce rapport resume les principales etudes qui ont ete faites sur l'element combustible 'Cristal de Neige' (a oxyde d'uranium, gaine d'aluminium) destine a remnlacer l'element actuel du reacteur experimental EL3, afin d'en augmenter la reactivite sans modifier le rapport flux neutronique-puissance thermique. (auteur)

Calibration of the hydraulic model of a full-scale activated sludge plant; Calibracion hidraulica a escala real de un reactor de lodos activados

Energy Technology Data Exchange (ETDEWEB)

Fall, Cheikh [Universidad Autonoma del Estado de Mexico (Mexico); Loaiza-Navia, Jimmy [Servicios de Agua y Drenaje de Monterrey (Mexico)

2008-04-15

When planning to simulate a wastewater treatment plant (WWTP) with the activated sludge model number 1 (ASM1), one of the first requirements is to determine the hydraulic model of the reactor. The aim of this study was to evaluate the hydrodynamic regime of the aeration tank of a municipal WWTP by using a rhodamine tracer test and the Aquasim simulation software. A pre-simulation was performed in order to quantify the appropriate colorant mass, set up a sampling plan and evaluate the anticipated visual impact of the tracer test in the river receiving the treated effluents. A tracer test and dynamic flow measurements were carried out, the results of which served to establish and calibrate the hydraulic model. The evaluated tank was physically built as a plug-flow reactor subdivided in 7 compartments, but the study revealed that it is best represented by a model with 5 virtual mixed reactors in series. Through the study, the approach of using a WWTP simulator for hydraulics calibration was shown to be a powerful and flexible tool for designing a tracer test and for identifying adequate tank-in-series models of full-scale activated sludge aeration tanks. [Spanish] Cuando se planea simular una planta de tratamiento con base en el modelo numero 1 de lodos activados (ASM1), uno de los primeros requisitos es determinar el modelo hidraulico del reactor. En este trabajo se estudio el regimen hidrodinamico del tanque de accion de una planta de tratamiento de aguas residuales municipales (PTAR), utilizando una prueba de trazador con rodamina y un programa de simulacion (Aquasim). Se realizo una prueba de trazador con el experimento, lo que permitio determinar la cantidad requerida de trazador, fijar los intervalos de muestreo y limitar el impacto visual anticipado del colorante sobre el rio que recibe el efluente tratado. Se llevaron a cabo la prueba de trazador y la medicion de los perfiles dinamicos de caudales, cuyos resultados sirvieron para establecer y calibrar el

Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

International Nuclear Information System (INIS)

Bussac, J.; Reuss, P.

1985-01-01

This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

TGF-beta1 expression in EL4 lymphoma cells overexpressing growth hormone.

Science.gov (United States)

Farmer, John T; Weigent, Douglas A

2006-03-01

Our previous studies show that growth hormone overexpression (GHo) upregulates the expression of the IGF-1R and IGF-2R resulting in the protection of the EL4 lymphoma cell line from apoptosis. In this study, we report that GHo also increases TGF-beta1 protein expression measured by luciferase promoter assay, Western analysis, and ELISA. Further, the data show that antibody to TGF-betaR2 decreases TGF-beta1 promoter activity to the level of vector alone control cells. GHo cells treated with (125)I-rh-latent TGF-beta1 showed increased activation of latent TGF-beta1 as measured by an increase in the active 24kDa, TGF-beta1 compared to vector alone control cells. The ability of endogenous GH to increase TGF-beta1 expression is blocked in EL4 cells by antisense but not sense oligodeoxynucleotides or in cells cultured with antibody to growth hormone (GH). The data suggest that endogenous GH may protect from apoptosis through the IGF-1R receptor while limiting cellular growth through increased expression and activation of TGF-beta1.

Supercritical Water Reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

2014-01-01

The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

4. generation sodium-cooled fast reactors. The ASTRID technological demonstrator

International Nuclear Information System (INIS)

2012-12-01

The sodium-cooled fast reactor (SFR) concept is one of the four fast neutron concepts selected by the Generation IV International Forum (GIF). SFRs have favourable technical characteristics and they are the sole type of reactor for which significant industrial experience feedback is available. After a discussion of the past experience gained on fast breeder reactors in the world (benefits, difficulties and problematics), the authors discuss the main improvement domains and the associated R and D advances (reactor safety, prevention and mitigation of severe accidents, the sodium-water risk, detection of sodium leaks, increased availability, instrumentation and inspection, control and repairability, assembly handling and washing). Then, they describe the technical requirements and safety objectives of the ASTRID experimental project, notably with its reactivity management, cooling management, and radiological containment management functions. They describe and discuss requirements to be met and choices made for Astrid, and the design options for its various components (core and fuels, nuclear heater, energy conversion system, fuel assembly handling, instrumentation and in-service inspection, control and command). They present the installations which are associated with the ASTRID cycle, evoke the development and use of simulations and codes, describe the industrial organization and the international collaboration about the ASTRID project, present the planning and cost definition

The development of octagon Zr-4 alloy tube for heating reactors

International Nuclear Information System (INIS)

Yang Fanglin; Yang Yingli; Wang Guangshen

1989-10-01

The asymmetrical octagon Zr-4 alloy tubes which are used for fuel assembly in the heating reactor have been developed. The thickness of tube wall is 1.5 mm and the length is 1725 mm. The long side of the octagon is 138.7 0.3 +0.2 mm, the short side is 93.1 ± 0.1 mm. To manufacture these tubes a stretch draw forming processing method is adopted. The process is divided into two phases. In the first phase, a short draw mould is used to stretch the Zr-4 alloy tube. In the second phase, a long draw mould, its length is equal to the end-produt length, is used to complete the final processing. The size accuracy and repeatability of this method are excellent and can fully meet the design requirements

Annex VII - Diagrams: 1. Reactor operation (1960-1977); 2. Mean daily reactor power density in 1977; 3. Monthly reactor power for 1977; 4. percent of utilization of experimental space in 1977; Prilog VII - Dijagrami: 1. Rad reaktora (MWh) po godinama (1960-1977); 2. Srednja dnevna snaga reaktora u 1977. godini; 3. Rad reaktora (MWh) po mesecima za 1977. godinu i 4. Procenat iskoriscenja eksperimentalnog prostora u 1977. godini

Energy Technology Data Exchange (ETDEWEB)

NONE

1977-12-15

This Annex includes the following diagrams: 1. Annual Reactor RA power production (MWh) for the period from 1960-1977; 2. Mean daily reactor power density MW in 1977; 3. Monthly reactor power production (MWh) for 1977; 4. percent of utilization of experimental space in 1977. [Serbo-Croat] Ovaj prilog sadrzi dijagrame: 1. Rad reaktora (MWh) po godinama (1960-1977); 2. Srednja dnevna snaga reaktora u 1977. godini; 3. Rad reaktora (MWh) po mesecima za 1977. godinu i 4. Procenat iskoriscenja eksperimentalnog prostora u 1977. godini.

Evaluación del comportamiento hidrodinámico como herramienta para optimización de reactores anaerobios de crecimiento en medio fijo

Directory of Open Access Journals (Sweden)

Andrea Pérez

2008-01-01

Full Text Available Las condiciones de flujo no ideal en los reactores afectan su desempeño; las causas comunes son cortos circuitos, zonas muertas y recirculación interna por corrientes cinéticas y/o de densidad. En este estudio se optimizó el diseño de un filtro anaerobio a escala real que trata las aguas residuales del proceso de extracción de almidón de yuca, el cual presentaba problemas de represamiento y bajas eficiencias de remoción. La evaluación del comportamiento hidrodinámico inicial mostró la presencia de flujo dual (32% flujo pistón - FP y 37% mezcla completa - CM, zonas muertas (20% y ausencia de cortos circuitos; adicionalmente, la modelación del reactor indicó un grado de dispersión elevado y un comportamiento tendiente a un reactor CM en serie de dos unidades. Con base en estos resultados, se implementaron dos modificaciones en el diseño del reactor: falso fondo y tubería perforada para evacuación de biogás, las cuales permitieron incrementar la fracción de FP (44%, reducir la fracción de zonas muertas (15%, disminuir el Índice de Dispersión (ID e incrementar la tendencia del reactor a un CM en serie de tres unidades, lo que aumentó el tiempo de retención hidráulico (TRH real de 9,6 a 10,2 horas (TRH teórico 12 horas y las eficiencias teóricas de remoción de 73 a 78%.

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2012. Operation, utilization and technical development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

International Nuclear Information System (INIS)

Murayama, Yoji; Ishii, Tetsuro; Nakamura, Kiyoshi; Uno, Yuki; Ishikuro, Yasuhiro; Kawashima, Kazuhito; Ishizaki, Nobuhiro; Matsumura, Taichi; Nagahori, Kazuhisa; Odauchi, Shouji; Maruo, Takeshi

2014-03-01

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2012 and March 31, 2013. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator, (2) Utilization of research reactors and tandem accelerator, (3) Upgrading of utilization techniques of research reactors and tandem accelerator, (4) Safety administration for department of research reactor and tandem accelerator, (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, number of staff members dispatched to Fukushima for the technical assistance, outcomes in service and technical developments and so on. (author)

The molten salt reactor adventure

International Nuclear Information System (INIS)

MacPherson, H.G.

1985-01-01

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

Generation III+ Reactor Portfolio

International Nuclear Information System (INIS)

2010-03-01

While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

Fuel elements for pressurised-gas reactors; Elements combustibles des piles a gaz sous pression

Energy Technology Data Exchange (ETDEWEB)

Stohr, J A; Englander, M; Gauthron, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The design and fabrication of fuel elements for the first CO{sub 2} pressurized reactors have induced to investigate: various cladding materials, natural uranium base fuels, canning processes. The main analogical tests used in connection with the fuel element study are described. These various tests have enabled, among others, the fabrication of the fuel element for the EL2 reactor. Lastly, future solutions for electrical power producing reactors are foreseen. (author)Fren. [French] L'etude et la realisation d'elements combustibles pour les premieres piles a CO{sub 2} sous pression ont conduit a examiner: les divers materiaux de gaine, les combustibles a base d'uranium naturel, les modes de gainage. Les principaux essais analogiques ayant servi au cours de l'etude de la cartouche sont decrits. Ces divers essais ont notamment permis la realisation de la cartouche de la pile EL2. Enfin sont envisagees les solutions futures pour les piles productrices d'energie electrique. (auteur)

Tratamiento de las excretas de cerdo mediante un reactor anaeróbico SCFBR a nivel de banco Treatment of pig excreta using an SCFBR anaerobic reactor

Directory of Open Access Journals (Sweden)

Caicedo Luis A.

1999-06-01

Full Text Available Un nuevo reactor anaeróbico denominado Sludge Central Fixed Bed Reactor (SCFBR fue construido y evaluado para tratar los residuos líquidos de las granjas porcícolas. El SCFBR está constituido por tres zonas principales. Una zona inferior de lodos, seguida por un módulo empacado ubicado en forma concéntrica y, en la parte superior, una zona de separación sólido, líquido y gas. El reactor de 28,5 1 de volumen de reacción fue evaluado durante 210 días para tres cargas orgánicas de 0,548, 0,421 y 1,239 g DQO/ 1 día. El SCFBR fue alimentado inicialmente en forma discontinua con tiempos de retención hidráulicos (TRH de 10 y 10,7 días. Posteriormente el TRH fue disminuido a 3,87 días con una alimentación en continuo. Para las tres cargas orgánicas de 0,548, 0,421 y 1,239 g DQO/1 día se obtuvieron remociones en la demanda química de oxígeno (DQO de 68%, 81% y 73% y en los sólidos volátiles (SV, de 53,5%, 55,8% y 50,1%, respectivamente. El SCFBR presentó un buen desempeño, re-presentado en las eficiencias de remoción y en la estabilidad observada. Se presenta una microfotografía tomada de una muestra de lodo de la zona inferior del SCFBR, observándose una gran presencia de microorganismos del género Methanosaeta (Methanothrix.

A new anaerobic reactor called the Sludge Central Fixed Bed Reactor (SCFBR was built and evaluated for the treatment of liquid residue from the pig farms. The SCFBR has three main parts. The lower area is for sludge, the middle part consists of a concentrically

Impact of proposed research reactor standards on reactor operation

Energy Technology Data Exchange (ETDEWEB)

Ringle, J C; Johnson, A G; Anderson, T V [Oregon State University (United States)

1974-07-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

Impact of proposed research reactor standards on reactor operation

International Nuclear Information System (INIS)

Ringle, J.C.; Johnson, A.G.; Anderson, T.V.

1974-01-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Younggwang nuclear power plant unit1

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae [Korea Atomic Energy Research Institute, Taejeon (Korea)

2001-08-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 1 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.555E+18, 1.662E+19, 3.358E+19, and 4.521E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.859 for the 1st through 4th testing and the calculational uncertainty, 11.80% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.551E+19 n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.929E+19, 4.880E+19, 5.831E+19 and 6.782E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 1 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 4 refs., 41 figs., 35 tabs. (Author)

EL SISTEMA VESTIBULAR EN EL DESARROLLO DE LA PSICOMOTRICIDAD GRUESA EN LOS NIÑOS-AS DE 4-5 AÑOS DEL INSTITUTO SAN FRANCISCO DE ASÍS EN LA PROVINCIA DE CHIMBORAZO, CANTON RIOBAMBA, PARROQUIA JUAN DE VELASCO EN EL AÑO LECTIVO 2011-2012

OpenAIRE

Paguay Franco, Rosana Isabel

2016-01-01

El presente trabajo de investigación va orientado a analizar el sistema vestibular y el desarrollo de la psicomotricidad gruesa en niños-as de 4-5 años del Instituto San Francisco de Asís; se realizó con el objetivo de concienciar a docentes y padres de familia acerca de la importancia del sistema vestibular en el proceso de aprendizaje de toda persona. Es novedosa la contribución de la psicomotricidad gruesa pues tiene como principal propósito el desarrollar y recuperar el propio cuerpo. Se ...

Viabilidade económica da implementação de um reactor nuclear para a produção de energia eléctrica em Portugal

OpenAIRE

Pedro, Miguel António de Morais

2012-01-01

O presente trabalho tem como objectivo avaliar economicamente e determinar a viabilidade da implementação de um reactor nuclear para produção de energia eléctrica. Faz-se uma abordagem a aspectos da energia nuclear no mundo e em particular a energia nuclear na união europeia, faz-se uma análise sobre a estrutura do sector nuclear em Espanha e o futuro da energia no mundo. É realizada uma análise sobre a energia nuclear em Portugal, são abordados aspectos como o planeamento energético, a local...

Management of spent fuel from research reactors - Brazilian progress report (within the framework of Regional Project IAEA-RLA-4/018)

International Nuclear Information System (INIS)

Soares, A.J.; Silva, J.E.R.

2005-01-01

There are four research reactors in Brazil. For three of them, because of the low reactor power and low burn-up of the fuel, except for the concern about ageing, spent fuel storage is not a problem. However for one of the reactors, more specifically IEA-R1 research reactor, the storage of spent fuel is a major concern, because, according to the proposed operation schedule for the reactor, unless an action is taken, by the year 2009 there will be no more racks available to store its spent fuel. This paper gives a brief description of the type and amount of fuel elements utilized in each one of the Brazilian research reactors, with a short discussion about the storage capacity at each installation. It also gives a description of the activities developed by Brazilian engineers and researchers during the period between 2001 and 2004, within the framework of regional project 'RLA-4/018-Management of Spent Fuel from Research Reactors'. As a conclusion, we can say that the advances of the project, and the integration promoted among the engineers and researchers of the participant countries were of fundamental importance for Brazilian researchers and engineers to understand the problems related to the storage of spent fuel, and to make a clear definition about the most suitable alternatives for interim storage of the spent fuel from IEAR1 research reactor. (author)

Reactor Radiation Loops as Large Gamma Sources; Boucles d'irradiation des reacteurs nucleaires utilisees comme sources gamma intenses; Radiatsionnye kontury yadernykh reaktorov kak moshchnye gamma-istochniki; Empleo de circuitos de irradiacion de los reactores como fuentes gamma de gran intensidad

Energy Technology Data Exchange (ETDEWEB)

Ryabukhina, Yu. S.

1963-11-15

primer lugar, se eligieron para la realizacion de esos circuitos las aleaciones de indio, metal liquido a temperatura ambiente. Se estudio el comportamiento de dos eutecticos de indio frente a algunos materiales de construccion y a principios de 1960 se construyo el primer circuito de prueba de indio-galio. Como iesultado de estudios ulteriores, se instalaron modelos de circuitos de indio-galio, con una actividad en el irradiador equivalente a unos 100 g de Ra, en el reactor IRT de la Academia de Ciencias de la Republica Socialista Sovietica de Georgia, asi'como un circuito de prueba de indio-galio-estafio in el canal del reactor IRT del Instituto de Energia Atomica de la Academia de Ciencias de la Union Sovietica. Por ultimo, en 1962, se instalo un circuito de trabajo de indio-galio-estano en el reactor IRT de la Academia de Ciencias de la Republica Socialista Sovietica de Latvia para efectuar irradiaciones en escala semiindustrial. La actividad maxima en el irradiador equivale a 30 000 g de Ra. La memoria consta de las siguientes partes: 1. ''Calculo de los circuitos de irradiacion''; en esta parte se resena la labor realizada en materia de metodos de calculo de los circuitos de irradiacion. 2. ''Modelo de un circuito de irradiacion deindio-galiodelreactor IRT de Tbilisi''; se describe el funcionamiento de este circuito. 3. ''Circuito de irradiacion de indio-galio-estano del reactor IRT de la Academia de Ciencias de la Republica Socialista Sovietica de Latvia'' ; se describe el funcionamiento de este circuito. 4. ''Perspectivas de desarrollo de los circuitos de irradiacion''; se describen los experimentos y circuitos y se presentan calculos que sugieren la posibilidad de construir circuitos de manganeso solido y circuitos con aleaciones liquidas de indio. (author) [Russian] Nachinaya s 1957 g. v SSSR provodilis' raboty po izucheniyu radiatsionnykh konturov. Byli razrabotany metody rascheta takikh sistem, izucheny vozmozhnosti razlichnykh gamma-nositelej. Vnachale byli

What occurred in the reactors

International Nuclear Information System (INIS)

Kudo, Kazuhiko

2013-01-01

Described is what occurred in the reactors of Fukushima Daiichi Nuclear Power Plant at the Tohoku earthquake and tsunami (Mar. 11, 2011) from the aspect of engineering science. The tsunami attacked the Plant 1 hr after the quake. The Plant had reactors in buildings no.1-4 at 10 m height from the normal sea level which was flooded by 1.5-5.5 m high wave. All reactors in no.1-6 in the Plant were the boiling water type, and their core nuclear reactions were stopped within 3 sec due to the first quake by control rods inserted automatically. Reactors in no.1-5 lost their external AC power sources by the breakdown and subsequent submergence (no.1-4) of various equipments and in no.1, 2 and 4, the secondary DC power was then lost by the battery death. Although the isolation condenser started to cool the reactor in no.1 after DC cut, its valve was then kept closed to heat up the reactor, leading to the reaction of heated Zr in the fuel tube and water to yield H 2 which was accumulated in the building: the cause of hydrogen explosion on 12th. The reactor in no.2 had the reactor core isolation cooling system (RCIC) which operated normally for few hrs, then probably stopped to heat up the reactor, resulting in meltdown of the core but no explosion occurred because of the opened door of the blowout panel on the wall by the blast of no.1 explosion. The reactor in no.3 had RCIC and high pressure coolant injection system, but their works stopped to result in the core damage and H 2 accumulation leading to the explosion on 14th. The reactor in no.4 had not been operated because of its periodical annual examination, but was explored on 15th, of which cause was thought to be due to backward flow of H 2 from no.3. Finally, the author discusses about this accident from the industrial aspect of the design of safety level (defense in depth) on international views, and problems and tasks given. (T.T.)

The Role of Non-Destructive Testing in the Los Alamos Reactor Programme; Role des Essais Non Destructifs dans le Programme de Reacteurs de los Alamos; Rol' nedestruktivnykh ispytanij materialov v Los-Alamosskoj reaktornoj programme; Papel de los Metodos de Ensayo No Destructivo en el Programa de Reactores de Los Alamos

Energy Technology Data Exchange (ETDEWEB)

Tenney, G. H. [University of California, Los Alamos Scientific Laboratory, Los Alamos, NM (United States)

1965-10-15

temperature UHTREX, actuellement en construction, on a etudie par microiadiogiaphie et au moyen de microscopes electroniques des grains de carbure d'uranium enrobes de carbone pyrolytique, d'un diametre de 150 {mu}m, pour evaluer la translocation de l'uranium en fonction de la temperature. On determine la quantite et l'uniformite de la charge d'uranium dans les elements au graphite d'UHTREX au moyen de compteurs a scintillation specialement concus. Environ 90% des travaux effectues a ce sujet n'ont encore fait l'objet d'aucune publication. (author) [Spanish] El Laboratorio Cientifico de Los Alamos, explotado por la Universidad de California por encargo de la Comision de Energia Atomica de los Estados Unidos, viene ocupandose desde hace mas de veinte afios del proyecto, diseno y construccion de reactores nucleares de cuatro tipos generales; a saber, de investigacion, de potencia, de propulsion espacial y para conjuntos criticos. El llamado Grupo de ensayos no destructivos colabora practicamente en todas las actividades y proyectos del laboratorio. En la presente memoria se exponen algunos de los metodos de ensayo no destructivo y sus aplicaciones, establecidos para uso en el programa de reactores. El programa LAPRE (Los Alamos Power Reactor Experiment) se basa en el empleo de una solucion de fosfato de uranio a alta temperatura. La solucion es muy corrosiva y todas las piezas que entren en contacto con ella deben ir revestidas de oro. Durante el proceso de produccion de chapa de oro laminada a partir de lingotes, se han utilizado procedimientos radiograficos especiales para inspeccionar el metal. Las juntas soldadas se examinaron del mismo modo, y ademas se establecio un metodo para comprobar la presencia de impurezas incrustadas en la superficie de la chapa de oro. El concepto fundamental en que se basa el programa LAMPRE (Los Alamos Molten Plutonium Reactor Experiment) es la utilizacion como combustible de plutonio metalico liquido en vez de solido. El combustible esta

Compact torsatron reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

1988-05-01

Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R 0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R 0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

Scale-4 analysis of pressurized water reactor critical configurations: Volume 5, North Anna Unit 1 Cycle 5

International Nuclear Information System (INIS)

Bowman, S.M.; Suto, T.

1996-10-01

ANSI/ANS 8.1 requires that calculational methods for away-from- reactor (AFR) criticality safety analyses be validated against experiment. This report summarizes part of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial PWRs. Codes and data in the SCALE-4 code system were used. This volume documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. The KENO V.a criticality calculations for the North Anna 1 Cycle 5 beginning-of-cycle model yielded a value for k eff of 1. 0040±0.0005

Investigation of thermodynamic cycle for generic 1200 MW{sub el} pressure channel reactor with nuclear steam superheat

Energy Technology Data Exchange (ETDEWEB)

Vincze, A.; Sidawi, K.; Abdullah, R.; Baldock, M.; Saltanov, E.; Pioro, I., E-mail: andrei.vincze@uoit.net, E-mail: khalil.sidawi@uoit.net, E-mail: rand.abdullah@uoit.net, E-mail: matthew.baldock@uoit.net, E-mail: eugene.saltanov@uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada)

2014-07-01

Current Nuclear Power Plants (NPPs) play a significant role in energy production around the world. All NPPs operating today employ a Rankine steam cycle for the conversion of thermal power to electricity. This paper will examine the steam cycle arrangement an experimental pressure channel reactor using Nuclear Steam Superheat (NSS) and compare it to two advanced reactor designs, the Advanced CANDU Reactor 1000 (ACR-1000) and the Advanced Boiling Water Reactor (ABWR) designs. The thermodynamic cycle layout and thermal efficiencies of the three reactor types will be discussed. (author)

Fast reactors

International Nuclear Information System (INIS)

Vasile, A.

2001-01-01

Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

Comparative economic analysis of the Integral Molten Salt Reactor and an advanced PWR using the G4-ECONS methodology

International Nuclear Information System (INIS)

Samalova, Ludmila; Chvala, Ondrej; Maldonado, G. Ivan

2017-01-01

The assessment of economic viability of a new reactor concept is crucial particularly during the early stages of its concept development. The G4-ECONS methodology provides a standardized top-down estimate of electricity cost and parametric sensitivities, not specifically targeted toward an accurate prediction of the final cost when deployed, but rather seeking an approximation of cost variations relative to other systems. This study presents an analysis of the Integral Molten Salt Reactor (IMSR) concept in comparison with a consistent analysis of an advanced PWR reactor (represented by AP1000). Estimation of levelized unit electricity costs, as well as sensitivity analyses to the discount rate and uranium or SWU prices, are presented using this methodology.

Necessity of research reactors

International Nuclear Information System (INIS)

Ito, Tetsuo

2016-01-01

Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

The Use of Prestressed Concrete Vessels in the French Power Reactor Programme; Les caissons en beton precontraint dans le programme francais des reacteurs de puissance; Korpusy iz predvaritel'no napryazhennogo betona vo frantsuzskoj programme ehnergeticheskikh reaktorov; Empleo de recipientes de presion de hormigon pretensado en el programa frances de reactores de potencia

Energy Technology Data Exchange (ETDEWEB)

Conte, F. [Centre d' Etudes Nucleaires de Marcoule (France); Dambrine, C. [Centre d' Etudes Nucleaires de Fontenay-aux-Roses (France); Gaussot, D. [Electricite de France, Clamart (France)

1963-10-15

de la pression de service et de toutes manieres, une augmentation notable des dimensions, ce qui permet d'envisager des solutions du type integre. (author) [Spanish] G3 de Marcoule y en el reactor EDF3, en construccion en Chinon. Los reactores se encuentran en servicio desde 1959 y 1960, respectivamente; el Comissariat a l'energie atomique indica los problemas que ha planteado la construccion de los recipientes de presion y las observaciones efectuadas durante el funcionamiento de los reactores, que ponen de manifiesto la gran seguridad de los mencionados recipientes. La construccion del recipiente de presion del reactor EDF3, que comenzo en Chinon en el segundo semestre de 1961, prosigue actualmente y quedara terminada a fines de 1963. L'Electricite de France expone los motivos de la eleccion de este tipo de recipiente, los resultados de los calculos y de los ensayos efectuados con maquetas, asi como los problemas planteados por sir construccion. Se han llevado a cabo varios estudios sobre las perspectivas del empleo del hormigon pretensado en los reactores. Al parecer, este material permite obtener un aumento de la presion de trabajo y de todas maneras, un incremento notable de las dimensiones, lo que a su vez permite tomar en consideracion soluciones de tipo integrado. (author) [Russian] Izlagaetsya vopros o primenenii predvaritel'no napryazhennogo betona dlya reaktorov G.2 i G.3 v Markule i dlya stroyashchegosya v Shinone reaktora EDF.3. Reaktory dostigli mosnosti sootvetstvenno v 1959 i 1960 godakh; KAEH otmechaet problemy, kotorye voznikli v protsesse stroitel'stva korpusa reaktora, i izlagaet filosofiyu nablyudenij, kotorye prodemonstrirovali vysokuyu bezopasnost' ehtikh ustanovok. K stroitel'stvu korpusa reaktora EDF.3 v SHinone pristupili vo vtoroj polovine 1961 goda; stroitel'stvo budet zaversheno k kontsu 1963 goda. ''Ehlektrisite de Frans'' ob{sup y}asnyaet prichiny vybora takogo korpusa, privodit rezul'taty raschetov i provedennykh na makete ispytanij, a

Extrusion and drawing of zircaloy 2. Production of pressure tubes for EL-4; Filage et etirage du zircaloy 2. Realisation des tubes de force pour EL-4

Energy Technology Data Exchange (ETDEWEB)

Thevenet, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Buffet, J [Cefilac (France)

1964-07-01

The authors give briefly the physical mechanical and chemical properties of zircaloy 2, as far as the transformation of this alloy is concerned. Extrusion: After a few general remarks concerning the extrusion and co-extrusion, including a comparison of the deformation resistance of canning metals and of zircaloy 2, the following points are considered: - the difficulties occurring because of the use of this alloy: - atmosphere protection - adjustment on to the machine tools - low thermal conductivity - economy of the metal (price) - the factors affecting the quality of the extruded products extrusion under a copper can and under lubricant glass - fine grain structure - temperature homogeneity - working temperature The transformation cycle - '550 kg ingot - preliminary shape 'for drawing of EL-4 tubes (112 x 120 L 12 m)' - is described in detail (extrusion or forging of the {phi} = 340 ingot into {phi} = 220 billets, cutting into lengths and hot drilling at {phi} = 125, fixing into a copper can and rough extrusion). Drawing: The main difficulties are due to seizing of the tools and to the necessity of protecting the alloy from the atmosphere during annealings. A brief description is given of drawing out on a short mandrel, on a long mandrel, of laminating on a reducing machine and of the carrying out of an annealing, as well as of the production of EL-4 tubes ({phi} =107 x 113 L 430 m) by drawing out shapes having a size of 112 x 120 on long mandrels. Conclusion: It is possible by extrusion and drawing to produce zircaloy 2 tubes similar to those which may be obtained normally using stainless steel. (authors) [French] Les auteurs donnent un resume succint des proprietes physiques mecaniques et chimiques du zircaloy 2 en ce qui concerne la transformation de cet alliage. Filage: Apres quelques generalites sur le filage et le cofilage, dont une comparaison entre les resistances a la deformation des metaux de gainage et du zircaloy 2, on etudie successivement: - les

[Study of the immunological mechanism of anti-tumor effects of 5-FU by establishing EL4 tumor-bearing mouse models].

Science.gov (United States)

Li, Mo-Lin; Li, Chuan-Gang; Shu, Xiao-Hong; Li, Ming-Xia; Jia, Yu-Jie; Qin, Zhi-Hai

2007-11-01

To investigate the immunological mechanism of anti-tumor effect of 5-FU by establishing lymphoma EL4 tumor-bearing mouse models in wild type C57BL/6 mice and nude C57BL/6 mice, respectively. The mouse lymphoma EL4 cells were inoculated subcutaneously into wild type C57BL/6 mice (immune-competent mice). Twelve days later, 5-FU of different doses was administered intraperitoneally to treat these wild type C57BL/6 tumor-bearing mice. The size of tumors in the wild type C57BL/6 mice was observed and recorded to explore the minimal dose of 5-FU that could cure the tumor-bearing mice. Then the same amount of EL4 tumor cells was inoculated subcutaneously into wild type C57BL/6 mice and nude C57BL/6 mice (T cell-deficient mice) simultaneously, which had the same genetic background of C57BL/6. Twelve days later, 5-FU of the minimal dose was given intraperitoneally to treat both the wild type and nude C57BL/6 tumor-bearing mice. The size of tumors in the two different types of mice was observed and recorded. A single dose of 5-FU (75 mg/kg) cured both the EL4 tumor-bearing wild type C57BL/6 mice and the EL4 tumor-bearing nude C57BL/6 mice in the first week. Two weeks after 5-FU treatment, all of the nude mice died of tumor relapse while most of the wild type C57BL/6 mice were fully recovered. A single dose of 5-FU has marked anti-tumor effects on lymphoma EL4 tumor-bearing C57BL/6 mice with or without T lymphocytes. The relapse of tumors after 5-FU treatment might be related to the function of T lymphocytes.

Analysis of calculated neutron flux response at detectors of G.A. Siwabessy multipurpose reactor (RSG-GAS Reactor)

International Nuclear Information System (INIS)

Taryo, Taswanda

2002-01-01

Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect

The IAEA programme on research reactor safety

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

According to the research reactor database of IAEA (RRDB), 250 reactors are operating worldwide, 248 have been shut down and 170 have been decommissioned. Among the 248 reactors that do not run, some will resume their activities, others will be dismantled and the rest do not face a clear future. The analysis of reported incidents shows that the ageing process is a major cause of failures, more than two thirds of operating reactors are over 30 years old. It also appears that the lack of adequate regulations or safety standards for research reactors is an important issue concerning reactor safety particularly when reactors are facing re-starting or upgrading or modifications. The IAEA has launched a 4-axis program: 1) to set basic safety regulations and standards for research reactors, 2) to provide IAEA members with an efficient help for the application of these safety regulations to their reactors, 3) to foster international exchange of information on research reactor safety, and 4) to provide IAEA members with a help concerning safety issues linked to malicious acts or sabotage on research reactors

Reactor Materials Research

International Nuclear Information System (INIS)

Van Walle, E.

2001-01-01

The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

Reactor Materials Research

Energy Technology Data Exchange (ETDEWEB)

Van Walle, E

2001-04-01

The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

Tritium release from lithium silicate and lithium aluminate, in-reactor and out-of-reactor

International Nuclear Information System (INIS)

Johnson, A.B. Jr.

1976-09-01

Studies were conducted to determine the generation and evolution of tritium and helium in lithium aluminate (LiAlO 2 ) and lithium silicate (Li 2 SiO 3 ) by the reaction: Li 6 + n → 4 He + T. Targets were irradiated 4.4 days in the K-West Reactor snout facility. (Silicate GVR* approximately 2.0 cc/cc; aluminate GVR approximately 1.4 cc/cc.) Gas release in-reactor was determined by post-irradiation drilling experiments on aluminum ampoules containing silicate and aluminate targets. In-reactor tritium release (at approximately 100 0 C) was found to decrease linearly with increasing target density. Tritium released in-reactor was primarily in the noncondensible form (HT and T 2 ), while in laboratory extractions (300-1300 0 C), the tritium appeared primarily in the condensible form (HTO and T 2 O). Concentrations of HT (and presumably HTO) were relatively high, indicating moisture pickup in canning operations or by inleakage of moisture after the capsule was welded. Impurities in extracted gases included H 2 O, CO 2 , CO, O 2 , H 2 , NO, SO 2 , SiF 4 and traces of hydrocarbons

Can-rupture detection in gas-cooled nuclear reactors; La detection des ruptures de gaine dans les piles nucleaires refroidies par gaz

Energy Technology Data Exchange (ETDEWEB)

Roguin, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-07-01

Can-rupture detection (DRG) is one important aspect of pile safety, more particularly so in the case of gas-cooled reactors. A rapid and sure detection constitutes also an improvement as far as the efficiency of electricity-producing nuclear power stations are concerned. Among the numerous can-rupture detection methods, that based on the measurement of the concentration of short-lived fission gases in the heat-carrying fluid has proved to be the most sensitive and the most rapid. A systematic study of detectors based on the electrostatic collection of the daughter products of fission gases has been undertaken with a view to equip the reactors EL 2, G 3, EDF 1, EDF 2 and EDF 3, the gas loops of PEGASE and EL 4. The different parameters are studied in detail in order to obtain a maximum sensitivity and to make it possible to construct detection devices having the maximum operational reliability and requiring the minimum maintenance. The primary applications of these devices are examined in the case of the above-mentioned reactors. (author) [French] La Detection des Ruptures de Gaines (D. R. G.) est un aspect important de la securite des piles et plus particulierement des piles refroidies par un gaz. Une detection rapide et sure constitue aussi un element d'amelioration du rendement des centrales nucleaires productrices d'energie electrique. Parmi les nombreuses methodes de detection des ruptures de gaines, la mesure de la concentration dans le fluide caloporteur des gaz de fission a vie courte s'est revelee comme la plus sensible et la plus rapide. Une etude systematique des detecteurs a collection electrostatique des descendants des gaz de fission a ete entreprise en vue d'equiper les piles EL 2, G 3, EDF 1, EDF 2 et EDF 3, les boucles a gaz de la pile Pegase et la pile EL 4. Les divers parametres sont etudies en detail pour obtenir une sensibilite maximum et permettre la realisation de dispositifs de detection ayant le maximum de securite de fonctionnement et le

Can-rupture detection in gas-cooled nuclear reactors; La detection des ruptures de gaine dans les piles nucleaires refroidies par gaz

Energy Technology Data Exchange (ETDEWEB)

Roguin, A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-07-01

Can-rupture detection (DRG) is one important aspect of pile safety, more particularly so in the case of gas-cooled reactors. A rapid and sure detection constitutes also an improvement as far as the efficiency of electricity-producing nuclear power stations are concerned. Among the numerous can-rupture detection methods, that based on the measurement of the concentration of short-lived fission gases in the heat-carrying fluid has proved to be the most sensitive and the most rapid. A systematic study of detectors based on the electrostatic collection of the daughter products of fission gases has been undertaken with a view to equip the reactors EL 2, G 3, EDF 1, EDF 2 and EDF 3, the gas loops of PEGASE and EL 4. The different parameters are studied in detail in order to obtain a maximum sensitivity and to make it possible to construct detection devices having the maximum operational reliability and requiring the minimum maintenance. The primary applications of these devices are examined in the case of the above-mentioned reactors. (author) [French] La Detection des Ruptures de Gaines (D. R. G.) est un aspect important de la securite des piles et plus particulierement des piles refroidies par un gaz. Une detection rapide et sure constitue aussi un element d'amelioration du rendement des centrales nucleaires productrices d'energie electrique. Parmi les nombreuses methodes de detection des ruptures de gaines, la mesure de la concentration dans le fluide caloporteur des gaz de fission a vie courte s'est revelee comme la plus sensible et la plus rapide. Une etude systematique des detecteurs a collection electrostatique des descendants des gaz de fission a ete entreprise en vue d'equiper les piles EL 2, G 3, EDF 1, EDF 2 et EDF 3, les boucles a gaz de la pile Pegase et la pile EL 4. Les divers parametres sont etudies en detail pour obtenir une sensibilite maximum et permettre la realisation de dispositifs de detection ayant le maximum de securite de

Studies of conceptual spheromak fusion reactors

International Nuclear Information System (INIS)

Katsurai, M.; Yamada, M.

1982-01-01

Preliminary design studies are carried out for a spheromak fusion reactor. Simplified circuit theory is applied to obtain the characteristic relations among various parameters of the spheromak configuration for an aspect ratio of A >or approx. 1.6. These relations are used to calculate the parameters for the conceptual designs of three types of fusion reactor: (1) the DT reactor with two-component-type operation, (2) the ignited DT reactor, and (3) the ignited catalysed-type DD reactor. With a total wall loading of approx. 4 MW.m -2 , it is found that edge magnetic fields of only approx. 4 T (DT) and approx. 9 T (Cat. DD) are required for ignited reactors of 1 m plasma (minor) radius with output powers in the gigawatt range. An assessment of various schemes of generation, compression and translation of spheromak plasmas is presented. (author)

Effect of the empty fraction in a solar reactor of fluidized bed; Efecto de la fraccion vacia en un reactor solar de lecho fluidizado

Energy Technology Data Exchange (ETDEWEB)

Torres, Alejandro; Romero-Paredes, Hernando; Vazquez, Alejandro; Torijano, Eugenio; Ambriz, Juan J [Universidad Autonoma Metropilitana-Iztapalapa, Mexico, D.F. (Mexico)

2000-07-01

The objective of this paper is to obtain the temperature profiles and concentration of a solar reactor of fluidized bed that simultaneously serves as a solar receiver of a thermo-chemical storage system of solar energy. The complex phenomena that are inherent of these reactors make their sizing difficult for their solar application. For this reason, to model and to simulate its behavior without and with the chemical reaction helps to palliate this disadvantage. One of the present phenomena is the change of the empty fraction in which we concentrate our attention. In this paper an alternative is proposed in the modeling of these systems, considering local fluctuations of the empty fraction or porosity {epsilon}(x,y) in the bed. For this a probabilistic uniform distribution is proposed for all the nodes of the (x, y) mesh of the bed where local values of porosity for each node of the mesh are associated by means of a random generator where {epsilon}(x, y){epsilon} [0.1]. The hollow fraction plays a very important role because the penetration of the solar radiation in these systems of opaque bodies depends directly on the distribution of empty spaces in the trajectory of the incident radiation that affects its thermal and kinetic behavior. From the results the characteristic of non- isothermicity of the reactor can be incorporated which entails, once reached the reaction temperature, to a dispersed profile of concentrations. The empty fraction is a parameter that influences greatly in these profiles and that increasing the fluidization number is the way this time is diminished. In conclusion, the importance the empty fraction plays in the evolution of the temperature profiles as well as in the concentration profiles is emphasized. The behavior of the bed in the simulation becomes more precise in agreement with the experimental results previously obtained. [Spanish] El objetivo de este trabajo es obtener los perfiles de temperatura y concentracion de un reactor solar

Simulacion borrosa de un reactor con reaccion exotermica no lineal

Directory of Open Access Journals (Sweden)

MIGUEL ANGEL RODRIGUEZ BORROTO

2007-01-01

Full Text Available En el presente trabajo se desarrolla un modelo difuso basado en la estructura Takagi-Sugeno-Kang dinámica para un reactor continuo de tanque con agitador (RCTA con reacción química de primer orden exotérmico. A partir de datos experimentales obtenidos mediante simulación del proceso real, se obtiene la base de datos de las variables de entrada y salida del proceso y a partir de la misma se elaboran los archivos de datos de entrenamiento y de verificación del modelo borroso el cual es obtenido mediante la herramienta anfis de MATLAB. El modelo obtenido permite predecir la salida del sistema con errores de predicción muy bajos, por lo que el mismo sienta las bases para el diseño de un controlador predictivo no lineal del mismo en próximas etapas de la investigación

Site Investigation for Detection of KIJANG Reactor Core Center

Energy Technology Data Exchange (ETDEWEB)

Kwon, Tae-Hyun; Kim, Jun Yeon; Kim, Jeeyoung [KAERI, Daejeon (Korea, Republic of)

2015-05-15

It was planned for the end of March 2017 and extended to April 2018 according to the government budget adjustment. The KJRR project is intended for filling the self-sufficiency of RI demand including Mo-99, increasing the NTD capacity and developing technologies related to the research reactor. In project, site investigation is the first activity that defines seismologic and related geologic aspects of the site. Site investigation was carried out from Oct. 2012 to Jan. 2014 and this study is intended to describe detail procedures in locating the reactor core center. The location of the reactor core center was determined by collectively reviewing not only geological information but also information from architects engineering. EL 50m was selected as ground level by levering construction cost. Four recommended locations (R-1a - R-1d) are displayed for the reactor core center. R-1a was found optimal in consideration of medium rock contour, portion of medium rock covering reactor buildings, construction cost, physical protection and electrical resistivity. It is noted that engineering properties of the medium rock is TCR/RQD 100/53, elastic modulus 7,710 - 8,720MPa, permeability coefficient 2.92E-06cm/s, and S-wave velocity 1,380m/s, sound for foundations of reactor buildings.

The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Yonggwang nuclear power plant unit 2

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

2001-02-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.762E+18, 1.5391E+19, 3.5119E+19, and 4.2610E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.899 for the 1st through 4th testing and the calculational uncertainty, 12.3% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.357E+19 n/cm{sup 2} based on the end of 11th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.525E+19, 4.337E+19, 5.148E+19 and 5.960E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 42 tabs. (Author)

Neutron Spectrometry Work at the CNEN; La spectrometrie neutronique au CNEN; Raboty po nejtronnoj spektrometrii v nkyaeh; Trabajos de espectrometria neutronica realizados en el CNEN

Energy Technology Data Exchange (ETDEWEB)

Caglioti, G; Ascarelli, P [Comitato Nazionale per L' energia Nucleare Euratom, G. C. R. Ispra, Varese (Italy)

1963-01-15

In this communication the main features of the triple axis spectrometer recently installed at the Ispra-1 reactor are presented. Preliminary results in the quasi elastic angular distribution of 1.4A neutrons scattered in liquid bromine are shown and briefly discussed. (author) [French] Dans ce memoire, l'auteur decrit les principales caracteristiques du spectrometre triaxial installe recemment au reacteur Ispra-1. Il donne et examine brievement les resultats preliminaires des experiences sur la distribution angulaire quasi elastique de neutrons de 1.4A a diffuses dans du brome liquide. (author) [Spanish] La memoria describe las principales caracteristicas del espectrometro triaxial recientemente instalado en el reactor lspra-1. Asimismo, expone brevemente los resultados iniciales de las determinaciones de distribucion angular de neutrones de 1,4 A dispersados en forma cuasi elastica por bromo liquido. (author)

Paraoxon induces apoptosis in EL4 cells via activation of mitochondrial pathways.

Science.gov (United States)

Saleh, A M; Vijayasarathy, C; Masoud, L; Kumar, L; Shahin, A; Kambal, A

2003-07-01

The toxicity of organophosphorus compounds, such as paraoxon (POX), is due to their anticholinesterase action. Recently, we have shown that, at noncholinergic doses (1 to 10 nM), POX (the bioactive metabolite of parathion) causes apoptotic cell death in murine EL4 T-lymphocytic leukemia cell line through activation of caspase-3. In this study, by employing caspase-specific inhibitors, we extend our observations to elucidate the sequence of events involved in POX-stimulated apoptosis. Pretreatment of EL4 cells with the caspase-9-specific inhibitor zLEHD-fmk attenuated POX-induced apoptosis in a dose-dependent manner, whereas the caspase-8 inhibitor zIETD-fmk had no effect. Furthermore, the activation of caspase-9, -8, and -3 in response to POX treatment was completely inhibited in the presence of zLEHD-fmk, implicating the involvement of caspase 9-dependent mitochondrial pathways in POX-stimulated apoptosis. Indeed, under both in vitro and in vivo conditions, POX triggered a dose- and time-dependent translocation of cytochrome c from mitochondria into the cytosol, as assessed by Western blot analysis. Investigation of the mechanism of cytochrome c release revealed that POX disrupted mitochondrial transmembrane potential. Neither this effect nor cytchrome c release was dependent on caspase activation, since the general inhibitor of the caspase family zVAD-fmk did not influence both processes. Finally, POX treatment also resulted in a time-dependent up-regulation and translocation of the proapoptotic molecule Bax to mitochondria. Inhibition of this event by zVAD-fmk suggests that the activation and translocation of Bax to mitochondria is subsequent to activation of the caspase cascades. The results indicate that POX induces apoptosis in EL4 cells through a direct effect on mitochondria by disrupting its transmembrane potential, causing the release of cytochrome c into the cytosol and subsequent activation of caspase-9. Inhibition of this specific pathway might provide

Directory of Nuclear Research Reactors 1994

International Nuclear Information System (INIS)

1995-08-01

The Directory of Nuclear Research Reactors is an output of the Agency's computerized Research Reactor Data Base (RRDB). It contains administrative, technical and utilization information on research reactors known to the Agency at the end of December 1994. The data base converted from mainframe to PC is written in Clipper 5.0 and the publication generation system uses Excel 4. The information was collected by the Agency through questionnaires sent to research reactor owners. All data on research reactors, training reactors, test reactors, prototype reactors and critical assemblies are stored in the RRDB. This system contains all the information and data previously published in the Agency's publication, Directory of Nuclear Research Reactor, as well as updated information

Hythane (H2 and CH4) production from unsaturated polyester resin wastewater contaminated by 1,4-dioxane and heavy metals via up-flow anaerobic self-separation gases reactor

International Nuclear Information System (INIS)

Mahmoud, Mohamed; Elreedy, Ahmed; Pascal, Peu; Sophie, Le Roux; Tawfik, Ahmed

2017-01-01

Highlights: • Bio-hythane production from polyester wastewater via UASG reactor was assessed. • Impacts of influent contamination by 1,4-dioxane and heavy metals were discussed. • Maximum volumetric H 2 and CH 4 productions of 0.12 and 1.06 L/L/d were achieved. • Significant drop in CH 4 production was resulted at OLR up to 1.07 ± 0.06 gCOD/L/d. • Bioenergy recovery through UASG economically achieved a net profit of 10,231 $/y. - Abstract: A long-term evaluation of hythane generation from unsaturated polyester resin wastewater contaminated by 1,4-dioxane and heavy metals was investigated in a continuous up-flow anaerobic self- separation gases (UASG) reactor inoculated with mixed culture. The reactor was operated at constant hydraulic retention time (HRT) of 96 h and different organic loading rates (OLRs) of 0.31 ± 0.04, 0.71 ± 0.08 and 1.07 ± 0.06 gCOD/L/d. Available data showed that volumetric hythane production rate was substantially increased from 0.093 ± 0.021 to 0.245 ± 0.016 L/L/d at increasing OLR from 0.31 ± 0.04 to 0.71 ± 0.08 gCOD/L/d. However, at OLR exceeding 1.07 ± 0.06 gCOD/L/d, it was dropped to 0.114 ± 0.016 L/L/d. The reactor achieved 1,4-dioxane removal efficiencies of 51.8 ± 2.8, 35.9 ± 1.6 and 26.3 ± 1.6% at initial 1,4-dioxane concentrations of 1.14 ± 0.28, 1.97 ± 0.41 and 4.21 ± 0.30 mg/L, respectively. Moreover, the effect and potential removal of the contaminated by heavy metals (i.e., Cu 2+ , Mn 2+ , Cr 3+ , Fe 3+ and Ni 2+ ) were highlighted. Kinetic modelling and microbial community dynamics were studied, according to each OLR, to carefully describe the UASG performance. The economic analysis showed a stable operation for the anaerobic digestion of unsaturated polyester resin wastewater using UASG, and the maximum net profit was achieved at OLR of 0.71 ± 0.08 gCOD/L/d.

Repairing liner of the reactor; Reparacion del liner del reactor

Energy Technology Data Exchange (ETDEWEB)

Aguilar H, F [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2001-07-15

Due to the corrosion problems of the aluminum coating of the reactor pool, a periodic inspections program by ultrasound to evaluate the advance grade and the corrosion speed was settled down. This inspections have shown the necessity to repair some areas, in those that the slimming is significant, of not making it can arrive to the water escape of the reactor pool. The objective of the repair is to place patches of plates of 1/4 inch aluminum thickness in the areas of the reactor 'liner', in those that it has been detected by ultrasound a smaller thickness or similar to 3 mm. To carry out this the fuels are move (of the core and those that are decaying) to a temporary storage, the structure of the core is confined in a tank that this placed inside the pool of the reactor, a shield is placed in the thermal column and it is completely extracted the water for to leave uncover the 'liner' of the reactor. (Author)

Diseño y simulación de un reactor prototipo que soporte condiciones de hidrogenación para crudo pesado con una capacidad de 1 galón por segundo para Petroecuador

OpenAIRE

Checa Ramírez, Pablo Andrés; Andy Cerda, Amilkar Patricio

2010-01-01

El presente proyecto se plantea en cinco capítulos estrictamente relacionados: Procesos teóricos, Tipos de reactores, Diseño y cálculo del reactor prototipo, Método del cálculo de prueba y error, y, costos que genera el diseño

TRATAMIENTO DE AGUAS RESIDUALES MEDIANTE REACTORES ANAERÓBICOS DE PLACAS VERTICALES PARALELAS EN ACRÍLICO TRATAMENTO DE ÁGUAS RESIDUÁRIAS POR REATORES ANAERÓBIOS DE PLACAS VERTICAIS PARALELAS EM ACRÍLICO WASTEWATER TREATMENT BY ANAEROBIC REACTORS OF VERTICAL PARALLEL PLATES IN ACRYLIC

Directory of Open Access Journals (Sweden)

Guillermo Chaux F

2011-12-01

Full Text Available Algunos filtros anaeróbicos con lecho de piedra construidos en el departamento del Cauca (Colombia, están presentando problemas de colmatación. Si se reemplaza la piedra por placas verticales paralelas, se elimina el problema de obstrucción. Este documento presenta el desarrollo y resultados de una investigación que evaluó a escala de laboratorio el potencial de los reactores anaeróbicos de placas verticales paralelas en acrílico para remover contaminantes (materia orgánica y sólidos suspendidos. El reactor anaeróbico de placas paralelas en acrílico se desempeñó como tratamiento secundario; se alimentó con agua residual efluente de un Tanque Imhoff con concentraciones medias de 156 ± 14 mg/L de DB05, 438 ± 32 mg/L de DQO y 98 ±22 mg/L de sólidos suspendidos totales. Las remociones de DQO y DB05 en el reactor sobrepasan el 50% y la remoción de sólidos suspendidos sobrepasó el 60% para tiempos de detención de 24 horas. La facilidad en la operación del reactor lo hace viable como tratamiento biológico anaeróbico de aguas residuales previamente decantadasAlguns filtros anaeróbios com recheio de pedras construída no departamento de Cauca (Colombia estão apresentando problemas de obstrução. Se a pedra é substituída por placas verticais paralelas, evita o problema da obstrução. Este artigo apresenta o desenvolvimento e os resultados e no estudo realizado em escala de laboratório que avaliaram o potencial de reatores anaeróbios de placas verticais paralelas em acrílico para remover os contaminantes (sólidos suspensos e matéria orgânica. 0 reator anaeróbio de placas paralelas de acrílico serviu como tratamento secundário; foi alimentado com água residuária do efluente de um tanque Imhoff com concentrações médias de 156 ± 14 mg/L DB05, 438 ± 32 mg/L de DQO e 98 ±22 mg/L de sólidos suspensos totais. A remoção de DQO e DB05 no reator são mais de 50% ea remoção de sólidos em suspensão superior a 60

The G4-ECONS Economic Evaluation Tool for Generation IV Reactor Systems and its Proposed Application to Deliberately Small Reactor Systems and Proposed New Nuclear Fuel Cycle Facilities. Annex IX

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-12-15

At the outset of the international Generation IV programme, it was decided that the six candidate reactor systems will ultimately be evaluated on the basis of safety, sustainability, non-proliferation attributes, technical readiness and projected economics. It is likely that the same factors will influence the evaluation of deliberately small reactor systems1 and new fuel cycle facilities, such as reprocessing plants that are being considered under the more recent Global Nuclear Energy Partnership (GNEP). This annex describes how the development of an economic modelling system has evolved to address the issue of economic competitiveness for both the Generation IV and GNEP programmes. In 2004, the Generation IV Economic Modelling Working Group (EMWG) commissioned the development of a Microsoft Excel based model capable of calculating the levelized unit electricity cost (LUEC) in mills/kW.h (1 mill = $10{sup -3}) or $/MW.h for multiple types of reactor system being developed under the Generation IV programme. This overall modelling system is now called the Generation IV spreadsheet calculation of nuclear systems (G4-ECONS), and is being expanded to calculate costs of energy products in addition to electricity, such as hydrogen and desalinated water. A version has also been developed to evaluate the costs of products or services from fuel cycle facilities. The cost estimating methodology and algorithms are explained in detail in the Generation IV Cost Estimating Guidelines and in the G4-ECONS User's Manual. The model was constructed with relatively simple economic algorithms such that it could be used by almost any nation without regard to country specific taxation, cost accounting, depreciation or capital cost recovery methodologies. It was also designed with transparency to the user in mind (i.e. all algorithms and cell contents are visible to the user). A short description of version 1.0 G4-ECONS-R (reactor economics model) has also been published in the

The G4-ECONS Economic Evaluation Tool for Generation IV Reactor Systems and its Proposed Application to Deliberately Small Reactor Systems and Proposed New Nuclear Fuel Cycle Facilities. Annex IX

International Nuclear Information System (INIS)

2013-01-01

At the outset of the international Generation IV programme, it was decided that the six candidate reactor systems will ultimately be evaluated on the basis of safety, sustainability, non-proliferation attributes, technical readiness and projected economics. It is likely that the same factors will influence the evaluation of deliberately small reactor systems1 and new fuel cycle facilities, such as reprocessing plants that are being considered under the more recent Global Nuclear Energy Partnership (GNEP). This annex describes how the development of an economic modelling system has evolved to address the issue of economic competitiveness for both the Generation IV and GNEP programmes. In 2004, the Generation IV Economic Modelling Working Group (EMWG) commissioned the development of a Microsoft Excel based model capable of calculating the levelized unit electricity cost (LUEC) in mills/kW.h (1 mill = $10 -3 ) or $/MW.h for multiple types of reactor system being developed under the Generation IV programme. This overall modelling system is now called the Generation IV spreadsheet calculation of nuclear systems (G4-ECONS), and is being expanded to calculate costs of energy products in addition to electricity, such as hydrogen and desalinated water. A version has also been developed to evaluate the costs of products or services from fuel cycle facilities. The cost estimating methodology and algorithms are explained in detail in the Generation IV Cost Estimating Guidelines and in the G4-ECONS User's Manual. The model was constructed with relatively simple economic algorithms such that it could be used by almost any nation without regard to country specific taxation, cost accounting, depreciation or capital cost recovery methodologies. It was also designed with transparency to the user in mind (i.e. all algorithms and cell contents are visible to the user). A short description of version 1.0 G4-ECONS-R (reactor economics model) has also been published in the Proceedings of

Elemental analysis of soil and plant samples at El-Manzala lake neutron activation analysis technique. Vol. 4

Energy Technology Data Exchange (ETDEWEB)

Eissa, E A; Rofail, N B; Hassan, A M [Reactor and Neutron Physics Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt); Abd El-Haleem, A S [Radiactive Environment Pollution Department, Hot Laboratories Center, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt); El-Abbady, W H [Phsics Department, Faculty of Science , Al-Azhar University, (Egypt)

1996-03-01

Soil and plant samples were collected from from two locations, Bahr El-Baker, and Bahr kados at the manzala lake, where, where high pollution is expected. The samples were especially treated and prepared for investigation by thermal neutron activation analysis (NAA). The irradiation facilities of the first egyptian research reactor (ET-R R-1), and the hyper pure germanium (HPGe) detection system were used for analysis. Among the 34 identified Fe, Co, As, Cd, Te, La, Sm, Rb, Hg, Th, and U are of a special significance because of the their toxic deleterious impact on living organisms. This work is part of a research project concerning pollution studies on the river Nile and some lakes of egypt. The data obtained in the present work stands as a reference basic record for any future follow up of contamination level. 1 tab.

Elemental analysis of soil and plant samples at El-Manzala lake neutron activation analysis technique. Vol. 4

International Nuclear Information System (INIS)

Eissa, E.A.; Rofail, N.B.; Hassan, A.M.; Abd El-Haleem, A.S.; El-Abbady, W.H.

1996-01-01

Soil and plant samples were collected from from two locations, Bahr El-Baker, and Bahr kados at the manzala lake, where, where high pollution is expected. The samples were especially treated and prepared for investigation by thermal neutron activation analysis (NAA). The irradiation facilities of the first egyptian research reactor (ET-R R-1), and the hyper pure germanium (HPGe) detection system were used for analysis. Among the 34 identified Fe, Co, As, Cd, Te, La, Sm, Rb, Hg, Th, and U are of a special significance because of the their toxic deleterious impact on living organisms. This work is part of a research project concerning pollution studies on the river Nile and some lakes of egypt. The data obtained in the present work stands as a reference basic record for any future follow up of contamination level. 1 tab

Reactor safety in Eastern Europe

International Nuclear Information System (INIS)

1995-02-01

The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. All papers are indexed separately in report GRS--117. (HP)

The Integral Fast Reactor

International Nuclear Information System (INIS)

Chang, Y.I.

1988-01-01

The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1989-01-01

The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs

Reactor Physics Development for Advanced Gas-Cooled Reactors; Recherches en Physique des Reacteurs, pour des Reacteurs Perfectionnes Refroidis par un Gaz; Razrabotka metodov v oblasti reaktornoj fiziki dlya usovershenstvovannogo reaktora s gazovym okhlazhdeniem; Progresos de la Fisica de los Reactores de Tipo Avanzado Refrigerados por Gas

Energy Technology Data Exchange (ETDEWEB)

Moore, J. [United Kingdom Atomic Energy Authority (United Kingdom)

1964-04-15

verifier les methodes theoriques elaborees pour etudier des coeurs de reacteurs heterogenes. Ces methodes theoriques, utilisees jusqu'a ce jour, sont connues sous les noms dr 'hetrecontrol' et de 'FTD2'. Les experiences avaient pour but de verifier dans le detail les caracteristiques de ces methodes; on a analyse les mesures faites sur plusieurs coeurs de 'reacteur' de differentes dimensions dans les installations APEX et HERO pour determiner une serie coherente de constantes de reseau concordant avec les resultats des experiences. A ces constantes purement empiriques, on a applique ensuite les methodes <> et 'FTD2' pour preparer la mise en service sans accord d'AGR et le choix du regime de chargement de ce reacteur. Le memoire enumere les techniques experimentales qui ont ete essayees et celles qui ont ete elaborees pour resoudre certains problemes qui se presentaient. Particulierement interessantes sont les methodes ayant pour but de mesurer les effets sur la reactivite dans les installations APEX, HERO et AGR, et de determiner les donnees relatives a la structure fine ainsi que la repartition de la puissance dans les assemblages complexes. Les recherches theoriques actuelles et futures sont axees principalement sur la mise au point d'une methode capable de remplacer 'hetrecontrol' et 'FTD2' pour les etudes sur des coeurs de reacteur apres qu'une bonne partie du combustible a 'brule'. Le programme d'experiences avec l'installation HERO a pour but de verifier ces methodes au moyen de coeurs complexes contenant du plutonium. On compte obtenir des renseignements supplementaires sur l'effet du plutonium au cours du fonctionnement d'AGR et a la suite de mesures de physique sur le combustible irradie. (author) [Spanish] La memoria describe los trabajos experimentales y teoricos que se han ejecutado durante el diseno, el desarrollo y la puesta en marcha del reactor AGR de Windscale y para facilitar el desarrollo de nuevos tipos de reactores refrigerados por gas

Factors affecting biological reduction of CO{sub 2} into CH{sub 4} using a hydrogenotrophic methanogen in a fixed bed reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Hyung; Pak, Daewon [Seoul National University of Science and Technology, Seoul (Korea, Republic of); Chang, Won Seok [Korea District Heating Corp, Seongnam (Korea, Republic of)

2015-10-15

Biological conversion of CO{sub 2} was examined in a fixed bed reactor inoculated with anaerobic mixed culture to investigate influencing factors, the type of packing material and the composition of the feeding gas mixture. During the operation of the fixed bed reactor by feeding the gas mixture (80% H{sub 2} and 20% CO{sub 2} based on volume basis), the volumetric CO{sub 2} conversion rate was higher in the fixed bed reactor packed with sponge due to its large surface area and high mass transfer from gas to liquid phase compared with PS ball. Carbon dioxide loaded into the fixed bed reactor was not completely converted because some of H{sub 2} was used for biomass growth. When a mole ratio of H{sub 2} to CO{sub 2} in the feeding gas mixture increased from 4 to 5, CO{sub 2} was completely converted into CH{sub 4}. The packing material with large surface area is effective in treating gaseous substrate such as CO{sub 2} and H{sub 2}. H{sub 2}, electron donor, should be providing more than required according to stoichiometry because some of it is used for biomass growth.

Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

Energy Technology Data Exchange (ETDEWEB)

Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

2013-10-15

The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

Twenty years of Radiology in RP-10 nuclear reactor protection; Veinte anos de proteccion radiologica en el reactor nuclear RP-10

Energy Technology Data Exchange (ETDEWEB)

Zapata, Alejandro L.; Ramos, Fernando T.; Arrieta, Rolando W.B.; Vela Mora, Mariano, E-mail: lzapata@ipen.gob.pe, E-mail: framos@ipen.gob.pe, E-mail: rarrieta@ipen.gob.pe, E-mail: mvela@ipen.gob.pe [Instituto Peruano de Energia Nuclear (IPEN), Lima (Peru)

2013-07-01

In this report we present the results about radiation controls during 1990 - 2010, carried out in the Nuclear Reactor RP-10 of the Nuclear Center of Huarangal. These controls and radiological evaluation are of much utility for the radio personnel protection of this one and other reactors, since it allows to compares these variables with respect to the time. From the results obtained in monitoring and radiation controls, we conclude that in no case it has been reached the limits allowed by the Peruvian Regulating Authority. (author)

Tritium release from lithium silicate and lithium aluminate, in-reactor and out-of-reactor

Energy Technology Data Exchange (ETDEWEB)

Johnson, A.B. Jr.

1976-09-01

Studies were conducted to determine the generation and evolution of tritium and helium in lithium aluminate (LiAlO/sub 2/) and lithium silicate (Li/sub 2/SiO/sub 3/) by the reaction: Li/sup 6/ + n ..-->.. /sup 4/He + T. Targets were irradiated 4.4 days in the K-West Reactor snout facility. (Silicate GVR* approximately 2.0 cc/cc; aluminate GVR approximately 1.4 cc/cc.) Gas release in-reactor was determined by post-irradiation drilling experiments on aluminum ampoules containing silicate and aluminate targets. In-reactor tritium release (at approximately 100/sup 0/C) was found to decrease linearly with increasing target density. Tritium released in-reactor was primarily in the noncondensible form (HT and T/sub 2/), while in laboratory extractions (300-1300/sup 0/C), the tritium appeared primarily in the condensible form (HTO and T/sub 2/O). Concentrations of HT (and presumably HTO) were relatively high, indicating moisture pickup in canning operations or by inleakage of moisture after the capsule was welded. Impurities in extracted gases included H/sub 2/O, CO/sub 2/, CO, O/sub 2/, H/sub 2/, NO, SO/sub 2/, SiF/sub 4/ and traces of hydrocarbons.

Uncertainty analysis of the 35% reactor inlet header break in a CANDU 6 reactor using RELAP/SCDAPSIM/MOD4.0 with integrated uncertainty analysis option

International Nuclear Information System (INIS)

Dupleac, D.; Perez, M.; Reventos, F.; Allison, C.

2011-01-01

The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology Software Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses the publicly available RELAP5 and SCDAP models in combination with (a) advanced programming and numerical techniques, (b) advanced SDTP-member-developed models for LWR, HWR, and research reactor analysis, and (c) a variety of other member-developed computational packages. One such computational package is an integrated uncertainty analysis (IUA) package being developed jointly by the Technical University of Catalonia (UPC) and Innovative Systems Software (ISS). RELAP/SCDAPSIM/MOD4.0(IUA) follows the input-propagation approach using probability distribution functions to define the uncertainty of the input parameters. The main steps for this type of methodologies, often referred as to statistical approaches or Wilks’ methods, are the ones that follow: 1. Selection of the plant; 2. Selection of the scenario; 3. Selection of the safety criteria; 4. Identification and ranking of the relevant phenomena based on the safety criteria; 5. Selection of the appropriate code parameters to represent those phenomena; 6. Association of uncertainty by means of Probability Distribution Functions (PDFs) for each selected parameter; 7. Random sampling of the selected parameters according to its PDF and performing multiple computer runs to obtain uncertainty bands with a certain percentile and confidence level; 8. Processing the results of the multiple computer runs to estimate the uncertainty bands for the computed quantities associated with the selected safety criteria. RELAP/SCDAPSIM/MOD4.0(IUA) calculates the number of required code runs given the desired percentile and confidence level, performs the sampling process for the

Uncertainty analysis of the 35% reactor inlet header break in a CANDU 6 reactor using RELAP/SCDAPSIM/MOD4.0 with integrated uncertainty analysis option

Energy Technology Data Exchange (ETDEWEB)

Dupleac, D., E-mail: danieldu@cne.pub.ro [Politehnica Univ. of Bucharest (Romania); Perez, M.; Reventos, F., E-mail: marina.perez@upc.edu, E-mail: francesc.reventos@upc.edu [Technical Univ. of Catalonia (Spain); Allison, C., E-mail: iss@cableone.net [Innovative Systems Software (United States)

2011-07-01

The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology Software Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses the publicly available RELAP5 and SCDAP models in combination with (a) advanced programming and numerical techniques, (b) advanced SDTP-member-developed models for LWR, HWR, and research reactor analysis, and (c) a variety of other member-developed computational packages. One such computational package is an integrated uncertainty analysis (IUA) package being developed jointly by the Technical University of Catalonia (UPC) and Innovative Systems Software (ISS). RELAP/SCDAPSIM/MOD4.0(IUA) follows the input-propagation approach using probability distribution functions to define the uncertainty of the input parameters. The main steps for this type of methodologies, often referred as to statistical approaches or Wilks’ methods, are the ones that follow: 1. Selection of the plant; 2. Selection of the scenario; 3. Selection of the safety criteria; 4. Identification and ranking of the relevant phenomena based on the safety criteria; 5. Selection of the appropriate code parameters to represent those phenomena; 6. Association of uncertainty by means of Probability Distribution Functions (PDFs) for each selected parameter; 7. Random sampling of the selected parameters according to its PDF and performing multiple computer runs to obtain uncertainty bands with a certain percentile and confidence level; 8. Processing the results of the multiple computer runs to estimate the uncertainty bands for the computed quantities associated with the selected safety criteria. RELAP/SCDAPSIM/MOD4.0(IUA) calculates the number of required code runs given the desired percentile and confidence level, performs the sampling process for the

PERFORMA NEUTRONIK BAHAN BAKAR LiF-BeF2-ThF4-UF4 PADA SMALL MOBILE-MOLTEN SALT REACTOR

Directory of Open Access Journals (Sweden)

S. N. Rokhman

2015-04-01

Full Text Available Telah dilakukan analisis terhadap performa neutronik bahan bakar garam lebur LiF-BeF2-ThF4-UF4 pada Small Mobile-Molten Salt Reactor (SM-MSR. Penyesuaian konfigurasi teras dan temperatur operasi harus dilakukan untuk penggunaan bahan bakar baru tersebut agar mencapai keff > 1 dan CR (conversion ratio > 1 pada fraksi 0,5% 233U, 20% 232Th, 28% Li, 51,5% Be. Setelah didapat nilai keff ≈ 1 dan CR ≈ 1, dilakukan analisis pengaruh perubahan Th terhadap Be dan Be terhadap Li yang terlihat dalam perubahan parameter keff dan CR. Setelah itu fraksi 233U divariasi antara 0,5–0,46% untuk memperoleh keff > 1 dan CR > 1. Dalam perhitungan koefisien reaktifitas temperatur (αT, temperatur teras dinaikkan sebesar +25K dan +50K., dan untuk koefisien reaktifitas void (αV, densitas bahan bakar dikurangi hingga 90%. Hasil perhitungan menunjukkan bahwa pengurangan Th terhadap Be menyebabkan penurunan nilai CR dan naiknya keff akibat berkurangnya material fertil. Sebaliknya penambahan Be terhadap Li mengakibatkan terjadi kenaikan nilai keff dan menurunkan CR, akibat laju serapan Li lebih besar dari Be. Pada 5 (lima fraksi 233U dalam rentang 0,5–0,49%, hasil perhitungan keff dan CR masing-masing bervariasi dalam rentang 1,00001 - 1,00327 dan 1,00016 - 1,00731. Untuk faktor puncak daya (PPF, hasil perhitungan memberikan nilai dalam rentang 2,4311 -2,4714. Sedangkan untuk parameter keselamatan, koefisien reaktivitas temperatur (αT dan reaktivitas void (αV masingmasing bernilai negatif dalam rentang 4,972×10-5 - 5,909×10-5 dan 2,596×10-2- 2,8287×10-2 ∆k/k/K. Dapat disimpulkan bahwa teras SM-MSR memberikan nilai negatif di kedua koefisien reaktivitas tersebut untuk setiap fraksi,, sehingga memenuhi kriteria keselamatan dan keselamatan melekat. Kata kunci: SM-MSR (small mobile-molten salt reactor, bahan bakar LiF-BeF2-ThF4-UF4, keselamatan melekat, koefisien reaktivitas temperatur, koefisien reaktivitas void   The analysis of neutronic performance has

Spatial kinetics in nuclear reactor systems. Chapter 4

International Nuclear Information System (INIS)

Owens, D.H.

1980-01-01

The problem of constructing a low-order linear lumped-parameter model of xenon-induced spatial power oscillations in a large, cylindrical nuclear power reactor to replace an (assumed known) nonlinear distributed parameter model is examined. Model expansion and finite difference methods are used together to provide a successful solution to the problem. (U.K.)

Synthesis of superior fast charging-discharging nano-LiFePO4/C from nano-FePO4 generated using a confined area impinging jet reactor approach.

Science.gov (United States)

Liu, Xiao-min; Yan, Pen; Xie, Yin-Yin; Yang, Hui; Shen, Xiao-dong; Ma, Zi-Feng

2013-06-14

LiFePO4/C nanocomposites with excellent electrochemical performance is synthesized from nano-FePO4, generated by a novel method using a confined area impinging jet reactor (CIJR). When discharged at 80 C (13.6 Ag(-1)), the LiFePO4/C delivers a discharge capacity of 95 mA h g(-1), an energy density of 227 W h kg(-1) and a power density of 34 kW kg(-1).

Nuclear data usage for research reactors

International Nuclear Information System (INIS)

Nakano, Yoshihiro; Soyama, Kazuhiko; Amano, Toshio

1996-01-01

In the department of research reactor, many neutronics calculations have been performed to construct, to operate and to modify research reactors of JAERI with several kinds of nuclear data libraries. This paper presents latest two neutronic analyses on research reactors. First one is design work of a low enriched uranium (LEU) fuel for JRR-4 (Japan Research Reactor No.4). The other is design of a uranium silicon dispersion type (silicide) fuel of JRR-3M (Japan Research Reactor No.3 Modified). Before starting the design work, to estimate the accuracy of computer code and calculation method, experimental data are calculated with several nuclear data libraries. From both cases of calculations, it is confirmed that JENDL-3.2 gives about 1 %Δk/k higher excess reactivity than JENDL-3.1. (author)

Linking of FRAP-T, FRAPCON and RELAP-4 codes for transient analysis and accidents of light water reactors fuel rods

International Nuclear Information System (INIS)

Marra Neto, A.; Silva, A.T. e; Sabundjian, G.; Freitas, R.L.; Neves Conti, T. das.

1991-09-01

The computer codes FRAP-T, FRAPCON and RELAP-4 have been linked for the fuel rod behavior analysis under transients and hypothetical accidents in light water reactors. The results calculated by thermal hydraulic code RELAP-4 give input in file format into the transient fuel analysis code FRAP-T. If the effect of fuel burnup is taken into account, the fuel performance code FRAPCON should provide the initial steady state data for thhe transient analysis. With the thermal hydraulic boundary conditions provided by RELAP-4 (MOD3), FRAP-T6 is used to analyse pressurized water reactor fuel rod behavior during the blowdown phase under large break loss of coolant accident conditions. Two cases have been analysed: without and with initialization from FRAPCON-2 steady state data. (author)

Proceedings of the 4th international symposium on material testing reactors

International Nuclear Information System (INIS)

Ishihara, Masahiro; Suzuki, Masahide

2012-03-01

This report is the Proceedings of the fourth International Symposium on Material Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The first symposium was held on 2008, at the Oarai Research and Development Center of JAEA, the second, 2009, Idaho National Laboratory (INL) of United States and the third 2010, Nuclear Research Institute (NRI) in Czech Republic to exchange information for deep mutual understanding of material testing reactors. The fourth symposium was originally scheduled to be held INVAP in Argentina. However, the aftermath of volcanic explosion at Chili forced the symposium to change place. Total 111 participants attended from Argentina, Belgium, France, Germany, Indonesia, Malasia, Korea, South Africa, Switzerland, the United State and Japan. This symposium addressed the general topics of 'status and future plan of material testing reactors', 'advancement of irradiation technology', 'expansion of industry use(RI)', 'facility, upgrade, aging management', 'new generation MTR', 'advancement of PIE technology', 'development of advanced driver fuel', and 'nuclear human resource development(HRD) for next generation', and 39 presentations were made. Furthermore, three topics, 'Necessity of cooperation for Mo-99 production by (n,gamma) reaction', 'Necessity of standardization of irradiation technology' and 'Conceptual design of next generation materials testing reactor by collaboration', were selected and discussed. (author)

Proceedings of the 4th international symposium on material testing reactors

Energy Technology Data Exchange (ETDEWEB)

Ishihara, Masahiro; Suzuki, Masahide [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

2012-03-15

This report is the Proceedings of the fourth International Symposium on Material Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The first symposium was held on 2008, at the Oarai Research and Development Center of JAEA, the second, 2009, Idaho National Laboratory (INL) of United States and the third 2010, Nuclear Research Institute (NRI) in Czech Republic to exchange information for deep mutual understanding of material testing reactors. The fourth symposium was originally scheduled to be held INVAP in Argentina. However, the aftermath of volcanic explosion at Chili forced the symposium to change place. Total 111 participants attended from Argentina, Belgium, France, Germany, Indonesia, Malasia, Korea, South Africa, Switzerland, the United State and Japan. This symposium addressed the general topics of 'status and future plan of material testing reactors', 'advancement of irradiation technology', 'expansion of industry use(RI)', 'facility, upgrade, aging management', 'new generation MTR', 'advancement of PIE technology', 'development of advanced driver fuel', and 'nuclear human resource development(HRD) for next generation', and 39 presentations were made. Furthermore, three topics, 'Necessity of cooperation for Mo-99 production by (n,gamma) reaction', 'Necessity of standardization of irradiation technology' and 'Conceptual design of next generation materials testing reactor by collaboration', were selected and discussed. (author)

Analysis of gamma dose for 4,8 gU/cm3 density silicide core at the RSG-GAS reactor using MCNP code

International Nuclear Information System (INIS)

Ardani

2011-01-01

Radiation safety analysis should be done following of substitution of fuel density of 2.96 gU/cc to density of 4,8 gU/cc silicide fuels for the RSG-GAS reactor. MCNP-5 code has been used to perform gamma dose calculation of the RSG-GAS reactor. Gamma radiation source at reactor consists of capture gamma rays, prompt fission gamma rays, and gamma rays of decay of fission and activation products. The strength of the prompt fission gamma rays is obtained by gamma releases of fission process of U-235 and reactor power of 30 MWt., during 46,6 days operation. Radiation dose is calculated at the experimental hall by detection point at the surface of outer of biological shielding and the operation hall by detection point at the top of the pool. The calculation is conducted at reactor on the normal operation and on the worst postulated accident causing the water level at the pool decreases. Calculation result shows that the biggest source strength of gamma rays come from the decay process. The highest calculated dose at the experiment hall is 4,07x10 -3 μSv/h, far from the maximum external dose permitted 25 μSv/h. The highest calculated dose at the operation hall is 19.98 μSv/h. Even though the calculated dose is still acceptable but this is close to the maximum permitted dose for worker. It concluded that loading of 4,8 gU/cc silicide fuel for the RSG-GAS still safe. (author)

Pebble bed reactors simulation using MCNP: The Chinese HTR-10 reactor

Directory of Open Access Journals (Sweden)

SA Hosseini

2013-09-01

Full Text Available   Given the role of Gas-Graphite reactors as the fourth generation reactors and their recently renewed importance, in 2002 the IAEA proposed a set of Benchmarking problems. In this work, we propose a model both efficient in time and resources and exact to simulate the HTR-10 reactor using MCNP-4C code. During the present work, all of the pressing factors in PBM reactor design such as the inter-pebble leakage, fuel particle distribution and fuel pebble packing fraction effects have been taken into account to obtain an exact and easy to run model. Finally, the comparison between the results of the present work and other calculations made at INEEL proves the exactness of the proposed model.

Indium-Gallium Radiation Contour of the IRT Nuclear Reactor; Circuit d'activation d'indium-gallium dans le reacteur nucleaire IRT; Indij-gallievyj radiatsionnyj kontur yadernogo reaktora IRT; Circuito de radiaciones de indio-galio del reactor IRT

Energy Technology Data Exchange (ETDEWEB)

Breger, A K; Ryabukin, Y S; Tulkes, S G; Volkov, E N

1960-07-15

-industrielles. (author) [Spanish] Basandose en un trabajo teorico ya publicado, se preparo en el reactor IRT un circuito de radiaciones de indio-galio que constituye una nueva fuente de rayos gamma de elevada intensidad. El primer circuito de este tipo ''RK-1'' se construyo para el reactor IRT en el Instituto de Fisica de la Academia de Ciencias de la Republica Socialista Sovietica de Georgia. Este trabajo estudia los puntos siguientes: calculo de la activacion de la desintegracion del conjunto indio-galio; estructura del circuito RK-1 y su disposicion en el tanque del reactor y en la camara activa; dispositivo de admision de las sustancias liquidas y gaseosas en la zona de la irradiacion; transportador de las sustancias solidas sometidas a irradiacion. En el reactor IRT, cuya potencia es de 2 000 kW, la intensidad de irradiacion del circuito es igual a la de una fuente de radiacion gamma equivalente a 20 000 g de radio. En el trabajo se estudian las posibilidades de utilizacion de este circuito con fines semi-industriales y de investigacion. (author) [Russian] Osnovyvayas' na uzhe opublikovannoj teoreticheskoj rabote, byl podgotovlen indij-gallievyj radiatsionnyj kontur yadernogo reaktora IRT, kotoryj yavlyaetsya novym moshchnym istochnikom gamma-oblucheniya . Pervyj kontur ehtogo tipa RK-1 byl podgotovlen na reaktore IRT v Institute fiziki Akademii nauk Gruzinskoj SSR. V doklade dayutsya raschety aktivizatsij dlya indij-gallievogo splava, strukturnye kompanovki RK-1 i ikh raspolozhenie v bake reaktora i goryachej kamere, ustrojstvo podachi zhidkikh i gazoobraznykh veshchestv v zonu oblucheniya i konvejer dlya tverdykh veshchestv, kotorye podlezhat oblucheniyu. V reaktore IRT moshchnost'yu 2000 kW moshchnost' oblucheniya kontura ehkvivalentna moshchnosti oblucheniya gamma-izluchatelya, obladayushchego aktivnost'yu v 20000 g ehkv. radiya. V doklade obsuzhdayutsya perspektivy ispol'zovaniya indij-gallievogo radiatsionnogo kontura dlya issledovatel'skikh i polupromyshlennykh tselej

Differential immunotoxic effects of ethanol on murine EL-4 lymphoma and normal lymphocytes is mediated through increased ROS production and activation of p38MAPK.

Science.gov (United States)

Premachandran, Sudha; Khan, Nazir M; Thakur, Vikas S; Shukla, Jyoti; Poduval, T B

2012-08-01

Ethanol has been used to achieve thymic depletion in myasthenia gravis patients. Ethanol (95%) has also been used widely in the therapy of many tumors including hepatocellular carcinoma. In light of these findings, we delineated the differential immunotoxic behavior and mechanism of lower concentration of ethanol towards murine EL-4 lymphoma and its normal counterpart lymphocytes. EL-4 lymphoma and normal lymphocytes were cultured with ethanol (0%-5%) for 6 h and cytotoxicity was measured by various methods. EL-4 cells treated with ethanol showed concentration-dependent loss of viability at 2%-5% ethanol concentration and exhibit proliferative arrest at preG1 stage. Acridine-orange and ethidium-bromide staining indicated that ethanol induced death in EL-4 cells, by induction of both apoptosis and necrosis which was further supported by findings of DNA-fragmentation and trypan blue dye exclusion test. However, treatment of lymphocytes with similar concentration of ethanol did not show any death-associated parameters. Furthermore, ethanol induced significantly higher ROS generation in EL-4 cells as compared to lymphocytes and caused PARP cleavage and activation of apoptotic proteins like p53 and Bax, in EL-4 cells and not in normal lymphocytes. In addition, ethanol exposure to EL-4 cells led to phosphorylation of p38MAPK, and upregulation of death receptor Fas (CD95). Taken together, these results suggest that ethanol upto a concentration of 5% caused no significant immunotoxicity towards normal lymphocytes and induced cell death in EL-4 cells via phosphorylation of p38MAPK and regulation of p53 leading to further activation of both extrinsic (Fas) and intrinsic (Bax) apoptotic markers.

Biodegradation of 2,4,6-trichlorophenol in a packed-bed biofilm reactor equipped with an internal net draft tube riser for aeration and liquid circulation

Energy Technology Data Exchange (ETDEWEB)

Gomez-De Jesus, A.; Romano-Baez, F.J.; Leyva-Amezcua, L.; Juarez-Ramirez, C.; Ruiz-Ordaz, N. [Departamento de Ingenieria Bioquimica, Escuela Nacional de Ciencias Biologicas, IPN. Prol. Carpio y Plan de Ayala, Colonia Santo Tomas, s/n. CP 11340, Mexico, D.F. (Mexico); Galindez-Mayer, J. [Departamento de Ingenieria Bioquimica, Escuela Nacional de Ciencias Biologicas, IPN. Prol. Carpio y Plan de Ayala, Colonia Santo Tomas, s/n. CP 11340, Mexico, D.F. (Mexico)], E-mail: cmayer@encb.ipn.mx

2009-01-30

For the aerobic biodegradation of the fungicide and defoliant 2,4,6-trichlorophenol (2,4,6-TCP), a bench-scale packed-bed bioreactor equipped with a net draft tube riser for liquid circulation and oxygenation (PB-ALR) was constructed. To obtain a high packed-bed volume relative to the whole bioreactor volume, a high A{sub D}/A{sub R} ratio was used. Reactor's downcomer was packed with a porous support of volcanic stone fragments. PB-ALR hydrodynamics and oxygen mass transfer behavior was evaluated and compared to the observed behavior of the unpacked reactor operating as an internal airlift reactor (ALR). Overall gas holdup values {epsilon}{sub G}, and zonal oxygen mass transfer coefficients determined at various airflow rates in the PB-ALR, were higher than those obtained with the ALR. When comparing mixing time values obtained in both cases, a slight increment in mixing time was observed when reactor was operated as a PB-ALR. By using a mixed microbial community, the biofilm reactor was used to evaluate the aerobic biodegradation of 2,4,6-TCP. Three bacterial strains identified as Burkholderia sp., Burkholderia kururiensis and Stenotrophomonas sp. constituted the microbial consortium able to cometabolically degrade the 2,4,6-TCP, using phenol as primary substrate. This consortium removed 100% of phenol and near 99% of 2,4,6-TCP. Mineralization and dehalogenation of 2,4,6-TCP was evidenced by high COD removal efficiencies ({approx}95%), and by the stoichiometric release of chloride ions from the halogenated compound ({approx}80%). Finally, it was observed that the microbial consortium was also capable to metabolize 2,4,6-TCP without phenol as primary substrate, with high removal efficiencies (near 100% for 2,4,6-TCP, 92% for COD and 88% for chloride ions)

The liquid hydrogen cell in the EL3 Saclay reactor; Cellule a hydrogene liquide dans la pile EL3 de Saclay

Energy Technology Data Exchange (ETDEWEB)

Jacrot, B; Lacaze, A; Weil, L [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; [Grenoble-1 Univ., 38 (France)

1960-07-01

Description and results in connection with the liquid hydrogen cell, for obtaining slow neutrons, in the EL3. (author) [French] Description et resultats concernant la cellule a hydrogene liquide de EL3 utilisee pour obtenir des neutrons lents. (auteur)

Radiation protection at the RA Reactor in 1993, RA research reactor, Part

International Nuclear Information System (INIS)

Ninkovic, M.; Pavlovic, R.; Mandic, M.; Sipka, V.; Grsic, Z.

1993-01-01

Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry and radiation protection at the RA reactor; (2) decontamination, collecting and treatment of fluid effluents and solid wastes; (3) Radioactivity control in the vicinity of the reactor and (4)meteorology measurements; (3). Each of the category is described as a separate annex of this report [sr

ESTUDIO DEL EFECTO DE LA ISOTERMA EN EL PROCESO DE SÍNTESIS DEL COMPUESTO DE ESTRONCIO Sr4Al6O12SO4 POR REACCIÓN EN ESTADO SÓLIDO

Directory of Open Access Journals (Sweden)

José Rodríguez-García

2012-01-01

Full Text Available Se estudió el efecto de la isoterma en la síntesis del compuesto de estroncio Sr 4 Al 6 O 12 SO 4 por reacción en estado sólido. Se conformaron (100MPa gránulos a partir de la mezcla 3:3:1 molar de SrCO 3 , Al 2 O 3 y SrSO 4 , respectivamente. Las muestras fueron tratadas térmicamente a temperaturas de 800, 900, 1000, 1100 y 1200°C por periodos de 4, 6, 8 y 10 horas. Las muestras, en presentación de polvo y pastilla, fueron analizadas por ATG, ATD, DRX y MEB, desde temperatura ambiente ha sta los 1200°C. Los resultad os de análisis térmico señalan que el rango de formación del compuesto de estroncio Sr 4 Al 6 O 12 SO 4 es entre los 800 y 1200°C. Por otro lado, de acuerdo a los patrones de difr acción de rayos X, la formación del mismo se favorece al incrementarse la isoterma del tratamiento térmico. Existen dos fases meta-estables presentes durante el proceso antes mencionado, las mismas que son observadas en el MEB.

Effect of ionizing radiation in combination with 5-flurouracil on cell cycle uncoupling of EL-4 cell line

International Nuclear Information System (INIS)

Liu Yang; Sun Yanhong; Zhang Xuan; Gong Shouliang; Zhang Wei; Li Song

2009-01-01

Objective: To observe the dose-and time-effect of ionizing radiation in combination with 5-flurouracil(5-FU) on the cell cycle uncoupling of EL-4 cell line. Methods: EL-4 cells were collected after irradiation with 0,1.0,2.0 and 4.0 Gy X-irradiation and treatment with 5-FU(0.001,0.010,0.100 and 1.000 mg·L -1 ) for 0,4,8,16,24 and 48 h.The regularity in the polyloid cells was analyzed by flow cytometry(FCM) following staining cells with propidium iodide(PI). Results: As compared with sham-irradiation group,the percentage of diploid EL-4 cells increased significantly at 8-24 h and returned to normal level at 48 h after irradiation with 2.0 Gy X-rays(P -1 group, the percentage of diploid cells decreased obviously at 16-48 h after treatment with 0.100 mg·L -1 5-FU(P -1 group, the percentage of diploid cells decreased significantly 16 h after treatment with different doses 5-FU(P -1 ; the percentage of octoploid cells increased significantly after treatment with 0.010 and 0.100 mg·L -1 5-Fu(P -1 5-FU. (authors)

Determination of the flows profile in the role of power in the central thimble of TRIGA Mark III Reactor; Determinacion del perfil de flujos en funcion de la potencia en el dedal central del Reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Garcia F, A.

2010-07-01

The overall objective of the thesis project is to determine the flow profiles sub cadmic and epi cadmic in the central thimble to different powers and operation times of TRIGA Mark III Reactor, using activation foils as detectors. In the reactor operation, it is necessary to know the neutron flow profile for to realize other tasks as: the radioisotopes production, research in reactors physics and fuel burning. The distribution of the neutron flow, accurately reflects what is happening in the reactor core, plus the flows value in this distribution is directly related to the power generated. For this reason it is performed the sub cadmic flow measurement with energies between 0 and 0.4 eV (energy of the cadmium cut E{sub cd} approx 0.4 eV) and epi cadmic flow with energies greater than 0.4 eV, in the central thimble powers to the powers of 10, 100 W, 1, 10 100 Kw and 1 MW. The method used is known as flakes activation, which is to be arranged by placing flakes ( 3 mm of diameter and 0.0508 mm of thickness) of a given material (either Au, In, Cu, Mn, etc.) into an aluminum tube outside diameter equal to 6.35 mm, alternating flakes with lids covered and discovered of cadmium (3.4 mm of diameter and 0.508 mm of thickness) and separated by lucite pieces of 3 mm of diameter and 25.4 mm in length. After irradiating the flakes for some time, is measured the gamma activity of each of them, using a hyper pure germanium detector of high resolution. Already known gamma activity, proceed to calculate the epi cadmic and sub cadmic flows using a computer program in Fortran language, called Caflu. (Author)

Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (Capsule 2) of Ulchin Nuclear Power Plant Unit 4

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

2007-04-15

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin Unit 4 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.306E+18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.918 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.615E+18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 8.478E+18 and 1.673E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin Unit 4 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

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

Nuclear reactors to come

International Nuclear Information System (INIS)

Lung, M.

2002-01-01

The demand for nuclear energy will continue to grow at least till 2050 because of mainly 6 reasons: 1) the steady increase of the world population, 2) China, India and Indonesia will reach higher social standard and their energy consumption will consequently grow, 3) fossil energy resources are dwindling, 4) coal will be little by little banned because of its major contribution to the emission of green house effect gas, 5) renewable energies need important technological jumps to be really efficient and to take the lead, and 6) fusion energy is not yet ready to take over. All these reasons draw a promising future for nuclear energy. Today 450 nuclear reactors are operating throughout the world producing 17% of the total electrical power demand. In order to benefit fully of this future, nuclear industry has to improve some characteristics of reactors: 1) a more efficient use of uranium (it means higher burnups), 2) a simplification and automation of reprocessing-recycling chain of processes, 3) efficient measures against proliferation and against any misuse for terrorist purposes, and 4) an enhancement of safety for the next generation of reactors. The characteristics of fast reactors and of high-temperature reactors will likely make these kinds of reactors the best tools for energy production in the second half of this century. (A.C.)

Remoción de fósforo de diferentes aguas residuales en reactores aeróbios de lecho fluidizado trifásico con circulación interna

Directory of Open Access Journals (Sweden)

Gleyce Teixeira Correia

2013-01-01

Full Text Available El vertimiento de aguas residuales (AR produce impactos sobre los cuerpos de agua receptores. Nutrientes como P generan implicaciones en los sistemas lénticos pues aceleran los procesos de eutrofización. Se han utilizado diversas tecnologías para la remoción de P de las AR: sistemas de tratamiento físico químico con importantes efectos por adición de productos coagulantes; procesos biológicos basados en alternancia de condiciones anaerobias y aerobias con importantes implicaciones de volumen necesario; sistemas como lagunas de estabilización e irrigación requieren de áreas muy considerables y procesos de postratamiento. Los reactores aerobios de lecho fluidizado con circulación interna (RALFCI son opciones compactas que utilizan gran concentración de biomasa activa que han demostrado su capacidad para remover materia orgánica y N. Para AR domésticas provenientes de la estación de bombeo de Ilha Solteira y para los efluentes de un sistema de recirculación acuícola (SRA de cultivo semi-intensivo de tilapia se evaluó la eficiencia de remoción de P reactivo y P total en tres tipos de RALFCI con diámetro externo de 250 mm y diferentes diámetros de tubo interno (DTI, con dos medios de soporte y diferentes concentraciones en dos de los reactores. Las eficiencias medias de remoción de P reactivo en AR domésticas para un tiempo de retención hidráulica (TRH de 3 horas en el reactor con DTI 125 mm variaron entre 25,6 y 38,4% y en el reactor con DTI 150 mm entre 27,5 y 32,5%; la remoción de P total en el SRA para un TRH de 0,19 h y DTI 100 mm fue de 32,7%.

Experimental neutronic science and instrumentation: from hybrid reactors to fourth generation reactors

International Nuclear Information System (INIS)

Jammes, Ch.

2010-07-01

After an overview of his academic career and scientific and research activities, the author proposes a rather detailed synthesis and overview of his scientific activities in the fields of cross sections and Doppler effect (development and validation of a code), on the MUSE-4 hybrid reactor (experiments, static and dynamic measurements), on the TRADE hybrid reactor (experimental means, sub-critical reactivity measurement), on the RACE hybrid reactor (experimental results, modelling and interpretation), and on neutron detection (design and modelling of fission chamber, on-line measurement of the fast flow). The next part gives an overview of some research programs (neutron monitoring in sodium-cool fast reactors, research and development on fission chambers, improvement of effective delayed neutron measurements)

Mechanism of phorbol ester-mediated protein kinase C activation in EL4 thymoma cells

International Nuclear Information System (INIS)

Huang, F.L.; Arora, P.K.; Hanna, E.E.; Huang, K.P.

1987-01-01

Mouse thymoma EL4 cells respond to phorbol 12-myristate 13-acetate (PMA) in interleukin-2 secretion and growth inhibition. A rapid translocation of protein kinase C (PKC) from cytosol to the particulate fraction and followed by proteolytic degradation occur when EL4 cells are incubated with PMA. In the present study the translocated membrane-associated PKC (PP-PKC) was solubilized by buffer containing NP-40 and its behavior on column chromatography, molecular weight, and kinetic properties were compared to the cytosolic PKC (CS-PKC) from untreated cells. From DE-52 cellulose column, CS-PKC could be eluted by buffer containing 0.1 M KCl, whereas PP-PKC was eluted with buffer containing 0.25 M KCl and 0.2% NP-40. On gel filtration the partially purified PP-PKC from DE-52 was separated into two species: a high Mr species, which was a complex of 82KDa PKC, PMA, and lipid as evidenced by immunoblot analysis and labeling with [ 3 H]PMA and [ 3 H]myristic acid, and a 82KDa species, which was free of PMA and lipid. This 82KDa PP-PKC, though similar to the CS-PKC in molecular weight, is distinguishable from the CS-PKC in having lower Ka values for both Ca 2+ and PS and no longer requires diacylglycerol for maximum activation. These results indicate that upon PMA treatment of EL4 cells, the CS-PKC was modified through enhancing the kinase activity and affinity for membrane lipid. The modification results in the translocation and complexing of PKC with membrane lipid and PMA and subsequent degradation

The integral fast reactor concept

International Nuclear Information System (INIS)

Chang, Yoon I.; Marchaterre, J.F.

1987-01-01

The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) an integral fuel cycle, based on pyrometallurgical processing and injection-cast fuel fabrication, with the fuel cycle facility collocated with the reactor, if so desired. This paper gives a review of the IFR concept

Reactor safety in Eastern Europe. Proceedings

International Nuclear Information System (INIS)

1995-02-01

The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. (HP) [de

Study of the origin of elements of the uranium-235 family observed in excess in the vicinity of the experimental nuclear EL4 reactor under dismantling. Lessons got at this day and conclusions; Etude de l'origine des elements de la famille de l'uranium-235 observes en exces dans les environs du reacteur nucleaire experimental EL4 en cours de demantelement. Enseignements retires a ce jour et conclusion

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

This study resumes the discovery of an excess of actinium 227 found around by EL4 nuclear reactor actually in dismantling. The search for the origin of this excess revealed a real inquiry of investigation during three years. Because a nuclear reactor existed in this area a particular attention will have concerned this region. The doubt became the line of conduct to find the answer to the human or natural origin of this excess. Finally and against any evidence, it appears that the origin of this phenomenon was natural, consequence of the particular local geology. The detail of the different investigations is given: search of a possible correlation with the composition of elevations constituent of lanes, search (and underlining) of new sites in the surroundings of the Rusquec pond and the Plouenez station, study of the atmospheric deposits under winds of the nuclear power plant and in the east direction, search of a possible relationship with the gaseous effluents of the nuclear power plant in the past, historical study of radioactive effluents releases in the fifty last years by the analysis of the sedimentary deposits in the Saint-Herbiot reservoir, search of a possible correlation between the excess of actinium 227 and the nuclear power plant activity; search of a possible correlation with a human activity without any relationship with the nuclear activities, search of a correlation with the underground waters, search of a correlation with the geological context, collect of information on the possible transfers in direction of the food chain, determination of the radiological composition of the underground waters ( not perturbed by human activity), search of the cause of an excess of actinium 227 in the old channel of liquid effluents release of the nuclear power plant. The results are given and discussed. And contrary to all expectations the origin of the excess of actinium 227 is completely natural. (N.C.)

The physics design of EBR-II; Physique du reacteur EBR-II; Fizicheskij raschet ehksperimental'nogo reaktora - razmnozhitelya EVR-II; Aspectos fisicos del reactor EBR-II

Energy Technology Data Exchange (ETDEWEB)

Loewenstein, W. B. [Argonne National Laboratory, Argonne, IL (United States)

1962-03-15

notamment l'introduction de taux de reactivite normaux et anormaux, les consequences des effets supposes de reactivite, a partir du comportement physique de l'alliage combustible et de la structure du reacteur, ainsi que par extrapolation des experiences faites sur TREAT au systeme EBR-II. Il examine le probleme de la fusion du coeur de EBR-II. (author) [Spanish] La memoria informa sobre los calculos del comportamiento estatico, dinamico y a largo plazo de la reactividad del reactor reproductor experimental EBR-II, asi como sobre los resultados y analisis de los experimentos criticos en seco del EBR-II y de los experimentos simulados en el reactor de potencia cero ZPR-III. Insiste particularmente en los problemas de fisica del reactor que, en la elaboracion del proyecto, siguen a la eleccion del modelo, pero preceden a la construccion y puesta en marcha del reactor. Presenta diversos analisis del reactor desde el punto de vista de la seguridad y formula consideraciones sobre la evaluacion de los riesgos y su influencia sobre el diseno del reactor. El trabajo explica tambien la manera de emplear los datos obtenidos en los experimentos arriba citados. Estos experimentos, su analisis y sus predicciones teoricas constituyen la base para determinar el comportamiento fisico del reactor. La memoria estudia detalladamente las limitaciones inherentes a la aplicacion de los datos experimentales al funcionamiento del reactor de potencia. Ello incluye datos precisos sobre as dimensiones del cuerpo, el enriquecimiento de la aleacion combustible, o de ambos factores; el establecimiento de una reactividad adecuada para el reactor funcionando o detenido, la determinacion de la variacion de los coeficientes de reactividad en funcion de la temperatura de funcionamiento y de la potencia generadora y detalles de la distribucion de la potencia y del flujo en diversos puntos de la estructura del reactor. La memoria expone tambien el problema general que supone transferir a la verdadera geometria

Status of French reactors

International Nuclear Information System (INIS)

Ballagny, A.

1997-01-01

The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm 3 . The OSIRIS reactor has already been converted to LEU. It will use U 3 Si 2 as soon as its present stock of UO 2 fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU

El CRIT: el renacer Pijao

Directory of Open Access Journals (Sweden)

Daniel Santiago Roldán Zarazo

2016-01-01

Full Text Available El artículo reconstruye el periodo histórico previo a la creación del Consejo Regional Indígena del Tolima (CRIT, organización que agrupa una gran variedad de cabildos y colectivos principalmente del Sur del Tolima. Así, desde el marco de coyunturas y estructuras se desarrolla un análisis documental y de prensa que evidencian los imaginarios, disputas y enfrentamientos de los indígenas en la región. El periodo se divide en cuatro momentos: (1 las condiciones estructurales, políticas y económicas que atravesaba el país y los indígenas en la época; posteriormente (2, se concentra en quienes le conformaban, sus posibilidades y restricciones de movilización; en tercera medida (3 se exponen los debates, marcos de acción, estructuras de movilización y las conclusiones llegadas en los encuentros del movimiento; por último (4 se recogen los elementos generales que arroja el periodo estudiado junto con elementos de discusión y aportes a la teoría e historia política.

Fusion reactor materials

International Nuclear Information System (INIS)

Anon.

1977-01-01

The following topics are briefly discussed: (1) surface blistering studies on fusion reactor materials, (2) TFTR design support activities, (3) analysis of samples bombarded in-situ in PLT, (4) chemical sputtering effects, (5) modeling of surface behavior, (6) ion migration in glow discharge tube cathodes, (7) alloy development for irradiation performance, (8) dosimetry and damage analysis, and (9) development of tritium migration in fusion devices and reactors

Repairing liner of the reactor; Reparacion del liner del reactor

Energy Technology Data Exchange (ETDEWEB)

Aguilar H, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2001-07-15

Due to the corrosion problems of the aluminum coating of the reactor pool, a periodic inspections program by ultrasound to evaluate the advance grade and the corrosion speed was settled down. This inspections have shown the necessity to repair some areas, in those that the slimming is significant, of not making it can arrive to the water escape of the reactor pool. The objective of the repair is to place patches of plates of 1/4 inch aluminum thickness in the areas of the reactor 'liner', in those that it has been detected by ultrasound a smaller thickness or similar to 3 mm. To carry out this the fuels are move (of the core and those that are decaying) to a temporary storage, the structure of the core is confined in a tank that this placed inside the pool of the reactor, a shield is placed in the thermal column and it is completely extracted the water for to leave uncover the 'liner' of the reactor. (Author)

Mirror hybrid reactor optimization studies

International Nuclear Information System (INIS)

Bender, D.J.

1976-01-01

A system model of the mirror hybrid reactor has been developed. The major components of the model include (1) the reactor description, (2) a capital cost analysis, (3) various fuel management schemes, and (4) an economic analysis that includes the hybrid plus its associated fission burner reactors. The results presented describe the optimization of the mirror hybrid reactor, the objective being to minimize the cost of electricity from the hybrid fission-burner reactor complex. We have examined hybrid reactors with two types of blankets, one containing natural uranium, the other thorium. The major difference between the two optimized reactors is that the uranium hybrid is a significant net electrical power producer, whereas the thorium hybrid just about breaks even on electrical power. Our projected costs for fissile fuel production are approximately 50 $/g for 239 Pu and approximately 125 $/g for 233 U

Preparation of mandatory documentation before the start up of the RA-0 `zero power` nuclear reactor at Cordoba National University; Preparacion de la documentacion mandatoria para la puesta en marcha del reactor nuclear RA-0 en la Universidad Nacional de Cordoba

Energy Technology Data Exchange (ETDEWEB)

Martin, H R; Keil, W M; Pezzi, N

1992-12-31

Before the start up of the RA-0 `zero power` nuclear reactor installed at Cordoba National University, it was necessary to send to the Regulatory Authority the mandatory documentation which is required in the licensing process. With the previous papers existing for the operation in the first years of the `70, a work program for the future operational training personnel was elaborated. Based on the Authority`s applicable rules and the recommendations and with particular criteria originated in the working university conditions, the SAFETY report of RA-0 nuclear reactor was prepared. This paper describes the principal contents, items and documents involved in the safety report. (Author). [Espanol] Con motivo de la nueva puesta en servicio del REACTOR NUCLEAR RA-0 fue necesario elaborar la documentacion mandatoria requerida por la Autoridad Regulatoria Nacional. Siguiendo los lineamientos de las normas y recomendaciones vigentes e incluyendo criterios propios en lo que debia ser el contenido final de dicha documentacion, fue preparado lo que se ha denominado el INFORME DE SEGURIDAD DEL REACTOR NUCLEAR RA-0. Este documento que se describe en este trabajo, si bien contiene las habituales descripciones de todos los Informes de Seguridad, incluye otros aspectos que no siendo requeridos expresamente en el mismo, han dado una mayor coherencia a la conformacion de todos los aspectos que interrelacionan las areas de seguridad fisica, radiologica, nuclear y de control de materiales nucleares bajo salvaguardias. (Autor).

Neutrino-4 experiment on the search for a sterile neutrino at the SM-3 reactor

Energy Technology Data Exchange (ETDEWEB)

Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Chernyi, A. V.; Zherebtsov, O. M. [National Research Centre “Kurchatov Institute,”, Konstantinov Petersburg Nuclear Physics Institute (Russian Federation); Martemyanov, V. P.; Tsinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I. [National Research Centre “Kurchatov Institute,” (Russian Federation); Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K. [State Scientific Centre Research Institute of Atomic Reactors (Russian Federation); and others

2015-10-15

In view of the possibility of the existence of a sterile neutrino, test measurements of the dependence of the reactor antineutrino flux on the distance from the reactor core has been performed on SM-2 reactor with the Neutrino-2 detector model in the range of 6–11 m. Prospects of the search for reactor antineutrinos at short distances have been discussed.

Evaluation of the Cytotoxic Effect of the Brittle Star (Ophiocoma Erinaceus) Dichloromethane Extract and Doxorubicin on EL4 Cell Line.

Science.gov (United States)

Afzali, Mahbubeh; Baharara, Javad; Nezhad Shahrokhabadi, Khadijeh; Amini, Elaheh

2017-01-01

Leukemia is a blood disease that creates from inhibition of differentiation and increased proliferation rate. The nature has been known as a rich source of medically useful substances. High diversity of bioactive molecules, extracted from marine invertebrates, makes them as ideal candidates for cancer research. The study has been done to investigate cytotoxic effects of dichloromethane brittle star extract and doxorubicin on EL4 cancer cells. Blood cancer EL4 cells were cultured and treated at different concentrations of brittle star ( Ophiocoma erinaceus ) dichloromethane extract at 24, 48 and 72 h. Cell toxicity was studied using MTT assay. Cell morphology was examined using an invert microscope. Further, apoptosis was examined using Annexin V-FITC, propodium iodide, DAPI, and Acridine orange/propodium iodide staining. Eventually, the apoptosis pathways were analyzed using measurement of Caspase-3 and -9 activity. The statistical analysis was performed using SPSS, ANOVA software, and Tukey's test. P EL4 proliferation as IC 50 =32 µg/mL. All experiments related to apoptosis analysis confirmed that dichloromethane brittle star extract and doxorubicin have a cytotoxic effect on EL4 cells inIC 50 concentration. The study showed that dichloromethane brittle star extract is as an adjunct to doxorubicin in treatment of leukemia cells.

Detection of the contamination of air by tritiated water vapour around the reactor EL3

International Nuclear Information System (INIS)

Lebouleux, P.; Ardellier, A.; Valero, S.

1968-01-01

The authors describe the apparatus used for the detection of the tritiated water vapour contamination in the air around the reactor EL 3. The apparatus consists of two air-circulation ionisation chambers; the air in one of these is dried by passage through a silica-gel column. By carrying out a differential measurement of the ionization currents, it is possible to measure the tritiated water vapour concentration. A theoretical study of the response of the chambers is carried out for two types of emission of the tritiated water vapour: continuous, or in bursts. The experimental work comprises: calibration in the measurement range employed; study of the selectivity for other active gases; study of typical accidents; the interpretation of the results in the case of discontinuous emission, taking into account the desorption from the walls of the measurement chamber, a phenomenon which is observed during the emptying process. The authors give finally actual examples of how to use the results. The apparatus built makes it possible to detect, in less than ten minutes, contamination by tritiated water vapour in the presence of other active gases, in a measurement range of between 3 and 2200 MPC, and with an accuracy of about 25 per cent. A transposition to calculations of the risk to workers should be made with the utmost caution; an envelope of this risk can be drawn up more or less accurately depending on particular cases. (authors) [fr

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Martinc, R.; Cupac, S.; Stanic, A.

1990-01-01

RA reactor was not operated during the past five years due to the renewal and reconstruction of the reactor systems, which in underway. In the period from 1986-1990, reactor was operated only 144 MWh in 1986, for the need of testing the reactor systems and possibility of irradiating 125 I. Reactor will not be operated in 1991 because of the exchange of complete instrumentation which is planned to be finished by the end of 1991. It is expected to start operation in May 1992. That is why this annex includes the plan of reactor operation for period of nine months starting from from the moment of start-up. It is planned to operate the reactor at 0.02 MW power first three months, to increase the power gradually and reach 3.5 MW after 8 months of operation. It is foreseen to operate the reactor at 4.7 MW from the tenth month on [sr

Basic study on apoptosis induction into cancer cells U-937 and EL-4 by ultrasound exposure.

Science.gov (United States)

Takeuchi, Shinichi; Udagawa, Yoshiko; Oku, Yumiko; Fujii, Takuma; Nishimura, Hiroyuki; Kawashima, Norimichi

2006-12-22

Recently, the low invasive cancer treatments with small aftereffects have been considered. We are studying on the suppression methods of cancer cell proliferation with ultrasound. Cancer cells of mouse T lymphoma (EL-4) have been used in our study. The human histitocytic lymphoma cells (U-937) was used in this time. The cancer cells were cultured in a culture medium of RPMI1640. The standing wave acoustic field was formed in a water tank of our ultrasound exposure system by a vibrating plate driven with a Langevine type transducer. The U-937 and EL-4 were exposed to ultrasound in the acoustic field with spatial average acoustic intensity of 350 mW/cm(2) at 150 kHz. The viable rate of EL-4 decreased with the lapse of culture time after ultrasound exposure. U-937 did not show the remarkable decrease tendency. The proliferation of U-937 which exposed to ultrasound with 700 mW/cm(2) was suppressed. It can be thought that apoptosis was induced in the cancer cells in this condition. We observed the morphological change on the U-937 exposed to ultrasound with this condition. The morphological changes by apoptosis like the shrink of cells, formation of apoptotic bodies etc. can be observed with an optical microscope and a phase contrast microscope.

Reactor core structure

International Nuclear Information System (INIS)

Higashinakagawa, Emiko; Sato, Kanemitsu.

1992-01-01

Taking notice on the fact that Fe based alloys and Ni based alloys are corrosion resistant in a special atmosphere of a nuclear reactor, Fe or Ni based alloys are applied to reactor core structural components such as fuel cladding tubes, fuel channels, spacers, etc. On the other hand, the neutron absorption cross section of zirconium is 0.18 barn while that of iron is 2.52 barn and that of nickel is 4.6 barn, which amounts to 14 to 25 times compared with that of zirconium. Accordingly, if the reactor core structural components are constituted by the Fe or Ni based alloys, neutron economy is lowered. Since it is desirable that neutrons contribute to uranium fission with least absorption to the reactor core structural components, the reactor core structural components are constituted with the Fe or Ni based alloys of good corrosion resistance only at a portion in contact with reactor water, that is, at a surface portion, while the main body is constituted with zircalloy in the present invention. Accordingly, corrosion resistnace can be kept while keeping small neutron absorption cross section. (T.M.)