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Sample records for full reactor e-n

  1. Neutronics and thermohydraulics of the reactor C.E.N.E.-Part I; Analisis neutronico y termohidraulico del reactor C.E.N.E. Parte I

    Caro, R; Ahnert, C; Naudin, A E; Martinez Fanegas, R; Minguez, E; Rovira, A

    1976-07-01

    In this report the analysis of neutronics (both statics and kinetics), of the 10 MWt swimming pool reactor C.E.N.E, is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking, carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.

  2. Neutronics and thermohydraulics of the reactor C.E.N.E. Part II; Analisis neutronico y termohidraulico del reactor C.E.N.E. Parte II

    Caro, R

    1976-07-01

    In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.

  3. Neutronics and thermohydraulics of the reactor C.E.N.E. Pt. 1

    Caro, R.; Ahnert, C.; Esteban Naudin, A.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

    1976-01-01

    The analysis of neutronics (both statics and kinetics), of the 10 Mwt swimming pool reactor C.E.N.E. is included. A short description of the theoretical model used, along with the theoretical versus experimental cheking, carried out, whenever possible, with the reactors JEN-1 and JEN-2 of Junta de Energia Nuclear, is given in each of these chapters. (author) [es

  4. Neutronics and thermohydraulics of the reactor C.E.N.E. Part II

    Caro, R.

    1976-01-01

    In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs

  5. Neutronics and thermohydraulics of the reactor C.E.N.E.-Part I

    Caro, R.; Ahnert, C.; Naudin, A. E.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

    1976-01-01

    In this report the analysis of neutronics (both statics and kinetics), of the 10 MWt swimming pool reactor C.E.N.E, is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking, carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs

  6. Operation experience and maintenance at the TRIGA Mark II L.E.N.A. reactor

    Gngoli, F.; Berzero, A.; Lana, F.; Rosti, G.; Meloni, S.

    2008-01-01

    The TRIGA Mark II reactor of the University of Pavia was operated in the last two years on a routine basis, mostly for neutron activation analysis purposes. Moreover the reactor was completely shutdown in the first six months of this year to allow the dismantling of the NADIR experimental setup. The paper presents: - Reactor operation from July 1990 to June 1992; - Reactor users in the time period January 1990 - December 1991; - Specific activities of some radionuclides in the filling materials; - Specific activity of some radionuclides in thermal column materials. Operations related to dismantling of NADIR experimental facility are described. Finally the new thermal column configuration is presented. Starting from the end inside the reactor tank, a graphite layer (35 cm thick) was positioned, followed by a bismuth layer (10 cm thick) to reduce gamma-ray intensity. The old graphite rods were then positioned leaving in the central part, on the equatorial plane of the thermal column, a cavity whose vertical section has 40 cm width and 20 cm height. The bottom of the cavity, towards to the reactor tank, has been lined with additional layers of graphite (10 cm), bismuth (10 cm) and again graphite (1 cm). The new configuration allowed new experiments to be performed. The cavity in the central part has been created to allow the irradiation of large biological samples such as experimental animal and human livers. This is a peculiar step in a neutron capture boron therapy project to be carried out at the University of Pavia. In order to avoid an implemented 41 Ar production in the void space between shutters and the thermal column outer end, the external surface of the thermal column has been coated with boral sheets. The neutron flux profile, both thermal and epithermal, and cadmium ratio for gold are shown. The flux distribution appears to be adequate to proceed with the neutron capture boron therapy experiment. The LENA Health Physics Service has checked all phases of

  7. Density variations in a reactor during liquid full dimerization

    Golombok, M.; Bruijn, J.

    2000-01-01

    In a liquid full plug flow reactor during lower olefin dimerization, the assumption of constant density is not valid—the volume of a plug changes as it proceeds along the reactor. The observed kinetics depend on the density variation in the reactor as the conversion proceeds towards a distribution

  8. Tokamak fusion reactors with less than full tritium breeding

    Evans, K. Jr.; Gilligan, J.G.; Jung, J.

    1983-05-01

    A study of commercial, tokamak fusion reactors with tritium concentrations and tritium breeding ratios ranging from full deuterium-tritium operation to operation with no tritium breeding is presented. The design basis for these reactors is similar to those of STARFIRE and WILDCAT. Optimum operating temperatures, sizes, toroidal field strengths, and blanket/shield configurations are determined for a sequence of reactor designs spanning the range of tritium breeding, each having the same values of beta, thermal power, and first-wall heat load. Additional reactor parameters, tritium inventories and throughputs, and detailed costs are calculated for each reactor design. The disadvantages, advantages, implications, and ramifications of tritium-depleted operation are presented and discussed

  9. Full-scale leaching study of commercial reactor waste forms

    Kalb, P.D.; Colombo, P.

    1984-01-01

    This paper describes a full-scale leaching experiment which has been conducted at Brookhaven National Laboratory (BNL) to study the release of radionuclides from actual commercial reactor waste forms. While many studies characterizing the leaching behavior of simulated laboratory-scale waste forms have been performed, this program represents one of the first attempts in the United States to quantify activity releases for real, full-scale waste forms. 5 references, 5 figures, 1 table

  10. Feasibility study of full-reactor gas core demonstration test

    Kunze, J. F.; Lofthouse, J. H.; Shaffer, C. J.; Macbeth, P. J.

    1973-01-01

    Separate studies of nuclear criticality, flow patterns, and thermodynamics for the gas core reactor concept have all given positive indications of its feasibility. However, before serious design for a full scale gas core application can be made, feasibility must be shown for operation with full interaction of the nuclear, thermal, and hydraulic effects. A minimum sized, and hence minimum expense, test arrangement is considered for a full gas core configuration. It is shown that the hydrogen coolant scattering effects dominate the nuclear considerations at elevated temperatures. A cavity diameter of somewhat larger than 4 ft (122 cm) will be needed if temperatures high enough to vaporize uranium are to be achieved.

  11. Substitution of the old console with a fully computerized one. Major maintenance and enhancements of the ventilation and air conditioning system of the reactor building. Research activities at the TRIGA L.E.N.A. plant in Pavia, Italy

    Orvini, E.; Piazzoli, A.; Borio, A.

    2008-01-01

    pulse power monitor, the new ICS will improve the capability for the pulsing operational mode of the reactor. The new ACFS is also operated under a computer control. The technical specifications of the new system are basically the same of the old one, but for a double air filtering in suction and emission from the reactor building, instead of a simple filtering of the external air. The design philosophy of the new system assures a fairly good filtering of the external air inlet (>95%) and a very good filtering of the air released in the atmosphere, thanks to absolute filters with an efficiency higher that 99.95%. This new solution goes definitively in the direction of increasing the safety level of the reactor for what concerns the potential consequences on the population due to an accidental release of radioactive products in the atmosphere. Both the upgrading activities required to undergo the evaluation of the Italian National Regulatory Commission for the reactor licence modification. The following data are supplied in 8 tables, as follows: - Operation Data (July 1998 - June 2000); - Number of Irradiation and of Irradiated Samples in Reactor Facilities; - Users of the L.E.N.A. Reactor Irradiation Facilities; - Activities at the L.E.N.A. Reactor; - Utilization Data by different Users; - Utilization Data for different Activities; - Survey of Fuel Elements elongation (1965 - 2000); - Fuel Elements of L.E.N.A. TRIGA Reactor

  12. Fort St. Vrain reactor performance and operation to full power

    Simon, W.A.; Bramblett, G.C.

    1982-01-01

    The Fort St. Vrain Nuclear Generating Station, powered by a high-temperature gas-cooled reactor (HTGR), has now been tested to full thermal power. Testing was conducted for the dual purposes of demonstrating component and system capability as a part of the rise-to-power program and determining core fluctuation/redistribution behavior under full power conditions. Both objectives were met. Full power performance of all major components and the achievement of nearly all design objectives has been verified. In addition, the tests showed that the fluctuation phenomenon has been corrected. Core region outlet temperature redistributions have been characterized, related to a physical mechanism, and shown to be inconsequential for overall plant operation

  13. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    Nash, C.A.; Blake, J.E.; Rush, G.C.

    1990-01-01

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m 2 ) (1.1E+6 BTU/(ft 2 hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient

  14. Full-Scale Continuous Mini-Reactor Setup for Heterogeneous Grignard Alkylation of a Pharmaceutical Intermediate

    Pedersen, Michael Jønch; Holm, Thomas; Rahbek, Jesper P.

    2013-01-01

    A reactor setup consisting of two reactors in series has been implemented for a full-scale, heterogeneous Grignard alkylation. Solutions pass from a small filter reactor into a static mixer reactor with multiple side entries, thus combining continuous stirred tank reactor (CSTR) and plug flow...

  15. Development of a full scope reactor engineering simulator

    Venhuizen, J.R.; Laats, E.T.

    1988-01-01

    An engineering laboratory is pursuing the development of an engineering simulator for use by several agencies of the U.S. Government. According to the authors, this simulator will provide the highest fidelity simulation with initial objectives for studying augmented nuclear reactor operator training, and later for advanced concepts testing as applicable to control room accident diagnosis and management

  16. Full reactor coolant system chemical decontamination qualification programs

    Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1995-03-01

    Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

  17. Performance evaluation of full scale UASB reactor in treating stillage wastewater

    A.Mirsepasi , H. R. Honary , A. R. Mesdaghinia, A. H. Mahvi , H. Vahid , H. Karyab

    2006-04-01

    Full Text Available Upflow anaerobic sludge blanket (UASB reactors have been widely used for treatment of industrial wastewater. In this study two full-scale UASB reactors were investigated. Volume of each reactor was 420 m3. Conventional parameters such as pH, temperature and efficiency of COD, BOD, TOC removal in each reactor were investigated. Also several initial parameters in designing and operating of UASB reactors, such as upflow velocity, organic loading rate (OLR and hydraulic retention time were investigated. After modifying in operation conditions in UASB-2 reactor, average COD removal efficiency at OLR of 10–11 kg COD / m3 day was 55 percent. In order to prevent solids from settling, upflow velocity was increased to 0.35 m/h. Also to prevent solids from settling, the hydraulic retention time of wastewater in UASB-2 reactor was increased from 200 to 20 hours. This was expected that with good operation of UASB-2 reactor and with expanding of granules in the bed of the reactor, COD removal efficiency will be increased to more than 80 percent. But, because of deficiency on granulation and operation in UASB-2 reactor, this was not achieved. COD removal efficiency in the UASB-1 reactor was little. To enhance COD efficiency of UASB-1 reactor, several parameters were needed to be changed. These changes included enhancing of OLRs and upflow velocity, decreasing hydraulic retention time and operating with new sludge.

  18. Full core reactor analysis: Running Denovo on Jaguar

    Jarrell, J. J.; Godfrey, A. T.; Evans, T. M.; Davidson, G. G. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States)

    2012-07-01

    Fully-consistent, full-core, 3D, deterministic neutron transport simulations using the orthogonal mesh code Denovo were run on the massively parallel computing architecture Jaguar XT5. Using energy and spatial parallelization schemes, Denovo was able to efficiently scale to more than 160 k processors. Cell-homogenized cross sections were used with step-characteristics, linear-discontinuous finite element, and trilinear-discontinuous finite element spatial methods. It was determined that using the finite element methods gave considerably more accurate eigenvalue solutions for large-aspect ratio meshes than using step-characteristics. (authors)

  19. Characterization of the TRIGA Mark II reactor full-power steady state

    Cammi, Antonio, E-mail: antonio.cammi@polimi.it [Politecnico di Milano – Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via La Masa 34, 20156 Milano (Italy); Zanetti, Matteo [Politecnico di Milano – Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via La Masa 34, 20156 Milano (Italy); Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica [University of Milano-Bicocca, Physics Department “G. Occhialini” and INFN Section, Piazza dell’Ateneo Nuovo, 20126 Milan (Italy); Magrotti, Giovanni; Prata, Michele; Salvini, Andrea [University of Pavia, Applied Nuclear Energy Laboratory (L.E.N.A.), Via Gaspare Aselli 41, 27100 Pavia (Italy)

    2016-04-15

    Highlights: • Full-power steady state characterization of the TRIGA Mark II reactor. • Monte Carlo and Multiphysics simulation of the TRIGA Mark II reactor. • Sub-cooled boiling effects in the TRIGA Mark II reactor. • Thermal feedback effects in the TRIGA Mark II reactor. • Experimental data based validation. - Abstract: In this paper, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor at the University of Pavia is achieved by coupling the Monte Carlo (MC) simulation for neutronics with the “Multiphysics” model for thermal-hydraulics. Neutronic analyses have been carried out with a MCNP5 based MC model of the entire reactor system, already validated in fresh fuel and zero-power configurations (in which thermal effects are negligible) and using all available experimental data as a benchmark. In order to describe the full-power reactor configuration, the temperature distribution in the core must be established. To evaluate this, a thermal-hydraulic model has been developed, using the power distribution results from the MC simulation as input. The thermal-hydraulic model is focused on the core active region and takes into account sub-cooled boiling effects present at full reactor power. The obtained temperature distribution is then entered into the MC model and a benchmark analysis is carried out to validate the model in fresh fuel and full-power configurations. An acceptable correspondence between experimental data and simulation results concerning full-power reactor criticality proves the reliability of the adopted methodology of analysis, both from the perspective of neutronics and thermal-hydraulics.

  20. Design of a reactor core in the Oma Full MOX-ABWR

    Hama, Teruo

    1999-01-01

    The Electric Power Development Co., Ltd. has progressed a construction plan on an improved boiling-water reactor aiming at loading of MOX fuel in all reactor cores (full MOX-ABWR) at Oma-cho, Aomori prefecture, which is a last stage on application of approval on establishment at present. Here were described on outlines of reactor core in the full MOX-ABWR and its safety evaluation. For the full MOX-ABWR loading MOX fuel assembly into all reactor core, thermal and mechanical design analysis of fuel bars and core design analysis were conducted. As a result, it was confirmed that judgement standards in mixed core of MOX fuel and uranium fuel were also applicable as well as that in uranium fuel. (G.K.)

  1. Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

    Devan, K.; Bachchan, Abhitab; Riyas, A.; Sathiyasheela, T.; Mohanakrishnan, P.; Chetal, S.C.

    2011-01-01

    Highlights: → In this study we made physics designs of experimental metal fast reactor cores. → Aim is for full-scale demonstration of fuel assemblies in a commercial power reactor. → Minimum power with adequate safety is considered. → In addition, fuel sustainability is also considered in the design. → Sodium bonded U-Pu-6%Zr and mechanically bonded U-Pu alloys are used. - Abstract: Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.

  2. Searching for full power control rod patterns in a boiling water reactor using genetic algorithms

    Montes, Jose Luis [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jlmt@nuclear.inin.mx; Ortiz, Juan Jose [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jjortiz@nuclear.inin.mx; Requena, Ignacio [Departamento Ciencias Computacion e I.A. ETSII, Informatica, Universidad de Granada, C. Daniel Saucedo Aranda s/n. 18071 Granada (Spain)]. E-mail: requena@decsai.ugr.es; Perusquia, Raul [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: rpc@nuclear.inin.mx

    2004-11-01

    One of the most important questions related to both safety and economic aspects in a nuclear power reactor operation, is without any doubt its reactivity control. During normal operation of a boiling water reactor, the reactivity control of its core is strongly determined by control rods patterns efficiency. In this paper, GACRP system is proposed based on the concepts of genetic algorithms for full power control rod patterns search. This system was carried out using LVNPP transition cycle characteristics, being applied too to an equilibrium cycle. Several operation scenarios, including core water flow variation throughout the cycle and different target axial power distributions, are considered. Genetic algorithm fitness function includes reactor security parameters, such as MLHGR, MCPR, reactor k{sub eff} and axial power density.

  3. Safety assessment to support NUE fuel full core implementation in CANDU reactors

    Fan, H.Z.; Laurie, T.; Siddiqi, A.; Li, Z.P.; Rouben, D.; Zhu, W.; Lau, V.; Cottrell, C.M. [CANDU Energy Inc., Mississauga, Ontario (Canada)

    2013-07-01

    The Natural Uranium Equivalent (NUE) fuel contains a combination of recycled uranium and depleted uranium, in such a manner that the resulting mixture is similar to the natural uranium currently used in CANDU® reactors. Based on successful preliminary results of 24 bundles of NUE fuel demonstration irradiation in Qinshan CANDU 6 Unit 1, the NUE full core implementation program has been developed in cooperation with the Third Qinshan Nuclear Power Company and Candu Energy Inc, which has recently received Chinese government policy and funding support from their National-Level Energy Innovation program. This paper presents the safety assessment results to technically support NUE fuel full core implementation in CANDU reactors. (author)

  4. Microbial diversity in a full-scale anaerobic reactor treating high ...

    Microbial characteristics in the up-flow anaerobic sludge blanket reactor (UASB) of a full-scale high concentration cassava alcohol wastewater plant capable of anaerobic hydrocarbon removal were analyzed using cultivation-independent molecular methods. Forty-five bacterial operational taxonomic units (OTUs) and 24 ...

  5. Safety And Transient Analyses For Full Core Conversion Of The Dalat Nuclear Research Reactor

    Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

    2011-01-01

    Preparing for full core conversion of Dalat Nuclear Research Reactor (DNRR), safety and transient analyses were carried out to confirm about ability to operate safely of proposed Low Enriched Uranium (LEU) working core. The initial LEU core consisting 92 LEU fuel assemblies and 12 Beryllium rods was analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and fuel cladding fail. Working LEU core response were evaluated under these initial events based on RELAP/Mod3.2 computer code and other supported codes like ORIGEN, MCNP and MACCS2. Obtained results showed that safety of the reactor is maintained for all transients/accidents analyzed. (author)

  6. Full Core modeling techniques for research reactors with irregular geometries using Serpent and PARCS applied to the CROCUS reactor

    Siefman, Daniel J.; Girardin, Gaëtan; Rais, Adolfo; Pautz, Andreas; Hursin, Mathieu

    2015-01-01

    Highlights: • Modeling of research reactors. • Serpent and PARCS coupling. • Lattice physics codes modeling techniques. - Abstract: This paper summarizes the results of modeling methodologies developed for the zero-power (100 W) teaching and research reactor CROCUS located in the Laboratory for Reactor Physics and Systems Behavior (LRS) at the Swiss Federal Institute of Technology in Lausanne (EPFL). The study gives evidence that the Monte Carlo code Serpent can be used effectively as a lattice physics tool for small reactors. CROCUS’ core has an irregular geometry with two fuel zones of different lattice pitches. This and the reactor’s small size necessitate the use of nonstandard cross-section homogenization techniques when modeling the full core with a 3D nodal diffusion code (e.g. PARCS). The primary goal of this work is the development of these techniques for steady-state neutronics and future transient neutronics analyses of not only CROCUS, but research reactors in general. In addition, the modeling methods can provide useful insight for analyzing small modular reactor concepts based on light water technology. Static computational models of CROCUS with the codes Serpent and MCNP5 are presented and methodologies are analyzed for using Serpent and SerpentXS to prepare macroscopic homogenized group cross-sections for a pin-by-pin model of CROCUS with PARCS. The most accurate homogenization scheme lead to a difference in terms of k eff of 385 pcm between the Serpent and PARCS model, while the MCNP5 and Serpent models differed in terms of k eff by 13 pcm (within the statistical error of each simulation). Comparisons of the axial power profiles between the Serpent model as a reference and a set of PARCS models using different homogenization techniques showed a consistent root-mean-square deviation of ∼8%, indicating that the differences are not due to the homogenization technique but rather arise from the definition of the diffusion coefficients

  7. Methodology used to calculate moderator-system heat load at full power and during reactor transients in CANDU reactors

    Aydogdu, K.

    1998-01-01

    Nine components determine the moderator-system heat load during full-power operation and during a reactor power transient in a CANDU reactor. The components that contribute to the total moderator-system heat load at any time consist of the heat generated in the calandria tubes, guide tubes and reactivity mechanisms, moderator and reflector; the heat transferred from calandria shell, the inner tubesheets and the fuel channels; and the heat gained from moderator pumps and heat lost from piping. The contributions from each of these components will vary with time during a reactor transient. The sources of heat that arise from the deposition of nuclear energy can be divided into two categories, viz., a) the neutronic component (which is directly proportional to neutronic power), which includes neutron energy absorption, prompt-fission gamma absorption and capture gamma absorption; and b) the fission-product decay-gamma component, which also varies with time after initiation of the transient. An equation was derived to calculate transient heat loads to the moderator. The equation includes two independent variables that are the neutronic power and fission-product decay-gamma power fractions during the transient and a constant term that represents the heat gained from moderator pumps and heat lost from piping. The calculated heat load in the moderator during steady-state full-power operation for a CANDU 6 reactor was compared with available measurements from the Point Lepreau, Wolsong 1 and Gentilly-2 nuclear generating stations. The calculated and measured values were in reasonably good agreement. (author)

  8. Microbial Community Composition and Ultrastructure of Granules from a Full-Scale Anammox Reactor

    Gonzalez-Gil, Graciela

    2014-12-11

    Granules in anammox reactors contain besides anammox bacteria other microbial communities whose identity and relationship with the anammox bacteria are not well understood. High calcium concentrations are often supplied to anammox reactors to obtain sufficient bacterial aggregation and biomass retention. The aim of this study was to provide the first characterization of bacterial and archaeal communities in anammox granules from a full-scale anammox reactor and to explore on the possible role of calcium in such aggregates. High magnification imaging using backscattered electrons revealed that anammox bacteria may be embedded in calcium phosphate precipitates. Pyrosequencing of 16S rRNA gene fragments showed, besides anammox bacteria (Brocadiacea, 32 %), substantial numbers of heterotrophic bacteria Ignavibacteriacea (18 %) and Anaerolinea (7 %) along with heterotrophic denitrifiers Rhodocyclacea (9 %), Comamonadacea (3 %), and Shewanellacea (3 %) in the granules. It is hypothesized that these bacteria may form a network in which heterotrophic denitrifiers cooperate to achieve a well-functioning denitrification system as they can utilize the nitrate intrinsically produced by the anammox reaction. This network may provide a niche for the proliferation of archaea. Hydrogenotrophic methananogens, which scavenge the key fermentation product H2, were the most abundant archaea detected. Cells resembling the polygon-shaped denitrifying methanotroph Candidatus Methylomirabilis oxyfera were observed by electron microscopy. It is hypothesized that the anammox process in a full-scale reactor triggers various reactions overall leading to efficient denitrification and a sink of carbon as biomass in anammox granules.

  9. Microbial community composition and ultrastructure of granules from a full-scale anammox reactor.

    Gonzalez-Gil, Graciela; Sougrat, Rachid; Behzad, Ali R; Lens, Piet N L; Saikaly, Pascal E

    2015-07-01

    Granules in anammox reactors contain besides anammox bacteria other microbial communities whose identity and relationship with the anammox bacteria are not well understood. High calcium concentrations are often supplied to anammox reactors to obtain sufficient bacterial aggregation and biomass retention. The aim of this study was to provide the first characterization of bacterial and archaeal communities in anammox granules from a full-scale anammox reactor and to explore on the possible role of calcium in such aggregates. High magnification imaging using backscattered electrons revealed that anammox bacteria may be embedded in calcium phosphate precipitates. Pyrosequencing of 16S rRNA gene fragments showed, besides anammox bacteria (Brocadiacea, 32%), substantial numbers of heterotrophic bacteria Ignavibacteriacea (18%) and Anaerolinea (7%) along with heterotrophic denitrifiers Rhodocyclacea (9%), Comamonadacea (3%), and Shewanellacea (3%) in the granules. It is hypothesized that these bacteria may form a network in which heterotrophic denitrifiers cooperate to achieve a well-functioning denitrification system as they can utilize the nitrate intrinsically produced by the anammox reaction. This network may provide a niche for the proliferation of archaea. Hydrogenotrophic methananogens, which scavenge the key fermentation product H2, were the most abundant archaea detected. Cells resembling the polygon-shaped denitrifying methanotroph Candidatus Methylomirabilis oxyfera were observed by electron microscopy. It is hypothesized that the anammox process in a full-scale reactor triggers various reactions overall leading to efficient denitrification and a sink of carbon as biomass in anammox granules.

  10. Modeling requirements for full-scope reactor simulators of fission-product transport during severe accidents

    Ellison, P.G.; Monson, P.R.; Mitchell, H.A.

    1990-01-01

    This paper describes in the needs and requirements to properly and efficiently model fission product transport on full scope reactor simulators. Current LWR simulators can be easily adapted to model severe accident phenomena and the transport of radionuclides. Once adapted these simulators can be used as a training tool during operator training exercises for training on severe accident guidelines, for training on containment venting procedures, or as training tool during site wide emergency training exercises

  11. Aerobic Sludge Granulation in a Full-Scale Sequencing Batch Reactor

    Jun Li

    2014-01-01

    Full Text Available Aerobic granulation of activated sludge was successfully achieved in a full-scale sequencing batch reactor (SBR with 50,000 m3 d−1 for treating a town’s wastewater. After operation for 337 days, in this full-scale SBR, aerobic granules with an average SVI30 of 47.1 mL g−1, diameter of 0.5 mm, and settling velocity of 42 m h−1 were obtained. Compared to an anaerobic/oxic plug flow (A/O reactor and an oxidation ditch (OD being operated in this wastewater treatment plant, the sludge from full-scale SBR has more compact structure and excellent settling ability. Denaturing gradient gel electrophoresis (DGGE analysis indicated that Flavobacterium sp., uncultured beta proteobacterium, uncultured Aquabacterium sp., and uncultured Leptothrix sp. were just dominant in SBR, whereas uncultured bacteroidetes were only found in A/O and OD. Three kinds of sludge had a high content of protein in extracellular polymeric substances (EPS. X-ray fluorescence (XRF analysis revealed that metal ions and some inorganics from raw wastewater precipitated in sludge acted as core to enhance granulation. Raw wastewater characteristics had a positive effect on the granule formation, but the SBR mode operating with periodic feast-famine, shorter settling time, and no return sludge pump played a crucial role in aerobic sludge granulation.

  12. MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

    Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A. [AREVA - Tour AREVA, 1 Place Jean Millier, 92084 Paris La Defense (France)

    2013-07-01

    In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

  13. MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

    Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A.

    2013-01-01

    In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO 2 fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory

  14. Report and analysis on 'PR and PP evaluation. Example sodium fast reactor full system case study'

    Sagara, Hiroshi; Inoue, Naoko; Kawakubo, Yoko; Watahiki, Masaru

    2011-01-01

    The Generation IV (GEN IV) Nuclear Energy Systems International Forum (GIF) Proliferation Resistance and Physical Protection Working Group (PRPP WG) was established in December 2002 in order to develop the PR and valuation methodology for GEN IV nuclear energy systems. In the final report of 'PR and PP Evaluation: Example Sodium Fast Reactor (ESFR) Full System Case Study,' issued in October 2009, the demonstration study of PR and PP evaluation with the qualitative approach are summarized using ESFR with four scenario threats. The present paper reviews and analyzes some results of the ESFR case study, and identifies the challenges and direction for the PR and PP evaluation methodology with quantitative approach. (author)

  15. A computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors

    Hu, Rui, E-mail: rhu@anl.gov; Yu, Yiqi

    2016-11-15

    Highlights: • Developed a computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors. • Applied fully-coupled JFNK solution scheme to avoid the operator-splitting errors. • The accuracy and efficiency of the method is confirmed with a 7-assembly test problem. • The effects of different spatial discretization schemes are investigated and compared to the RANS-based CFD simulations. - Abstract: For efficient and accurate temperature predictions of sodium fast reactor structures, a 3-D full-core conjugate heat transfer modeling capability is developed for an advanced system analysis tool, SAM. The hexagon lattice core is modeled with 1-D parallel channels representing the subassembly flow, and 2-D duct walls and inter-assembly gaps. The six sides of the hexagon duct wall and near-wall coolant region are modeled separately to account for different temperatures and heat transfer between coolant flow and each side of the duct wall. The Jacobian Free Newton Krylov (JFNK) solution method is applied to solve the fluid and solid field simultaneously in a fully coupled fashion. The 3-D full-core conjugate heat transfer modeling capability in SAM has been demonstrated by a verification test problem with 7 fuel assemblies in a hexagon lattice layout. Additionally, the SAM simulation results are compared with RANS-based CFD simulations. Very good agreements have been achieved between the results of the two approaches.

  16. International Intercomparison of Calorimeters. A Report of Measurements Made in the Reactor Melusine at C.E.N. Grenoble March 1970 by Eight Reactor Calorimetry Groups from Six Member States of the IAEA

    Boyd, A.W.

    1970-01-01

    The Working Group on Reactor Radiation Measurements has recommended an international intercomparison of calorimeters through its Subgroup 5 on calorimetric methods. Through the cooperation of the IAEA and the Centre d'Etudes Nucléaires de Grenoble of the Commissariat à l'Energie Atomique of France, the intercomparison was arranged. Representatives of eight dosimetry groups from six countries took part in this intercomparison which was carried out during the period 9-17 March 1970. This document will help workers in the field of absorbed dose measurements in reactors in choosing consistent and accurate procedures of experimentation and reporting data

  17. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  18. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-01-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  19. Characterization of the full cone pressure swirl spray nozzles for the nuclear reactor containment spray system

    Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); John, Benny [Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

    2014-07-01

    Highlights: • Full cone spray pressure swirl nozzle with X-Vane is studied. • Laser illuminated imaging technique is used. • Correlations for coefficient of discharge, spray cone angle and SMD are suggested. • Droplet size and mass fraction distribution is measured. • Inviscid theory predicts the coefficient of discharge. - Abstract: The objective of the present study is to characterize a full cone pressure swirl nozzle for the Containment Spray System (CSS) of Indian Pressurized heavy Water reactors (IPHWR). The influence of Reynolds number and geometric parameters on the coefficient of discharge, spray cone angle, mass flux density distribution, droplet size distribution, Sauter mean diameter (SMD is studied for full cone pressure swirl full cone nozzles. The nozzles of orifice diameter range from 1.3 to 7.2 mm are studied. Experiments are conducted with water at room temperature as the working medium. The nozzles are operated with the pressure ranging from 1 to 8 bar. The measurements of the drop size distributions are performed with laser illuminated imaging technique. The spray cone-angle of the full cone nozzles is measured by the evaluation of images recorded with a camera using IMAGE J software. Correlations for coefficient of discharge, spray cone angle and Sauter mean diameter are suggested on the basis of the experimental results. Rosin–Rammler model and Nukiyama–Tanasawa distributions predict the mass fraction distribution reasonably well. However, the droplet size distribution is predicted by Nukiyama-Tanasawa model only.

  20. A full scope nuclear power plant simulator for multiple reactor types with virtual control panels

    Yonezawa, Hisanori; Ueda, Hiroki; Kato, Takahisa

    2017-01-01

    This paper summarizes a full scope nuclear power plant simulator for multiple reactor types with virtual control panels which Toshiba developed and delivered. After the Fukushima DAIICHI nuclear power plants accident, it is required that all the people who are engaged in the design, manufacturing, operation, maintenance, management and regulation for the nuclear power plant should learn the wide and deep knowledge about the nuclear power plant design including the severe accident. For this purpose, the training with a full scope simulator is one of the most suitable ways. However the existing full scope simulators which are consist of the control panels replica of the referenced plants are costly and they are hard to remodel to fit to the real plant of the latest condition. That's why Toshiba developed and delivered the new concept simulator system which covers multiple referenced plants even though they have different design like BWR and PWR. The control panels of the simulator are made by combining 69 large Liquid Crystal Display (LCD) panels with touch screen instead of a control panel replica of referenced plant. The screen size of the each panel is 42 inches and 3 displays are arranged in tandem for one unit and 23 units are connected together. Each panel displays switches, indicators, recorders and lamps with the Computer Graphics (CG) and trainees operate them with touch operations. The simulator includes a BWR and a PWR simulator model, which enable trainees to learn the wide and deep knowledge about the nuclear power plant of BWR and PWR reactor types. (author)

  1. Characters of neutron noise in full-size molten salt reactor

    Wang, Jiangmeng; Cao, Xinrong

    2015-01-01

    Highlights: • The larger system size makes full-size MSR deviate from point kinetic behavior. • The increasing velocity has non-monotonic effect on the effective delayed neutron fraction. • The amplitude of Green’s function at low frequencies is inversely proportional to the effective delayed neutron fraction. • The range of plateau region is smaller due to the more prominent point kinetic effect. - Abstract: In the present paper, the frequency-dependent and space-dependent behavior of the neutron noise in a full-size Molten Salt Reactor (MSR) is investigated. The theoretical models considering the fuel circulation are established based on one-group neutron diffusion theory. Green’s function of the neutron noise induced by a propagating perturbation is introduced with linear noise theory, due to the small perturbation. The equations are numerically calculated by developing a code, in which the eigenfunction expansion method is adopted. The static results show that the effective delayed neutron fraction changes non-monotonically with the increasing fuel velocity. In the dynamic case, the results are compared to those obtained in 1D MSR and a traditional reactor, in order to figure out the effects of both the fuel circulation and the system size. It is found that there is no difference in 1D and 3D MSR systems from the view of fuel circulation, i.e., the fuel circulation enhances the spatial neutronic coupling and leads to the stronger point kinetic effect. The more prominent space-dependent effect founded in 3D traditional reactors is also observed in the MSR, due to the looser neutronic coupling and the unique singularity of Green’s function in the location of the perturbation. Another interesting finding is that Green’s function for low frequencies changes non-monotonically with increasing velocity. The largest magnitude of Green’s function is observed at the velocity where the effective delayed neutron fraction reaches its minimum. Finally, the

  2. Application of assembly module to high-temperature gas-cooled reactor full-scope simulation system

    Li Sifeng; Li Fu; Ma Yuanle; Shi Lei

    2007-01-01

    According to the circumstances that exist in the reactor full-scope simulators development as long development cycle, very difficult upgrade and narrow range of applicability, a kind of new model was developed based on assembly module which root in Linux kernel and successfully applied to the design of high-temperature gas-cooled reactor full-scope simulator system. The simulation results are coincident with the experimental ones, and it indicates that the new model based on assembly module is feasible to design of high-temperature gas cooled reactor simulation system. (authors)

  3. Full system decontamination for dose reduction at the preventive maintenance work of the reactor core internals

    Sato, Y.; Inami, I.; Suzuki, N.; Fujimori, A.; Wille, H.

    2000-01-01

    At the Fukushima Dai-ichi Nuclear Power Station unit 3 and unit 2 of Tokyo Electric Power Company (TEPCO), the replacement of the core shroud and internals have been conducted respectively in the FY 1997 outage and in the FY 1998 outage. The replacement of the welded core internals in operating BWR plants is the first time in the world as complete countermeasure to improve SCC resistance. At present both units are operating smoothly. The developed technology concept is to restore those internals in air inside the reactor pressure vessel. To reduce the radiation dose rate inside the RPV, not only a shielding method was applied to cut the radiation from the irradiated structures but also a chemical decontamination method was applied to dissolve the radioactive crud deposit on the surface by using chemical agents. The CORD UV process was applied for this Full System Decontamination including operating the reactor recirculation pumps. The critical pass time required was approximately 7 days for each unit. In both units the radioactivity of 10 TBq (280 Ci) and the Fe, Ni, Cr crud of 60-70 kg as metal in total was dissolved and removed by 5 m 3 (175 ft 3 ) ion exchange resins as only waste generated. The obtained decontamination factor (DF) at the RPV bottom reached 40-100. As result, the dose rate decreased to approximately 0.1 mSv/h under water. Before and after the installation of the in-vessel shielding, a mechanical cleaning was extensively applied inside the RPV to remove the residual crud as well as the cutting particles. As result, the RPV bottom dose rate decreased further to 0.03 mSv/h under water and 0.2 mSv/h in air. A better working environment for human access than expected was established inside the RPV, resulting the 70, 140 man*Sv saving respectively at unit 3 (1F-3) and unit 2 (1F-2). (author)

  4. Neutral beam energy and power requirements for expanding radius and full bore startup of tokamak reactors

    Houlberg, W.A.; Mense, A.T.; Attenberger, S.E.

    1979-09-01

    Natural beam power and energy requirements are compared for full density full bore and expanding radius startup scenarios in an elongated plasma, The Next Step (TNS), as a function of beam pulse time and plasma density. Because of the similarity of parameters, the results should also be applicable to Engineering Test Facility (ETF) and International Tokamak Reactor (INTOR) studies. A transport model consisting of neoclassical ion conduction and anomalous electron conduction and diffusion based on ALCATOR scaling leads to average densities in the range approx. 0.8 to 1.2 x 10 14 cm -3 being sufficient for ignition. Neutral deuterium beam energies in the range 120 to 180 keV are adequate for penetration, with the required power injected into the plasma decreasing with increasing beam energy. The neutral beam power decreases strongly with increasing beam pulse length b/sub b/ until t/sub b/ exceeds a few total energy confinement times, yielding b/sub b/ approx. = 4 to 6 s for the TNS plasma

  5. Team training using full-scale reactor coolant pump seal mock-ups

    McDonald, T.J.; Hamill, R.W.

    1987-01-01

    The use of full-scale reactor coolant pump (RCP) seal mock-ups has greatly enhanced Northeast Utilities' ability to effectively utilize the team training approach to technical training. With the advent of the Institute of Nuclear Power Operations accreditation come a new emphasis and standards for the integrated training of plant engineering personnel, maintenance mechanics, quality control personnel, and health physics personnel. The results of purchasing full-scale RCP mock-ups to pilot the concept of team training have far exceeded expectations and cost-limiting factors. The initial training program analysis identified RCP seal maintenance as a task that required training for maintenance department personnel. Due to radiation exposure considerations and the unavailability of actual plant equipment for training purposes, the decision was made to procure a mock-up of an RCP seal assembly and housing. This mock-up was designed to facilitate seal cartridge removal, disassembly, assembly, and installation, duplicating all internal components of the seal cartridge and housing area in exact detail

  6. MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

    Ishii, M.; Revankar, S. T.; Downar, T.; Xu, Y.; Yoon, H. J.; Tinkler, D.; Rohatgi, U. S.

    2003-01-01

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral

  7. Traveling-wave reactors: A truly sustainable and full-scale resource for global energy needs

    Ellis, T.; Petroski, R.; Hejzlar, P.; Zimmerman, G.; McAlees, D.; Whitmer, C.; Touran, N.; Hejzlar, J.; Weave, K.; Walter, J. C.; McWhirter, J.; Ahlfeld, C.; Burke, T.; Odedra, A.; Hyde, R.; Gilleland, J.; Ishikawa, Y.; Wood, L.; Myhrvold, N.; Gates Iii, W. H.

    2010-01-01

    Rising environmental and economic concerns have signaled a desire to reduce dependence on hydrocarbon fuels. These concerns have brought the world to an inflection point and decisions made today will dictate what the global energy landscape will look like for the next half century or more. An optimal energy technology for the future must meet stricter standards than in the past; in addition to being economically attractive, it now must also be environmentally benign, sustainable and scalable to global use. For stationary energy, only one existing resource comes close to fitting all of the societal requirements for an optimal energy source: nuclear energy. Its demonstrated economic performance, power density, and emissions-free benefits significantly elevate nuclear electricity generation above other energy sources. However, the current nuclear fuel cycle has some attributes that make it challenging to expand on a global scale. Traveling-wave reactor (TWR) technology, being developed by TerraPower, LLC, represents a potential solution to these limitations by offering a nuclear energy resource which is truly sustainable at full global scale for the indefinite future and is deployable in the near-term. TWRs are capable of offering a ∼40-fold gain in fuel utilization efficiency compared to conventional light-water reactors burning enriched fuel. Such high fuel efficiency, combined with an ability to use uranium recovered from river water or sea-water (which has been recently demonstrated to be technically and economically feasible) suggests that enough fuel is readily available for TWRs to generate electricity for 10 billion people at United States per capita levels for million-year time-scales. Interestingly, the Earth's rivers carry into the ocean a flux of uranium several times greater than that required to replace the implied rate-of-consumption, so that the Earth's slowly-eroding crust will provide a readily-accessible flow of uranium sufficient for all of

  8. Two-dimensional full-core transport theory Benchmarks for the WWER reactors

    Petkov, P.T.

    2002-01-01

    Several two-dimensional full-core real geometry many-group steady-state problems for the WWER-440 and WWER-1000 reactors have been solved by the MARIKO code, based on the method of characteristics. The reference transport theory solutions include assembly-wise and pin-wise power distributions. Homogenized two-group diffusion parameters and discontinuity factors have been calculated by MARIKO for each assembly type both for the whole assembly and for each cell in the smallest sector of symmetry, using the B1 method for calculation of the critical spectrum. Accurate albedo-type boundary conditions have been calculated by MARIKO for the core-reflector and core-absorber boundaries, both for each outer assembly face and for each outer cell face. Comparison with the reference solutions of the two-group nodal diffusion code SPPS-1.6 and the few-group fine-mesh diffusion codes HEX2DA and HEX2DB are presented (Authors)

  9. Three dimensional numerical simulation of a full scale CANDU reactor moderator to study temperature fluctuations

    Sarchami, Araz; Ashgriz, Nasser; Kwee, Marc

    2014-01-01

    Highlights: • 3D model of a Candu reactor is modeled to investigate flow distribution. • The results show the temperature distribution is not symmetrical. • Temperature contours show the hot regions at the top left-hand side of the tank. • Interactions of momentum flows and buoyancy flows create circulation zones. • The results indicate that the moderator tank operates in the buoyancy driven mode. -- Abstract: Three dimensional numerical simulations are conducted on a full scale CANDU Moderator and transient variations of the temperature and velocity distributions inside the tank are determined. The results show that the flow and temperature distributions inside the moderator tank are three dimensional and no symmetry plane can be identified. Competition between the upward moving buoyancy driven flows and the downward moving momentum driven flows in the center region of the tank, results in the formation of circulation zones. The moderator tank operates in the buoyancy driven mode and any small disturbances in the flow or temperature makes the system unstable and asymmetric. Different types of temperature fluctuations are noted inside the tank: (i) large amplitude are at the boundaries between the hot and cold; (ii) low amplitude are in the core of the tank; (iii) high frequency fluctuations are in the regions with high velocities and (iv) low frequency fluctuations are in the regions with lower velocities

  10. Reactor control and protection of full scope simulator for Qinshan 300 MW Nuclear Power Unit

    Zhu Jinping; Sun Jiliang

    1996-01-01

    The control and protection simulation of Qinshan 300 MW Nuclear Power Unit, including the nuclear control, the pressurizer pressure control, the pressurizer level control, the rod control, the reactor shutdown protection and engineered safety feature etc are briefly introduced

  11. Study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP

    Soares, Alexandre de Souza

    2014-01-01

    The present paper proposes a study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP. To this end, it considered a loss of coolant accident with 160 cm 2 breaking area in cold leg of 20 circuit of the reactor cooling system of nuclear power plant Angra 2, with the reactor operating in stationary condition, to 100% power. It considered that occurred at the same time the loss of External Power Supply and the availability of emergency cooling system was not full. The results obtained are quite relevant and with the possibility of being used in the planning of future activities, given that the construction of Angra 3 is underway and resembles the Angra 2. (author)

  12. Three-dimensional reactor model for the Paks NPP full-scope simulator

    Gyori, C.; Hegyi, G.; Hozer, Z.; Kereszturi, A.; Maraczy, C.

    1993-01-01

    The reactor model includes thermohydraulic and neutron-physical components. The thermohydraulic model is based on the SMABRE code developed at the Technical Research Centre of Finland for the analysis of loss-of-coolant transients in PWRs. The fuel rod model will be replaced by a new software module providing a comprehensive description of the behavior of fuel rods during reactor transients and hypothetical accidents. The calculation is performed in four individual models: fuel rod temperature model, fuel rod internal pressure model, fuel rod deformation model and fuel rod failure model. In the neutron-physical model the core is calculated with nodes for all of the 349 fuel assemblies, and each assembly is calculated in ten layers. (Z.S.) 1 fig., 5 refs

  13. Assessment of uncertainty in full core reactor physics calculations using statistical methods

    McEwan, C.

    2012-01-01

    The best estimate method of safety analysis involves choosing a realistic set of input parameters for a proposed safety case and evaluating the uncertainty in the results. Determining the uncertainty in code outputs remains a challenge and is the subject of a benchmarking exercise proposed by the Organization for Economic Cooperation and Development. The work proposed in this paper will contribute to this benchmark by assessing the uncertainty in a depletion calculation of the final nuclide concentrations for an experiment performed in the Fukushima-2 reactor. This will be done using lattice transport code DRAGON and a tool known as DINOSAUR. (author)

  14. Assessment of uncertainty in full core reactor physics calculations using statistical methods

    McEwan, C., E-mail: mcewac2@mcmaster.ca [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    The best estimate method of safety analysis involves choosing a realistic set of input parameters for a proposed safety case and evaluating the uncertainty in the results. Determining the uncertainty in code outputs remains a challenge and is the subject of a benchmarking exercise proposed by the Organization for Economic Cooperation and Development. The work proposed in this paper will contribute to this benchmark by assessing the uncertainty in a depletion calculation of the final nuclide concentrations for an experiment performed in the Fukushima-2 reactor. This will be done using lattice transport code DRAGON and a tool known as DINOSAUR. (author)

  15. Kinetic modelling and characterization of microbial community present in a full-scale UASB reactor treating brewery effluent.

    Enitan, Abimbola M; Kumari, Sheena; Swalaha, Feroz M; Adeyemo, J; Ramdhani, Nishani; Bux, Faizal

    2014-02-01

    The performance of a full-scale upflow anaerobic sludge blanket (UASB) reactor treating brewery wastewater was investigated by microbial analysis and kinetic modelling. The microbial community present in the granular sludge was detected using fluorescent in situ hybridization (FISH) and further confirmed using polymerase chain reaction. A group of 16S rRNA based fluorescent probes and primers targeting Archaea and Eubacteria were selected for microbial analysis. FISH results indicated the presence and dominance of a significant amount of Eubacteria and diverse group of methanogenic Archaea belonging to the order Methanococcales, Methanobacteriales, and Methanomicrobiales within in the UASB reactor. The influent brewery wastewater had a relatively high amount of volatile fatty acids chemical oxygen demand (COD), 2005 mg/l and the final COD concentration of the reactor was 457 mg/l. The biogas analysis showed 60-69% of methane, confirming the presence and activities of methanogens within the reactor. Biokinetics of the degradable organic substrate present in the brewery wastewater was further explored using Stover and Kincannon kinetic model, with the aim of predicting the final effluent quality. The maximum utilization rate constant U max and the saturation constant (K(B)) in the model were estimated as 18.51 and 13.64 g/l/day, respectively. The model showed an excellent fit between the predicted and the observed effluent COD concentrations. Applicability of this model to predict the effluent quality of the UASB reactor treating brewery wastewater was evident from the regression analysis (R(2) = 0.957) which could be used for optimizing the reactor performance.

  16. Self-Cleaning Boudouard Reactor for Full Oxygen Recovery from Carbon Dioxide

    Coutts, Janelle; Hintze, Paul E.; Muscatello, Anthony C.; Gibson, Tracy L.; Captain, James G.; Lunn, Griffin M.; Devor, Robert W.; Bauer, Brint; Parks, Steve

    2016-01-01

    Oxygen recovery from respiratory carbon dioxide is an important aspect of human spaceflight. Methods exist to sequester the carbon dioxide, but production of oxygen needs further development. The current International Space Station Carbon Dioxide Reduction System (CRS) uses the Sabatier reaction to produce water (and ultimately breathing air). Oxygen recovery is limited to 50 because half of the hydrogen used in the Sabatier reactor is lost as methane, which is vented overboard. The Bosch reaction, which converts carbon dioxide to oxygen and solid carbon is capable of recovering all the oxygen from carbon dioxide, and is the only real alternative to the Sabatier reaction. However, the last reaction in the cycle, the Boudouard reaction, produces solid carbon and the resulting carbon buildup will eventually foul the nickel or iron catalyst, reducing reactor life and increasing consumables. To minimize this fouling and increase efficiency, a number of self-cleaning catalyst designs have been created. This paper will describe recent results evaluating one of the designs.

  17. Full scale reactor safety experiments performed in the Marviken Power Station Sweden

    Thoren, H.G.; Ericson, L.

    1977-01-01

    Since 1972 experiments oriented towards increasing the understanding of reactor safety processes have been performed at the Marviken Power Station. This was originally built as a direct cycle BHWR but was never taken into nuclear operation. In addition to Sweden, the countries represented in these experiments are Denmark, the Federal Republic of Germany, Finland, Norway, the United States, the Netherlands, France and Japan. The first series of sixteen experiments included studies of the response of the PS-containment to simulated ruptures in the pipe systems that are connected to the pressure vessel. These tests were completed in 1973 and also included experimental studies of iodine transport, containment leakage, the behaviour of auxiliary components under accident conditions and pressure fluctuations in the wetwell water pool. One of the more essential findings of the tests was that the containment performance was in accordance with the pre-test calculations. A second series of eight blowdown tests was begun in February 1976. The main purpose of these tests is to provide additional information as to the characteristics of the pressure oscillations inside the containment and primarily in the wetwell water pool under different conditions. These oscillations were observed in the first series of blowdowns but only low frequencies could then be detected due to limitations in the measurement system. The measurement system was therefore substantially extended for this second series of experiments. A summary of the results from these two sets of blowdown tests are given in the paper. In 1976 preparations for a new test program were initiated. The objective of these tests is to improve the understanding of critical flow in the low quality and subcooled flow regions through short length, large diameter pipes. Extensive modifications of the test facility will be necessary in order to allow a discharge flow through openings which are up to 500 mm in diameter. Advanced plans

  18. Practical guide to dosimetry as applied in the research reactors of the Saclay and Grenoble nuclear research centers; Guide pratique de la dosimetrie mise en oeuvre dans les reacteurs de recherche du C.E.N./G et du C.E.N./S

    NONE

    1968-07-01

    Since the problems concerning neutron and gamma flux measurements which arise during irradiation experiments in the reactors in the Grenoble and Saclay Centres are of the same type, and since the solutions found are very often adopted in common, we have attempted to describe the methods we use at the present time. A brief description is given of the production of the detectors, the electronic apparatus; the formulae usually used for the interpretation of the measurements are given. A series of technical data cards give the most commonly used detector characteristics. These cards give the physical characteristics of the detectors, their nuclear constants, if any, the most suitable counting methods and the field of application. (authors) [French] Les problemes de mesures de flux de neutrons et de flux gamma qui se posent pour les experiences irradiees dans les reacteurs des Centres de Grenoble et de Saclay etant du meme type et les solutions trouvees, tres souvent adoptees en commun, nous avons cherche a decrire les methodes que nous pratiquons actuellement. On decrit tres brievement la fabrication des detecteurs, l'appareillage electronique; on rappelle les formules usuelles qui servent dans l'interpretation des mesures. Une serie de fiches techniques rassemble les caracteristiques des detecteurs les plus couramment utilises. Ces fiches indiquent les caracteristiques physiques des detecteurs, leurs constantes nucleaires s'il y a lieu, les methodes de comptage les mieux adaptees et le domaine d'utilisation. (auteurs)

  19. National demonstration of full reactor coolant system (RCS) chemical decontamination at Indian Point 2

    Trovato, S.A.; Parry, J.O. [Consolidated Edison Co., New York, NY (United States)

    1995-03-01

    Key to the safe and efficient operation of the nation`s civilian nuclear power plants is the performance of maintenance activities within regulations and guidelines for personnel radiation exposure. However, maintenance activities, often performed in areas of relatively high radiation fields, will increase as the nation`s plant age. With the Nuclear Regulatory Commission (NRC) lowering the allowable radiation exposure to plant workers in 1994 and considering further reductions and regulations in the future, it is imperative that new techniques be developed and applied to reduce personnel exposure. Full primary system chemical decontamination technology offers the potential to be single most effective method of maintaining workers exposure {open_quotes}as low as reasonably achievable{close_quotes} (ALARA) while greatly reducing plant operation and maintenance (O&M) costs. A three-phase program underway since 1987, has as its goal to demonstrate that full RCS decontamination is a visible technology to reduce general plant radiation levels without threatening the long term reliability and operability of a plant. This paper discusses research leading to and plans for a National Demonstration of Full RCS Chemical Decontamination at Indian Point 2 nuclear generating station in 1995.

  20. Probabilistic methods of optimization of scheduled tests for heat equipment of safety systems of reactor at full power

    Bilej, D.V.; Fridman, N.A.; Kolykhanov, V.N.; Skalozubov, V.I.

    2004-01-01

    This article generalises the basic results of a long-term teamwork with respect to a scientific and technical substantiation of perfection of the regulations of safe operation power units with VVER. This perfection is concerning a periodicity and volumes of tests of safety systems when a reactor works at full power. The article shows that the application of the probabilistic approaches connected to minimisation of a risk criterion function is an effective methodical base for the optimisation. For certain safety systems of serial power units with VVER 1000 the results of calculated substantiations are presented

  1. Effect of operating conditions and reactor configuration on efficiency of full-scale biogas plants

    Angelidaki, Irini; Boe, Kanokwan; Ellegaard, L.

    2005-01-01

    A study on 18 full-scale centralized biogas plants was carried out in order to find significant operational factors influencing productivity and stability of the plants. It was found that the most plants were operating relatively stable with volatile fatty acids (VFA) concentration below 1.5 g....../l. VFA concentration increase was observed in occasions with dramatic overloading or other disturbances such as operational temperature changes. Ammonia was found to be a significant factor for stability. A correlation between increased residual biogas production and high ammonia was found. When ammonia...

  2. Analysis Of Core Management For The Transition Cores Of RSG-GAS Reactor To Full-Silicide Core

    Malem Sembiring, Tagor; Suparlina, Lily; Tukiran

    2001-01-01

    The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 g U/cc is still doing. At the end of 2000, the reactor has been operated for 3 transition cores which is the mixed core of oxide-silicide. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 10 transition cores to achieve a full-silicide core. The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters such as excess reactivity and shutdown margin. The measurement of the core parameters was carried out using the method of compensation of couple control rods. The experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safety to a full-silicide core

  3. Coupled full core neutron transport/CFD simulations of pressurized water reactors

    Kochunas, B.; Stimpson, S.; Collins, B.; Downar, T.; Brewster, R.; Baglietto, E.; Yan, J.

    2012-01-01

    Recently as part of the CASL project, a capability to perform 3D whole-core coupled neutron transport and computational fluid dynamics (CFD) calculations was demonstrated. This work uses the 2D/1D transport code DeCART and the commercial CFD code STAR-CCM+. It builds on previous CASL work demonstrating coupling for smaller spatial domains. The coupling methodology is described along with the problem simulated and results are presented for fresh hot full power conditions. An additional comparison is made to an equivalent model that uses lower order T/H feedback to assess the importance and cost of high fidelity feedback to the neutronics problem. A simulation of a quarter core Combustion Engineering (CE) PWR core was performed with the coupled codes using a Fixed Point Gauss-Seidel iteration technique. The total approximate calculation requirements are nearly 10,000 CPU hours and 1 TB of memory. The problem took 6 coupled iterations to converge. The CFD coupled model and low order T/H feedback model compared well for global solution parameters, with a difference in the critical boron concentration and average outlet temperature of 14 ppm B and 0.94 deg. C, respectively. Differences in the power distribution were more significant with maximum relative differences in the core-wide pin peaking factor (Fq) of 5.37% and average relative differences in flat flux region power of 11.54%. Future work will focus on analyzing problems more relevant to CASL using models with less approximations. (authors)

  4. Domain Decomposition strategy for pin-wise full-core Monte Carlo depletion calculation with the reactor Monte Carlo Code

    Liang, Jingang; Wang, Kan; Qiu, Yishu [Dept. of Engineering Physics, LiuQing Building, Tsinghua University, Beijing (China); Chai, Xiao Ming; Qiang, Sheng Long [Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu (China)

    2016-06-15

    Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC) codes in accomplishing pin-wise three-dimensional (3D) full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC) code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.

  5. Methanogenic population dynamics during startup of a full-scale anaerobic sequencing batch reactor treating swine waste.

    Angenent, Largus T; Sung, Shihwu; Raskin, Lutgarde

    2002-11-01

    Changes in methanogenic population levels were followed during startup of a full-scale, farm-based anaerobic sequencing batch reactor (ASBR) and these changes were linked to operational and performance data. The ASBR was inoculated with anaerobic digester sludge from a municipal wastewater treatment facility. During an acclimation period of approximately 3 months, the ASBR content was diluted to maintain a total ammonia-N level of approximately 2000mg l(-1). After this acclimation period, the volatile solids loading rate was increased to its design value of 1.7g l(-1) day(-1) with a 15-day hydraulic retention time, which increased the total ammonia-N level in the ASBR to approximately 3,600 mg l(-1). The 16S ribosomal RNA (rRNA) levels of the acetate-utilizing methanogens of the genus Methanosarcina decreased from 3.8% to 1.2% (expressed as a percentage of the total 16S rRNA levels) during this period, while the 16S rRNA levels of Methanosaeta concilii remained low (below 2.2%). Methane production and reactor performance were not affected as the 16S rRNA levels of the hydrogen-utilizing methanogens of the order Methanomicrobiales increased from 2.3% to 7.0%. Hence, it is likely that during operation with high ammonia levels, the major route of methane production is through a syntrophic relationship between acetate-oxidizing bacteria and hydrogen-utilizing methanogens. Anaerobic digestion at total ammonia-N levels exceeding 3500mg l(-1) was sustainable apparently due to the acclimation of hydrogen-utilizing methanogens to high ammonia levels.

  6. Minimizing N2O emissions and carbon footprint on a full-scale activated sludge sequencing batch reactor.

    Rodriguez-Caballero, A; Aymerich, I; Marques, Ricardo; Poch, M; Pijuan, M

    2015-03-15

    A continuous, on-line quantification of the nitrous oxide (N2O) emissions from a full-scale sequencing batch reactor (SBR) placed in a municipal wastewater treatment plant (WWTP) was performed in this study. In general, N2O emissions from the biological wastewater treatment system were 97.1 ± 6.9 g N2O-N/Kg [Formula: see text] consumed or 6.8% of the influent [Formula: see text] load. In the WWTP of this study, N2O emissions accounted for over 60% of the total carbon footprint of the facility, on average. Different cycle configurations were implemented in the SBR aiming at reaching acceptable effluent values. Each cycle configuration consisted of sequences of aerated and non-aerated phases of different time length being controlled by the ammonium set-point fixed. Cycles with long aerated phases showed the largest N2O emissions, with the consequent increase in carbon footprint. Cycle configurations with intermittent aeration (aerated phases up to 20-30 min followed by short anoxic phases) were proven to effectively reduce N2O emissions, without compromising nitrification performance or increasing electricity consumption. This is the first study in which a successful operational strategy for N2O mitigation is identified at full-scale. Copyright © 2014 Elsevier Ltd. All rights reserved.

  7. Numerical Simulation of Three-Dimensional Flow Through Full Passage and Performance Prediction of Nuclear Reactor Coolant Pump

    Li Ying; Zhou Wenxia; Zhang Jige; Wang Dezhong

    2009-01-01

    In order to achieve the level of self-design and domestic manufacture of the reactor coolant pump (nuclear main pump), the software FLUENT was used to simulate the three-dimensional flow through full passage of one nuclear main pump basing on RNG κ-ε turbulence model and SIMPLE algorithm. The distribution of pressure and velocity of the flow in the impeller's surface was analyzed in different working conditions. Moreover, the performance of the pump was predicted based on the simulation results. The results show that the distributions of pressure and velocity are reasonable in both the working and back face of the blade in the steady working condition. The pressure of the flow is increased from the inlet to the outlet of the pump, and shows the maximal value in the impeller region. Comparatively satisfactory efficiency and head value were obtained in the condition of the pump design. The shaft power of the nuclear main pump is gradually increased with the increase of the flow flux. These results are helpful in understanding the change of the internal flow field in the nuclear main pump, which is of some importance for the pre-exploration and theoretical research on the domestic manufacture of the nuclear main pump. (authors)

  8. NMR and MALDI-TOF MS based characterization of exopolysaccharides in anaerobic microbial aggregates from full-scale reactors

    Gonzalez-Gil, Graciela

    2015-09-22

    Anaerobic granular sludge is composed of multispecies microbial aggregates embedded in a matrix of extracellular polymeric substances (EPS). Here we characterized the chemical fingerprint of the polysaccharide fraction of EPS in anaerobic granules obtained from full-scale reactors treating different types of wastewater. Nuclear magnetic resonance (NMR) signals of the polysaccharide region from the granules were very complex, likely as a result of the diverse microbial population in the granules. Using nonmetric multidimensional scaling (NMDS), the 1H NMR signals of reference polysaccharides (gellan, xanthan, alginate) and those of the anaerobic granules revealed that there were similarities between the polysaccharides extracted from granules and the reference polysaccharide alginate. Further analysis of the exopolysaccharides from anaerobic granules, and reference polysaccharides using matrix-assisted laser desorption/ionization time-of-flight mass spectrometry (MALDI-TOF MS) revealed that exopolysaccharides from two of the anaerobic granular sludges studied exhibited spectra similar to that of alginate. The presence of sequences related to the synthesis of alginate was confirmed in the metagenomes of the granules. Collectively these results suggest that alginate-like exopolysaccharides are constituents of the EPS matrix in anaerobic granular sludge treating different industrial wastewater. This finding expands the engineered environments where alginate has been found as EPS constituent of microbial aggregates.

  9. Technical, hygiene, economic, and life cycle assessment of full-scale moving bed biofilm reactors for wastewater treatment in India.

    Singh, Anju; Kamble, Sheetal Jaisingh; Sawant, Megha; Chakravarthy, Yogita; Kazmi, Absar; Aymerich, Enrique; Starkl, Markus; Ghangrekar, Makarand; Philip, Ligy

    2018-01-01

    Moving bed biofilm reactor (MBBR) is a highly effective biological treatment process applied to treat both urban and industrial wastewaters in developing countries. The present study investigated the technical performance of ten full-scale MBBR systems located across India. The biochemical oxygen demand, chemical oxygen demand, total suspended solid, pathogens, and nutrient removal efficiencies were low as compared to the values claimed in literature. Plant 1 was considered for evaluation of environmental impacts using life cycle assessment approach. CML 2 baseline 2000 methodology was adopted, in which 11 impact categories were considered. The life cycle impact assessment results revealed that the main environmental hot spot of this system was energy consumption. Additionally, two scenarios were compared: scenario 1 (direct discharge of treated effluent, i.e., no reuse) and scenario 2 (effluent reuse and tap water replacement). The results showed that scenario 2 significantly reduce the environmental impact in all the categories ultimately decreasing the environmental burden. Moreover, significant economic and environmental benefits can be obtained in scenario 2 by replacing the freshwater demand for non-potable uses. To enhance the performance of wastewater treatment plant (WWTP), there is a need to optimize energy consumption and increase wastewater collection efficiency to maximize the operating capacity of plant and minimize overall environmental footprint. It was concluded that MBBR can be a good alternative for upgrading and optimizing existing municipal wastewater treatment plants with appropriate tertiary treatment. Graphical abstract ᅟ.

  10. NMR and MALDI-TOF MS based characterization of exopolysaccharides in anaerobic microbial aggregates from full-scale reactors

    Gonzalez-Gil, Graciela; Thomas, Ludivine; Emwas, Abdul-Hamid M.; Lens, Piet N. L.; Saikaly, Pascal

    2015-01-01

    Anaerobic granular sludge is composed of multispecies microbial aggregates embedded in a matrix of extracellular polymeric substances (EPS). Here we characterized the chemical fingerprint of the polysaccharide fraction of EPS in anaerobic granules obtained from full-scale reactors treating different types of wastewater. Nuclear magnetic resonance (NMR) signals of the polysaccharide region from the granules were very complex, likely as a result of the diverse microbial population in the granules. Using nonmetric multidimensional scaling (NMDS), the 1H NMR signals of reference polysaccharides (gellan, xanthan, alginate) and those of the anaerobic granules revealed that there were similarities between the polysaccharides extracted from granules and the reference polysaccharide alginate. Further analysis of the exopolysaccharides from anaerobic granules, and reference polysaccharides using matrix-assisted laser desorption/ionization time-of-flight mass spectrometry (MALDI-TOF MS) revealed that exopolysaccharides from two of the anaerobic granular sludges studied exhibited spectra similar to that of alginate. The presence of sequences related to the synthesis of alginate was confirmed in the metagenomes of the granules. Collectively these results suggest that alginate-like exopolysaccharides are constituents of the EPS matrix in anaerobic granular sludge treating different industrial wastewater. This finding expands the engineered environments where alginate has been found as EPS constituent of microbial aggregates.

  11. Reactor

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  12. Evaluation of materials' corrosion and chemistry issues for advanced gas cooled reactor steam generators using full scale plant simulations

    Woolsey, I.S.; Rudge, A.J.; Vincent, D.J.

    1998-01-01

    Advanced Gas Cooled Reactors (AGRS) employ once-through steam Generators of unique design to provide steam at approximately 530 degrees C and 155 bar to steam turbines of similar design to those of fossil plants. The steam generators are highly compact, and have either a serpentine or helical tube geometry. The tubes are heated on the outside by hot C0 2 gas, and steam is generated on the inside of the tubes. Each individual steam generator tube consists of a carbon steel feed and primary economiser section, a 9%Cr steel secondary economiser, evaporator and primary superheater, and a Type 316L austenitic stainless steel secondary superheater, all within a single tube pass. The multi-material nature of the individual tube passes, the need to maintain specific thermohydraulic conditions within the different material sections, and the difficulties of steam generator inspection and repair, have required extensive corrosion-chemistry test programmes to ensure waterside corrosion does not present a challenge to their integrity. A major part of these programmes has been the use of a full scale steam generator test facility capable of simulating all aspects of the waterside conditions which exist in the plant. This facility has been used to address a wide variety of possible plant drainage/degradation processes. These include; single- and two-phase flow accelerated corrosion of carbon steel, superheat margins requirements and the stress-corrosion behaviour of the austenitic superheaters, on-load corrosion of the evaporator materials, and iron transport and oxide deposition behaviour. The paper outlines a number of these, and indicates how they have been of value in helping to maintain reliable operation of the plant. (author)

  13. Reactor

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  14. Reactor

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  15. Monte Carlo simulation of irradiation of MTR fuel plates in the BR2 reactor using a full-scale 3-d model with inclined channels

    Kuzminov, V. V; Koonen, E.; Ponsard, B.

    2002-01-01

    A three-dimensional full-scale Monte Carlo model of the BR2 reactor has been developed for simulation of irradiation conditions of materials and fuel loaded in various irradiation devices. This new reactor model includes a detailed geometrical description of the inclined reactor channels, the irradiation devices loaded in these channels including the materials to be tested/loaded in these devices, the burn-up of the BR2 fuel elements and the poisoning of the beryllium matrix. Recently a benchmark irradiation of new irradiation device for testing and qualification of MTR fuel plates has been performed. For this purpose the detailed irradiation conditions of fuel plates had to be predetermined. Monte Carlo calculations of neutron fluxes and heat load distributions in irradiated MTR fuel plates were performed taking into account the contents of all loaded experimental devices in the reactor channels. A comparison of the calculated and measured values of neutron fluxes and of heat loads in the BR2 reactor is presented in this paper. The comparison is part of the validation process of the new reactor model. It also serves to establish the capability to conduct a fuel plate irradiation program under requested and well- known irradiation conditions. (author)

  16. Nuclear data sensitivity and uncertainty for the Canadian supercritical water-cooled reactor II: Full core analysis

    Langton, S.E.; Buijs, A.; Pencer, J.

    2015-01-01

    Highlights: • H-2, Pu-239, and Th-232 make large contributions to SCWR modelling sensitivity. • H-2, Pu-239, and Th-232 make large contributions to SCWR modelling uncertainty. • Isotopes of Zr make large contributions to SCWR modelling uncertainty. - Abstract: Uncertainties in nuclear data are a fundamental source of uncertainty in reactor physics calculations. To determine their contribution to uncertainties in calculated reactor physics parameters, a nuclear data sensitivity and uncertainty study is performed on the Canadian supercritical water reactor (SCWR) concept. The nuclear data uncertainty contributions to the neutron multiplication factor k eff are 6.31 mk for the SCWR at the beginning of cycle (BOC) and 6.99 mk at the end of cycle (EOC). Both of these uncertainties have a statistical uncertainty of 0.02 mk. The nuclear data uncertainty contributions to Coolant Void Reactivity (CVR) are 1.0 mk and 0.9 mk for BOC and EOC, respectively, both with statistical uncertainties of 0.1 mk. The nuclear data uncertainty contributions to other reactivity parameters range from as low as 3% of to as high as ten times the values of the reactivity coefficients. The largest contributors to the uncertainties in the reactor physics parameters are Pu-239, Th-232, H-2, and isotopes of zirconium

  17. Electrical resistivity tomography to quantify in situ liquid content in a full-scale dry anaerobic digestion reactor.

    André, L; Lamy, E; Lutz, P; Pernier, M; Lespinard, O; Pauss, A; Ribeiro, T

    2016-02-01

    The electrical resistivity tomography (ERT) method is a non-intrusive method widely used in landfills to detect and locate liquid content. An experimental set-up was performed on a dry batch anaerobic digestion reactor to investigate liquid repartition in process and to map spatial distribution of inoculum. Two array electrodes were used: pole-dipole and gradient arrays. A technical adaptation of ERT method was necessary. Measured resistivity data were inverted and modeled by RES2DINV software to get resistivity sections. Continuous calibration along resistivity section was necessary to understand data involving sampling and physicochemical analysis. Samples were analyzed performing both biochemical methane potential and fiber quantification. Correlations were established between the protocol of reactor preparation, resistivity values, liquid content, methane potential and fiber content representing liquid repartition, high methane potential zones and degradations zones. ERT method showed a strong relevance to monitor and to optimize the dry batch anaerobic digestion process. Copyright © 2015 Elsevier Ltd. All rights reserved.

  18. Full-fluence tests of experimental thermosetting fuel rods for the high-temperature gas-cooled reactor

    Bullock, R.E.

    1981-01-01

    The irradiation performance of injected thermosetting fuel rods is compared to that of standard pitch-temperature gas-cooled reactor requirements. The primary objective of the experiments reported here was to obtain additional irradiation data at higher fluences for resin-based rods with intermediate binder char contents within the 15 to 30 wt% ''window of acceptability'' that had been previously established. 12 refs

  19. Full integrated system of real-time monitoring based on distributed architecture for the high temperature engineering test reactor (HTTR)

    Subekti, Muhammad; Ohno, Tomio; Kudo, Kazuhiko; Takamatsu, Kuniyoshi; Nabeshima, Kunihiko

    2005-01-01

    A new monitoring system scheme based on distributed architecture for the High Temperature Engineering Test Reactor (HTTR) is proposed to assure consistency of the real-time process of expanded system. A distributed monitoring task on client PCs as an alternative architecture maximizes the throughput and capabilities of the system even if the monitoring tasks suffer a shortage of bandwidth. The prototype of the on-line monitoring system has been developed successfully and will be tested at the actual HTTR site. (author)

  20. Full scale seismic simulation of a nuclear reactor with parallel finite element analysis code for assembled structure

    Yamada, Tomonori

    2010-01-01

    The safety requirement of nuclear power plant attracts much attention nowadays. With the growing computing power, numerical simulation is one of key technologies to meet this safety requirement. Center for Computational Science and e-Systems of Japan Atomic Energy Agency has been developing a finite element analysis code for assembled structure to accurately evaluate the structural integrity of nuclear power plant in its entirety under seismic events. Because nuclear power plant is very huge assembled structure with tens of millions of mechanical components, the finite element model of each component is assembled into one structure and non-conforming meshes of mechanical components are bonded together inside the code. The main technique to bond these mechanical components is triple sparse matrix multiplication with multiple point constrains and global stiffness matrix. In our code, this procedure is conducted in a component by component manner, so that the working memory size and computing time for this multiplication are available on the current computing environment. As an illustrative example, seismic simulation of a real nuclear reactor of High Temperature engineering Test Reactor, which is located at the O-arai research and development center of JAEA, with 80 major mechanical components was conducted. Consequently, our code successfully simulated detailed elasto-plastic deformation of nuclear reactor and its computational performance was investigated. (author)

  1. Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

    Porter, D.L., E-mail: Douglas.Porter@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Tsai Hanchung [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439-4803 (United States)

    2012-08-15

    The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt.%) metallic fast reactor test with commercial-length (91.4-cm active fuel-column length) conducted to date. With few remaining test reactors, there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning-of-life (BOL) peak cladding temperature of the hottest pin was 608 Degree-Sign C, cooling to 522 Degree-Sign C at end-of-life (EOL). Selected fuel pins were examined non-destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta-gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3-cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at {approx}0.7 X/L axial location along the fuel column. This resulted from a higher production of rare-earth fission products at this location and a higher {Delta}T between fuel center and cladding than at core center, together providing more rare earths at the cladding and

  2. ZOCO VI - a computer code to calculate the time- and space-dependent pressure distribution in full pressure containments of water-cooled reactors

    Mansfeld, G.

    1974-12-01

    ZOCO VI is a computer code to investigate the time and space dependent pressure distribution in full pressure containment of water cooled nuclear power reactors following a loss-of-coolant accident, which is caused by the rupture of a main coolant or steam line. ZOCO VI is an improved version of the computer code ZOCO V with enlarged description of condensing events. (orig.) [de

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

    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

  4. Comparative analysis of taxonomic, functional, and metabolic patterns of microbiomes from 14 full-scale biogas reactors by metagenomic sequencing and radioisotopic analysis.

    Luo, Gang; Fotidis, Ioannis A; Angelidaki, Irini

    2016-01-01

    Biogas production is a very complex process due to the high complexity in diversity and interactions of the microorganisms mediating it, and only limited and diffuse knowledge exists about the variation of taxonomic and functional patterns of microbiomes across different biogas reactors, and their relationships with the metabolic patterns. The present study used metagenomic sequencing and radioisotopic analysis to assess the taxonomic, functional, and metabolic patterns of microbiomes from 14 full-scale biogas reactors operated under various conditions treating either sludge or manure. The results from metagenomic analysis showed that the dominant methanogenic pathway revealed by radioisotopic analysis was not always correlated with the taxonomic and functional compositions. It was found by radioisotopic experiments that the aceticlastic methanogenic pathway was dominant, while metagenomics analysis showed higher relative abundance of hydrogenotrophic methanogens. Principal coordinates analysis showed the sludge-based samples were clearly distinct from the manure-based samples for both taxonomic and functional patterns, and canonical correspondence analysis showed that the both temperature and free ammonia were crucial environmental variables shaping the taxonomic and functional patterns. The study further the overall patterns of functional genes were strongly correlated with overall patterns of taxonomic composition across different biogas reactors. The discrepancy between the metabolic patterns determined by metagenomic analysis and metabolic pathways determined by radioisotopic analysis was found. Besides, a clear correlation between taxonomic and functional patterns was demonstrated for biogas reactors, and also the environmental factors that shaping both taxonomic and functional genes patterns were identified.

  5. COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

    ALEKSEY. L. IZHUTOV; VALERIY. V. IAKOVLEV; ANDREY. E. NOVOSELOV; VLADIMIR. A. STARKOV; ALEKSEY. A. SHELDYAKOV; VALERIY. YU. SHISHIN; VLADIMIR. M. KOSENKOV; ALEKSANDR. V. VATULIN; IRINA. V. DOBRIKOVA; VLADIMIR. B. SUPRUN; GENNADIY. V. KULAKOV

    2013-01-01

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; th...

  6. Reactors

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  7. Antibiotic Resistance Genes and Correlations with Microbial Community and Metal Resistance Genes in Full-Scale Biogas Reactors As Revealed by Metagenomic Analysis

    Luo, Gang; Li, Bing; Li, Li-Guan

    2017-01-01

    resistance genes (MRGs). The total abundance of ARGs in all the samples varied from 7 × 10-3 to 1.08 × 10-1 copy of ARG/copy of 16S-rRNA gene, and the samples obtained from thermophilic biogas reactors had a lower total abundance of ARGs, indicating the superiority of thermophilic anaerobic digestion......Digested residues from biogas plants are often used as biofertilizers for agricultural crops cultivation. The antibiotic resistance genes (ARGs) in digested residues pose a high risk to public health due to their potential spread to the disease-causing microorganisms and thus reduce...... the susceptibility of disease-causing microorganisms to antibiotics in medical treatment. A high-throughput sequencing (HTS)-based metagenomic approach was used in the present study to investigate the variations of ARGs in full-scale biogas reactors and the correlations of ARGs with microbial communities and metal...

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

    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

  9. Prokaryotic diversity and dynamics in a full-scale municipal solid waste anaerobic reactor from start-up to steady-state conditions.

    Cardinali-Rezende, Juliana; Colturato, Luís F D B; Colturato, Thiago D B; Chartone-Souza, Edmar; Nascimento, Andréa M A; Sanz, José L

    2012-09-01

    The prokaryotic diversity of an anaerobic reactor for the treatment of municipal solid waste was investigated over the course of 2 years with the use of 16S rDNA-targeted molecular approaches. The fermentative Bacteroidetes and Firmicutes predominated, and Proteobacteria, Actinobacteria, Tenericutes and the candidate division WWE1 were also identified. Methane production was dominated by the hydrogenotrophic Methanomicrobiales (Methanoculleus sp.) and their syntrophic association with acetate-utilizing and propionate-oxidizing bacteria. qPCR demonstrated the predominance of the hydrogenotrophic over aceticlastic Methanosarcinaceae (Methanosarcina sp. and Methanimicrococcus sp.), and Methanosaetaceae (Methanosaeta sp.) were measured in low numbers in the reactor. According to the FISH and CARD-FISH analyses, Bacteria and Archaea accounted for 85% and 15% of the cells, respectively. Different cell counts for these domains were obtained by qPCR versus FISH analyses. The use of several molecular tools increases our knowledge of the prokaryotic community dynamics from start-up to steady-state conditions in a full-scale MSW reactor. Copyright © 2012 Elsevier Ltd. All rights reserved.

  10. Strategy to identify the causes and to solve a sludge granulation problem in methanogenic reactors: application to a full-scale plant treating cheese wastewater.

    Macarie, Hervé; Esquivel, Maricela; Laguna, Acela; Baron, Olivier; El Mamouni, Rachid; Guiot, Serge R; Monroy, Oscar

    2017-08-26

    Granulation of biomass is at the basis of the operation of the most successful anaerobic systems (UASB, EGSB and IC reactors) applied worldwide for wastewater treatment. Despite of decades of studies of the biomass granulation process, it is still not fully understood and controlled. "Degranulation/lack of granulation" is a problem that occurs sometimes in anaerobic systems resulting often in heavy loss of biomass and poor treatment efficiencies or even complete reactor failure. Such a problem occurred in Mexico in two full-scale UASB reactors treating cheese wastewater. A close follow-up of the plant was performed to try to identify the factors responsible for the phenomenon. Basically, the list of possible causes to a granulation problem that were investigated can be classified amongst nutritional, i.e. related to wastewater composition (e.g. deficiency or excess of macronutrients or micronutrients, too high COD proportion due to proteins or volatile fatty acids, high ammonium, sulphate or fat concentrations), operational (excessive loading rate, sub- or over-optimal water upflow velocity) and structural (poor hydraulic design of the plant). Despite of an intensive search, the causes of the granulation problems could not be identified. The present case remains however an example of the strategy that must be followed to identify these causes and could be used as a guide for plant operators or consultants who are confronted with a similar situation independently of the type of wastewater. According to a large literature based on successful experiments at lab scale, an attempt to artificially granulate the industrial reactor biomass through the dosage of a cationic polymer was also tested but equally failed. Instead of promoting granulation, the dosage caused a heavy sludge flotation. This shows that the scaling of such a procedure from lab to real scale cannot be advised right away unless its operability at such a scale can be demonstrated.

  11. Evaluation of In-Core Fuel Management for the Transition Cores of RSG-GAS Reactor to Full-Silicide Core

    S, Tukiran; MS, Tagor; P, Surian

    2003-01-01

    The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 gU/cc has been done. The core-of RSG-GAS reactor has been operated full core of silicide fuels which is started with the mixed core of oxide-silicide start from core 36. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 9 transition cores (core 36 - 44) to achieve a full-silicide core (core 45). The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters. Conversion core was achieved by transition cores mixed oxide-silicide fuels. Each transition core is calculated and measured core parameter such as, excess reactivity and shutdown margin. Calculation done by Batan-EQUIL-2D code and measurement of the core parameters was carried out using the method of compensation of couple control rods. The results of calculation and experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safely to a full-silicide core

  12. Experimental evaluation of the oxygen transfer in bubble aeration systems. Full scale experiences in lengthened activated sludge reactors

    Andreottola, G.; Ragazzi, M.; Tatano, F.

    1999-01-01

    The results of some full-scale oxygen transfer measurements conduced at the lengthened activate sludge tanks of two WWTPs of Trentino Region, are presented and discussed. As far at the tests in clean water are concerned, the non-liner regression method seems non accurate; important conclusion on the correlation between oxygen transfer process and typical parameters (i.e., fine-bubble diffusers, specific air flux) are derived. As far as the test in the wastewater is concerned, an increase of α-value from the inlet to the end of aeration tanks has been observed in the 'Andalo' WWTP [it

  13. A full-scale UASB reactor for treatment of pig and cattle slaughterhouse wastewater with a high oil and grease content

    L. A. S. Miranda

    2005-12-01

    Full Text Available This paper discusses the performance of an 800m³ full-scale UASB reactor in treating meat-packing plant and slaughterhouse effluents containing high concentrations of oil and grease (O&G (413-645 mg/L, resulting in a COD/O&G ratio of 26-32%. Those macromolecules were considered responsible for the unbalance of the system resulting in a total washout of the biomass. The removal of O&G from the influent using a physicochemical system (coagulation-flocculation improved the physical characteristics of the anaerobic sludge, controlling the biomass washout. Reactor performance was significantly improved when the COD/O&G ratio influent was maintained in the 10%. The COD and O&G removal rates obtained after implantation of the physicochemical system were 70-92% and 27-58%, respectively. The specific methanogenic activity (SMA of the biomass shows towards a tendency stabilisation and adaptation to the substrate influent. Pretreatment of the influent allowed the maximum organic load to be increased (1.46 to 2.43 Kg COD/m³.d and improved the quality of the effluent.

  14. Relative power density distribution calculations of the Kori unit 1 pressurized water reactor with full-scope explicit modeling of monte carlo simulation

    Kim, J. O.; Kim, J. K.

    1997-01-01

    Relative power density distributions of the Kori unit 1 pressurized water reactor calculated by Monte Carlo modeling with the MCNP code. The Kori unit 1 core is modeled on a three-dimensional representation of the one-eighth of the reactor in-vessel component with reflective boundaries at 0 and 45 degrees. The axial core model is based on half core symmetry and is divided into four axial segments. Fission reaction density in each rod is calculated by following 100 cycles with 5,000 test neutrons in each cycle after starting with a localized neutron source and ten noncontributing settle cycles. Relative assembly power distributions are calculated from fission reaction densities of rods in assembly. After 100 cycle calculations, the system coverages to a κ value of 1.00039 ≥ 0.00084. Relative assembly power distribution is nearly the same with that of the Kori unit 1 FSAR. Applicability of the full-scope Monte Carlo simulation in the power distribution calculation is examined by the relative root mean square error of 2.159%. (author)

  15. Two-Dimensional Mapping of the Calculated Fission Power for the Full-Size Fuel Plate Experiment Irradiated in the Advanced Test Reactor

    Chang, G. S.; Lillo, M. A.

    2009-08-01

    The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y

  16. Full Core Criticality Modeling of Gas-Cooled Fast Reactor Using the SCALE6.0 and MCNP5 Code Packages

    Matijevic, M.; Jecmenica, R.; Pevec, D.; Trontl, K.

    2012-01-01

    The Gas-Cooled Fast Reactor (GFR) is one of the reactor concepts selected by the Generation IV International Forum (GIF) for the next generation of innovative nuclear energy systems. It was selected among a group of more than 100 prototypes and his commercial availability is expected by 2030. GFR has common goals of the rest GIF advanced reactor types: economy, safety, proliferation resistance, availability and sustainability. Several GFR fuel design concepts such as plates, rod pins and pebbles are currently being investigated in order to meet the high temperature constraints characteristic for a GFR working enviroment. In the previous study we have compared the fuel depletion results for heterogeneous GFR fuel assembly (FA), obtained with TRITON6 sequence of SCALE6.0 code system, with the MCNPX-CINDER90 and TRIPOLI-4-D codes. Present work is a continuation of neutronic criticality analysis of heterogeneous FA and full core configurations of a GFR concept using 3-D Monte Carlo codes KENO-VI/SCALE6.0 and MCNP5. The FA is based on a hexagonal mesh of fuel rods (uranium and plutonium carbide fuel, silicon carbide clad, helium gas coolant) with axial reflector thickness being varied for the purpose of optimization. Three reflector materials were analysed: zirconium carbide (ZrC), silicon carbide (SiC) and natural uranium. ZrC has been selected as a reflector material, having the best contribution to the neutron economy and to the reactivity of the core. The core safety parameters were also analysed: a negative temperature coefficient of reactivity was verified for the heavy metal fuel and coolant density loss. Criticality calculations of different FA active heights were performed and the reflector thickness was also adjusted. Finally, GFR full core criticality calculations using different active fuel rod heights and fixed ZrC reflector height were done to find the optimal height of the core. The Shannon entropy of the GFR core fission distribution was proved to be

  17. Appling hydrolysis acidification-anoxic–oxic process in the treatment of petrochemical wastewater: From bench scale reactor to full scale wastewater treatment plant

    Wu, Changyong; Zhou, Yuexi; Sun, Qingliang; Fu, Liya; Xi, Hongbo; Yu, Yin; Yu, Ruozhen

    2016-01-01

    Highlights: • Hydrolysis acidification-anoxic–oxic process can be used to treat petrochemical wastewater. • The toxicity and treatability changed significantly after hydrolysis acidification. • The type and concentration of organics reduced greatly after treatment. • The effluent shows low acute toxicity by luminescent bacteria assay. • Advanced treatment is recommended for the effluent. - Abstract: A hydrolysis acidification (HA)-anoxic–oxic (A/O) process was adopted to treat a petrochemical wastewater. The operation optimization was carried out firstly by a bench scale experimental reactor. Then a full scale petrochemical wastewater treatment plant (PCWWTP, 6500 m 3 h −1 ) was operated with the same parameters. The results showed that the BOD 5 /COD of the wastewater increased from 0.30 to 0.43 by HA. The effluent COD was 54.4 mg L −1 for bench scale reactor and 60.9 mg L −1 for PCWWTP when the influent COD was about 480 mg L −1 on optimized conditions. The organics measured by gas chromatography-mass spectrometry (GC–MS) reduced obviously and the total concentration of the 5 organics (1,3-dioxolane, 2-pentanone, ethylbenzene, 2-chloromethyl-1,3-dioxolane and indene) detected in the effluent was only 0.24 mg L −1 . There was no obvious toxicity of the effluent. However, low acute toxicity of the effluent could be detected by the luminescent bacteria assay, indicating the advanced treatment is needed. The clone library profiling analysis showed that the dominant bacteria in the system were Acidobacteria, Proteobacteria and Bacteriodetes. HA-A/O process is suitable for the petrochemical wastewater treatment.

  18. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-12-15

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%{sup 235}U; the mini-rods were irradiated to an average burnup of ∼ 85%{sup 235}U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  19. Optimization aspects of the biological nitrogen removal process in a full-scale twin sequencing batch reactor (SBR) system in series treating landfill leachate.

    Remmas, Nikolaos; Ntougias, Spyridon; Chatzopoulou, Marianna; Melidis, Paraschos

    2018-03-29

    Despite the fact that biological nitrogen removal (BNR) process has been studied in detail in laboratory- and pilot-scale sequencing batch reactor (SBR) systems treating landfill leachate, a limited number of research works have been performed in full-scale SBR plants regarding nitrification and denitrification. In the current study, a full-scale twin SBR system in series of 700 m 3 (350 m 3 each) treating medium-age landfill leachate was evaluated in terms of its carbon and nitrogen removal efficiency in the absence and presence of external carbon source, i.e., glycerol from biodiesel production. Both biodegradable organic carbon and ammonia were highly oxidized [biochemical oxygen demand (BOD 5 ) and total Kjehldahl nitrogen (TKN) removal efficiencies above 90%], whereas chemical oxygen demand (COD) removal efficiency was slightly above 40%, which is within the range reported in the literature for pilot-scale SBRs. As the consequence of the high recalcitrant organic fraction of the landfill leachate, dissimilatory nitrate reduction was restricted in the absence of crude glycerol, although denitrification was improved by electron donor addition, resulting in TN removal efficiencies above 70%. Experimental data revealed that the second SBR negligibly contributed to BNR process, since carbon and ammonia oxidation completion was achieved in the first SBR. On the other hand, the low VSS/SS ratio, due to the lack of primary sedimentation, highly improved sludge settleability, resulting in sludge volume indices (SVI) below 30 mL g -1 .

  20. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-01-01

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60% 235 U; the mini-rods were irradiated to an average burnup of ∼ 85% 235 U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%

  1. Nuclear reactors

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  2. Full-scale model development of the WWER-440 reactor fuel rod bundle for core temperature regime study under reflooding conditions

    Bezrukov, Yu.A.; Logvinov, S.A.; Levchuk, S.V.; Nakladnov, V.D.; Onshin, V.P.; Sokolov, A.S.

    1982-01-01

    Consideration is given to the issues of a full scale WWER-440 fuel rod bundle imitation. An imitator contains a molybdenum heating rod inclosed in stainless steel shell. The shell diameter is 9 mm, the heated length is 2500 mm, the total len.o.th is 2855 mm. 125 fuel rod imitators are set in the bundle mock-up. The experiments were run on a test facility imitating the WWER-440 reactor primary loop, providing the conditions of the loop breaking. The mock-up thermal hydraulics has been studied during the refloodino. stage. The mock-up was heated up to predetermined initial temperature at a low power level with saturated steam cooling. Then the steam input was stopped, the power level rarapidly rised up to a given value and the cooling water injected. Simultaneously with water injection all the measured parameters monitoring was started. Both at the top spraying and combined cooling temperature oscillations in the upper and middle parts of the mock-up were observed. At the bottom reflooding the mock-up cooling down took more time, thereat temperature inthe upper part first slowly rised during reflooding then decreased and then dropped abruptly at thefront coming up [ru

  3. Full-Scale Modeling Explaining Large Spatial Variations of Nitrous Oxide Fluxes in a Step-Feed Plug-Flow Wastewater Treatment Reactor.

    Ni, Bing-Jie; Pan, Yuting; van den Akker, Ben; Ye, Liu; Yuan, Zhiguo

    2015-08-04

    Nitrous oxide (N2O) emission data collected from wastewater treatment plants (WWTPs) show huge variations between plants and within one plant (both spatially and temporarily). Such variations and the relative contributions of various N2O production pathways are not fully understood. This study applied a previously established N2O model incorporating two currently known N2O production pathways by ammonia-oxidizing bacteria (AOB) (namely the AOB denitrification and the hydroxylamine pathways) and the N2O production pathway by heterotrophic denitrifiers to describe and provide insights into the large spatial variations of N2O fluxes in a step-feed full-scale activated sludge plant. The model was calibrated and validated by comparing simulation results with 40 days of N2O emission monitoring data as well as other water quality parameters from the plant. The model demonstrated that the relatively high biomass specific nitrogen loading rate in the Second Step of the reactor was responsible for the much higher N2O fluxes from this section. The results further revealed the AOB denitrification pathway decreased and the NH2OH oxidation pathway increased along the path of both Steps due to the increasing dissolved oxygen concentration. The overall N2O emission from this step-feed WWTP would be largely mitigated if 30% of the returned sludge were returned to the Second Step to reduce its biomass nitrogen loading rate.

  4. Experimental studies and computational benchmark on heavy liquid metal natural circulation in a full height-scale test loop for small modular reactors

    Shin, Yong-Hoon, E-mail: chaotics@snu.ac.kr [Department of Energy Systems Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Cho, Jaehyun [Korea Atomic Energy Research Institute, 111 Daedeok-daero, 989 Beon-gil, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Lee, Jueun; Ju, Heejae; Sohn, Sungjune; Kim, Yeji; Noh, Hyunyub; Hwang, Il Soon [Department of Energy Systems Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of)

    2017-05-15

    Highlights: • Experimental studies on natural circulation for lead-bismuth eutectic were conducted. • Adiabatic wall boundaries conditions were established by compensating heat loss. • Computational benchmark with a system thermal-hydraulics code was performed. • Numerical simulation and experiment showed good agreement in mass flow rate. • An empirical relation was formulated for mass flow rate with experimental data. - Abstract: In order to test the enhanced safety of small lead-cooled fast reactors, lead-bismuth eutectic (LBE) natural circulation characteristics have been studied. We present results of experiments with LBE non-isothermal natural circulation in a full-height scale test loop, HELIOS (heavy eutectic liquid metal loop for integral test of operability and safety of PEACER), and the validation of a system thermal-hydraulics code. The experimental studies on LBE were conducted under steady state as a function of core power conditions from 9.8 kW to 33.6 kW. Local surface heaters on the main loop were activated and finely tuned by trial-and-error approach to make adiabatic wall boundary conditions. A thermal-hydraulic system code MARS-LBE was validated by using the well-defined benchmark data. It was found that the predictions were mostly in good agreement with the experimental data in terms of mass flow rate and temperature difference that were both within 7%, respectively. With experiment results, an empirical relation predicting mass flow rate at a non-isothermal, adiabatic condition in HELIOS was derived.

  5. S.E.N.S.I.B. project

    2006-01-01

    This report presents the state of progress of all the studies which constitute at present the S.E.N.S.I.B. project. The S.E.N.S.I.B. project receives a financial participation of the Ademe. The different chapters treat the following questions: the sensitivity of territories in the deposit; the sensitivity of grounds; the sensitivity of the banks of rivers; the sensitivity of the agricultural productions; the anthropological sensitivity of territories; comparative study of the global sensitivity of two sites; uncertainties, communication, perception and representation; assessment of the contributions to the S.E.N.S.I.B. project in 2005. (N.C.)

  6. arXiv Performance of a full scale prototype detector at the BR2 reactor for the SoLid experiment

    Abreu, Y.; Arnold, L.; Ban, G.; Beaumont, W.; Bongrand, M.; Boursette, D.; Castle, B.C.; Clark, K.; Coupé, B.; Cussans, D.; De Roeck, A.; D'Hondt, J.; Durand, D.; Fallot, M.; Ghys, L.; Giot, L.; Guillon, B.; Ihantola, S.; Janssen, X.; Kalcheva, S.; Kalousis, L.N.; Koonen, E.; Labare, M.; Lehaut, G.; Manzanillas, L.; Mermans, J.; Michiels, I.; Moortgat, C.; Newbold, D.; Park, J.; Pestel, V.; Petridis, K.; Piñera, I.; Pommery, G.; Popescu, L.; Pronost, G.; Rademacker, J.; Ryckbosch, D.; Ryder, N.; Saunders, D.; Schune, M.-H.; Simard, L.; Vacheret, A.; Van Dyck, S.; Van Mulders, P.; van Remortel, N.; Vercaemer, S.; Verstraeten, M.; Weber, A.; Yermia, F.

    2018-05-03

    The SoLid collaboration has developed a new detector technology to detect electron anti-neutrinos at close proximity to the Belgian BR2 reactor at surface level. A 288 kg prototype detector was deployed in 2015 and collected data during the operational period of the reactor and during reactor shut-down. Dedicated calibration campaigns were also performed with gamma and neutron sources. This paper describes the construction of the prototype detector with a high control on its proton content and the stability of its operation over a period of several months after deployment at the BR2 reactor site. All detector cells provide sufficient light yields to achieve a target energy resolution of better than 20%/√E(MeV). The capability of the detector to track muons is exploited to equalize the light response of a large number of channels to a precision of 3% and to demonstrate the stability of the energy scale over time. Particle identification based on pulse-shape discrimination is demonstrated with calibration so...

  7. Determination and Variation of Core Bacterial Community in a Two-Stage Full-Scale Anaerobic Reactor Treating High-Strength Pharmaceutical Wastewater.

    Ma, Haijun; Ye, Lin; Hu, Haidong; Zhang, Lulu; Ding, Lili; Ren, Hongqiang

    2017-10-28

    Knowledge on the functional characteristics and temporal variation of anaerobic bacterial populations is important for better understanding of the microbial process of two-stage anaerobic reactors. However, owing to the high diversity of anaerobic bacteria, close attention should be prioritized to the frequently abundant bacteria that were defined as core bacteria and putatively functionally important. In this study, using MiSeq sequencing technology, the core bacterial community of 98 operational taxonomic units (OTUs) was determined in a two-stage upflow blanket filter reactor treating pharmaceutical wastewater. The core bacterial community accounted for 61.66% of the total sequences and accurately predicted the sample location in the principal coordinates analysis scatter plot as the total bacterial OTUs did. The core bacterial community in the first-stage (FS) and second-stage (SS) reactors were generally distinct, in that the FS core bacterial community was indicated to be more related to a higher-level fermentation process, and the SS core bacterial community contained more microbes in syntrophic cooperation with methanogens. Moreover, the different responses of the FS and SS core bacterial communities to the temperature shock and influent disturbance caused by solid contamination were fully investigated. Co-occurring analysis at the Order level implied that Bacteroidales, Selenomonadales, Anaerolineales, Syneristales, and Thermotogales might play key roles in anaerobic digestion due to their high abundance and tight correlation with other microbes. These findings advance our knowledge about the core bacterial community and its temporal variability for future comparative research and improvement of the two-stage anaerobic system operation.

  8. Technical-and-economic analysis and optimization of the full flow charts of processing of radioactive wastes on a polyfunctional plant of pyrochemical processing of the spent nuclear fuel of fast reactors

    Gupalo, V. S.; Chistyakov, V. N.; Kormilitsyn, M. V.; Kormilitsyna, L. A.; Osipenko, A. G.

    2015-12-01

    When considering the full flow charts of processing of radioactive wastes (RAW) on a polyfunctional plant of pyrochemical processing of the spent nuclear fuel of NIIAR fast reactors, we corroborate optimum technical solutions for the preparation of RAW for burial from a standpoint of heat release, dose formation, and technological storage time with allowance for technical-and-economic and ecological indices during the implementation of the analyzed technologies and equipment for processing of all RAW fluxes.

  9. Phenotypic properties and microbial diversity of methanogenic granules from a full-scale upflow anaerobic sludge bed reactor treating brewery wastewater.

    Díaz, Emiliano E; Stams, Alfons J M; Amils, Ricardo; Sanz, José L

    2006-07-01

    Methanogenic granules from an anaerobic bioreactor that treated wastewater of a beer brewery consisted of different morphological types of granules. In this study, the microbial compositions of the different granules were analyzed by molecular microbiological techniques: cloning, denaturing gradient gel electrophoresis and fluorescent in situ hybridization (FISH), and scanning and transmission electron microscopy. We propose here that the different types of granules reflect the different stages in the life cycle of granules. Young granules were small, black, and compact and harbored active cells. Gray granules were the most abundant granules. These granules have a multilayer structure with channels and void areas. The core was composed of dead or starving cells with low activity. The brown granules, which were the largest granules, showed a loose and amorphous structure with big channels that resulted in fractured zones and corresponded to the older granules. Firmicutes (as determined by FISH) and Nitrospira and Deferribacteres (as determined by cloning and sequencing) were the predominant Bacteria. Remarkably, Firmicutes could not be detected in the brown granules. The methanogenic Archaea identified were Methanosaeta concilii (70 to 90% by FISH and cloning), Methanosarcina mazei, and Methanospirillum spp. The phenotypic appearance of the granules reflected the physiological condition of the granules. This may be valuable to easily select appropriate seed sludges to start up other reactors.

  10. Color coherent effects in (e,e'N) and (e,e'N,N(h)) processes at CEBAF

    Frankfurt, L.L.; Sargsyan, M.M.; Strikman, M.I.

    1994-01-01

    The options for investigating color coherent effects and competing nuclear effects of nucleon-nucleon correlations in nuclei, nuclear shell effects in (e, e'N) and (e, e'NN(h)) reactions are considered. They argue that extension of CEBAF energies to reach Q 2 = 10 GeV 2 will allow systematical investigations of color coherent effects in nonperturbative regime of QCD and their interplay with nuclear effects

  11. On the transient pressure build-up in the full pressure safety shell of watercooled nuclear reactors after a loss of coolant accident

    Mansfeld, G.

    1979-08-01

    The thermo-and fluid-dynamic processes in a multichamber full pressure safety containment during a loss of coolant accident have been investigated. Comparison of the calculations carried out with the computer programs, in which ZOCO VI was used as being representative of similar programs, with the experimental results pointed out discrepancies in the determination of time dependent pressure, pressure difference and temperature curves. This led to the development of a new theoretical model and a program COFLOW which pays particular attention to the fluid dynamic processes in the initial phase of a loss of coolant accident. It can also be used to determine the maximum containment pressure towards the end of a loss of coolant accident. Comparison of the COFLOW results with experiments has shown that COFLOW provides a model and a procedure by which the physical processes in a multichamber full pressure safety containment can be simulated satisfactorily

  12. COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

    ALEKSEY. L. IZHUTOV

    2013-12-01

    The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; the mini-rods were irradiated to an average burnup of ∼ 85%235U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  13. Ambiguities in the reaction mechanism for (e,e'N)

    Forest, T. de Jr.

    1984-01-01

    The primary motivation for performing quasi-elastic (e,e'N) experiments revolves around the plane wave impulse approximation (PWIA) description of this reaction. Since the PWIA is an approximation, corrections are necessary in practice in order to extract the desired nuclear structure. Unless one understands the physics behind these corrections this introduces ambiguities. In fact the PWIA itself is an ambiguous prescription since a 'free' off-shell cross section is not a well-defined concept. It is these ambiguities which are discussed in this talk. Most of the paper is devoted to the ambiguities associated with the electromagnetic interaction. The author concentrates on four topics: (1) the interaction of the electron with the nucleus in general; (2) ambiguities in the application of the impulse approximation; (3) the sigma-omega model; and (4) the Coulomb sum rule. (Auth.)

  14. Nuclear reactor coolant channels

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  15. A fuel management study and cycle nuclear design for PW reactors

    Minguez, E.; Ahnert, C.; Aragones, J. M.; Corella, M. R.

    1975-01-01

    A reference reactor was chosen to do a general study involving Fuel Management Evaluations of several cycles, and Design Calculations of cycles already performed, according to a calculation scheme set up in the Reactor Technology Division of the J.E.N., using some computer codes acquired to foreign sources and other ones developed in the J.E.N. (Author) 5 refs

  16. A fuel management study and cycle nuclear design for PW reactors

    Minguez, E; Ahnert, C; Aragones, J M; Corella, M R

    1975-07-01

    A reference reactor was chosen to do a general study involving Fuel Management Evaluations of several cycles, and Design Calculations of cycles already performed, according to a calculation scheme set up in the Reactor Technology Division of the J.E.N., using some computer codes acquired to foreign sources and other ones developed in the J.E.N. (Author) 5 refs.

  17. Nuclear reactors

    Prescott, R F; George, B V; Baglin, C J

    1978-05-10

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given.

  18. Nuclear reactors

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1978-01-01

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

  19. (E-N-(3,3-Diphenylallylidene-3-nitroaniline

    Joo Hwan Cha

    2012-10-01

    Full Text Available In the title compound, C21H16N2O2, the 3-nitrophenyl and two phenyl rings are twisted from the mean plane of the enimino fragment by 44.4 (1, 37.2 (1 and 74.1 (1°, respectively. The crystal packing exhibits no classical intermolecular contacts.

  20. Full system decontamination feasibility studies

    Denault, R.P.; LeSurf, J.E.; Walschot, F.W.

    1988-01-01

    Many chemical decontaminations have been performed on subsystems in light water reactors (BWRs and PWRs) but none on the full system (including the fuel) of large, (>500 MWe) investor owned reactors. Full system decontaminations on pressure-tubed reactors have been shown to facilitate maintenance, inspection, repair and replacement of reactor components. Further advantages are increased reactor availability and plant life extension. A conceptual study has been performed for EPRI (for PWRs) and Commonwealth Edison Co (for BWRs) into the applicability and cost benefit of full system decontaminations (FSD). The joint study showed that FSDs in both PWRs and BWRs, with or without the fuel included in the decontamination, are feasible and cost beneficial provided a large amount of work is to be done following the decontamination. The large amounts of radioactive waste generated can be managed using current technologies. Considerable improvements in waste handling, and consequent cost savings, can be obtained if new techniques which are now reaching commercial application are used. (author)

  1. (E-N-[Cyclopentyl(morpholin-4-ylmethylidene]-4-fluorobenzenesulfonamide

    Ilya Efimov

    2016-01-01

    Full Text Available The title compound, C16H21FN2O3S, was obtained from the reaction between sulfonyl azide, cyclohexanone and morpholine. The bond lengths at the amidine N—C—N grouping are similar [1.326 (3 and 1.338 (3 Å], indicating significant conjugation. The cyclopentyl moiety displays disorder of one of the methylene groups into two orientations with occupancy coefficients 0.75/0.25. No shortened intermolecular contacts in the crystal are observed.

  2. Traditional M e n Caps o f Kyrgyzstan

    Serap TANRISEVER

    2015-07-01

    Full Text Available Kyrgyzstan has an extraordinarily rich mosaic of folk culture. There is a great importance of wear clothing, crafts and decorative arts in this mosaic. Traditional caps and clothes of Kyrgyzstan are the best elements that reflect the presence of culture, t raditions, geography inhabited, climatic conditions that they already live in. Kyrgyzstan has been trying to keep alive the traditions, styles of lives by using the required crafts and handmade products although the global fashion, time and technological development. According to research, it has still been seen that the caps and clothes of Kyrgyzstan are being used less in daily life. Caps, Kalpak , Tebetey, Malakay, are some of the best known extant that come to this age related with the traditions of ca p s of Kyrgyzstan’s men . In this research, it has been worked of decorations of Kirghiz men caps , specialties of color and design, techniques of sewing, characteristics of mold and fabrics, Moreover the place and the meaning of the titles has been trying to determine in the cultural life. The scan method has been used by doing regional research.

  3. Nuclear reactors

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  4. S.E.N.S.I.B. project; Projet SENSIB. Rapport d'avancement 2005

    NONE

    2006-07-01

    This report presents the state of progress of all the studies which constitute at present the S.E.N.S.I.B. project. The S.E.N.S.I.B. project receives a financial participation of the Ademe. The different chapters treat the following questions: the sensitivity of territories in the deposit; the sensitivity of grounds; the sensitivity of the banks of rivers; the sensitivity of the agricultural productions; the anthropological sensitivity of territories; comparative study of the global sensitivity of two sites; uncertainties, communication, perception and representation; assessment of the contributions to the S.E.N.S.I.B. project in 2005. (N.C.)

  5. S.E.N.S.I.B. project. Progress report 2006

    2007-01-01

    This report presents the state of progress of all the studies which establish at present the S.E.N.S.I.B. project. For year 2006, the progress of the project is globally in compliance with the general schedule of realization of the S.E.N.S.I.B. project and with the perspectives announced in 2005. Factors of sensitivity were identified in diverse circles ( thematic studies) and methods and specific tools of the project are developed. 14 publications (reviews and congress) and 9 I.R.S.N. reports were produced. An international work group was launched to the I.R.S.N. initiative. The web site of S.E.N.S.I.B. on the I.R.S.N. scientific site was built. The S.E.N.S.I.B. project receives a financial participation of the Ademe. (N.C.)

  6. Benthic megafauna at deep-sea study areas W-N and E-N

    Keller, C.H.

    1985-01-01

    Photographs were utilized determine what megafaunal organisms live at areas E-N and W-N and in what densities. Life history characteristics gleaned from the literature as well as from the photographs are presented to establish these organisms' roles in oceanic food chains. In area W-N, the most abundant megafaunal group is the echinoderms. These are the asteroids, crinoids, echinoids, holothuroids, and ophiuroids. Dominant amongst the echinoderms are the ophiuroids, which account for about 46% of the total density. Taxonomic and photographic resolution at area E-N are both poor. Numbers of megafaunal organisms cannot be directly compared between areas E-N and W-N. Differing photographic techniques severely limit comparisons between the two areas. Nevertheless, megafauna seen at area E-N is apparently less dense and probably less diverse than at area W-N. 30 references, 8 figures, 8 tables

  7. The drift velocity of electrons in water vapour at low values of E/N

    Cheung, B.; Elford, M.T.

    1990-01-01

    The drift velocity of electrons in water vapour at 294 K has been measured over the E/N range 1.4 to 40 Td with an error estimated to be 35 Td. The present data show that μN decreases monotonically with decreasing E/N at low E/N values as observed by Wilson et al. (1975) and does not become independent of E/N as indicated by Lowke and Rees (1963). The present values although lower than those of Lowke and Rees, lie within the combined error limits, except for values below 2 Td. The present data suggested that the momentum transfer cross section at low energies is approximately 10% larger than that obtained by Pack et al. (1962) from their drift velocity measurements. 13 refs., 2 tabs., 5 figs

  8. RA Reactor applications, Annex A

    Cupac, S.; Vukadin, Z.

    2000-01-01

    Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  9. RA Reactor applications, Annex A

    Cupac, S.; Vukadin, Z.

    1998-01-01

    Full text: In 1998 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  10. RA Reactor applications, Annex A

    Cupac, S.; Vukadin, Z.

    1996-01-01

    Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  11. Software é desenvolvido, e não fabricado como geladeira e fogão - Gerenciamento é essencial

    Antonio Mendes Silva Filho

    2012-10-01

    Full Text Available Software não é fabricado como geladeira e fogão que são montados. Software é um produto desenvolvido, de modo sistemático, através de um conjunto de atividades bem definidas, seguindo a execução disciplinada do gerenciamento de projeto. Software requer um processo de desenvolvimento e não um processo de ‘fabricação’, implicando que ele exigirá o uso da engenharia de software. Nesse sentido, este artigo explora a importancia da estimativa de tamanho de projeto para o desenvolvimento de software.

  12. H Reactor

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  13. Full page fax print

    user

    prepared by reducing the iron oxide m agnetite (Fe3 O 4 ) is used. In 1910 C arlB osch ... W ar,w hich w as then used to produce the m uch needed am m onium nitrate,for use in explosives. A t that tim e, all th e n atu ral .... hydrogens actually add up one by one to the nitrogen to generate ammonia. should not change as one ...

  14. Politics in the Western Maya Region (I: Ajawil/Ajawlel and Ch'e'n

    Péter Bíró

    2011-01-01

    Full Text Available In a series of articles I reflect on the use of various expressions which are connected to what we call the political in the inscriptions of the Classic Maya Western Region. These words express concepts which help to understand the intricate details of the interactions between the political entities and their internal organisations in the Classic Maya Lowlands. Words such as 7ajawil, 7ajawlel, the kennings built on the base of ch'e7n (cave, pond, the emblem glyphs and titles will be examined in light of what they tell us about the functioning of the political organisation of the Classic Period in a constrained region.En una serie de artículos como éste investigo el uso de varias palabras en las inscripciones mayas de la época Clásica de la Región Occidental vinculadas con lo que nosotros llamamos "política". Estas palabras expresan conceptos que ayudan a entender los matices de las relaciones entre las entidades políticas de las Tierras Bajas Mayas y su organización interna. Términos como ajaw'ü I ajawlel, los difrasismos con base ch'e'n (cueva, pozo, los glifos emblemas y los títulos serán examinados tomando en cuenta la información que nos proporcionan sobre el funcionamiento de la organización política de la época Clásica en una región restringida.

  15. Reactor Physics

    Ait Abderrahim, A.

    2002-01-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the 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 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  16. Reactor Physics

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA 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 of reactor fuel. This expertise is applied in the 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 by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  17. Reactor Physics

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the 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 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  18. Reactor Physics

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA 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 of reactor fuel. This expertise is applied in the 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 by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  19. Reactor operation

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  20. Reactor safeguards

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  1. The transport parameters of an electron swarm in nitrogen at elevated E/N

    Blevin, H.A.; Fletcher, J.; Reid, I.D.

    1980-06-01

    Values of the electron drift velocity, the longitudonal diffusion coefficient and the transverse diffusion coefficient for electron swarms in nitrogen over the range 50 Td <= E/N <500 Td have been determined by a method of counting and analysing the photons produced by such an electron swarm in a drift tube. A measure of the Nitrogen E state deactivation rate is presented

  2. ADAPTIVE CONTROL SYSTEM OF INDUSTRIAL REACTORS

    Vyacheslav K. Mayevski

    2014-01-01

    Full Text Available This paper describes a mathematical model of an industrial chemical reactor for production of synthetic rubber. During reactor operation the model parameters vary considerably. To create a control algorithm performed transformation of mathematical model of the reactor in order to obtain a dependency that can be used to determine the model parameters are changing during reactor operation.

  3. Reactor decommissioning

    Lawton, H.

    1984-01-01

    A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)

  4. Heat transfer tests conducted on full-scale model, to investigate cooling conditions of EL.3 experimental reactor; Essais de transmission de chaleur sur maquette pour l'etude du refroidissement de la pile EL 3

    Raievski, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Butzbach, M; Domenjoud, M [Alsthom, 75 - Paris (France); Bousquet, M [Chantiers de l' Atlantique (France); Braudeau, M; Milliat, M [Electricite de France (EDF), 75 - Paris (France)

    1958-07-01

    For such high heat flux density as is released in the channels of EL3 reactor (2.10{sup 6} kcal/m{sup 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)Fren. [French] A la valeur elevee prevue pour la densite de flux de chaleur (2.10{sup 6} kcal/m{sup 2}h au point le plus chaud) il est verifie que le refroidissement de la pile s'effectue normalement (sans ebullition de paroi). Les essais sont menes sur la maquette grandeur nature d'un canal d'EL3. Les dispositions relatives a la conduite des essais et a la technique des mesures (de temperature en particulier) sont precisees. Deux domaines sont etudies; pour t{sub p} < T{sub sat} (convection forcee sans ebullition de paroi) on constate un bon accord avec la formule de Colburn, avec toutefois l'influence du rapport L/D. Pour t{sub p} < T{sub sat} (debut d'ebullition) l'augmentation prevue du coefficient de transmission presente un certain retard. (auteur)

  5. RA Reactor

    1978-02-01

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

  6. Multiregion reactors

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

  7. Multi-purpose reactor

    1991-05-01

    The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MW t h, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co 60 ) production capacity is 50000 Ci/yr, 200 Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N 16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

  8. Nuclear reactor

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  9. Full MOX core design in ABWR

    Ihara, Toshiteru; Mochida, Takaaki; Izutsu, Sadayuki; Fujimaki, Shingo

    2003-01-01

    Electric Power Development Co., Ltd. (EPDC) has been investigating an ABWR plant for construction at Oma-machi in Aomori Prefecture. The reactor, termed FULL MOX-ABWR will have its reactor core eventually loaded entirely with mixed-oxide (MOX) fuel. Extended use of MOX fuel in the plant is expected to play important roles in the country's nuclear fuel recycling policy. MOX fuel bundles will initially be loaded only to less than one-third of the reactor, but will be increased to cover its entire core eventually. The number of MOX fuel bundles in the core thus varies anywhere from 0 to 264 for the initial cycle and, 0 to 872 for equilibrium cycles. The safety design of the FULL MOX-ABWR briefly stated next considers any probable MOX loading combinations out of such MOX bundle usage scheme, starting from full UO 2 to full MOX cores. (author)

  10. Nuclear power reactors

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  11. Research reactors

    Merchie, Francois

    2015-10-01

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  12. Reactor physics and reactor computations

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  13. Bathymetry and acoustic echo character lower continental rise study area, E-N2

    Laine, E.P.; Friedrich, N.E.

    1985-01-01

    A bathymetric map of area E-N2 was constructed using navigational and bathymetric data supplied by the Defense Mapping Agency and navigational and bathymetric data obtained during Endeavor cruises. E-N2 contains two dominant echo-types: an area of submarine canyons, and an area of levee deposits and debris flow deposits. The canyon system is complex, characterized by many channels and inter-channel levee deposits. The second echo province, characterized by levee and debris flow deposits, is a region suitable for the proposed operations (low-level radioactive waste disposal). Given a favorable analysis of data collected on cruise EN-084, attention will be focused at shallower depths to determine slope stability and other seabed conditions. 4 figures

  14. Research reactors

    Kowarski, L.

    1955-01-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  15. The replacement research reactor

    Cameron, R.; Horlock, K.

    2001-01-01

    The contract for the design, construction and commissioning of the Replacement Research Reactor was signed in July 2000. This was followed by the completion of the detailed design and an application for a construction licence was made in May 2001. This paper will describe the main elements of the design and their relation to the proposed applications of the reactor. The future stages in the project leading to full operation are also described

  16. Probing deformed orbitals with vector A( vector e, e' N)B reactions

    Garrido, E.; Caballero, J.A.; Moya de Guerra, E.; Sarriguren, P.; Udias, J.M.

    1995-01-01

    We present results for response functions and asymmetries in the nuclear reactions 37 vector Ar( vector e, e' n) 36 Ar and 37 vector K( vector e,e' p) 36 Ar at quasifree kinematics. We compare PWIA results obtained using deformed HF wave functions with PWIA and DWIA results obtained assuming a spherical mean field. We show that the complex structure of the deformed orbitals can be probed by coincidence measurements with polarized beam and targets. ((orig.))

  17. Full MOX high burn-up PWR

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Araya, Fumimasa; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of conceptual investigation on advanced light water reactors for the future, a light water reactor with the high burn-up of 100 GWd/t, the long cycle operation of 3 years and the full MOX core is being studied, aiming at the improvement on economical aspects, the reduction of the spent fuel production, the utilization of Plutonium and so forth. The present report summarizes investigation on PWR-type reactors. The core with the increased moderation of the moderator-to-fuel volume ratio of 2.6 {approx} 3.0 has been proposed be such a core that accomplishes requirements mentioned above. Through the neutronic and the thermo-hydrodynamic evaluation, the performances of the core have been evaluated. Also, the safety designing is underway considering the reactor system with the passive safety features. (author)

  18. A practical method for determining γ-ray full-energy peak efficiency considering coincidence-summing and self-absorption corrections for the measurement of environmental samples after the Fukushima reactor accident

    Shizuma, Kiyoshi, E-mail: shizuma@hiroshima-u.ac.jp [Graduate School of Engineering, Hiroshima University, Higashi-Hiroshima 739-8527 (Japan); Oba, Yurika; Takada, Momo [Graduate School of Integrated Arts and Sciences, Hiroshima University, Higashi-Hiroshima 739-8521 (Japan)

    2016-09-15

    A method for determining the γ-ray full-energy peak efficiency at positions close to three Ge detectors and at the well port of a well-type detector was developed for measuring environmental volume samples containing {sup 137}Cs, {sup 134}Cs and {sup 40}K. The efficiency was estimated by considering two correction factors: coincidence-summing and self-absorption corrections. The coincidence-summing correction for a cascade transition nuclide was estimated by an experimental method involving measuring a sample at the far and close positions of a detector. The derived coincidence-summing correction factors were compared with those of analytical and Monte Carlo simulation methods and good agreements were obtained. Differences in the matrix of the calibration source and the environmental sample resulted in an increase or decrease of the full-energy peak counts due to the self-absorption of γ-rays in the sample. The correction factor was derived as a function of the densities of several matrix materials. The present method was applied to the measurement of environmental samples and also low-level radioactivity measurements of water samples using the well-type detector.

  19. The integral fast reactor

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two dramatic demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the Integral Fast Reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics and also makes possible a simplified closed fuel cycle and waste process improvements

  20. The integral fast reactor

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the integral fast reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics also makes possible a simplified close fuel cycle and waste process improvements. The paper describes the IFR concept, the inherent safety, tests, and status of IFR development today

  1. Reactor container

    Naruse, Yoshihiro.

    1990-01-01

    The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

  2. Hybrid reactors

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  3. Nuclear reactor

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

  4. Nuclear reactor

    Batheja, P.; Huber, R.; Rau, P.

    1985-01-01

    Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

  5. Heterogeneous reactors

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

  6. Reactor BR2. Introduction

    Gubel, P.

    2002-01-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system

  7. Reactor BR2. Introduction

    Gubel, P

    2002-04-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

  8. Heterogeneous gas core reactor

    Han, K.I.

    1977-01-01

    Preliminary investigations of a heterogeneous gas core reactor (HGCR) concept suggest that this potential power reactor offers distinct advantages over other existing or conceptual reactor power plants. One of the most favorable features of the HGCR is the flexibility of the power producing system which allows it to be efficiently designed to conform to a desired optimum condition without major conceptual changes. The arrangement of bundles of moderator/coolant channels in a fissionable gas or mixture of gases makes a truly heterogeneous nuclear reactor core. It is this full heterogeneity for a gas-fueled reactor core which accounts for the novelty of the heterogeneous gas core reactor concept and leads to noted significant advantages over previous gas core systems with respect to neutron and fuel economy, power density, and heat transfer characteristics. The purpose of this work is to provide an insight into the design, operating characteristics, and safety of a heterogeneous gas core reactor system. The studies consist mainly of neutronic, energetic and kinetic analyses of the power producing and conversion systems as a preliminary assessment of the heterogeneous gas core reactor concept and basic design. The results of the conducted research indicate a high potential for the heterogeneous gas core reactor system as an electrical power generating unit (either large or small), with an overall efficiency as high as 40 to 45%. The HGCR system is found to be stable and safe, under the conditions imposed upon the analyses conducted in this work, due to the inherent safety of ann expanding gaseous fuel and the intrinsic feedback effects of the gas and water coolant

  9. New fast reactor installation concept

    Anon.

    1976-01-01

    The large size and complexity of fast reactor installations are emphasised and these difficulties will be increased with the advent of fast reactors of higher power. In this connection a new concept of fast reactor installation is described with a view to reducing the size of the installation and enabling most components, including even the primary vessel, to be constructed within the confines of a workshop. Full constructional details are given. (U.K.)

  10. Elbereth. A soubroutine library for graphic representation from J.E.N. Fusion Division

    Guasp, J.

    1981-01-01

    A library for graphic representation, named Elbereth, has been built, all the subroutines have been written in Fortran-V for the Univac-1100/80 computer from J.E.N. and are able to produce drawing on the Calcomp-936 plotter, as well as with the printers and on the UTS-400 terminal screen. The library can yield two or three-dimensional plots as well as level lines, all of them under free format or in several standard ones either. It allows to be used on a simplified way by direct reading from a file. It has been constructed too with a flexible structure to allow forthcoming extensions. (author)

  11. Actualization of the Tandem-E N Accelerator of the Nuclear Centre of Mexico

    Villasenor S, P.; Aguilera R, E.; Aspiazu F, J.; Fernandez A, J.; Fernandez B, M.; Garcia R, B.; Lopez M, J.; Martinez Q, E.; Mendez G, B.; Moreno B, E.; Murillo O, G.; Policroniades R, R.; Ramirez T, J.; Reynoso V, R.; Varela G, A.; Vega C, J.

    2004-01-01

    In this work, the activities are described carried out to change the tubes accelerators and original resistances of the accelerator Tandem-E N of the Nuclear Center, for tubes DOWLISH and resistances again design, both donated ones for ORNL. This way same, the problem is described that imply this changes, and like it was solved by the personnel of the laboratory, without having to appeal to external services, what there is redounded in a considerable increment in the costs. In form preliminary the improvements are described observed after the rehabilitation of the Accelerator. (Author)

  12. Desenvolvimento de mudas de catingueira em diferentes substratos e níveis de luminosidade

    Antunes,Cimille Gabrielle Cardoso; Souza,Cíntia Luiza Mascarenhas de; Gomes,Hugo Leonardo Rodrigues; Souza,Jonaicon Vieira de; Barroso,Natália dos Santos; Castro,Renato Delmondez de; Pelacani,Claudinéia Regina

    2014-01-01

    Poincianella pyramidalis Tul. (catingueira) é uma leguminosa endêmica do semiárido nordestino que se enquadra bem no contexto de espécies florestais brasileiras que ainda carecem de estudos quanto às condições ideais para seu desenvolvimento e estabelecimento em campo. Dessa forma, objetivou-se com este estudo avaliar o crescimento inicial de mudas da referida espécie submetidas a diferentes substratos e níveis de luminosidade. Plântulas de catingueira foram cultivadas em três condições de lu...

  13. Slurry reactors

    Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

    1979-08-01

    Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

  14. Reactor safety

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  15. Nuclear reactor

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  16. Reactor container

    Kato, Masami; Nishio, Masahide.

    1987-01-01

    Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

  17. RA Reactor

    1989-01-01

    This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels [sr

  18. Reactor physics

    Ait Abderrahim, H.

    1998-01-01

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  19. The P-A-C-I-E-N-T-E Protocol: An instrument for breaking bad news adapted to the Brazilian medical reality

    Carolina Rebello Pereira

    Full Text Available Summary Objective: There are plenty of published tools for breaking bad medical news; however, none of them is culturally appropriate to our reality or published in the Brazilian literature. This study proposes a genuinely Brazilian communication tool and evaluates its acceptance among doctors and nurses. Method: This was a prospective study. The data were collected after specific training of doctors and nurses on the bad news communication techniques based on the P-A-C-I-E-N-T-E ("patient," in Portuguese Protocol. This instrument is in accordance with the Brazilian reality and was based on the SPIKES communication tool. Results: The worst task to be performed during communication is "talking about death" followed by "discussing the end of curative treatment attempts" and "diagnosis" itself. Among the respondents, 48% reported they did not receive formal training for communicating. Also, 52% of respondents do not use any systematic approach in their daily practice when communicating with patients, but 97% considered the proposed P-A-C-I-E-N-T-E Protocol as a useful and appropriate communication tool. Conclusion: The P-A-C-I-E-N-T-E Protocol proved to be suitable to the Brazilian context.

  20. Reactor core

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  1. Reactor container

    Fukazawa, Masanori.

    1991-01-01

    A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

  2. Reactor container

    Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

    1997-01-01

    In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

  3. Nuclear reactor

    Tilliette, Z.

    1975-01-01

    A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

  4. TU Electric reactor physics model verification: Power reactor benchmark

    Willingham, C.E.; Killgore, M.R.

    1988-01-01

    Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

  5. Charged particle transport in gaseous nitrogen at intermediate E/N using the voltage transient method

    Purdie, P.H.; Fletcher, J.

    1992-01-01

    A pulsed swarm of charged particles crossing an inter-electrode gap under the influence of an applied electric field E will produce a pulsed current in the external circuit which, when integrated over time, will result in a transient voltage pulse, the shape and magnitude of which is characteristic of the number of type of charged particles. This voltage transient technique has been used to investigate a gas discharge in nitrogen gas at values of E/N (the ratio of applied electric field to gas number density), such that ionization is non-negligible. The voltage transients have been subjected to a theoretical analysis, which has previously been reported, which includes not only cathode and anode image terms but also both electron and ion diffusion terms. Electron transport parameters are reported for E/N ≤ 350 Td (1 Td = 10 -17 V cm 2 ). Data are also obtained for the drift velocities and diffusion coefficients of the ions operative within the nitrogen discharge. An estimate is obtained for the collisional decay rate of N 2 + . 21 refs., 7 figs

  6. Analysis of Quasi-Elastic e-n and e-p Scattering from Deuterium

    Balsamo, Alexander; Gilfoyle, Gerard; CLAS12 Collaboration

    2017-09-01

    One of Jefferson Lab's goals is to unravel the quark-gluon structure of nuclei. We will use the ratio, R, of electron-neutron to electron-proton scattering on deuterium to probe the magnetic form factor of the neutron. We have developed an end-to-end analysis from simulation to extraction of R in quasi-elastic kinematics for an approved experiment with the CLAS12 detector. We focus on neutrons detected in the CLAS12 calorimeters and protons measured with the CLAS12 forward detector. Events were generated with the Quasi-Elastic Event Generator (QUEEG) and passed through the Monte Carlo code gemc to simulate the CLAS12 response. These simulated events were reconstructed using the latest CLAS12 Common Tools. We first match the solid angle for e-n and e-p events. The electron information is used to predict the path of both a neutron and proton through CLAS12. If both particles interact in CLAS12 the e-n and e-p events have the same solid angle. We select QE events by searching for nuclei near the predicted position. An angular cut between the predicted 3-momentum of the nucleon and the measured value, θpq, separates QE and inelastic events. We will show the simulated R as a function of the four-momentum transfer Q2. Work supported by the University of Richmond and the US Department of Energy.

  7. Full Scale Alternative Catalyst Testing for Bosch Reactor Optimization

    Barton, Katherine; Abney, Morgan B.

    2011-01-01

    Current air revitalization technology onboard the International Space Station (ISS) cannot provide complete closure of the oxygen and hydrogen loops. This makes re-supply necessary, which is possible for missions in low Earth orbit (LEO) like the ISS, but unviable for long term space missions outside LEO. In comparison, Bosch technology reduces carbon dioxide with hydrogen, traditionally over a steel wool catalyst, to create water and solid carbon. The Bosch product water can then be fed to the oxygen generation assembly to produce oxygen for crew members and hydrogen necessary to reduce more carbon dioxide. Bosch technology can achieve complete oxygen loop closure, but has many undesirable factors that result in a high energy, mass, and volume system. Finding a different catalyst with an equal reaction rate at lower temperatures with less catalyst mass and longer lifespan would make a Bosch flight system more feasible. Developmental testing of alternative catalysts for the Bosch has been performed using the Horizontal Bosch Test Stand. Nickel foam, nickel shavings, and cobalt shavings were tested at 500 C and compared to the original catalyst, steel wool. This paper presents data and analysis on the performance of each catalyst tested at comparable temperatures and recycle flow rates.

  8. Full autonomous monitoring tools inside nuclear reactor building

    Giraud, A.; Veau, J.

    2009-01-01

    In this paper, we define, design and test a radiation tolerant autonomous monitoring tool for nuclear embedded applications. The goal of the instrumentation system was to record the values of some parameters such as dose, temperature or vibrations appearing inside the containment building of nuclear power plants. The knowledge of these parameters will be a good help for predictive maintenance of the power plant components. For the design of the monitoring tool, we rely on commercial-off-the-shelf (COTS) low power electronic components to use battery-supplied power. A large amount of components starting from discrete transistors or logic units to memories and micro-controllers was associated to define and design a prototype. We then confirm the environment conditions tolerance estimated to up to 2 kGy of total dose and 80 C for temperature by on-line irradiation experiments for individual components and functions and prototypes. Two different sets of about 60 systems were realised as industrial products and then installed in EDF facilities. They were exploited during NPPs operating times of about eighteen months. This large scale experiment was concluded with the inspection of some withdrawn systems. No major degradations were observed showing the efficiency of the hardening method allowing regular use of COTS

  9. Reactor building

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

    In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

  10. Reactor container

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

    Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

  11. Chernobyl reactor transient simulation study

    Gaber, F.A.; El Messiry, A.M.

    1988-01-01

    This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

  12. Backfitting of the FRG reactors

    Krull, W.

    1990-01-01

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U 3 Si 2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

  13. Review of Past Reactor Activities in Italy, April 1975

    Pierantoni, P.

    1975-01-01

    Since the last meeting of the International Working Group on Fast Reactors the natural vocation of the italian program to a cooperation, illustrated on many occasions, has found factual confirmation. As well-known, a broad partnership with France has been entered into. Four agreements have been concluded, towards May-June 1974 - between C.E.A. and C.N.E.N. for the ulterior development of the French filière and the construction and use of PEC; - between C.E.A. and NIRA for the license on the system; - between C.E.A. and AGIP-NUCLEARE for the fabrication of fuel for fast reactors; - between TECHNICATOME and NIRA for the preparation of the offer and the later construction of Super-Phenix. Those agreements are in the full logic of the previous one passed between the producers E.d.F., ENEL and R.W.E. for the construction of the first two commercial-size plants, extrapolated from Phenix and S.N.R. 300. In particular the NERSA society, acting as client for Super-Phenix, has been founded by those same producers and will receive the offer now prepared by a joint team. The new orientation of the programme and the validity of the previous agreements for its implementation have been confirmed at governmental level by the Interministerial Economic Planning Committee (CIPE), which has also approved the third CNEN's five years Plan

  14. Nuclear reactor

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  15. Full closure strategic analysis.

    2014-07-01

    The full closure strategic analysis was conducted to create a decision process whereby full roadway : closures for construction and maintenance activities can be evaluated and approved or denied by CDOT : Traffic personnel. The study reviewed current...

  16. Imaging Fukushima Daiichi reactors with muons

    Haruo Miyadera

    2013-05-01

    Full Text Available A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

  17. S.E.N.S.I.B. project. Progress report 2006; Projet SENSIB. Rapport d'avancement 2006

    NONE

    2007-07-01

    This report presents the state of progress of all the studies which establish at present the S.E.N.S.I.B. project. For year 2006, the progress of the project is globally in compliance with the general schedule of realization of the S.E.N.S.I.B. project and with the perspectives announced in 2005. Factors of sensitivity were identified in diverse circles ( thematic studies) and methods and specific tools of the project are developed. 14 publications (reviews and congress) and 9 I.R.S.N. reports were produced. An international work group was launched to the I.R.S.N. initiative. The web site of S.E.N.S.I.B. on the I.R.S.N. scientific site was built. The S.E.N.S.I.B. project receives a financial participation of the Ademe. (N.C.)

  18. Nuclear reactors

    Prescott, R.F.

    1976-01-01

    A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates

  19. Nuclear reactor

    Miyashita, Akio.

    1981-01-01

    Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

  20. Reactor container

    Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.

    1987-01-01

    Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)

  1. REACTOR SHIELD

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  2. NUCLEAR REACTOR

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  3. Breeder reactors

    Gollion, H.

    1977-01-01

    The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components [fr

  4. Nuclear reactor

    Schulze, I.; Gutscher, E.

    1980-01-01

    The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de

  5. Reactor technology

    Erdoes, P.

    1977-01-01

    This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

  6. Inclusive (e,e'N), (e,e'NN), (e,e'π),... reactions in nuclei

    Gil, A.; Oset, E.

    1997-01-01

    We study the inclusive (e,e'N), (e,e'NN), (e,e'π), (e,e'πN) reactions in nuclei using a Monte Carlo simulation method to treat the multichannel problem of the final state. The input consists of reaction probabilities for the different steps evaluated using microscopical many body methods. We obtain a good agreement with experiment in some channels where there is data and make predictions for other channels which are presently under investigation in several electron laboratories. The comparison of the theoretical results with experiment for several kinematical conditions and diverse channels can serve to learn about different physical processes occurring in the reaction. The potential of this theoretical tool to make prospections for possible experiments, aiming at pinning down certain reaction probabilities, is also emphasized. (orig.)

  7. Reservoir characterization and performance predictions for the E.N. Woods lease

    Aka-Milan, Francis A.

    2000-07-07

    The task of this work was to evaluate the past performance of the E.N. WOODS Unit and to forecast its future economic performance by taking into consideration the geology, petrophysics and production history of the reservoir. The Decline Curve Analysis feature of the Appraisal of Petroleum Properties including Taxation Systems (EDAPT) software along with the Production Management Systems (PMS) software were used to evaluate the original volume of hydrocarbon in place and estimate the reserve. The Black Oil Simulator (BOAST II) was then used to model the waterflooding operation and estimate the incremental oil production attributable to the water injection. BOAST II was also used to predict future performance of the reservoir.

  8. Experience at J.E.N. with electrochemical cells for measurement of oxygen activity

    La Torre, M.de; Lapena, J.; Couchoud, M.

    1981-01-01

    The experience gained at the J.E.N. with oxygen meters since 1974 till 1980 is presented. Thirteen oxygen meters were tested. Eight with Cu/Cu/ 2 O reference electrode and the rest with Sn/SnO 2 , and two types of electrolyte tube produced by zircoa under specifications development by UNC and HEDL. The cells equiped with Cu/Cu 2 O showed an anomalous performance giving an e.m.f. higher than the theoretical value, and one of them was in close agreement to cells using air as reference electrode. An explanation is given. The performance of the cells with Sn/SnO 2 is in good agreement with those obtained in others laboratories. To calculate the theoretical value, it has derived a correlation colubility for oxygen with 262 data obtained by the vacuum distillation method. Various recommendations are pointed out on the future development of the oxygen meters to improve its performance. (author)

  9. Reactor containment

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

    A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

  10. Fast reactors

    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.)

  11. Perda da voz em professores e não professores Voice loss in teachers and non-teachers

    Kelly Park

    2009-01-01

    Full Text Available OBJETIVO: Verificar a percepção de professores e não-professores sobre as implicações de uma eventual perda de voz. MÉTODOS: Participaram 205 indivíduos sendo 105 professores e 100 não professores entre 23 a 65 anos, 106 mulheres e 99 homens. Foi aplicado um questionário contendo quatro perguntas referentes a uma eventual perda de visão, audição, voz e deambulação e o grau de impacto inferido (de 0 a 4. RESULTADOS: Para o grupo de professores, não enxergar gerou o maior impacto negativo (média de 3,8, seguido por não andar (média de 3,7, não ter voz (média de 3,7 e não ouvir (média de 3,6. Para o grupo de não-professores, não enxergar também gerou o maior impacto negativo (média de 3,4, seguido por não andar (média de 3,0; não ouvir (média de 2,2 e não ter voz (média de 2,0. Em relação ao maior impacto de uma eventual perda da voz, professores indicaram prejuízos no trabalho, relacionamento social e atividades rotineiras e, no grupo de não professores, nas atividades rotineiras, trabalho, relacionamento social e manifestações das emoções. CONCLUSÕES: Os professores valorizam sua voz de modo diverso dos não-professores e ambos os grupos avaliam a perda da voz como algo que não acarreta consequências negativas. Apesar de o professor perceber mais o impacto de um eventual problema de voz do que o não-professor, os sentimentos em relação à perda da voz foram muito semelhantes nos dois grupos.PURPOSE: To investigate teachers' and non-teachers' perception regarding the implications of an eventual loss of voice. METHODS: Participated in the study 205 individuals (106 women and 99 men, 105 teachers and 100 non-teachers, with ages varying from 23 to 65 years old. The participants were asked to answer a questionnaire with four questions regarding an eventual loss of vision, hearing, voice and deambulation, and the inferred impact degree (from 0 to 4. RESULTS: For the teachers group, not being able to see

  12. Full page insight

    Cortsen, Rikke Platz

    2014-01-01

    Alan Moore and his collaborating artists often manipulate time and space by drawing upon the formal elements of comics and making alternative constellations. This article looks at an element that is used frequently in comics of all kinds – the full page – and discusses how it helps shape spatio......, something that it shares with the full page in comics. Through an analysis of several full pages from Moore titles like Swamp Thing, From Hell, Watchmen and Promethea, it is made clear why the full page provides an apt vehicle for an apocalypse in comics....

  13. Generation IV reactors: reactor concepts

    Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

    2003-01-01

    Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

  14. Research reactors - an overview

    West, C.D.

    1997-01-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

  15. Cardiac response and anxiety levels in psychopathic murderers Resposta cardíaca e nível de ansiedade em homicidas psicopatas

    Antonio de Pádua Serafim

    2009-09-01

    Full Text Available OBJECTIVE: To compare the emotional response and level of anxiety of psychopathic murderers, non-psychopathic murderers, and nonpsychopathic non-criminals. METHOD: 110 male individuals aged over 18 years were divided into three groups: psychopathic murderers (n = 38; non-psychopathic murderers (n = 37 serving sentences for murder convictions in Maximum Security Prisons in the State of Sao Paulo; and non-criminal, non-psychopathic individuals (n = 35 according to the Psychopathy Checklist-Revised. The emotional response of subjects was assessed by heart rate variation and anxiety level (State-Trait Anxiety Inventory after viewing standardized pictures depicting pleasant, unpleasant and neutral content from the International Affective Picture System. RESULTS: Psychopathic murderers presented lower anxiety levels and smaller heart rate variations when exposed to pleasant and unpleasant stimuli than nonpsychopathic murderers or non-psychopathic non-criminals. The results also demonstrated that the higher the score for factor 1 on the Psychopathy Checklist-Revised, the lower the heart rate variation and anxiety level. CONCLUSION: The results suggest that psychopathic murderers do not present variation in emotional response to different visual stimuli. Although the non-psychopathic murderers had committed the same type of crime as the psychopathic murderers, the former tended to respond with a higher level of anxiety and heart rate variation.OBJETIVO: Comparar a atividade cardíaca e nível de ansiedade de homicidas psicopatas e não psicopatas e não criminosos não psicopatas. MÉTODO: 110 homens com idade superior a 18 anos, divididos em três grupos: homicidas psicopatas (n = 38, homicidas não psicopatas (n = 37 cumprindo pena por homicídio em Prisões de Segurança Máxima do Estado de São Paulo e não criminosos e não psicopatas (n = 35 de acordo com a Escala de Avaliação de Psicopatia. A resposta emocional foi avaliada pela variação da

  16. Backfitting swimming pool reactors

    Roebert, G.A.

    1978-01-01

    Calculations based on measurements in a critical assembly, and experiments to disclose fuel element surface temperatures in case of accidents like stopping of primary coolant flow during full power operation, have shown that the power of the swimming pool type research reactor FRG-2 (15 MW, operating since 1967) might be raised to 21 MW within the present rules of science and technology, without major alterations of the pool buildings and the cooling systems. A backfitting program is carried through to adjust the reactor control systems of FRG-2 and FRG-1 (5 MW, housed in the same reactor hall) to the present safety rules and recommendations, to ensure FRG-2 operation at 21 MW for the next decade. (author)

  17. Progress report of the S.E.N.S.I.B. project 2005-2008

    Baumont, G.; Bourcier, L.; Briand, B.; Chojnacki, E.; Dufeu, A.; Duffa, C.; Durand, V.; Eyrolle, F.; Larue, C.; Levain, A.; Masson, O.; Mercat, C.; Metivier, J.M.; Pey, B.; Pourcelot, L.; Renaud, Ph.; Roussel-Debet, S.; Thebault, H.

    2008-01-01

    The aim of the project is to define a method for classifying territories by their sensitivity to a radioactive pollution, on the basis of their environmental and social characteristics, i.e. those characteristics associated with the human use of the environment. For example, the sensitivity of an urban environment will depend on the population density, the area covered, the urban services and their response to the contamination. The sensitivity of an agricultural area will be determined by the food chain and the way in which it is affected by contamination of the soil, crops or livestock. The end result of the S.E.N.S.I.B. project will be a standardised tool capable of representing and comparing the sensitivity of an area to radioactive pollution using a single measurement scale. This standardized measurement of the environmental and population characteristics will be useful in risk assessment and management, and in all the operational stages of a nuclear installation including commissioning, normal operation, accident and post-accident situations, and during decommissioning. The project will provide mappings of the sensitivity to a range of concrete scenarios, and of the vulnerability of each area once the probability of pollution occurring has been taken into account. The concept of sensitivity is shared by other fields besides radioecology and can be used to assess vulnerability to chemical pollution, climate change, or population pressures. The results obtained from the S.E.N.S.I.B. project will be applicable to other fields, especially that of chemical pollution, though a range of collaborations and partnerships. The S.E.N.S.I.B. project is divided into five stages. These are: Identification of sensitivity factors, i.e. all the characteristics of a territory likely to make the environment sensitive to a given type of pollution. Characterization of their ranges of values. Conversion of each range of values into a range of scores. Allocation of a weighting to be

  18. Power reactor noise

    Thie, J.A.

    1981-01-01

    This book concentrates on the different types of noise present in power reactors and how the analysis of this noise can be used as a tool for reactor monitoring and diagnostics. Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions thus preventing further complications by alerting operators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussion of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

  19. Full Service Leasing

    Richter, Ján

    2009-01-01

    Aim of this master thesis is to describe the service of Full Service Leasing, as a modern form of financing and management of assets, primarily automobile fleet. Description of full service leasing is designed as a comprehensive and complete guide to support reader's position when deciding to finance and manage a fleet by this service. Whether the reader is an entrepreneur, CFO, fleet manager, new employee of leasing company, or anyone who is interested in this service, this master thesis wil...

  20. Bioconversion reactor

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  1. Nuclear reactor

    Scholz, M.

    1976-01-01

    An improvement of the accessibility of that part of a nuclear reactor serving for biological shield is proposed. It is intended to provide within the biological shield, distributed around the circumference of the reactor pressure vessel, several shielding chambers filled with shielding material, which are isolated gastight from the outside by means of glass panes with a given bursting strength. It is advantageous that, on the one hand, inspection and maintenance will be possible without great effort and, on the other, a large relief cross section will be at desposal if required. (UWI) [de

  2. NEUTRONIC REACTOR

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  3. BR2 Reactor: Introduction

    Moons, F.

    2007-01-01

    The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3 He, 6 Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

  4. Compressive full waveform lidar

    Yang, Weiyi; Ke, Jun

    2017-05-01

    To avoid high bandwidth detector, fast speed A/D converter, and large size memory disk, a compressive full waveform LIDAR system, which uses a temporally modulated laser instead of a pulsed laser, is studied in this paper. Full waveform data from NEON (National Ecological Observatory Network) are used. Random binary patterns are used to modulate the source. To achieve 0.15 m ranging resolution, a 100 MSPS A/D converter is assumed to make measurements. SPIRAL algorithm with canonical basis is employed when Poisson noise is considered in the low illuminated condition.

  5. Full faith in myself

    Lawrence

    Full faith in myself. Meenakshi Banerjee. 12. Ihad my schooling at the Irish Convent, Loreto, in Asansol,. West Bengal. Perhaps the earliest memories I have are of myself as a very determined child with a deep appreciation of and inquisitiveness regarding nature although not understanding most of it at that tender age.

  6. Fast breeder reactor research

    1975-01-01

    Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power

  7. Neutronic reactor

    Wende, C.W.J.

    1976-01-01

    The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures

  8. Neutronic reactor

    Wende, C.W.J.

    1976-01-01

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield

  9. Reactor facility

    Suzuki, Hiroaki; Murase, Michio; Yokomizo, Osamu.

    1997-01-01

    The present invention provides a BWR type reactor facility capable of suppressing the amount of steams generated by the mutual effect of a failed reactor core and coolants upon occurrence of an imaginal accident, and not requiring spacial countermeasures for enhancing the pressure resistance of the container vessel. Namely, a means for supplying cooling water at a temperature not lower by 30degC than the saturated temperature corresponding to the inner pressure of the containing vessel upon occurrence of an accident is disposed to a lower dry well below the pressure vessel. As a result, upon occurrence of such an accident that the reactor core should be melted and flown downward of the pressure vessel, when cooling water at a temperature not lower than the saturated temperature, for example, cooling water at 100degC or higher is supplied to the lower dry well, abrupt generation of steams by the mutual effect of the failed reactor core and cooling water is scarcely caused compared with a case of supplying cooling water at a temperature lower than the saturation temperature by 30degC or more. Accordingly, the amount of steams to be generated can be suppressed, and special countermeasure is no more necessary for enhancing the pressure resistance of the container vessel is no more necessary. (I.S.)

  10. Nuclear reactor

    Gilroy, J.E.

    1980-01-01

    An improved cover structure for liquid metal cooled fast breeder type reactors is described which it is claimed reduces the temperature differential across the intermediate grid plate of the core cover structure and thereby reduces its subjection to thermal stresses. (UK)

  11. Reactor licensing

    Harvie, J.D.

    2002-01-01

    This presentation discusses reactor licensing and includes the legislative basis for licensing, other relevant legislation , the purpose of the Nuclear Safety and Control Act, important regulations, regulatory document, policies, and standards. It also discusses the role of the CNSC, its mandate and safety philosophy

  12. Nuclear reactor

    Hattori, Sadao; Sekine, Katsuhisa.

    1987-01-01

    Purpose: To decrease the thickness of a reactor container and reduce the height and the height and plate thickness of a roof slab without using mechanical vibration stoppers. Constitution: Earthquake proofness is improved by filling fluids such as liquid metal between a reactor container and a secondary container and connecting the outer surface of the reactor container with the inner surface of the secondary container by means of bellows. That is, for the horizontal seismic vibrations, horizontal loads can be supported by the secondary container without providing mechanical vibration stoppers to the reactor container and the wall thickness can be reduced thereby enabling to simplify thermal insulation structure for the reduction of thermal stresses. Further, for the vertical seismic vibrations, verical loads can be transmitted to the secondary container thereby enabling to reduce the wall thickness in the same manner as for the horizontal load. By the effect of transferring the point of action of the container load applied to the roof slab to the outer circumferential portion, the intended purpose can be attained and, in addition, the radiation dose rate at the upper surface of the roof slab can be decreased. (Kamimura, M.)

  13. Reactor system

    Miyano, Hiroshi; Narabayashi, Naoshi.

    1990-01-01

    The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

  14. Reactor core

    Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

    1990-01-01

    Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

  15. Oscillatory flow chemical reactors

    Slavnić Danijela S.

    2014-01-01

    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  16. New about research reactors

    Egorenkov, P.M.

    2001-01-01

    The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru

  17. Morfologia do útero de cutias nulíparas e não nulíparas

    L.L. Martins

    2011-04-01

    Full Text Available Descreveu-se a morfologia do útero de cutias nulíparas e não nulíparas por meio de ovariossalpingoisterectomia. A análise macroscópica do útero foi realizada in loco e na peça retirada. Segmentos das porções proximal, média e distal do órgão foram colhidas, fixadas e preparadas histologicamente, sendo as amostras analisadas à microscopia de luz, e realizada a histomorfometria das camadas uterinas. Topograficamente, o útero localiza-se na região sublombar, caudalmente aos rins, em continuação aos ovários e às tubas uterinas, estendendo-se até à entrada da pelve, onde se posiciona dorsalmente à bexiga. Caracteriza-se como do tipo duplo, embora culmine em apenas um óstio cervical externo. Microscopicamente, a mucosa uterina é formada por variações do epitélio, de cilíndrico a pseudoestratificado, que se apoia em tecido conjuntivo frouxo, onde se observam glândulas endometriais revestidas por epitélio cilíndrico, além da presença de vasos sanguíneos. A camada muscular subdivide-se em interna ou submucosa, média ou vascular e externa ou subserosa. A camada serosa é composta por tecido conjuntivo e mesotélio. Na histomorfometria, verificou-se que a espessura uterina total e a espessura da camada mucosa, em média, foram maiores nas fêmeas não nulíparas.

  18. The timing of reactor dismantling

    Roberts, P.

    2000-01-01

    Work has been progressing across the world for the decommissioning of nuclear reactors. The initial work focused on the early, complete dismantling but this was associated with small size reactors and was done for experimental or demonstration purposes. The situation now is that an increasing number of full size power reactors are being shutdown and decision are being made as to the decommissioning strategy to be applied, e.g. with respect to the appropriate timing of reactor dismantling. There are two basic approaches to the timing of reactor dismantling, which are to either proceed with dismantling on an early time scale or to delay it for a period of years. There are a number of examples worldwide of both approaches being taken but one common feature of the approach taken by most countries is that decisions are made on a case by case basis, taking account of relevant factors, and as a result the strategy can vary from reactor to reactor and from country to country. Decisions on timing take account of the following main factors: safety, radioactive decay, financial factors, radioactive waste, reactor type, technology, repository availability, site re-use, regulatory standards, plant knowledge/records, other issues

  19. Plate Full of Color

    The Eagle Books are a series of four books that are brought to life by wise animal characters - Mr. Eagle, Miss Rabbit, and Coyote - who engage Rain That Dances and his young friends in the joy of physical activity, eating healthy foods, and learning from their elders about health and diabetes prevention. Plate Full of Color teaches the value of eating a variety of colorful and healthy foods.

  20. Estudo comparativo das habilidades sociais de dependentes e não dependentes de álcool Comparative study of social skills among alcohol abusers and non abusers

    Poliana Patrício Aliane

    2006-04-01

    Full Text Available Este trabalho teve como objetivo principal avaliar e comparar diferenças nas habilidades sociais (HS de dependentes e não dependentes de álcool. Foram pesquisados 80 sujeitos, entre dependentes e não-dependentes, usuários do SUS, e usados dois instrumentos para coleta dos dados: Inventário de Habilidades Sociais (IHS e AUDIT (Alcohol Use Disorder Identification Test. Foi realizado um estudo comparativo das HS entre os grupos de dependentes e não-dependentes de álcool e álcool e outras drogas (AOD. Os resultados obtidos mostraram não existir diferença no escore do IHS entre dependentes e não-dependentes. Os homens obtiveram maior média nas habilidades de conversação e desenvoltura social e autocontrole da agressividade que as mulheres e, na amostra masculina, dependentes de álcool apresentaram maior média na habilidade de autocontrole da agressividade que dependentes de AOD. Apesar de a literatura sobre o tema discutir os déficits de HS entre dependentes, este estudo não confirmou esta hipótese.This research aims to evaluate and compare differences in Social Skills of alcohol abusers and non-abusers. Eighty people were interviewed, among alcohol abusers and non-abusers, patients of a Health Care Center (SUS. Two instruments were used to collect the data: Social Skills Inventory (ISS and AUDIT (Alcohol Use Disorder Identification Test. A Comparative study between the alcohol abusers and non-abusers including other drugs was carried out. The results showed that there is no difference between the alcohol abusers and non-abusers ISS scores. Men showed a better score in conversation and social abilities also in self-control of the aggressiveness than women did. In addiction, men addicted to alcohol showed a better score in self-control of the aggressiveness than men addicted to alcohol and other drugs. Although literature about this topic shows that the alcohol abuse have an decrease of these social skills, this research does not

  1. Agreement on economic and technological cooperation between the Federal Republic of Germany and the GDR. Project part 3.2, ''NDT and QA''. Project task 2.11. Experiments with the full-size vessel in Stuttgart for selection of practice-relevant non-destructive testing methods for evaluation of the value and performance of recurrent inspections of reactor components. Final report

    Betzold, K.; Brinette, R.; Bonitz, F.

    1992-01-01

    The efficiency of NDT methods such as ALOK, SAFT, EMUS, LLT, phased array, and multi-frequency eddy current testing which are generally used for reactor components recurrent inspection has been verified with experiments using two test specimens. These are a section of a main coolant pipe and the full-size vessel installed at MPA-Stuttgart, furnished with PWR test bodies with artificial defects and artificially applied natural defects. The defects have been detected with commercial probes as well as with probes optimized for the NDT methods EMUS, LLT, phased array, and multi-frequency eddy current testing. Type, location, orientation and geometry of the defects have been measured, also recording the influence of type of defect on the efficiency of the NDT methods, in order to reveal problems linked with the various methods as well as their advantages. Further tests have been made for evaluation of a combination of ALOK and SAFT using novel, specifically developed test probes, and a combination of ALOK and phased array testing. (orig.) [de

  2. TU electric reactor model verification

    Willingham, C.E.; Killgore, M.R.

    1989-01-01

    Power reactor benchmark calculations using the code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles include gadolinia as a burnable absorber, natural uranium axial blankets, and increased water-to-fuel ratio. The calculated results for both low-power physics tests (boron end points, control rod worths, and isothermal temperature coefficients) and full-power operation (power distributions and boron letdown) are compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important physics parameters for power reactors

  3. Reactor core of nuclear reactor

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  4. Plate Full of Color

    2008-08-04

    The Eagle Books are a series of four books that are brought to life by wise animal characters - Mr. Eagle, Miss Rabbit, and Coyote - who engage Rain That Dances and his young friends in the joy of physical activity, eating healthy foods, and learning from their elders about health and diabetes prevention. Plate Full of Color teaches the value of eating a variety of colorful and healthy foods.  Created: 8/4/2008 by National Center for Chronic Disease Prevention and Health Promotion (NCCDPHP).   Date Released: 8/5/2008.

  5. Efeitos de contingências de punição sobre os desempenhos verbal e não verbal

    Elisa Tavares Sanabio

    Full Text Available O presente estudo investigou a influência de contingências de punição sobre os desempenhos verbal (resposta de relato e não verbal (resposta de escolha. Estudantes universitários foram expostos a uma tarefa de escolha de acordo com o modelo. Após a resposta de escolha, a pergunta "Você acertou?" era apresentada e o participante deveria emitir a resposta de relato, indicando uma dentre duas alternativas, "SIM" e "NÃO". No Experimento 1, o feedback "Incorreto. Você perdeu 1 ponto" foi programado para a resposta de relato e, no Experimento 2, o feedback foi contingente à resposta de escolha. No Experimento 1, o feedback exerceu funções punitivas sobre a resposta de relato e, na maioria das condições, não afetou a resposta de escolha. No Experimento 2, o feedback não exerceu controle sobre a resposta de escolha, embora tenha influenciado as respostas de relato. Esses resultados sugerem que os desempenhos verbal e não verbal são funcionalmente independentes. O presente estudo evidencia o caráter operante do relato e, ao identificar possíveis variáveis de controle desse comportamento, contribui para uma maior fidedignidade dos dados obtidos por meio de auto-relatos.

  6. Reactor instrumentation and control

    Wach, D.; Beraha, D.

    1980-01-01

    The methods for measuring radiation are shortly reviewed. The instrumentation for neutron flux measurement is classified into out-of-core and in-core instrumentation. The out-of-core instrumentation monitors the operational range from the subcritical reactor to full power. This large range is covered by several measurement channels which derive their signals from counter tubes and ionization chambers. The in-core instrumentation provides more detailed information on the power distribution in the core. The self-powered neutron detectors and the aeroball system in PWR reactors are discussed. Temperature and pressure measurement devices are briefly discussed. The different methods for leak detection are described. In concluding the plant instrumentation part some new monitoring systems and analysis methods are presented: early failure detection methods by noise analysis, acoustic monitoring and vibration monitoring. The presentation of the control starts from an qualitative assessment of the reactor dynamics. The chosen control strategy leads to the definition of the part-load diagram, which provides the set-points for the different control systems. The tasks and the functions of these control systems are described. In additiion to the control, a number of limiting systems is employed to keep the reactor in a safe operating region. Finally, an outlook is given on future developments in control, concerning mainly the increased application of process computers. (orig./RW)

  7. Nuclear reactor

    Gibbons, J.F.; McLaughlin, D.J.

    1978-01-01

    In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de

  8. Nuclear reactor

    Sasaki, Tomozo.

    1987-01-01

    Purpose: To improve the nuclear reactor availability by enabling to continuously exchange fuels in the natural-slightly enriched uranium region during operation. Constitution: A control rod is withdrawn to the midway of a highly enriched uranium region by means of control rod drives and the highly enriched uranium region is burnt to maintain the nuclear reactor always at a critical state. At the same time, fresh uranium-slightly enriched uranium is continuously supplied gravitationally from a fresh fuel reservoir through fuel reservoir to each of fuel pipes in the natural-slightly enriched uranium region. Then, spent fuels reduced with the reactivity by the burn up are successively taken out from the bottom of each of the fuel pipes through an exit duct and a solenoid valve to the inside of a spent fuel reservoir and the burn up in the natural-slightly enriched uranium region is conducted continuously. (Kawakami, Y.)

  9. Nuclear reactor

    Sakurai, Mikio; Yamauchi, Koki.

    1983-01-01

    Purpose: To improve the channel stability and the reactor core stability in a spontaneous circulation state of coolants. Constitution: A reactor core stabilizing device comprising a differential pressure automatic ON-OFF valve is disposed between each of a plurality of jet pumps arranged on a pump deck. The stabilizing device comprises a piston exerted with a pressure on the lower side of the pump deck by way of a pipeway and a valve for flowing coolants through the bypass opening disposed to the pump deck by the opening and closure of the valve ON-OFF. In a case where the jet pumps are stopped, since the differential pressure between the upper and the lower sides of the pump deck is removed, the valve lowers gravitationally into an opened state, whereby the coolants flow through the bypass opening to increase the spontaneous circulation amount thereby improve the stability. (Yoshino, Y.)

  10. Nuclear reactor

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  11. Reactor container

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  12. Reactor container

    Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

    1979-01-01

    Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

  13. Emptiness and Fullness

    Bregnbæk, Susanne; Bunkenborg, Mikkel

    As critical voices question the quality, authenticity, and value of people, goods, and words in post-Mao China, accusations of emptiness render things open to new investments of meaning, substance, and value. Exploring the production of lack and desire through fine-grained ethnography, this volume...... examines how diagnoses of emptiness operate in a range of very different domains in contemporary China: In the ostensibly meritocratic exam system and the rhetoric of officials, in underground churches, housing bubbles, and nationalist fantasies, in bodies possessed by spirits and evaluations of jade......, there is a pervasive concern with states of lack and emptiness and the contributions suggest that this play of emptiness and fullness is crucial to ongoing constructions of quality, value, and subjectivity in China....

  14. Backfitting of the FRG reactors

    Krull, W [GKSS-Forschungszentrum Geesthacht GmbH, Geesthacht (Germany)

    1990-05-01

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U{sub 3}Si{sub 2} fuel. Both cooling towers were repaired. Replacement of instrumentation is planned.

  15. Neutronic reactor

    Lewis, W.R.

    1978-01-01

    Disclosed is a graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels

  16. Nuclear reactors

    Humphreys, P.; Davidson, D.F.; Thatcher, G.

    1980-01-01

    The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)

  17. Nuclear reactor

    Jungmann, A.

    1975-01-01

    Between a PWR's reactor pressure vessel made of steel and the biological shield made of concrete there is a gap. This gap is filled up with a heat insulation facting the reactor pressure vessel, for example with insulating concrete segments jacketed with sheet steel and with an additional layer. This layer serves for smooth absorption of compressive forces originating in radial direction from the reactor pressure vessel. It consists of cylinder-segment shaped bricks made of on situ concrete, for instance. The bricks have cooling agent ports in one or several rows which run parallel to the wall of the pressure vessel and in alignment with superposed bricks. Between the layer of bricks and the biological shield or rather the heat insulation, there are joints which are filled, however, with injected mortar. That guarantees a smooth series of connected components resistant tom compression. Besides, a slip foil can be set between the heat insulation and the joining joint filled with mortar for the reduction of the friction at thermal expansions. (TK) [de

  18. Reactor building

    Ebata, Sakae.

    1990-01-01

    At least one valve rack is disposed in a reactor building, on which pipeways to a main closure valve, valves and bypasses of turbines are placed and contained. The valve rack is fixed to the main body of the building or to a base mat. Since the reactor building is designed as class A earthquake-proofness and for maintaining the S 1 function, the valve rack can be fixed to the building main body or to the base mat. With such a constitution, the portions for maintaining the S 1 function are concentrated to the reactor building. As a result, the dispersion of structures of earthquake-proof portion corresponding to the reference earthquake vibration S 1 can be prevented. Accordingly, the conditions for the earthquake-proof design of the turbine building and the turbine/electric generator supporting rack are defined as only the class B earthquake-proof design conditions. In view of the above, the amount of building materials can be saved and the time for construction can be shortened. (I.S.)

  19. Nuclear reactors

    Yoshioka, Michiko.

    1985-01-01

    Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

  20. MLR reactor

    Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.

    1998-01-01

    The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)

  1. Nuclear reactor

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)

  2. Reactor container

    Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.

    1990-01-01

    In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)

  3. Reactor core in FBR type reactor

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  4. Full MOX core for PWRs

    Puill, A.; Aniel-Buchheit, S.

    1997-01-01

    Plutonium management is a major problem of the back end of the fuel cycle. Fabrication costs must be reduced and plant operation simplified. The design of a full MOX PWR core would enable the number of reactors devoted to plutonium recycling to be reduced and fuel zoning to be eliminated. This paper is a contribution to the feasibility studies for achieving such a core without fundamental modification of the current design. In view of the differences observed between uranium and plutonium characteristics it seems necessary to reconsider the safety of a MOX-fuelled PWR. Reduction of the control worth and modification of the moderator density coefficient are the main consequences of using MOX fuel in a PWR. The core reactivity change during a draining or a cooling is thus of prime interest. The study of core global draining leads to the following conclusion: only plutonium fuels of very poor quality (i.e. with low fissile content) cannot be used in a 900 MWe PWR because of a positive global voiding reactivity effect. During a cooling accident, like an spurious opening of a secondary-side valve, the hypothetical return to criticality of a 100% MOX core controlled by means of 57 control rod clusters (made of hafnium-clad B 4 C rods with a 90% 10 B content) depends on the isotopic plutonium composition. But safety criteria can be complied with for all isotopic compositions provided the 10 B content of the soluble boron is increased to a value of 40%. Core global draining and cooling accidents do not present any major obstacle to the feasibility of a 100% MOX PWR, only minor hardware modifications will be required. (author)

  5. Full metal jacket!

    Laëtitia Pedroso

    2011-01-01

    Ten years ago, standard issue clothing only gave CERN firemen partial protection but today our fire-fighters are equipped with state-of-the-art, full personal protective equipment.   CERN's Fire Brigade team. For many years, the members of CERN's Fire Brigade went on call-outs clad in their work trousers and fire-rescue coats, which only afforded them partial protection. Today, textile manufacturing techniques have moved on a long way and CERN's firemen are now kitted out with state-of-the-art personal protective equipment. The coat and trousers are three-layered, comprising fire-resistant aramide, a protective membrane and a thermal lining. The CERN Fire Brigade' new state-of-the-art personal protection equipment. "This equipment is fully compliant with the standards in force and is therefore resistant to cuts, abrasion, electrical arcs with thermal effects and, of course, fire," explains Patrick Berlinghi, the CERN Fire Brigade's Logistics Officer. You might think that su...

  6. Policies for full employment

    de Koning, Jaap; Layard, Richard; Nickel, Stephen

    European unemployment is too high, and employment is too low. Over 7½ per cent of Europe's workforce is unemployed, and only two thirds of people aged 15-64 are in work. At the Lisbon summit two years ago the heads of government set the target that by 2010 the employment rate should rise from 64...... per cent to at least 70 per cent. And for older workers between 55 and 64 the employment rate should rise from 38 per cent to at least one half. These are ambitious targets. They will require two big changes: more people must seek work, and among those seeking work a higher proportion must get a job....... So we need higher participation, and (for full employment) we need a much lower unemployment rate. Can it be done? A mere glance at the experience of different European countries shows that it can. As Table 1 shows, four E.U. countries already exceed the overall target for 2010 (Britain, Denmark...

  7. Simulation of a pool type research reactor

    Oliveira, Andre Felipe da Silva de; Moreira, Maria de Lourdes

    2011-01-01

    Computational fluid dynamic is used to simulate natural circulation condition after a research reactor shutdown. A benchmark problem was used to test the viability of usage such code to simulate the reactor model. A model which contains the core, the pool, the reflector tank, the circulation pipes and chimney was simulated. The reactor core contained in the full scale model was represented by a porous media. The parameters of porous media were obtained from a separate CFD analysis of the full core model. Results demonstrate that such studies can be carried out for research and test of reactors design. (author)

  8. Molten salt reactors: reactor cores

    1983-01-01

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

  9. Overview of JT-60U results toward high integrated performance in reactor-relevant regime

    Fujita, T.

    2002-01-01

    Toward steady sustainment of high integrated performance, we have developed weak magnetic shear (high β p mode) and reversed magnetic shear plasmas. As a large-sized tokamak equipped with a variety of devices for heating, current drive and profile/shape control, JT-60U has high ability to approach the conditions required in reactors: low values of normalized Larmor radius and collisionality, high temperatures with T e > or approx. T i , etc. This paper reviews recent JT-60U results with the emphasis on the projection to the reactor-relevant regime. Full non-inductive current drive has been achieved in a 1.8 MA high β p H-mode plasma with β N 2:4, HH y2 =1.2 and high fusion triple product (3 x 10 20 m -3 keVs) owing to increased N-NB power. In a reversed shear plasma, HH y2 =1.4 at n e /n GW 0.8 under the full non-inductive current drive has been achieved with injection of LHRF and N-NB. In box-type ITBs with reversed shear, barriers for ions and electrons were sustained in a regime with T e > or approx. T i . The pedestal pressure was doubled with increased total poloidal beta in pellet-injected high triangularity plasmas with type I and II ELMs. Stable existence of current hole was demonstrated. (author)

  10. Increased SRP reactor power

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

  11. Nuclear research reactors

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  12. Nuclear reactor physics course for reactor operators

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  13. Nuclear reactor

    Jolly, R.

    1979-01-01

    The support grid for the fuel rods of a liquid metal cooled fast breeder reactor has a regular hexagonal contour and contains a large number of unit cells arranged honeycomb fashion. The totality of these cells make up a hexagonal shape. The grid contains a number of strips of material, and there is a window in each of three sidewalls staggered by one sidewall. The other sidewalls have embossed protrusions, thus generating a guide lining or guide bead. The windows reduce the rigidity of the areas in the middle between the ends of the cells. (DG) [de

  14. Nuclear reactor

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  15. Modelos lineares e não lineares da curva de Phillips para previsão da taxa de inflação no Brasil

    Elano Ferreira Arruda

    2011-09-01

    Full Text Available Este trabalho compara previsões da taxa de inflação mensal brasileira a partir de diferentes modelos lineares e não lineares de séries temporais e da curva de Phillips. Em geral, os modelos não lineares apresentaram um melhor desempenho preditivo. Um modelo VAR produziu o menor erro quadrático médio de previsão (EQM entre os modelos lineares, enquanto as melhores previsões, entre todos os modelos, foram geradas pela curva de Phillips ampliada com threshold, a qual apresentou um EQM 20% menor do que a do modelo VAR. Essa diferença é significante de acordo com o teste de Diebold e Mariano (1995.

  16. Advanced boiling water reactor

    Nishimura, N.; Nakai, H.; Ross, M.A.

    1999-01-01

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  17. Unique features of space reactors

    Buden, D.

    1990-01-01

    This paper reports on space reactors that are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K

  18. Nuclear reactors

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1979-01-01

    In a nuclear reactor (e.g. one having coolant down-flow through a core to a hearth below) thermal insulation (e.g. of a floor of the hearth) comprises a layer of bricks and a layer of tiles thereon, with smaller clearances between the tiles than between the bricks but with the bricks being of reduced cross-section immediately adjacent the tiles so as to be surrounded by interconnected passages, of relatively large dimensions, constituting a continuous chamber extending behind the layer of tiles. By this arrangement, lateral coolant flow in the inter-brick clearances is much reduced. The reactor core is preferably formed of hexagonal columns, supported on diamond-shaped plates each supported on a pillar resting on one of the hearth-floor tiles. Each plate has an internal duct, four upper channels connecting the duct with coolant ducts in four core columns supported by the plate, and lower channels connecting the duct to a downwardly-open recess common to three plates, grouped to form a hexagon, at their mutually-adjacent corners. This provides mixing, and temperature-averaging, of coolant from twelve columns

  19. Reactor container

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-01-01

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

  20. Reactor monitor

    Takada, Tamotsu.

    1992-01-01

    The device of the present invention monitors a reactor so that each of the operations for the relocation of fuel assemblies and the withdrawal and the insertion of control rods upon exchange of fuel assemblies and control rods in the reactor. That is, when an operator conducts relocating operation by way of a fuel assembly operation section, the device of the present invention judges whether the operation indication is adequate or not, based on the information of control rod arrangement in a control rod memory section. When the operation indication is wrong, a stop signal is sent to a fuel assembly relocating device. Further, when the operator conducts control rod operation by way of a control rod operation section, the device of the present invention judges in the control rod withdrawal judging section, as to whether the operation indication given by the operator is adequate or not by comparing it with fuel assembly arrangement information. When the operation indication is wrong, a stop signal is sent to control rod drives. With such procedures, increase of nuclear heating upon occurrence of erroneous operation can be prevented. (I.S.)

  1. Nuclear reactors

    Matheson, J.E.

    1983-01-01

    A nuclear reactor has an upper and a lower grid plate. Protrusions project from the upper grid plate. Fuel assemblies having end fittings fit between the grid plates. An arrangement is provided for accepting axial forces generated during the operation of the nuclear reactor by the flow of the cooling medium and thermal expansion and irradiation-induced growth of the fuel assembly, which comprises rods. Each fuel assembly rests on the lower grid plate and its upper end is elastically supported against the upper grid plate by the above-mentioned arrangement. The arrangement comprises four (for example) torsion springs each having a torsion tube and a torsion bar nested within the torsion tube and connected at one end thereto. The other end of the torsion bar is connected to an associated one of four lever arms. The torsion tube is rigidly connected to the other end fitting and the springs are disposed such that the lever arms are biassed against the protrusions. (author)

  2. Reactor core fuel management

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  3. Hybrid adsorptive membrane reactor

    Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  4. Reactor outage schedule (tentative)

    Walton, R.P.

    1969-11-01

    This single page document is the November 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the production reactor.

  5. Reactor outage schedule (tentative)

    Walton, R.P.

    1969-10-01

    This single page document is the October 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  6. Reactor outage schedule (tentative)

    Walton, R.P.

    1969-10-15

    This single page document is the October 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  7. Reactor outage schedule (tentative)

    Walton, R.P.

    1969-09-15

    This single page document is the September 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  8. Reactor outage schedule (tentative)

    Walton, R.P.

    1969-12-15

    This single page document is the December 16, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  9. Reactor outage schedule (tentative)

    Walton, R.P.

    1969-12-01

    This single page document is the December 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  10. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  11. Irradiation of full size UMo plates

    Vacelet, H.; Lavastre, Y.; Grasse, M.; Sacristan, P.; Languille, A.

    1999-01-01

    An important development program for a UMo MTR fuel has been launched in France. The goal of the French working group is to develop a high performing and reprocessable fuel before the end of the US return policy. This paper is focussed on the fabrication of full-sized UMo plates with LEU (Low Enriched Enrichment) and their irradiation in OSIRIS reactor which was started on the 22nd of September. The results of the plates inspection are presented here as well as the irradiation conditions. (author)

  12. The CEA research reactors

    Schwartz, J.P.

    1993-01-01

    Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

  13. Atomic reactor thermal engineering

    Kim, Gwang Ryong

    1983-02-01

    This book starts the introduction of atomic reactor thermal engineering including atomic reaction, chemical reaction, nuclear reaction neutron energy and soon. It explains heat transfer, heat production in the atomic reactor, heat transfer of fuel element in atomic reactor, heat transfer and flow of cooler, thermal design of atomic reactor, design of thermodynamics of atomic reactor and various. This deals with the basic knowledge of thermal engineering for atomic reactor.

  14. Water simulation of sodium reactors

    Grewal, S.S.; Gluekler, E.L.

    1981-01-01

    The thermal hydraulic simulation of a large sodium reactor by a scaled water model is examined. The Richardson Number, friction coefficient and the Peclet Number can be closely matched with the water system at full power and the similarity is retained for buoyancy driven flows. The simulation of thermal-hydraulic conditions in a reactor vessel provided by a scaled water experiment is better than that by a scaled sodium test. Results from a correctly scaled water test can be tentatively extrapolated to a full size sodium system

  15. Nuclear reactor types

    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.)

  16. Fast-acting nuclear reactor control device

    Kotlyar, O.M.; West, P.B.

    1993-01-01

    A fast-acting nuclear reactor control device is described for controlling a safety control rod within the core of a nuclear reactor, the reactor controlled by a reactor control system, the device comprising: a safety control rod drive shaft and an electromagnetic clutch co-axial with the drive shaft operatively connected to the safety control rod for driving and positioning the safety control rod within or without the reactor core during reactor operation, the safety rod being oriented in a substantially vertical position to allow the rod to fall into the reactor core under the influence of gravity during shutdown of the reactor; the safety control rod drive shaft further operatively connected to a hydraulic pump such that operation of the drive shaft simultaneously drives and positions the safety control rod and operates the hydraulic pump such that a hydraulic fluid is forced into an accumulator, filling the accumulator with oil for the storage and supply of primary potential energy for safety control rod insertion such that the release of potential energy in the accumulator causes hydraulic fluid to flow through the hydraulic pump, converting the hydraulic pump to a hydraulic motor having speed and power capable of full length insertion and high speed driving of the safety control rod into the reactor core; a solenoid valve interposed between the hydraulic pump and the accumulator, said solenoid valve being a normally open valve, actuated to close when the safety control rod is out of the reactor during reactor operation; and further wherein said solenoid opens in response to a signal from the reactor control system calling for shutdown of the reactor and rapid insertion of the safety control rod into the reactor core, such that the opening of the solenoid releases the potential energy in the accumulator to place the safety control rod in a safe shutdown position

  17. Resultados da correção cirúrgica da exotropia permanente em pacientes amblíopes e não-amblíopes

    Adriana Valim Portes

    2011-08-01

    Full Text Available A exotropia permanente (XT acomete cerca de 1 a 2% da população. Seu tratamento é clínico: antiambliogênico e correção dos erros refrativos, e cirúrgico. O objetivo do tratamento cirúrgico é alinhar os olhos na posição primária do olhar, proporcionando melhor resultado estético. Há muito tempo diversos autores estudam os fatores pré, per e pós-operatórios relacionados ao resultado cirúrgico, uma vez que a taxa de sucesso varia de 60 a 80%. Ainda são poucos os estudos que comparam a presença de ambliopia como fator de influência no resultado final. OBJETIVO: Comparar o resultado cirúrgico dos pacientes amblíopes e não-amblíopes submetidos à cirurgia de correção de XT. MÉTODOS: Análise retrospectiva de 37 prontuários de pacientes amblíopes (Grupo A e não-amblíopes (Grupo B submetidos à correção cirúrgica de XT por retrocessoressecção monocular, sendo avaliados os registros pós-operatórios imediatos e tardios. Idade: grupo A 24,7 ± 14,2 anos, grupo B 22,6 ±18,6 anos; Desvio pré-operatório: grupo A 29,1± 7,2Δ, grupo B 28,4 ± 6,8Δ. RESULTADOS: A taxa de sucesso foi de 60% e 100% (p<0,05, no pós-operatório imediato e 50% e 82,3% (p=0,082, no pós-operatório final, nos grupos A e B, respectivamente. Não houve diferença significante quanto aos desvios pós-operatórios imediatos, tardios e variação do desvio. CONCLUSÃO: Pode-se concluir que o grupo B mostrou melhor resultado no pós-operatório imediato; porém não houve diferença no resultado cirúrgico de correção de exotropia permanente entre pacientes amblíopes e não-amblíopes no período pós-operatório de seis meses.

  18. Associação entre estado nutricional, hábitos alimentares e nível de atividade física em escolares

    Lorene G. Coelho

    2012-10-01

    Full Text Available OBJETIVO: Analisar a relação entre estado nutricional, hábitos alimentares e nível de atividade física em escolares. MÉTODOS: Realizou-se estudo transversal com 661 escolares de 6-14 anos de idade, matriculados nas escolas públicas e privadas da área urbana da cidade. Um questionário semiestruturado foi aplicado para coleta das variáveis demográficas, socioeconômicas, antropométricas e nível de atividade física. Os escolares foram classificados segundo critérios para índice de massa corporal por idade, percentual de gordura corporal e nível de atividade física, respectivamente. Os dados dietéticos foram obtidos por meio de questionário de frequência alimentar validado e específico para crianças e adolescentes, a partir do qual o consumo alimentar foi avaliado segundo o Recommended Foods Score adaptado, cuja finalidade é a avaliação da qualidade global da dieta com enfoque no consumo de alimentos saudáveis. RESULTADOS: Elevadas frequências de excesso de peso segundo índice de massa corporal por idade (20,1% e percentual de gordura corporal (22,8%, bem como inatividade física (80,3%, foram observadas entre os escolares. Baixas pontuações no escore de consumo foram encontradas para 77,2% dos escolares. Não foi observada associação significativa entre o excesso de peso e inatividade física ou consumo alimentar (p > 0,05. CONCLUSÃO: Os escolares apresentaram elevada frequência de excesso de peso, que de forma isolada não foi associada a alta inatividade física e baixa qualidade da dieta. Assim, mais estudos devem ser realizados para identificar outros fatores além dos descritos que possam estar influenciando a presença do excesso de peso nessa população.

  19. Advanced reactor experimental facilities

    Amri, A.; Papin, J.; Uhle, J.; Vitanza, C.

    2010-01-01

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  20. Reactor safety

    Meneley, D.A.

    The people of Ontario have begun to receive the benefits of a low cost, assured supply of electrical energy from CANDU nuclear stations. This indigenous energy source also has excellent safety characteristics. Safety has been one of the central themes of the CANDU development program from its very beginning. A great deal of work has been done to establish that public risks are small. However, safety design criteria are now undergoing extensive review, with a real prospect of more stringent requirements being applied in the future. Considering the newness of the technology it is not surprising that a consensus does not yet exist; this makes it imperative to discuss the issues. It is time to examine the policies and practice of reactor safety management in Canada to decide whether or not further restrictions are justified in the light of current knowledge

  1. Nuclear reactor

    Schabert, H.P.; Weber, R.; Bauer, A.

    1975-01-01

    The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

  2. Nuclear reactor

    Schweiger, F.; Glahe, E.

    1976-01-01

    In a nuclear reactor of the kind which is charged with spherical reaction elements and in which control rods are arranged to be thrust directly into the charge, each control rod has at least one screw thread on its external surface so that as the rod is thrust into the charge it is caused to rotate and thus make penetration easier. The length of each control rod may have two distinct portions, a latter portion which carries a screw thread and a lead-in portion which is shorter than the latter portion and which may carry a thread of greater pitch than that on the latter portion or may have a number of axially extending ribs instead of a thread

  3. Reactor container

    Furukawa, Hideyasu; Oyamada, Osamu; Uozumi, Hiroto.

    1976-01-01

    Purpose: To provide a container for a reactor provided with a pressure suppressing chamber pool which can prevent bubble vibrating load, particularly negative pressure generated at the time of starting to release exhaust from a main steam escape-safety valve from being transmitted to a lower liner plate of the container. Constitution: This arrangement is characterized in that a safety valve exhaust pool for main steam escape, in which a pressure suppressing chamber pool is separated and intercepted from pool water in the pressure suppressing chamber pool, a safety valve exhaust pipe is open into said safety valve exhaust pool, and an isolator member, which isolates the bottom liner plate in the pressure suppressing chamber pool from the pool water, is disposed on the bottom of the safety valve exhaust pool. (Nakamura, S.)

  4. Nuclear reactors

    Prescott, R.F.

    1976-01-01

    In the system described the fuel elements are arranged vertically in groups and are supported in such a manner as to tend to tilt them towards the center of the respective group, the fuel elements being urged laterally into abutment with one another. The elements have interlocking bearing pads, whereby lateral movement of adjacent elements is resisted; this improves the stability of the reactor core during refuelling operations. Fuel elements may comprise clusters of parallel fuel pins enclosed in a wrapper of hexagonal cross section, with bearing pads in the form of spline-like ribs located on each side of the wrapper and extending parallel to the longitudinal axis of the fuel element, being interlockable with ribs on pads of adjacent fuel elements. The arrangement is applicable to a reactor core in which fuel elements and control rod guide tubes are arranged in modules each of which comprises a cluster of at least three fuel elements, one of which is rigidly supported whilst the others are resiliently tilted towards the center of the cluster so as to lean on the rigidly supported element. It is also applicable to modules comprising a cluster of six fuel elements, each resiliently tilted towards a central void to form a circular arch. The modules may include additional fuel elements located outside the clusters and also resiliently tilted towards the central voids, the latter being used to accommodate control rod guide tubes. The need for separate structural members to act as leaning posts is thus avoided. Such structural members are liable to irradiation embrittlement, that could lead to core failure. (U.K.)

  5. Power reactor noise

    Thie, J.A.

    1981-01-01

    Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions, thus preventing further complications by alerting opeators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Dr. Thie hopes to further, through detailed explanations and over 70 illustrations, the acceptance of the use of noise analysis by the nuclear utility industry. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussions of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

  6. Crystal structure of (E-N-phenyl-N′-[1-(thiophen-2-ylethylidene]formohydrazide

    C. S. Dileep

    2014-09-01

    Full Text Available In the title compound, C13H12N2OS, the planes of the thiophene and phenyl rings are nearly perpendicular to each other, making a dihedral angle of 86.42 (12°. In the crystal, molecules are linked by C—H...O hydrogen bonds, forming a helical chain along the b-axis direction.

  7. Improvement of pressure-vessel surveillance of a PWR-power plant of the Societe d'Energie Nucleaire Franco-Belge des Ardennes (S.E.N.A.)

    Bevilacqua, A.; Lloret, R.; Riehl, R.

    1984-01-01

    This paper describes a new dosimetry, installed inside and outside the Pressure Vessel of CHOOZ Nuclear Power Plant of the Societe d'Energie Nucleaire Franco-Belge des Ardennes (S.E.N.A.), during its 1982-83 operation cycle. The inner dosimetry deals with a simulated capsule located under the reactor plate, and includes copper, nickel, iron, niobium, copper-cobalt, neptunium and uranium dosimeters. Its aim is to qualify the information given by the existing copper dosimetry. The spectrum used with these measurements is obtained by the 1 D ANISN Code and BIP-N 2 library. The outer dosimety is the fluence determination along the outer wall of the vessel. Two tubes, equiped by neutron dosimeters, seven meters long, were fixed along the vessel. On the median plane, the results are compared to a 2 D DOT transport calculation. Preliminary results are given which improve the vessel and specimens neutronic characterisation. (Auth.)

  8. Nuclear reactor neutron shielding

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  9. Environmental aspects of fusion reactors

    Coffman, F.E.; Williams, J.M.

    1975-01-01

    With the continued depletion of fossil and uranium resources in the coming decades, the U. S. will be forced to look more toward renewable energy resources (e.g., wind, tidal, geothermal, and solar power) and toward such longer-term and nondepletable energy resources as fissile fast breeder reactors and fusion power. Several reference reactor designs have been completed for full-scale fusion power reactors that indicate that the environmental impacts from construction, operation, and eventual decommissioning of fusion reactors will be quite small. The principal environmental impact from fusion reactor operation will be from thermal discharges. Some of the safety and environmental characteristics that make fusion reactors appear attractive include an effectively infinite fuel supply at low cost, inherent incapability for a ''nuclear explosion'' or a ''nuclear runaway,'' the absence of fission products, the flexibility of selecting low neutron-cross-section structural materials so that emergency core cooling for a loss-of-coolant or other accident will not be necesary, and the absence of special nuclear materials such as 235 U or 239 Pu, so that diversion of nuclear weapons materials will not be possible and nuclear blackmail will not be a serious concern

  10. Tritium management in fusion reactors

    Galloway, T.R.

    1978-05-01

    This is a review paper covering the key environmental and safety issues and how they have been handled in the various magnetic and inertial confinement concepts and reference designs. The issues treated include: tritium accident analyses, tritium process control, occupational safety, HTO formation rate from the gas-phase, disposal of tritium contaminated wastes, and environmental impact--each covering the Joint European Tokamak (J.E.T. experiment), Tokamak Fusion Test Reactor (TFTR), Russian T-20, The Next Step (TNS) designs by Westinghouse/ORNL and General Atomic/ANL, the ANL and ORNL EPR's, the G.A. Doublet Demonstration Reactor, the Italian Fintor-D and the ORNL Demo Studies. There are also the following full scale plant reference designs: UWMAK-III, LASL's Theta Pinch Reactor Design (RTPR), Mirror Fusion Reactor (MFR), Tandem Mirror Reactor (TMR), and the Mirror Hybrid Reactor (MHR). There are four laser device breakeven experiments, SHIVA-NOVA, LLL reference designs, ORNL Laser Fusion power plant, the German ''Saturn,'' and LLL's Laser Fusion EPR I and II

  11. BEAVRS full core burnup calculation in hot full power condition by RMC code

    Liu, Shichang; Liang, Jingang; Wu, Qu; Guo, JuanJuan; Huang, Shanfang; Tang, Xiao; Li, Zeguang; Wang, Kan

    2017-01-01

    Highlights: • TMS and thermal scattering interpolation were developed to treat cross sections OTF. • Hybrid coupling system was developed for HFP burnup calculation of BEAVRS benchmark. • Domain decomposition was applied to handle memory problem of full core burnup. • Critical boron concentration with burnup by RMC agrees with the benchmark results. • RMC is capable of multi-physics coupling for simulations of nuclear reactors in HFP. - Abstract: Monte Carlo method can provide high fidelity neutronics analysis of different types of nuclear reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. However, nuclear reactors are complex systems with multi-physics interacting and coupling. MC codes can couple with depletion solver and thermal-hydraulics (T/H) codes simultaneously for the “transport-burnup-thermal-hydraulics” coupling calculations. MIT BEAVRS is a typical “transport-burnup-thermal-hydraulics” coupling benchmark. In this paper, RMC was coupled with sub-channel code COBRA, equipped with on-the-fly temperature-dependent cross section treatment and large-scale detailed burnup calculation based on domain decomposition. Then RMC was applied to the full core burnup calculations of BEAVRS benchmark in hot full power (HFP) condition. The numerical tests show that domain decomposition method can achieve the consistent results compared with original version of RMC while enlarging the computational burnup regions. The results of HFP by RMC agree well with the reference values of BEAVRS benchmark and also agree well with those of MC21. This work proves the feasibility and accuracy of RMC in multi-physics coupling and lifecycle simulations of nuclear reactors.

  12. Advances in light water reactor technologies

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  13. Efeitos da governança corporativa e da performance empresarial sobre o turnover de executivos no Brasil: comparando empresas familiares e não-familiares.

    Wesley Mendes-da-Silva

    2006-03-01

    Full Text Available A análise da governança corporativa, incluindo-se a transparência da administração, tem se tornado um fator chave para conhecer as grandes empresas. Ao mesmo tempo, vem sendo apontada como um indicador de confiança dos investidores nas decisões tomadas pela administração das empresas listadas em bolsas de valores. Uma das formas de avaliar as estruturas e práticas de governança corporativa é verificar a sensibilidade do turnover de executivos ao desempenho da firma (VOLPIN, 2002. Neste artigo, objetiva-se verificar a existência de diferenças significativas de desempenho, valor e estruturas de governança corporativa entre empresas familiares e não-familiares; e, ainda, a existência de associações entre performance e turnover de executivos, tendo em vista a atuação da governança corporativa. A análise é conduzida tomando como referência um conjunto de 176 empresas industriais listadas na Bovespa, segmentadas por tipo (familiar ou não-familiar, entre 1997 e 2001. A contribuição do estudo se dá pelas peculiaridades que envolvem as empresas familiares, que não têm sido consideradas em pesquisas brasileiras, e pelo oferecimento de elementos para refletir teoricamente o desempenho das empresas familiares e suas estruturas de governança. Os resultados alcançados revelam a existência de diferenças significativas de desempenho, valor e estruturas de governança corporativa entre empresas familiares e não-familiares. É possível verificar ainda que, nas empresas familiares, quanto maior a rentabilidade, menores são os níveis de turnover.

  14. Desigualdades na mortalidade por suicídio entre indígenas e não indígenas no estado do Amazonas, Brasil

    Maximiliano Loiola Ponte de Souza

    2013-12-01

    Full Text Available OBJETIVO: Analisar as taxas e algumas características da mortalidade por suicídio entre indígenas e não indígenas no Amazonas. MÉTODO: Estudo de coorte retrospectiva, em que os dados de óbito foram obtidos no Sistema de Informações sobre Mortalidade e os populacionais no Instituto Brasileiro de Geografia e Estatística. Foram utilizados todos os registros de suicídio do período de 2006-2010 e a interpolação geométrica anual para a estimação das subpopulações. RESULTADOS: Ocorreram 688 suicídios no Amazonas, dos quais 19,0% em indígenas. A taxa ajustada de mortalidade por suicídio (TAMS nos indígenas, de 18,4/100 mil, foi 4,4 vezes superior a dos não indígenas. A TAMS em indígenas aumentou 1,6 vez em 2010 em relação a 2006. Nos municípios de Tabatinga e São Gabriel da Cachoeira, as TAMS foram muito altas, 75,8 e 41,9/100 mil, respectivamente. CONCLUSÕES: Evidenciou-se o comportamento desigual das taxas de mortalidade por suicídio entre indígenas e não indígenas, expondo não só sua importância local, como também sua invisibilidade como problema de saúde pública, principalmente entre jovens 15 e 24 anos.

  15. Tipos de divulgação da informação financeira e não financeira de responsabilidade social empresarial

    Maria José da Silva Faria

    Full Text Available Resumo A informação de responsabilidade social empresarial assumiu, nas últimas décadas, uma importância elevada para preparadores e utilizadores e intensificou-se desde 2007, com a crise financeira americana, espelhada em Portugal desde 2008. O crescente volume de informação apresenta várias formas de divulgação de ações que tenham impacto no desempenho das empresas e reflexo na visão dos stakeholders acerca da empresa na comunidade em que se insere. O objetivo deste artigo é elencar diferentes formas de divulgação de informação de responsabilidade social pelas empresas e destacar os canais usados estrategicamente para o efeito. Cumulativamente, são listadas vantagens e desvantagens dos meios utilizados para contribuir para o questionamento dos atuais recursos em uso para a divulgação da informação socialmente responsável. A metodologia adotada foi a análise bibliográfica de estudos recentes para levantar o estado da arte sobre o tema. Trata-se de pesquisa descritiva realizada em empresas nacionais cotadas na bolsa de valores de Lisboa no 1º semestre de 2016. Conclui-se que existe uma multiplicidade de informações divulgadas pelas empresas acerca das suas ações de responsabilidade social: escrita e não escrita, financeira e não financeira e interna e externa. As páginas web e os relatórios tradicionais parecem ser mais usados, apesar de subsistirem dificuldades de medida no relato quantitativo em comparação com o qualitativo.

  16. Análise de desempenho de medidas estruturais e não estruturais de controle de inundações em uma bacia urbana

    Thiago Galvão Tiradentes Decina

    2016-03-01

    Full Text Available RESUMO Este trabalho teve como objetivo analisar, por meio de simulações hidrológicas e hidráulicas, o desempenho de algumas medidas de controle de inundação, estruturais e não estruturais. Para tanto, os mapas e demais informações espaciais da bacia hidrográfica do Córrego do Gregório, em São Carlos (SP, foram digitalizadas com auxílio do software de geoprocessamento ArcGIS(r e imagens de satélite de alta resolução. Com base nessas informações, foram propostos cenários alternativos que incorporaram as medidas de controle. Através da utilização dos softwares HEC-HMS e HEC-RAS realizaram-se as modelações hidráulicas e hidrológicas, utilizando tempos de retorno de 25, 50 e 100 anos para a chuva de projeto. Com os resultados obtidos (hidrogramas de cheia e manchas de inundação referentes a cada tempo de retorno, foi possível analisar as medidas, comparando-se os diferentes cenários. Concluiu-se, para este caso, que os melhores resultados correspondem à associação das medidas estruturais e não estruturais. Contudo, também foi verificado que, mesmo com a concretização do cenário mais favorável, o problema das inundações não seria satisfatoriamente equacionado, o que revela a necessidade de se considerar outras medidas para minimizar os prejuízos decorrentes das inundações, tais como medidas de controle na fonte, seguros contra inundações e sistemas de alerta antecipados.

  17. Método para avaliação da potencialidade ornamental de flores e folhagens de corte nativas e não convencionais.

    Elisabeth R. Tempel Stumpf

    2007-06-01

    Full Text Available Para garantir o êxito no mercado de novos produtos da Floricultura, é preciso estabelecer critérios objetivos de escolha das características ornamentais desejáveis. Este trabalho tem como objetivo propor e validar um método para a avaliação do potencial ornamental de plantas nativas e não-convencionais, para uso como flores e folhagens de corte. A partir de uma planilha de avaliação especialmente elaborada para tal fim, foi testado o grau de potencialidade ornamental de 11 espécies nativas do Rio Grande do Sul, resultante do somatório das notas dadas para cada uma das dez características escolhidas por sua importância para a arte floral e para o mercado. Para a validação da eficácia do método, foi efetuada uma pesquisa de opinião on line. Andropogon bicornis, Baccharis articulata, B. usterii, Eryngium ebracteatum, E. eriophorum, E. sanguisorba, Hypericum connatum, Limonium brasiliense, Myrsine umbellata, Schinus lentiscifolius e S. terebinthifolius receberam pontuação superior a 70 pontos, sendo categorizadas como de alta potencialidade ornamental. A eficácia da escolha destas espécies com base em atributos estéticos visuais ficou comprovada nos resultados obtidos pela aplicação da planilha e posteriormente validados pela pesquisa de opinião. Os resultados indicam que o método aqui proposto é uma ferramenta eficaz para avaliar o grau de potencialidade ornamental de flores e folhagens de corte nativas e não-convencionais

  18. Estudo comparativo entre o trauma em idosos e não idosos atendidos em um Hospital Universitário de Curitiba

    César Augusto Broska Júnior

    Full Text Available OBJETIVO: comparar e identificar diferenças no perfil de pacientes idosos e não idosos atendidos por trauma. MÉTODOS: estudo transversal comparativo quantitativo retrospectivo realizado com 3112 pacientes entre 25/11/2010 e 25/02/2011 que foram classificados em GI idosos (60 anos ou mais e GA não idosos (13 a 59 anos. Foram coletadas informações sobre mecanismo, lesões e fatores associados ao acidente que foram comparados entre os grupos através da utilização dos testes χ2, t de Student e proporções. RESULTADOS: as quedas foram mais frequentes em GI em destaque as de própria altura enquanto que violência, quedas e acidentes de trânsito (em especial os com moto foram os mais importantes em GA. Ambos os grupos possuíram traumas leves (contusões, entorses como os mais frequentes, seguido por traumatismo crânio-encefálico (TCE e fraturas. Nos membros inferiores as fraturas de fêmur foram mais comuns em idosos enquanto que fraturas de mão e punho nos não idosos. Os idosos ficaram menos dias internados e não houve diferença quanto a presença de complicações, necessidade de UTI ou mortalidade entre os grupos. CONCLUSÃO: Quedas de mesmo nível são mais frequentes em idosos e acidentes de moto, violência interpessoal em não idosos. Ambos os grupos apresentaram traumas superficiais, TCE e fraturas como principais lesões decorrentes do trauma. Os idosos ficaram menos dias internados. Não houve diferença quanto a necessidade de UTI, presença de complicações e mortalidade.

  19. US DOE Idaho national laboratory reactor decommissioning

    Szilagyi, Andrew

    2012-01-01

    The United States Department of Energy (DOE) primary contractor, CH2M-WG Idaho was awarded the cleanup and deactivation and decommissioning contract in May 2005 for the Idaho National Lab (INL). The scope of this work included dispositioning over 200 Facilities and 3 Reactors Complexes (Engineering Test Reactor (ETR), Materials Test Reactor (MTR) and Power Burst Facility (PBF) Reactor). Two additional reactors were added to the scope of the contract during the period of performance. The Zero Power Physics Reactor (ZPPR) disposition was added under a separate subcontractor with the INL lab contractor and the Experimental Breeder Reactor II (EBR-II) disposition was added through American Recovery and Reinvestment Act (ARRA) Funding. All of the reactors have been removed and disposed of with the exception of EBR-II which is scheduled for disposition approximately March of 2012. A brief synopsis of the 5 reactors is provided. For the purpose of this paper the ZPPR reactor due to its unique design as compared to the other four reactors, and the fact that is was relatively lightly contaminated and irradiated will not be discussed with the other four reactors. The ZPPR reactor was readily accessible and was a relatively non-complex removal as compared to the other reactors. Additionally the EBR-II reactor is currently undergoing D and D and will have limited mention in this paper. Prior to decommissioning the reactors, a risk based closure model was applied. This model exercised through the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA), Non-Time Critical Removal Action (NTCRA) Process which evaluated several options. The options included; No further action - maintain as is, long term stewardship and monitoring (mothball), entombment in place and reactor removal. Prior to commencing full scale D and D, hazardous constituents were removed including cadmium, beryllium, sodium (passivated and elemental), PCB oils and electrical components, lead

  20. CANDU reactor - supporting the nuclear renaissance

    Oberth, R.

    2010-01-01

    'Full text:' The CANDU reactor has proven to be a strong performer in both the Canada, with 22 units constructed in Ontario, New Brunswick and Quebec, as well as in Argentina, Korea, Romania and China where a further nine units are operating and two in the planning stage. The average lifetime capacity factor of the CANDU reactor fleet is 89%. The last seven CANDU projects in Korea, China, and Romania have been completed on budget and on schedule. CANDU reactors have the highest uranium utilization efficiency measures as electricity output per ton of uranium mined. The CANDU fuel channel design using on-power fuelling and a heavy water moderator enables flexible fueling options - from the current natural uranium option to burning uranium recovered from used LWR reactor fuel and even a thorium-based fuel. AECL and the CANDU reactor are poised to participate in the worldwide construction at least 250 new reactors over the next 20 years. (author)

  1. Sulfide toxicity kinetics of a uasb reactor

    D. R. Paula Jr.

    2009-12-01

    Full Text Available The effect of sulfide toxicity on kinetic parameters of anaerobic organic matter removal in a UASB (up-flow anaerobic sludge blanket reactor is presented. Two lab-scale UASB reactors (10.5 L were operated continuously during 12 months. The reactors were fed with synthetic wastes prepared daily using glucose, ammonium acetate, methanol and nutrient solution. One of the reactors also received increasing concentrations of sodium sulfide. For both reactors, the flow rate of 16 L.d-1 was held constant throughout the experiment, corresponding to a hydraulic retention time of 15.6 hours. The classic model for non-competitive sulfide inhibition was applied to the experimental data for determining the overall kinetic parameter of specific substrate utilization (q and the sulfide inhibition coefficient (Ki. The application of the kinetic parameters determined allows prediction of methanogenesis inhibition and thus the adoption of operating parameters to minimize sulfide toxicity in UASB reactors.

  2. Application of modified R.E.N.A.L. nephrometry score system in evaluating the retroperitoneal partial nephrectomy for T1 renal cell carcinoma.

    Wang, Qinzhang; Qian, Biao; Li, Qiang; Ni, Zhao; Li, Yinglong; Wang, Xinmin

    2015-01-01

    This study aims to investigate the application of the modified R.E.N.A.L. nephrometry score system in evaluating the operation difficulty of retroperitoneal partial nephrectomy in T1 renal cell carcinoma patients. A total of 52 patients with T1 renal cell carcinoma were enrolled. They all had retroperitoneal partial nephrectomy. Their clinical data was retrospectively analyzed. R.E.N.A.L. nephrometry score system was modified based on the features of retroperitoneal partial nephrectomy. The specificity, sensitivity and Youden index were compared between R.E.N.A.L. nephrometry score system and the modified R.E.N.A.L. nephrometry score system. The effect of the modified R.E.N.A.L. nephrometry score system on perioperative outcomes was analyzed. Three degrees of operation difficulty were defined by the modified R.E.N.A.L. nephrometry score system, which included the low, medium and high degree of operation difficulty. The specificity, sensitivity and Youden index of the modified R.E.N.A.L. nephrometry score system were better than those of the original R.E.N.A.L. nephrometry score system. Compared with low degree of operation difficulty, patients with medium and high degree of operation difficulty had significantly higher levels of operative time, warm ischemia time, and intraoperative blood loss (P system has a good effect in evaluating the operation difficulty of retroperitoneal partial nephrectomy.

  3. Solution of heat removal from nuclear reactors by natural convection

    Zitek Pavel

    2014-03-01

    Full Text Available This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR.The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

  4. FBR type reactor

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  5. The prototype fast reactor

    Broomfield, A.M.

    1985-01-01

    The paper concerns the Prototype Fast Reactor (PFR), which is a liquid metal cooled fast reactor power station, situated at Dounreay, Scotland. The principal design features of a Fast Reactor and the PFR are given, along with key points of operating history, and health and safety features. The role of the PFR in the development programme for commercial reactors is discussed. (U.K.)

  6. Department of reactor technology

    1980-01-01

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  7. NCSU Reactor Sharing Program

    Perez, P.B.

    1993-01-01

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities

  8. Reactor safety method

    Vachon, L.J.

    1980-01-01

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature

  9. ROSA full-core and DNBR capabilities

    Gibcus, H.P.M.; Verhagen, F.C.M.; Wakker, P.H.

    2013-01-01

    The latest developments of the ROSA (Reloading Optimization by Simulated Annealing) code system with an emphasis on the first full-core version and the minimum DNBR (Departure from Nucleate Boiling Ratio) as a new optimization parameter are presented. Designing the core loading pattern of nuclear power plants is becoming a more and more complex task. This task becomes even more complicated if asymmetries in the core loading pattern arise, for instance due to damaged fuel assemblies. For over almost 2 decades ROSA, NRG's (Nuclear Research and consultancy Group) loading pattern optimization code system for PWRs, has proven to be a valuable tool to reactor operators in accomplishing this task. To improve the use of ROSA for designing asymmetric loading patterns, NRG has developed a full-core version of ROSA besides the original quarter-core version which requires rotational symmetry in the computational domain. The extension of ROSA with DNBR as an optimization parameter is part of ROSA's continuous development. (orig.)

  10. Full-length fuel rod behavior under severe accident conditions

    Lombardo, N.J.; Lanning, D.D.; Panisko, F.E.

    1992-12-01

    This document presents an assessment of the severe accident phenomena observed from four Full-Length High-Temperature (FLHT) tests that were performed by the Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. These tests were conducted for the US Nuclear Regulatory Commission (NRC) as part of the Severe Accident Research Program. The objectives of the test were to simulate conditions and provide information on the behavior of full-length fuel rods during hypothetical, small-break, loss-of-coolant severe accidents, in commercial light water reactors

  11. Inteligência artificial é mesmo de Spielberg e não de Kubrick

    Opermann , Alvaro

    2001-01-01

    Full Text Available Steven Spielberg, como já se escreveu inúmeras vezes, é o mais proeminente representante da Nova Hollywood, a geração que também reúne Coppola, Scorsese, Paul Schrader, George Lucas, Michael Cimino, Brian De Palma, Peter Bogdanovich, William Friedkin e Terrence Malick. É a geração que, para o bem e para o mal, substituiu os remanescentes da Velha Hollywood (Robert Aldrich, Robert Wise, Mark Robson, John Huston, Billy Wilder, Don Siegel.

  12. Inteligência artificial é mesmo de Spielberg e não de Kubrick

    Opermann , Alvaro

    2001-01-01

    Full Text Available Steven Spielberg, como já se escreveu inúmeras vezes, é o mais proeminente representante da Nova Hollywood, a geração que também reúne Coppola, Scorsese, Paul Schrader, George Lucas, Michael Cimino, Brian De Palma, Peter Bogdanovich, William Friedkin e Terrence Malick. É a geração que, para o bem e para o mal, substituiu os remanescentes da Velha Hollywood (Robert Aldrich, Robert Wise, Mark Robson, John Huston, Billy Wilder, Don Siegel

  13. Physics of nuclear reactors

    Baeten, Peter

    2006-01-01

    This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

  14. Reactor core and initially loaded reactor core of nuclear reactor

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

    In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

  15. Nuclear reactors. Introduction

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

  16. What is the future for fast reactor technology?

    Kraev, Kamen

    2017-01-01

    NucNet spoke to Vladimir Kriventsev, team leader for fast reactor technology development at the International Atomic Energy Agency (IAEA), about the possibilities and challenges of technology development in the fast reactor sector. Today, the field of fast reactors is vibrant and full of fascinating developments, some which will have an impact in the nearer term and others in the longer term.

  17. What is the future for fast reactor technology?

    Kraev, Kamen [NucNet, Brussels (Belgium). The Independent Global Nuclear News Agency

    2017-08-15

    NucNet spoke to Vladimir Kriventsev, team leader for fast reactor technology development at the International Atomic Energy Agency (IAEA), about the possibilities and challenges of technology development in the fast reactor sector. Today, the field of fast reactors is vibrant and full of fascinating developments, some which will have an impact in the nearer term and others in the longer term.

  18. Thermonuclear reactor

    Yasutomi, Yoshiyuki; Nakagawa, Moroo; Sawai, Yuichi; Chiba, Akio; Suzuki, Yasutaka.

    1997-01-01

    Silicon composited with reinforcing metals is used for a divertor cooling substrate having an effect as a cooling tube to provide a silicon base composite material having increased electric resistance and toughness. The blending ratio of reinforcing materials in the form of granules, whiskers or long fibers is controlled in order to control heat conductivity, electric resistivity and mechanical performances. The divertor cooling substrate comprising the silicon base composite material is integrated with a plasma facing material. The production method therefor includes ordinary metal matrix composite forming methods such as powder metallurgy, melting penetration method, high pressure solidification casting method, centrifugal casting method and vacuum casting method. Since the cooling plate is constituted with the light metal and highly electric resistant metal base composite material, sharing force due to eddy current can be reduced, and radiation exposure can be minimized. Accordingly, a cooling structure for a thermonuclear reactor effective for the improvement of environmental problems caused by waste disposal can be attained. (N.H.)

  19. Nuclear reactor

    Irion, L.; Tautz, J.; Ulrych, G.

    1976-01-01

    This additional patent complements the arrangement of non-return valves to prevent loss of cooling water on fracture of external tubes in the main coolant circuit (according to PS 24 24 427.7) by ensuring that the easily movable valves only operate in case of a fault, but do not flutter in operation, because the direction of flow is not the same at each location where they are installed. The remedy for this undesirable effect consists of allocating 1 non-return valve unit with 5 to 10 valves to each (of several) ducts for the cooling water intake. These units are installed in the annular space between the reactor vessel and the pressure vessel below the inlet of the ducts. Due to flow guidance surfaces in the same space, the incoming cooling water is deflected downwards and as the guiding surfaces are closed at the sides, must pass parallel to the valves of the non-return valve unit. On fracture of the external cooling water inlet pipe concerned, all valves of this unit close due to reversal of flow on the outlet side. (TK) [de

  20. Nuclear reactors

    Pearson, K.G.

    1977-01-01

    Reference is made to auxiliary means of cooling the nuclear fuel clusters used in light or heavy water cooled nuclear reactors. One method is to provide one or more spray cooling tubes. From holes in the side walls of those tubes coolant water may be sprayed laterally into the cluster against the rods. The flow of main coolant may thus be supplemented or even replaced by the auxiliary coolant. A difficulty, however, is that only those fuel rods close to a spray cooling tube can readily be reached by the auxiliary coolant. In the arrangement described, where the fuel rods are spaced apart by transverse grids, at least one of the interspaces between the grids is provided with an axially extending auxiliary coolant conduit having lateral holes through which an auxiliary coolant is sprayed into the cluster. A deflector is provided that extends from a transverse grid into a position in front of the holes and deflects auxiliary coolant on to parts of the fuel rods otherwise inaccessible to the auxiliary coolant. The construction of the deflector is described. (U.K.)

  1. Fast breeder reactors

    Heinzel, V.

    1975-01-01

    The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

  2. Material test reactor fuel research at the BR2 reactor

    Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

    2012-03-15

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  3. Bis[(E-N-(pyridin-3-ylmethylidenehydroxylamine-κN1]silver(I perchlorate

    Jing Xu

    2012-06-01

    Full Text Available Each of the ions in the title salt, [Ag(C6H6N2O2]ClO4, is completed by the application of crystallographic twofold symmetry. The AgI atom is coordinated by two pyridine N atoms in an almost linear fashion [N—Ag—N = 170.0 (2°], with the T-shaped coordination geometry being completed by a weakly associated perchlorate-O atom. Supramolecular zigzag chains along [100] mediated by O—H...N hydrogen bonds [as parts of R22(6 loops] feature in the crystal packing. The perchlorate O atoms are disordered over two sets of sites in a statistical ratio.

  4. Incerteza e não ergodicidade: crítica aos neoclássicos

    Marcelo Mallet Siqueira Campos

    2014-06-01

    Full Text Available The starting point of this essay is to show that, in our view, the problem of the traditional economics is not in the deductive method nor the mathematical methods used, but to attribute to economic agents "power" on the future and prescribe the existence of ergodic stochastic processes in their economic analyzes. Thus, building a theory on the ground whose bases are not able to sustain a proper understanding of the world, mainstream economics has difficulties in using the modeling for establishing deductions and conclusions that help understanding the system. Thus, the logical-mathematical rigor in economic models and deduction can be used with appropriate axioms, which is not the case of mainstream economics. Our hypothesis is that the inability of the mainstream in predicting economic crisis is due to the non-recognition of some principles that best describe the dynamics of financialized contemporary capitalism, as the principles of non-ergodicity and Keynesian uncertainty.

  5. Analfabetismo e níveis de letramento no Brasil: o que dizem os censos?

    Ferraro Alceu Ravanello

    2002-01-01

    Full Text Available O texto começa discutindo a velha questão da qualidade das estatísticas educacionais e alguns aspectos metodológicos relacionados com a utilização destas na pesquisa em educação. Analisa a seguir o analfabetismo, focalizando: a a sua emergência como problema político no final do período imperial; b a evolução do conceito; c a tendência secular, em números porcentuais e absolutos, desde o primeiro censo em 1872 até o Censo 2000. Por último, com base no Censo 2000, classifica a população em diferentes níveis de letramento.

  6. Conservadorismo contábil nas companhias abertas familiares e não-familiares no mercado brasileiro

    Iana Izadora Souza Lapa de Melo Paulo

    2015-05-01

    observadas nesta pesquisa apontam que as empresas familiares apresentam maior reversão de perdas contábeis do que as firmas não-familiares, sugerindo que elas têm práticas contábeis mais agressivas. Porém, com base na análise estatística proposta, não se pode confirmar que o conservadorismo contábil é significativamente diferente entre as empresas familiares e não-familiares.

  7. [Outstanding Soviet zoologist and parasitologist E. N. Pavlovsky--the creator of the theory of natural foci of disease].

    Pavlovskyĭ, L N

    2011-01-01

    The article presents information on the outstanding Soviet Zoology and Parasitology, Academician of the Academy of Sciences of the USSR Academy of Medical Sciences of the USSR, Hero of Socialist Labour, Lieutenant-General of the Medical Service E. N. Pavlovsky, the author of more than 1500 scientific papers, the founder of scientific school, one of the few scholars the twentieth century, approaching the level of scientists and encyclopedists. Considered its contribution to the study of natural foci of diseases has promoted the development of environmental trends in parasitology.

  8. Reconhecimento da prole por operárias companheiras e não companheiras de ninho em Acromyrmex laticeps nigrosetosus Forel, 1908 (Hymenoptera, Formicidae Brood recognition by workers of Acromyrmex laticeps nigrosetosus Forel, 1908 (Hymenoptera, Formicidae

    Danival José de Souza

    2003-02-01

    Full Text Available Estudou-se a capacidade de discriminação de formas jovens de Acromyrmex laticeps nigrosetosus por operárias adultas da mesma subespécie. Eram oferecidas, na área de forrageamento, larvas e pupas companheiras e não companheiras de ninho, sendo quantificado o comportamento frente a essas formas jovens. Foram utilizadas colônias oriundas do município de Paraopeba, MG, Brasil, mantidas em condições de laboratório. Os resultados evidenciaram que essa subespécie não é capaz de discriminar formas jovens companheiras e não companheiras de ninho, ou seja, transportaram indiscriminadamente as formas jovens oferecidas para o interior do ninho. Também não se observou diferença significativa para o tempo de resposta de aceitação de prole companheira e não companheira de ninho.This study investigated the behavioral response (acception or rejection of Acromyrmex laticeps nigrosetosus to their brood and to brood from different colonies of this subespecies. The four colonies used in the bioassays came from Paraopeba, MG, Brazil. Workers accepted either brood from their colonies or from different colonies. There was no significant difference on the time for brood acceptance (transport to the interior of the nest among nestmates and non-nestmates.

  9. As antinomias metaéticas entre cegos e não-cegos e o problema do realismo moral = The antinomies metaethical between the blind and not-blind and the problem of moral realism

    Peruzzo Júnior, Léo

    2016-01-01

    Full Text Available O presente artigo apresenta o cognitivismo moral pragmático como forma de superar as antinomias clássicas entre cegos e não-cegos e a (impossibilidade de representação mental do conteúdo moral. Esta posição é sustentada a partir da reconstrução do debate metaético instaurado entre cognitivistas e não-cognitivistas, especialmente nas interpretações fornecidas pelo realismo moral de McDowell e pelo quase-realismo de Blackburn. Assim, o pressuposto comum a essas duas teorias metaéticas [cognitivismo e não-cognitivismo] pretende ser superado pelo cognitivismo pragmático, uma vez que suas falsas dicotomias partem da tese de que todo conhecimento moral é proposicional. Para sanar esse paradoxo, evitamos uma dicotomia radical entre fatos e valores, isto é, não há ações que podem ser valoradas extrinsecamente ao contexto do seu uso

  10. Advanced Test Reactor probabilistic risk assessment

    Atkinson, S.A.; Eide, S.A.; Khericha, S.T.; Thatcher, T.A.

    1993-01-01

    This report discusses Level 1 probabilistic risk assessment (PRA) incorporating a full-scope external events analysis which has been completed for the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory

  11. Reactor Physics Training

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  12. (E-N,N-Diethyl-2,6-diisopropyl-4-[2-(4-nitrophenylethenyl]aniline

    Heiner Detert

    2013-12-01

    Full Text Available The title compound, C24H32N2O2, was prepared by Horner olefination of 4-diethylamino-3,5-diisopropylbenzaldehyde and diethyl p-nitrobenzylphosphonate. There are two independent molecules (A and B in the asymmetric unit. Their main axes, defined by the line connecting the N atoms of the nitro and amino groups, open an angle of 79.42 (3°. Steric hindrance around the amino group is reflected in a long aryl C—N bond [1.434 (3 Å for molecule A and 1.440 (3 Å for molecule B], a pyramidal geometry [angle sum = 350.0 (2° for molecule A and 349.6 (2° for molecule B], and dihedral angles between the phenylene group and the plane defined by the CH2—N—CH2 unit of 86.9 (3° for molecule A and 88.3 (3° for molecule B. This gives structural support for the electronic decoupling of the amino group from the nearly planar nitrostilbene moiety (r.m.s. deviation for C, N and O atoms = 0.097 for molecule A and 0.107 Å for molecule B.

  13. Variações sobre ver e não ver: dois relatos de casos

    Marcia Oliveira Moraes

    2011-04-01

    Full Text Available This paper presents two cases report referring to the subject of visual disability. On one hand, we discuss the western modernity conception of disability, that conceives it as a tragedy, as a difference in relation to a norm. On the other hand, we underline that considering the experience of blind people it is possible to affirm that there are other ways to live without vision, and this is far from the conception of disability as a deficit. Institutional Analysis is the theoretical basis that we adopt in order to discuss these concepts. In the cases reported we pointed to the crisis events which allows us to question the conception of visual disability as a deficit and which enact new ways to experience the lack of one sense. We conclude indicating that to deal with inclusion it is necessary to question the western conception of disability. We also pointed to the possibilities of a psychological intervention which is able to bring into being new ways to live without seeing.

  14. Variações sobre ver e não ver: dois relatos de casos

    Marcia Oliveira Moraes

    2011-01-01

    Full Text Available El artículo presenta dos informes de casos de discapacidad visual. Por un lado, hemos analizado el concepto de discapacidad que marca la modernidad occidental: lo que dice la discapacidad como una desviación. Por otro lado, hemos señalado que en la experiencia de las personas ciegas se mantienen otras formas de existir sin visión, lejos de los de déficit que marcan el pensamiento occidental. El análisis institucional es la herramienta teórico que nos permite analizar los conceptos en discussion. En los casos reportados se señalan los momentos de crisis que puso en jaque a la institución de discapacidade visual tal como la incapacidad y que producem nuevas formas de vivir a la falta de un sentido. El concepto de "subjetividad deficiente" se encuentra en análisis. Finalizamos señalando las vías para tratar la inclusión que apuntan a la problematisation de las nociones modernas de la discapacidad y que ponen de relieve las posibilidades de intervención en psicología capaces de potenciar la producción de nuevas formas de vida sin ver.

  15. The fast breeder reactor

    Collier, J.

    1990-01-01

    The arguments for and against the fast breeder reactor are debated. The case for the fast reactor is that the world energy demand will increase due to increasing population over the next forty years and that the damage to the global environment from burning fossil fuels which contribute to the greenhouse effect. Nuclear fission is the only large scale energy source which can achieve a cut in the use of carbon based fuels although energy conservation and renewable sources will also be important. Fast reactors produce more energy from uranium than other types of (thermal) reactors such as AGRs and PWRs. Fast reactors would be important from about 2020 onwards especially as by then many thermal reactors will need to be replaced. Fast reactors are also safer than normal reactors. The arguments against fast reactors are largely economic. The cost, especially the capital cost is very high. The viability of the technology is also questioned. (UK)

  16. Education and Training on ISIS Research Reactor

    Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X.

    2013-01-01

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions

  17. Education and Training on ISIS Research Reactor

    Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X. [French Atomic Energy and Alternative Energies Commission, Paris (France)

    2013-07-01

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions.

  18. Verification of reactor safety codes

    Murley, T.E.

    1978-01-01

    The safety evaluation of nuclear power plants requires the investigation of wide range of potential accidents that could be postulated to occur. Many of these accidents deal with phenomena that are outside the range of normal engineering experience. Because of the expense and difficulty of full scale tests covering the complete range of accident conditions, it is necessary to rely on complex computer codes to assess these accidents. The central role that computer codes play in safety analyses requires that the codes be verified, or tested, by comparing the code predictions with a wide range of experimental data chosen to span the physical phenomena expected under potential accident conditions. This paper discusses the plans of the Nuclear Regulatory Commission for verifying the reactor safety codes being developed by NRC to assess the safety of light water reactors and fast breeder reactors. (author)

  19. Nuclear reactor instrumentation at research reactor renewal

    Baers, B.; Pellionisz, P.

    1981-10-01

    The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

  20. Safeguarding research reactors

    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

  1. Guide to power reactors

    NONE

    1959-07-15

    The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

  2. Sede social de H A Z E N Madrid – España

    Gamboa, P.

    1976-05-01

    Full Text Available For the greatly renowned firm in the field of musical Instruments, Hazen, a building has been constructed in the outskirts of Madrid to house a permanent exhibition of pianos as well as the head offices of the firm. It comprises basement and two storeys and on the ground floor there is a big exhibition hall with adjoining rooms for assembling the pianos. The packed Instruments are stored in the basement which further contains a garage, two small offices and the living quarters of the porter. The management section and the main offices are situated on the second floor. Externally, there is a sharp contrast between the light weight of the enclosure of the ground floor, made of metal frames and glass, and that of the top floor which is a solid volume, which hides the roof and has Windows only on the sides. The construction is surrounded by big gardens.Para la firma Hazen, de gran prestigio en el campo de los instrumentos musicales, se ha construido un edificio en las afueras de Madrid destinado a albergar una exposición permanente de pianos y sus oficinas centrales. Consta de un sótano y dos plantas sobre el terreno, situándose en la planta noble la gran sala de exposición, con un local anexo para el montaje de los pianos. En el sótano se almacenan los instrumentos embalados; completan este nivel un garaje, dos pequeñas oficinas y una vivienda para el conserje. La zona principal de oficinas y los despachos de dirección se distribuyen en la planta superior. Exteriormente, contrasta de manera clara la ligereza del cerramiento de la planta baja, a base de carpintería metálica con acristalamiento total, y la planta superior, constituida por un volumen macizo que oculta la cubierta y que solamente tiene huecos en sus fachadas laterales. Rodeando la construcción hay una amplia zona verde cuidadosamente ajardinada.

  3. Pesquisando no perigo: etnografias voluntárias e não acidentais

    Alba Zaluar

    2009-10-01

    Full Text Available O artigo discute a questão do relativismo cultural que assume caráter polêmico na passagem do milênio, principalmente nas divergências surgidas sobre o que fazer diante dos dilemas postos pelos novos sujeitos-objetos e dos impasses advindos da difusão planetária de posturas universalistas. Novos problemas e dilemas éticos se antepuseram ao fazer antropológico: o que são direitos humanos e qual o seu alcance entre nativos não ocidentais; quem conta como pessoa humana e quem deve ser ouvido nas sociedades estudadas; como lidar com a opressão de pessoas no interior das "sociedade dos nativos" quando os opressores são também nativos. Na etnografia do crime, é possível e preciso o mergulho no mundo do outro a ponto de virar nativo? A antropologia do contemporâneo, praticada em sociedades cada vez mais diferenciadas e em um mundo que nunca foi tão globalizado, exige a multiplicidade de fontes de dados, de planos de abordagem e de perspectivas teóricas, averiguando o que acontece em outras disciplinas, sendo cauteloso e paciente na invenção de novos conceitos e novas pesquisas.The article discusses the question of cultural relativism, an issue that has become increasingly polemical at the turn of the millennium, especially with the diverging opinions on how to respond to the dilemmas posed by the new subjects-objects and the impasses generated by the global diffusion of universalist stances. New ethical problems and dilemmas prevene anthropological practice: what are human rights and what are their reach among non-western natives; who counts as a human person and who should be heard in the societies under study; and how to deal with the oppression of people within the 'native societies' when the oppressors are themselves natives? In the ethnography of crime, is it possible or necessary to submerge oneself in the world of the other to the point of turning native? The anthropology of the contemporary, practiced in increasingly

  4. Reactor core of FBR type reactor

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

    A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

  5. Reactor core for LMFBR type reactors

    Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

    1987-01-01

    Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

  6. Research reactor fuel - an update

    Finlay, M.R.; Ripley, M.I.

    2003-01-01

    In the two years since the last ANA conference there have been marked changes in the research reactor fuel scene. A new low-enriched uranium (LEU) fuel, 'monolithic' uranium molybdenum, has shown such promise in initial trials that it may be suitable to meet the objectives of the Joint Declaration signed by Presidents Bush and Putin to commit to converting all US and Russian research reactors to LEU by 2012. Development of more conventional aluminium dispersion UMo LEU fuel has continued in the meantime and is entering the final qualification stage of multiple full sized element irradiations. Despite this progress, the original 2005 timetable for UMo fuel qualification has slipped and research reactors, including the RRR, may not convert from silicide to UMo fuel before 2007. The operators of the Swedish R2 reactor have been forced to pursue the direct route of qualifying a UMo lead test assembly (LTA) in order to meet spent fuel disposal requirements of the Swedish law. The LTA has recently been fabricated and is expected to be loaded shortly into the R2 reactor. We present an update of our previous ANA paper and details of the qualification process for UMo fuel

  7. Reactor operations at SAFARI-1

    Vlok, J.W.H.

    2003-01-01

    A vigorous commercial programme of isotope production and other radiation services has been followed by the SAFARI-1 research reactor over the past ten years - superimposed on the original purpose of the reactor to provide a basic tool for nuclear research, development and education to the country at an institutional level. A combination of the binding nature of the resulting contractual obligations and tighter regulatory control has demanded an equally vigorous programme of upgrading, replacement and renovation of many systems in order to improve the safety and reliability of the reactor. Not least among these changes is the more effective training and deployment of operations personnel that has been necessitated as the operational demands on the reactor evolved from five days per week to twenty four hours per day, seven days per week, with more than 300 days per year at full power. This paper briefly sketches the operational history of SAFARI-1 and then focuses on the training and structuring currently in place to meet the operational needs. There is a detailed step-by-step look at the operator?s career plan and pre-defined milestones. Shift work, especially the shift cycle, has a negative influence on the operator's career path development, especially due to his unavailability for training. Methods utilised to minimise this influence are presented. The increase of responsibilities regarding the operation of the reactor, ancillaries and experimental facilities as the operator progresses with his career are discussed. (author)

  8. Reactors set for mini market

    Knox, Richard.

    1988-01-01

    Commercial nuclear power generation on a large-scale has an uncertain future. However, it is hoped that a small nuclear reactor could form the basis for providing heating, cooling or electricity in large buildings. Based on the Slowpoke research reactor, the Slowpoke energy system concept is simple. The concept and the way in which the small-scale reactor would work are discussed. The system consists of a stainless steel tank surrounded by reinforced concrete and let into the ground. The tank is full of light water which is heated to about 90 deg C by a central core of 2.4 percent enriched uranium fuel. The resulting natural circulation causes the water to pass through a heat exchanger. The water from the heat exchanger can be used for building or district heating, to operate air-conditioners or to generate small quantities of electricity. It is hoped to automate the operation of the reactor so that continuous supervision by a team of engineers would be unnecessary. A single operator on call in the building would be able to take control actions if the reactor's safety system failed. (UK)

  9. Tokamak reactor studies

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

  10. Full Core Multiphysics Simulation with Offline Mesh Deformation

    Merzari, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Obabko, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Jain, Rajeev [Argonne National Lab. (ANL), Argonne, IL (United States); Mahadevan, Vijay [Argonne National Lab. (ANL), Argonne, IL (United States); Solberg, Jerome [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ferencz, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Whitesides, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-12-21

    In this report, building on previous reports issued in FY13 we describe our continued efforts to integrate thermal/hydraulics, neutronics, and structural mechanics modeling codes to perform coupled analysis of a representative fast sodium-cooled reactor core. The focus of the present report is a full core simulation with off-line mesh deformation.

  11. Fast reactors and nonproliferation

    Orlov, V.V.

    1997-01-01

    1.Three aspects of nonproliferation relevant to nuclear power are: Pu buildup in NPP spent fuel cooling ponds (∼ 104 t in case of consumption of ∼ 107 t cheap uranium). Danger of illegal radiochemical extraction of Pu for weapons production; Pu extraction from NPP fuel at the plants available in nuclear countries, its burning along with weapon-grade Pu in NPP reactors or in special-purpose burners; increased hazard of nuclear weapons sprawl with breeders and closed fuel cycle technology spreading all over the world. 2.The latter is one of major obstacles to creation of large-scale nuclear power. 3.Nuclear power of the first stage using 235 U will be able to meet the demands of certain fuel-deficient countries and regions, replacing ∼ 5-10% of conventional fuels in the global consumption for a number of decades. 4.Fast reactors of the first generation and the currently employed fuel technology are far from exhausting their potential for solving economic problems and meeting the challenges of safety, radioactive waste and nonproliferation. Development of large-scale nuclear power will become an option accepted by society for solving energy problems in the following century, provided a breeder technology is elaborated and demonstrated in the next 15-20 years, which would comply with the totality of the following requirement: full internal Pu breeding deterministic elimination of severe accidents involving fuel damage and high radioactivity releases: fast runaway, loss of coolant, fires, steam and hydrogen explosions, etc.; reaching a balance between radioactive wastes disposed of and uranium mined in terms of radiation hazard; technology of closed fuel cycle preventing its use for Pu extraction and permitting physical protection from fuel thefts;economic competitiveness of nuclear power for most of countries and regions, i.e. primarily the cost of NPPs with fat reactors is to be below the cost of modern LWR plants, etc

  12. Full-scope training simulators

    Ugedo, E.

    1986-01-01

    The following topics to be covered in this report are: Reasons justifying the use of full-scope simulators for operator qualification. Full-scope simulator description: the control room, the physical models, the computer complex, the instructor's console. Main features of full-scope simulators. Merits of simulator training. The role of full-scope simulators in the training programs. The process of ordering and acquiring a full-scope simulator. Maintaining and updating simulator capabilities. (orig./GL)

  13. Control of SHARON reactor for autotrophic nitrogen removal in two-reactor configuration

    Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

    2012-01-01

    With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work explores the control design for a SHARON reactor. With this aim, a full model is developed, including the pH dependency, in order to simulate the reactor and determine the optimal operating...

  14. Reactor Physics Programme

    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

  15. Reactor Physics Programme

    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.

  16. Full instantaneous traversal rupture of the primary loop pipeline

    Baytelesov, S.A.; Kungurov, F.R.

    2010-01-01

    Accident, reflecting full immediate cross rupture of primary loop pipe of WWR-SM research reactor of INP AS RUz is observed in this paper. Calculations for accident situation and analysis for different reactor cores, formed from fully IRT-3M type high enriched fuel (36% enrichment on 235 U), first mixed core, compiled from 16 IRT-3M fuel assemblies and 4 IRT-4M type fuel assemblies with low enriched fuel (19,7% enrichment on 235 U) and the core fully formed from low enriched fuel are carried out

  17. Instrumentation and control strategies for an integral pressurized water reactor

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  18. Development of a system model for advanced small modular reactors.

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  19. Power generation costs for alternate reactor fuel cycles

    Smolen, G.R.; Delene, J.G.

    1980-09-01

    The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U 3 O 8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

  20. First preliminary design of an experimental fusion reactor

    1977-09-01

    A preliminary design of a tokamak experimental fusion reactor to be built in the near future is under way. The goals of the reactor are to achieve reactor-level plasma conditions for a sufficiently long operation period and to obtain design, construction and operational experience for the main components of full-scale power reactors. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics, shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel circulating system, reactor cooling system, tritium recovery system and maintenance scheme. The main design parameters are as follows: the reactor fusion power 100 MW, torus radius 6.75 m, plasma radius 1.5 m, first wall radius 1.75 m, toroidal magnet field on axis 6 T, blanket fertile material Li 2 O, coolant He, structural material 316SS and tritium breeding ratio 0.9. (auth.)

  1. Reactor assessments of advanced bumpy torus configurations

    Uckan, N.A.; Owen, L.W.; Spong, D.A.; Miller, R.L.; Ard, W.B.; Pipkins, J.F.; Schmitt, R.J.

    1984-02-01

    Recently, several innovative approaches were introduced for enhancing the performance of the basic ELMO Bumpy Torus (EBT) concept and for improving its reactor potential. These include planar racetrack and square geometries, Andreoletti coil systems, and bumpy torus-stellarator hybrids (which include twisted racetrack and helical axis stellarator - snakey torus). Preliminary evaluations of reactor implications of each approach have been carried out based on magnetics (vacuum) calculations, transport and scaling relationships, and stability properties deduced from provisional configurations that implement the approach but are not necessarily optimized. Further optimization is needed in all cases to evaluate the full potential of each approach. Results of these studies indicate favorable reactor projections with a significant reduction in reactor physical size as compared to conventional EBT reactor designs carried out in the past

  2. Undergraduate reactor control experiment

    Edwards, R.M.; Power, M.A.; Bryan, M.

    1992-01-01

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  3. Reactor System Design

    Chi, S. K.; Kim, G. K.; Yeo, J. W.

    2006-08-01

    SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

  4. Fusion reactor design studies

    Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

    1990-01-01

    This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources

  5. Nuclear reactors; graphical symbols

    1987-11-01

    This standard contains graphical symbols that reveal the type of nuclear reactor and is used to design graphical and technical presentations. Distinguishing features for nuclear reactors are laid down in graphical symbols. (orig.) [de

  6. Control for nuclear reactor

    Ash, E.B.; Bernath, L.; Facha, J.V.

    1980-01-01

    A nuclear reactor is provided with several hydraulically-supported spherical bodies having a high neutron absorption cross section, which fall by gravity into the core region of the reactor when the flow of supporting fluid is shut off. (auth)

  7. Hybrid plasmachemical reactor

    Lelevkin, V. M., E-mail: lelevkin44@mail.ru; Smirnova, Yu. G.; Tokarev, A. V. [Kyrgyz-Russian Slavic University (Kyrgyzstan)

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  8. Ship propulsion reactors technology

    Fribourg, Ch.

    2002-01-01

    This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

  9. Guidebook to nuclear reactors

    Nero, A.V. Jr.

    1976-05-01

    A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen

  10. continuous stirred tank reactor (CSTR)

    AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search ... stirred tank reactor (CSTR) and the small and large intestines as plug flow reactor (PFR) ... from the two equations are used for the reactor sizing of the modeled reactors.

  11. Nuclear reactor instrumentation

    Duncombe, E.; McGonigal, G.

    1976-01-01

    Reference is made to the instrumentation of liquid metal cooled fast reactors. In order to ensure the safe operation of such reactors it is necessary to constantly monitor the coolant flowing through the fuel assemblies for temperature and rate of flow, requiring a large number of sensors. An improved and simplified arrangement is claimed in which the fuel assemblies feed a fraction of coolant to three instrument units arranged to sense the temperature and rate of flow of samples of coolant. Each instrument unit comprises a sleeve housing a sensing unit and has a number of inlet ducts arranged for receiving coolant from a fuel assembly together with a single outlet. The sensing unit has three thermocouple hot junctions connected in series, the hot junctions and inlet ducts being arranged in pairs. Electromagnetic windings around an inductive core are arranged to sense variation in flow of liquid metal by flux distortion. Fission product sensing means may also be provided. Full constructional details are given. (U.K.)

  12. Nuclear reactors safety issues

    Barre, Francois; Seiler, Nathalie

    2008-01-01

    Full text of publication follows: Since the seventies, economic incentives have led the utilities to drive a permanent evolution of the light water reactor (LWR). The evolution deals with the reactor designs as well as the way to operate them in a more flexible manner. It is for instance related to the fuel technologies and management. On the one hand, the technologies are in continuous evolution, such as the fuel pellets (MOX, Gd fuel, or Cr doped fuels..) as well as advanced cladding materials (M5 TM , MDA or ZIRLO). On the other hand, the fuel management is also subject to continuous evolution in particular in terms of increasing the level of burn-up, the reactor (core) power, the enrichment, as well as the duration of reactor cycles. For instance, in a few years in France, the burn-up has raised beyond the value of 39 GWj/t, initially authorized up to 52 GWj/t for the UO 2 fuel. In the near future, utilities foreseen to reach fuel burn-up of 60 GWj/t for MOX fuel and 70 GWj/t for UO 2 fuel. Furthermore, the future reactor of fourth generation will use new fuels of advanced conception. Furthermore with the objective of improving the safety margins, methods and calculation tools used by the utilities in the elaboration of their safety demonstrations submitted to the Safety Authority, are in movement. The margin evaluation methodologies often consist of a calculation chain of best-estimate multi-field simulations (e.g. various codes being coupled to simulate in a realistic way the evolution of the thermohydraulic, neutronic and mechanic state of the reactor). The statistical methods are more and more sophisticated and the computer codes are integrating ever-complex physical models (e.g. three-dimensional at fine scale). Following this evolution, the Institute of Radioprotection and Nuclear Safety (IRSN), whose one of the roles is to examine the safety records and to rend a technical expertise, considers the necessity of reevaluating the safety issues for advanced

  13. Reactor physics aspects of CANDU reactors

    Critoph, E.

    1980-01-01

    These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

  14. Utilization of nuclear research reactors

    1980-01-01

    Full text: Report on an IAEA interregional training course, Budapest, Hungary, 5-30 November 1979. The course was attended by 19 participants from 16 Member States. Among the 28 training courses which the International Atomic Energy Agency organized within its 1979 programme of technical assistance was the Interregional Training Course on the Utilization of Nuclear Research Reactors. This course was held at the Nuclear Training Reactor (a low-power pool-type reactor) of the Technical University, Budapest, Hungary, from 5 to 30 November 1979 and it was complemented by a one-week Study Tour to the Nuclear Research Centre in Rossendorf near Dresden, German Democratic Republic. The training course was very successful, with 19 participants attending from 16 Member States - Bangladesh, Bolivia, Czechoslovakia, Ecuador, Egypt, India, Iraq, Korean Democratic People's Republic, Morocco, Peru, Philippines, Spain, Thailand, Turkey, Vietnam and Yugoslavia. Selected invited lecturers were recruited from the USA and Finland, as well as local scientists from Hungarian institutions. During the past two decades or so, many research reactors have been put into operation around the world, and the demand for well qualified personnel to run and fully utilize these facilities has increased accordingly. Several developing countries have already acquired small- and medium-size research reactors mainly for isotope production, research in various fields, and training, while others are presently at different stages of planning and installation. Through different sources of information, such as requests to the IAEA for fellowship awards and experts, it became apparent that many research reactors and their associated facilities are not being utilized to their full potential in many of the developing countries. One reason for this is the lack of a sufficient number of trained professionals who are well acquainted with all the capabilities that a research reactor can offer, both in research and

  15. Distribution of E/N and N sub e in a cross-flow electric discharge laser

    Dunning, J. W., Jr.; Lancashire, R. B.; Manista, E. J.

    1976-01-01

    The spatial distribution of the ratio of electric field to neutral gas density on a flowing gas, multiple pin-to-plane discharge was measured in a high-power, closed loop laser. The laser was operated at a pressure of 140 torr (1:7:20, CO2, N2, He) with typically a 100 meter/second velocity in the 5 x 8 x 135 centimeter discharge volume. E/N ratios ranged from 2.7 x 10 to the minus 16th power to 1.4 x 10 to the minus 16th power volts/cu cm along the discharge while the electron density ranged from 2.8 x 10 to the 10th power to 1.2 x 10 to the 10th power cm/3.

  16. The charge form factor of the neutron from sup 2 H-vector, (e-vector, e' n)p

    Passchier, I; Szczerba, D; Alarcon, R; Bauer, T S; Boersma, D J; Van der Brand, J F J; Bulten, H J; Ferro-Luzzi, M; Higinbotham, D W; Jager, C W D; Klous, S; Kolster, H; Lang, J; Nikolenko, D M; Nooren, G J; Norum, B E; Poolman, H R; Rachek, Igor A; Simani, M C; Six, E; Vries, H D; Wang, K; Zhou, Z L

    2000-01-01

    We report on the first measurement of spin-correlation parameters in quasifree electron scattering from vector-polarized deuterium. Polarized electrons were injected into an electron storage ring at a beam energy of 720 MeV. A Siberian snake was employed to preserve longitudinal polarization at the interaction point. Vector-polarized deuterium was produced by an atomic beam source and injected into an open-ended cylindrical cell, internal to the electron storage ring. The spin correlation parameter A sup V sub e sub d was measured for the reaction sup 2 H-vector, (e-vector, e'n)p at a four-momentum transfer squared of 0.21 (GeV/c) sup 2 from which a value for the charge form factor of the neutron was extracted.

  17. Measurement of the Electric Form Factor of the Neutron through d-vector(e-vector, e'n)p

    Hongguo Zhu; Abdellah Ahmidouch; Anklin, H.; Arenhoevel, H.; Chris Armstrong; C. Bernet; Werner Boeglin; Herbert Breuer; Paul Brindza; D. Brown; S. Bueltmann; Roger Carlini; Nicholas Chant; Anthony Cowley; Donald Crabb; Samuel Danagoulian; D. B. Day; Tom Eden; Rolf Ent; Yusuf Farah; Renee Fatimi; Kenneth Garrow; Chris Harris; Markus Hauger; A. Honegger; J. Jourdan; M. Kaufmann; Mahbub Khandaker; Gunther Kubon; Jechiel Lichtenstadt; Richard Lindgren; Robert Lourie; Allison Lung; David Mack; Sudhir Malik; Pete Markowitz; Kenneth McFarlane; Paul McKee; Dustin McNulty; Geoffrey Milanovich; Joseph Mitchell; Hamlet Mkrtchyan; M. Muehlbauer; Thomas Petitjean; Yelena Prok; Daniela Rohe; Emmanuel Rollinde; Oscar Rondon-Aramayo; Philip Roos; Reyad Sawafta; Ingo Sick; C. Smith; Tim Southern; Michael Steinacher; Stepan Stepanyan; Vardan Tadevosyan; Raphael Tieulent; Al Tobias; Bill Vulcan; Glen Warren; H. Woehrle; Stephen Wood; Chen Yan; Marko Zeier; Jianguo Zhao; Beni Zihlmann

    2001-01-01

    We report the first measurement using a solid polarized target of the neutron electric form factor G n E via d-vector(e-vector, e'n)p. G n E was determined from the beam-target asymmetry in the scattering of longitudinally polarized electrons from polarized deuterated ammonia ( 15 ND 3 ). The measurement was performed in Hall C at Thomas Jefferson National Accelerator Facility in quasifree kinematics with the target polarization perpendicular to the momentum transfer. The electrons were detected in a magnetic spectrometer in coincidence with neutrons in a large solid angle segmented detector. We find G n E = 0.04632±0.00616(stat)±0.00341(syst) at Q 2 = 0.495 (GeV/c) 2

  18. Inclusion of electron correlation for the target wave function in low- to intermediate-energy e-N2 scattering

    Weatherford, C. A.; Brown, F. B.; Temkin, A.

    1987-01-01

    In a recent calculation, an exact exchange method was developed for use in the partial-differential-equation approach to electron-molecule scattering and was applied to e-N2 scattering in the fixed-nuclei approximation with an adiabatic polarization potential at low energies (0-10 eV). Integrated elastic cross sections were calculated and found to be lower than experiment at energies both below and above the Pi(g) resonance. It was speculated at that time that improved experimental agreement could be obtained if a correlated target representation were used in place of the uncorrelated one. The present paper implements this suggestion and demonstrates the improved agreement. These calculations are also extended to higher energies (0-30 eV) so asd to include the Sigma(u) resonance. Some discrepancies among the experiments and between experiment and the various calculations at very low energy are noted.

  19. P300 com estímulo verbal e não verbal em adultos normo-ouvintes

    Massa,Camila Gonçalves Polo; Rabelo,Camila Maia; Matas,Carla Gentile; Schochat,Eliane; Samelli,Alessandra Giannella

    2011-01-01

    O P300 é resultado da focalização da atenção a estímulos raros, dentre outros estímulos frequentes, e investiga a atenção e a memória recente, ambos dependentes da discriminação entre os estímulos, quer sejam verbais ou não verbais. OBJETIVO: Comparar o P300 com estímulos verbais e não verbais em indivíduos adultos com audição normal. MATERIAL E MÉTODO: Estudo prospectivo, no qual foram avaliados 15 sujeitos, com idades entre 22 e 55 anos, do gênero masculino, sem queixas auditivas. Foram sub...

  20. Molten salt breeder reactor

    1977-01-01

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

  1. Zero energy reactor 'RB'

    Popovic, D; Takac, S; Markovic, H; Raisic, N; Zdravkovic, Z; Radanovic, Lj [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1959-03-15

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  2. Reactor utilization, Annex A

    Martinc, R.; Stanic, A.

    1984-01-01

    Reactor was operated until August 1984 due to prohibition issued by the Ministry since the reactor does not have the emergency cooling system nor special filters in the ventilation system yet. This means that the operation plan was fulfilled by 69%. This annex includes detailed tables containing data about utilization of reactor experimental channels, irradiated samples, as well as interruptions of operation. Detailed data about reactor power during this period are shown as well

  3. PUSPATI TRIGA Reactor

    Masood, Z.

    2016-01-01

    The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

  4. Nuclear reactor shutdown system

    Mangus, J.D.; Cooper, M.H.

    1982-01-01

    An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

  5. Full Employment in Industrialized Countries.

    Britton, Andrew

    1997-01-01

    Argues that full employment must be acceptable on both social and economic grounds. Examines profound changes in industrialized economies since the 1970s and the diversity of employment contracts. Suggests that difficult policy decisions surround full employment. (SK)

  6. The modification of the Rossendorf Research Reactor

    Gehre, G.; Hieronymus, W.; Kampf, T.; Ringel, V.; Robbander, W.

    1990-01-01

    The Rossendorf Research Reactor is of the WWR-SM type. It is a heterogeneous water moderated and cooled tank reactor with a thermal power of 10 MW, which was in operation from 1957 to 1986. It was shut down in 1987 for comprehensive modifications to increase its safety and to improve the efficiency of irradiation and experimentals. The modifications will be implemented in two steps. The first one to be finished in 1989 comprises: 1) the replacement of the reactor tank and its components, the reactor cooling system, the ventilation system and the electric power installation; 2) the construction of a new reactor control room and of filtering equipment; 3) the renewal of process instrumentation and control engineering equipment for reactor operation, equipment for radiation protection monitoring, and reactor operation and safety documentation. The second step, to be implemented in the nineties, is to comprise: 1) the enlargement of the capacity for storage of spent fuel; 2) the modernization of reactor operations by computer-aided control; 3) the installation of an automated measuring systems for accident and environmental monitoring. Two objects of the modification, the replacement of the reactor tank and the design of a new and safer one as well as the increase of the redundancy of the core emergency cooling system are described in detail. For the tank replacement the exposure data are also given. Furthermore, the licensing procedures based on national ordinances and standards as well as on international standards and recommendations and the mutual responsibilities and activities of the licensing authority and of the reactor manager are presented. Finally, the present state of the modifications and the schedule up to the reactor recommissioning and test operation at full power is outlined

  7. Insatisfação corporal de adolescentes atletas e não atletas de ginástica artística

    Clara Mockdece Neves

    2016-03-01

    Full Text Available DOI: http://dx.doi.org/10.5007/1980-0037.2016v18n1p82   A avaliação de fatores psicológicos pode auxiliar na prevenção de transtornos alimentares em atletas. Objetivou-se verificar possíveis relações entre a insatisfação corporal e influência da mídia, perfeccionismo, estado de humor e comportamento de risco para transtorno alimentar, bem como comparar as variáveis de estudo entre praticantes de ginástica artística de base e de alto rendimento e não atletas. Participaram da pesquisa 413 adolescentes de ambos os sexos (atletas de alto rendimento e de base e não atletas, com idade entre 10 e 18 (média de 12,86±1,80 anos. Foram aplicados: BSQ, EAT-26, SATAQ-3, MPS e BRUMS para avaliar insatisfação corporal, comportamento de risco para transtorno alimentar, influência midiática, perfeccionismo e estado de humor, respectivamente. Foram coletados dados antropométricos para estimar o Índice de Massa Corporal (IMC e o Percentual de Gordura (%G. Foram realizadas análises estatísticas descritivas, comparativas e correlacionais. Somente o EAT-26 influenciou o BSQ nos três grupos avaliados. Em não atletas, IMC, BRUMS e SATAQ-3 acrescentaram explicação ao modelo. Para os atletas de base, %G, influência da mídia e estado de humor também foram preditores para a insatisfação corporal. Os atletas de alto rendimento tiveram menores valores de IMC e %G do que os demais. Os não atletas apresentaram menores escores para o perfeccionismo, enquanto os atletas de base menores pontuações para o estado de humor. Conclui-se que o comportamento de risco para transtorno alimentar é o fator que se relaciona mais fortemente com a insatisfação corporal em atletas de base, de alto rendimento e em não atletas.

  8. Atividade basal de acetilcolinesterase e níveis plasmáticos de serotonina não se associam ao delirium em pacientes gravemente enfermos

    Cristiane Damiani Tomasi

    2015-06-01

    Full Text Available RESUMO Objetivo: Investigar se os níveis plasmáticos de serotonina e atividade de acetilcolinesterase determinados por ocasião da admissão à unidade de terapia intensiva preveem a ocorrência de disfunção cerebral aguda em pacientes internados em unidade de terapia intensiva. Métodos: Foi conduzido no período entre maio de 2009 e setembro de 2010 um estudo prospectivo de coorte em uma amostra com 77 pacientes não consecutivos. A ocorrência de delirium foi determinada utilizando a ferramenta Confusion Assessment Method for the Intensive Care Unit, tendo sido determinadas as avaliações de acetilcolinesterase e serotonina em amostras de sangue coletadas até um máximo de 24 horas após admissão do paciente à unidade de terapia intensiva. Resultados: No presente estudo, 38 pacientes (49,6% desenvolveram delirium durante sua permanência na unidade de terapia intensiva. Nem os níveis de atividade de acetilcolinesterase nem os de serotonina tiveram associação independente com delirium. Não se observaram correlações significantes entre atividade de acetilcolinesterase e níveis de serotonina com o número de dias livres de delirium/coma, porém, em pacientes que desenvolveram delirium, ocorreu uma forte correlação negativa entre níveis de acetilcolinesterase e número de dias livres de delirium/coma, demonstrando que níveis mais elevados de acetilcolinesterase se associaram com menos dias de vida sem delirium e coma. Nenhuma associação foi identificada entre os biomarcadores e mortalidade. Conclusão: Nem a atividade de acetilcolinesterase nem os níveis séricos de serotonina se associaram com delirium ou disfunção cerebral aguda em pacientes gravemente enfermos. A ocorrência de sepse não modificou esse relacionamento.

  9. SUGESTÃO DE UM CRITÉRIO DE VIGILÂNCIA DE ESTADO NUTRICIONAL (C.V.E.N. PARA PRÊ-ESCOLARES

    Esther Moraes

    Full Text Available Sugestão de um critério de vigilância de estado nutricional (C. V. E. N. para pré-escolares O C. V. E. N. consiste no registro das deficiências de crescimento, com base nos conceitos de cronicidade e severidade de WATERLOW & ALLEYNE (1971 e na classificação da altura das crianças, de acordo com a orientação de MARCONDES et al. (1974; põe em evidência a possível influência do tempo e da intensidade de carência alimentar nas deficiências de crescimento apresentadas pelas crianças. Estas deficiências são expressas em números absolutos, ao invés de porcentagens ou índices. A aplicação do Critério de Vigilância de Estado Nutricio-nal mostrou que a sensibilidade deste critério para detectar deficiências de crescimento aproxima-se à sensibilidade da classificação de GOMEZ (1956 — 80% e 84,2%, respectivamente. A equivalência na determinação das deficiências de crescimento entre estes dois critérios é a seguinte: crianças que apresentaram severidade e cronicidade, e que foram classificadas como desnutridas de I e II grau, apresentaram respectivamente, deficiências médias de 0,72 ± 0,57 kg e 6,4 ± 1,38 cm e de 1,40 ± 0,73 kg e 9,6 ± 2,30 cm, significativamente diferentes, ao nível de 0,01%; crianças que apresentaram severidade, apenas, e que foram consideradas como desnutridas de I e II grau apresentaram, respectivamente, deficiências médias de 1,10 ± 0,52 kg e de 2,41 ± 0,42 kg, significativamente diferentes, ao nível de 0,01%.

  10. Mortes por suicídio: diferenças de gênero e nível socioeconômico

    Leticia Marín-León

    2003-06-01

    Full Text Available OBJETIVO: Descrever a tendência da mortalidade por suicídio e o perfil sociodemográfico, identificando diferenças de sexo e nível socioeconômico. MÉTODOS: Analisou-se a tendência das taxas brutas de suicídio, em Campinas, SP, no período 1976-2001, segundo o sexo. Para a análise sociodemográfica dos óbitos, no período de 1996-2001, foi utilizado o Banco de Dados de Óbitos de Campinas. Para a análise das diferenças socioeconômicas utilizou-se abordagem ecológica, em que as 42 áreas de abrangência das unidades básicas dos serviços de saúde foram agrupadas em quatro estratos homogêneos. Calcularam-se taxas padronizadas por idade (método direto. RESULTADOS: Comparando a outros países, a mortalidade por suicídio no município foi baixa (<5 ób/100.000 hab. A sobremortalidade masculina foi superior a 2,7 suicídios masculinos para cada suicídio feminino. Em 1980-1985 as maiores taxas foram observadas nos adultos de 55 anos e mais; já em 1997-2001 as taxas são mais elevadas nos adultos de 35-54 anos. Entre os homens, os meios mais utilizados são o enforcamento (36,4% e as armas de fogo (31,8%. Entre as mulheres predomina o envenenamento (24,2%, seguido pelas armas de fogo e enforcamento (21,2% cada; este último ocorreu predominantemente no domicílio (75,7%; já as mortes por arma de fogo e envenenamento ocorreram em maior proporção em hospitais. Diferentemente dos homicídios, os suicídios não apresentam aumento progressivo das taxas com a diminuição do nível socioeconômico. CONCLUSÕES: As taxas são baixas, oscilando com aumentos e declínios sucessivos, sem tendência continua de crescimento ou redução. Os riscos de morte por suicídio são maiores nos homens e não aumentam com a redução do nível socioeconômico.

  11. Perfil epidemiológico e nível de conhecimento de pacientes diabéticos sobre diabetes e retinopatia diabética

    Alana Ferreira Gomes Dias

    2010-10-01

    Full Text Available OBJETIVO: Conhecer o perfil epidemiológico e nível de conhecimento de pacientes diabéticos sobre diabetes e retinopatia diabética (RD. MÉTODOS: Estudo transversal com pacientes atendidos no Ambulatório de Retina e Vítreo do Hospital do Servidor Público Estadual de São Paulo, os quais foram divididos em dois grupos: pacientes diabéticos encaminhados para primeira avaliação oftalmológica (G1 e pacientes já acompanhados no Ambulatório (G2. Os pacientes responderam questionário e submeteram-se a exame oftalmológico. Foram utilizados os testes x², exato de Fisher e não paramétricos de Mann-Whitney, presumindo nível de significância de 5%. RESULTADOS: A amostra total foi composta por 357 pacientes (109 no G1 e 248 no G2. A maioria dos pacientes era do sexo feminino, casada, com ensino fundamental incompleto, com média de idade de 63,3 anos e afirmou saber o que é diabetes. Entretanto, 53,2% não sabiam qual o seu tipo de diabetes. As complicações visuais do diabetes são as mais conhecidas. Menos de um terço dos pacientes já tinha ouvido falar em RD e 77,3% não sabiam se eram acometidos. A maioria dos pacientes nunca havia recebido alguma explicação ou algum tipo de material escrito sobre diabetes ou RD. Somente 3,6% dos pacientes participavam de algum programa de educação sobre diabetes. A AV média na amostra, em logMAR, foi de 0,57 no OD e 0,51 no OE. Metade dos pacientes não tinha RD. CONCLUSÃO: A maioria dos pacientes, apesar de receber acompanhamento multidisciplinar, apresentou pouco conhecimento sobre o diabetes e suas complicações.

  12. E agora ... de quem, cuidarei?”: o cuidar na percepção de idosas institucionalizadas e não institucionalizadas

    Camila Vianna Duarte

    Full Text Available O perfil demográfico brasileiro tem mudado as características do envelhecimento. Este trabalho objetiva investigar a percepção de idosas sobre o cuidado consigo e com o outro, comparando-a com suas necessidades atuais de cuidados. Participaram dez mulheres, entre 65 e 89 anos, subdivididas em: institucionalizadas e não-institucionalizadas. Realizaram-se entrevistas semi-estruturadas, analisadas qualitativamente. Observou-se que o papel de “cuidadora” cria um vácuo na concepção dominante quando essas mulheres acedem à condição de objeto do cuidado do outro. Entretanto, parecem assumir um maior comprometimento com o autocuidado. Conclui-se que a família, programas comunitários e religião são as principais redes de suporte psicossocial. A ocupação e a sociabilidade influem na percepção do cuidado.

  13. Avaliação de antibiose e não-preferência em cultivares de milho selecionados com resistência à lagarta-do-cartucho

    VIANA PAULO AFONSO

    2000-01-01

    Full Text Available Dentre os fatores que contribuem para reduzir a produtividade das lavouras de milho estão as pragas, destacando-se a lagarta-do-cartucho, Spodoptera frugiperda, a qual pode reduzir a produção em até 34%. Um dos métodos de controle de pragas, cujo custo é reduzido e não causa efeitos indesejáveis ao ambiente, é o desenvolvimento de cultivares resistentes. Diversas fontes de resistência foram identificadas, linhagens desenvolvidas, registradas e liberadas ao público; entretanto, pouca informação tem sido veiculada a respeito dos mecanismos e causas dessa resistência. O objetivo do presente trabalho foi estudar os mecanismos de resistência, não-preferência e antibiose em genótipos de milho selecionados, com resistência a S. frugiperda. Os ensaios foram conduzidos em laboratório e em casa de vegetação da EMBRAPA/CNPMS. O genótipo CMS 14C apresentou antibiose influindo negativamente na biologia de S. frugiperda. 'CMS 24' e 'CMS 23' apresentaram também esse mecanismo, porém em menor intensidade. Foi constatada não-preferência alimentar das lagartas para 'Zapalote Chico' e 'BR 201'. Não-preferência para oviposição foi constatada em 'CMS 14C' e 'Zapalote Chico'.

  14. Qualidade de vida, condições e autopercepçãoda saúde entre idosos hipertensos e não hipertensos

    Dóris Firmino Rabelo

    2011-03-01

    Full Text Available O objetivo foi investigar o estilo de vida, as condições e a auto-percepção da saúde e da qualidade de vida entre idosos hipertensos e não hipertensos do município de Patos de Minas (MG. Participaram 364 idosos, sendo 54,7% hipertensos. Utilizaram-se: anamnese (histórico de saúde, engajamento social, estilo de vida e saúde percebida; Escala de Lawton; Mini-Exame do Estado Mental; Escala de Qualidade de Vida de Flanagan. Foram feitas estatísticas descritivas, comparativas de grupos, correlação e regressão linear. Os resultados indicaram que a qualidade de vida foi influenciada pelas atividades física e religiosa, e pela saúde percebida, comparada a contemporâneos. A saúde percebida hoje foi influenciada pelas mesmas variáveis e pela hipertensão arterial. Considerou-se que a continuidade de estudos que investiguem a qualidade de vida e a saúde percebida é necessária para a consolidação de modelos teóricos do desenvolvimento psicossocial de idosos saudáveis ou não.

  15. Equilíbrio corporal em crianças e adolescentes asmáticos e não asmáticos

    Marta Cristina Rodrigues da Silva

    2013-06-01

    Full Text Available O objetivo foi analisar e comparar o equilíbrio corporal em crianças e adolescentes asmáticos e não asmáticos. Fizeram parte do grupo de estudos 24 sujeitos com idades de 7 a 14 anos divididos em dois grupos: grupo asmático e grupo controle. Para avaliação do equilíbrio corporal utilizou-se uma plataforma de força. Foram utilizadas as condições, olhos abertos e fechados com três tentativas aleatórias, com duração de 30 segundos cada uma. Os resultados apontaram diferença significativa entre os grupos, no teste de equilíbrio com olhos abertos apresentando maior amplitude de deslocamento na direção ântero-posterior (COPap (p = 0,04, e médio lateral (COPml (p = 0,02 no grupo asmático. Enquanto que no teste com olhos fechados a diferença foi significante apenas na amplitude de deslocamento ântero-posterior (COPap (p = 0,02 e Área de Elipse (p=0,03. Desse modo, a asma com suas limitações e consequências parece influenciar negativamente no equilíbrio corporal de seus portadores quando comparados com crianças sem a patologia e da mesma faixa etária.

  16. Bis{(E-3-[2-(hydroxyiminopropanamido]-2,2-dimethylpropan-1-aminium} bis[μ-(E-N-(3-amino-2,2-dimethylpropyl-2-(hydroxyiminopropanamido(2−]bis{[(E-N-(3-amino-2,2-dimethylpropyl-2-(hydroxyiminopropanamide]copper(II} bis((E-{3-[2-(hydroxyiminopropanamido]-2,2-dimethylpropyl}carbamate acetonitrile disolvate

    Andrii I. Buvailo

    2012-12-01

    Full Text Available The reaction between copper(II nitrate and (E-N-(3-amino-2,2-dimethylpropyl-2-(hydroxyiminopropanamide led to the formation of the dinuclear centrosymmetric copper(II title complex, (C8H18N3O22[Cu2(C8H15N3O22(C8H17N3O22](C9H16N3O42·2CH3CN, in which an inversion center is located at the midpoint of the Cu2 unit in the center of the neutral [Cu2(C8H15N3O22(C8H17N3O22] complex fragment. The Cu2+ ions are connected by two N—O bridging groups [Cu...Cu separation = 4.0608 (5 Å] while the CuII ions are five-coordinated in a square-pyramidal N4O coordination environment. The complex molecule co-crystallizes with two molecules of acetonitrile, two molecules of the protonated ligand (E-3-[2-(hydroxyiminopropanamido]-2,2-dimethylpropan-1-aminium and two negatively charged (E-{3-[2-(hydroxyiminopropanamido]-2,2-dimethylpropyl}carbamate anions, which were probably formed as a result of condensation between (E-N-(3-amino-2,2-dimethylpropyl-2-(hydroxyiminopropanamide and hydrogencarbonate anions. In the crystal, the complex fragment [Cu2(C8H15N3O22(C8H17N3O22] and the ion pair C8H18N3O2+.C9H16N3O4− are connected via an extended system of hydrogen bonds.

  17. Saúde e mercado de trabalho no Brasil: diferenciais entre ocupados agrícolas e não agrícolas

    Alexandre Gori Maia

    2010-12-01

    Full Text Available Este trabalho analisa as diferenças no estado de saúde entre e dentro dos grupos de ocupados agrícolas e não agrícolas no Brasil, utilizando medidas de autoavaliação do estado de saúde captadas no suplemento da PNAD (Pesquisa Nacional por Amostra de Domicílios de 2008. As análises baseiam-se na composição socioeconômica dos ocupados agrícolas e não agrícolas e nas estimativas de equações simultâneas para captar as relações não lineares entre saúde, renda e jornada de trabalho. Um dos pressupostos do trabalho é que a baixa prevalência de saudáveis entre os ocupados agrícolas estaria associada, sobretudo, à composição socioeconômica desse grupo, e não à maior insalubridade a que os ocupados das atividades agrícolas estariam submetidos em condições socioeconômicas semelhantes às dos não agrícolas. Analogamente, a desigualdade no estado de saúde seria menor entre os ocupados agrícolas devido aos hábitos mais homogêneos, não só no que se refere ao tipo de atividade, mas também em relação à qualidade de vida e ao acesso a uma série de itens de consumo, típico das localidades menos desenvolvidas onde as atividades agrícolas prevalecem.This paper analyzes differences of health conditions between and within agricultural and non-agricultural workers in Brazil. Self-reported health measures of the Health Supplement of Pesquisa Nacional por Amostra de Domicílios (PNAD, from 2008, are used. Results are based on the description of the socioeconomic characteristics of each group of employees and on estimates of a simultaneous equation to measure non-linear relations between health conditions, income and work hours. The main hypothesis is that the lower prevalence of workers with good health among agricultural workers is mainly due to socioeconomic characteristics of this group and not necessarily to higher insalubrious conditions that such workers would be submitted to in equivalent socioeconomic conditions

  18. The fast reactor

    1980-02-01

    The subject is discussed as follows: brief description of fast reactors; advantage in conserving uranium resources; experience, in UK and elsewhere, in fast reactor design, construction and operation; safety; production of plutonium, security aspects; consideration of future UK fast reactor programme. (U.K.)

  19. Mirror fusion reactor design

    Neef, W.S. Jr.; Carlson, G.A.

    1979-01-01

    Recent conceptual reactor designs based on mirror confinement are described. Four components of mirror reactors for which materials considerations and structural mechanics analysis must play an important role in successful design are discussed. The reactor components are: (a) first-wall and thermal conversion blanket, (b) superconducting magnets and their force restraining structure, (c) neutral beam injectors, and (d) plasma direct energy converters

  20. Towards nuclear fusion reactors

    1993-11-01

    The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)

  1. Rotating reactors : a review

    Visscher, F.; Schaaf, van der J.; Nijhuis, T.A.; Schouten, J.C.

    2013-01-01

    This review-perspective paper describes the current state-of-the-art in the field of rotating reactors. The paper has a focus on rotating reactor technology with applications at lab scale, pilot scale and industrial scale. Rotating reactors are classified and discussed according to their geometry:

  2. Refuelling nuclear reactors

    Stacey, J.; Webb, J.; White, W.P.; McLaren, N.H.

    1981-01-01

    An improved nuclear reactor refuelling machine is described which can be left in the reactor vault to reduce the off-load refuelling time for the reactor. The system comprises a gripper device rangeable within a tubular chute, the gripper device being movable by a pantograph. (U.K.)

  3. Reactor power measuring device

    Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

    1993-01-01

    The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

  4. Ulysse, mentor reactor

    Bouquin, B.; Rio, I.; Safieh, J.

    1997-01-01

    On July 23, 1961, the ULYSSE reactor began its first power rise. Designed at that time to train nuclear engineering students and reactor operators, this reactor still remains an indispensable tool for nuclear teaching and a choice instrument for scientists. (author)

  5. Mechanical spectral shift reactor

    Sherwood, D.G.; Wilson, J.F.; Salton, R.B.; Fensterer, H.F.

    1981-01-01

    A mechanical spectral shift reactor comprises apparatus for inserting and withdrawing water displacer elements from the reactor core for selectively changing the water-moderator volume in the core thereby changing the reactivity of the core. The apparatus includes drivemechanisms for moving the displacer elements relative to the core and guide mechanisms for guiding the displayer rods through the reactor vessel

  6. Mechanical spectral shift reactor

    Sherwood, D.G.; Wilson, J.F.; Salton, R.B.; Fensterer, H.F.

    1982-01-01

    A mechanical spectral shift reactor comprises apparatus for inserting and withdrawing water displacer elements from the reactor core for selectively changing the water-moderator volume in the core thereby changing the reactivity of the core. The apparatus includes drive mechanisms for moving the displacer elements relative to the core and guide mechanisms for guiding the displacer rods through the reactor vessel. (author)

  7. Design of an Organic Simplified Nuclear Reactor

    Koroush Shirvan

    2016-08-01

    Full Text Available Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

  8. An automatic device for sample insertion and extraction to/from reactor irradiation facilities

    Alloni, L.; Venturelli, A.; Meloni, S.

    1990-01-01

    At the previous European Triga Users Conference in Vienna,a paper was given describing a new handling tool for irradiated samples at the L.E.N.A plant. This tool was the first part of an automatic device for the management of samples to be irradiated in the TRIGA MARK ii reactor and successively extracted and stored. So far sample insertion and extraction to/from irradiation facilities available on reactor top (central thimble,rotatory specimen rack and channel f),has been carried out manually by reactor and health-physics operators using the ''traditional'' fishing pole provided by General Atomic, thus exposing reactor personnel to ''unjustified'' radiation doses. The present paper describes the design and the operation of a new device, a ''robot''type machine,which, remotely operated, takes care of sample insertion into the different irradiation facilities,sample extraction after irradiation and connection to the storage pits already described. The extraction of irradiated sample does not require the presence of reactor personnel on the reactor top and,therefore,radiation doses are strongly reduced. All work from design to construction has been carried out by the personnel of the electronic group of the L.E.N.A plant. (orig.)

  9. Neutron behavior, reactor control, and reactor heat transfer. Volume four

    Anon.

    1986-01-01

    Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

  10. Halden reactor project

    1981-01-01

    Accomplishments realized during 1981 are summarized in this report. Reactor safety considerations continue to be the prime motivation for the Halden fuel programme. A major part of the experimental efforts deal with effects of loss-of-coolant accidents (LOCA) on fuel thermal and dimensional response. Fuel defect mechanisms and probabilities, both during safety related accident sequences and in response to operational transients, are also extensively studied. The programme encompasses current fuel designs as well as design modifications expected to improve performance. The Halden Project is expanding its work on process computer applications with emphasis on operator-process communication and operator guidance systems. This includes human factors experiments, control room layout and design, alarm handling systems and core surveillance. The principal tool being developed for performing realistic experiments in this field is the new control room with a full scope training simulator model of a nuclear power plant

  11. Aspects regarding the fuel management for PHWR nuclear reactors

    Dragusin, O.; Bobolea, A.; Voicu, A.

    2001-01-01

    Fuel management for PHWR nuclear reactors is completely different from the PWR reactors fuel management. PHWR reactor fuel loading procedures are repeated after an interval of time, as defined and specified in the project documentation, using a fuel machine that can be attached to the terminal fittings of horizontal pressure tubes while the reactor is a full power. Another aspect of fuel management policy is related to the possibility of bi-directional loading of the reactor, with the primary advantage of uniform and symmetrical characteristics. (authors)

  12. Reactor as furnace and reactor as lamp

    Goldanskii, V.I.

    1992-01-01

    There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

  13. Nuclear reactor physics

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  14. Light water reactor safety

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  15. Fundamentals of reactor chemistry

    Akatsu, Eiko

    1981-12-01

    In the Nuclear Engineering School of JAERI, many courses are presented for the people working in and around the nuclear reactors. The curricula of the courses contain also the subject material of chemistry. With reference to the foreign curricula, a plan of educational subject material of chemistry in the Nuclear Engineering School of JAERI was considered, and the fundamental part of reactor chemistry was reviewed in this report. Since the students of the Nuclear Engineering School are not chemists, the knowledge necessary in and around the nuclear reactors was emphasized in order to familiarize the students with the reactor chemistry. The teaching experience of the fundamentals of reactor chemistry is also given. (author)

  16. How Full is Full Employment? : How Tools and Not Theory Explained Full Employment

    Rodenburg, P.

    2016-01-01

    The post-war debate on full employment policy was blurred and unclear since the concept of full employment itself was theoretically unclear and un-operational. Unable to theoretically determine the unemployment level of full employment, economists tried to find more empirically based ways to

  17. A study of reactor vessel integrity assessment

    Lee, Jae Hoon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Kim, Jong Kyung; Shin, Chang Ho; Seo, Bo Kyun [Hanyang Univ., Seoul (Korea, Republic of)

    1999-02-15

    The fast neutron fluence at the Reactor Pressure Vessel(RPV) of KNGR designed for 60 years lifetime was calculated by full-scope Monte Carlo simulation for reactor vessel integrity assessment. KNGR core geometry was modeled on a three-dimensional representation of the one-sixteenth of the reactor in-vessel component. Each fuel assemblies were modeled explicitly, and each fuel pins were axially divided into 5 segments. The maximum flux of 4.3 x 10{sup 10} neutrons/cm{sup 2}. sec at the RPV was obtained by tallying neutrons crossing the beltline of inner surface of the RPV.

  18. ADVANCED CONTROL FOR A ETHYLENE REACTOR

    Dumitru POPESCU

    2017-06-01

    Full Text Available The main objective of this work is the design and implementation of control solutions for petrochemical processes, namely the control and optimization of a pyrolysis reactor, the key-installation in the petrochemical industry. We present the technological characteristics of this petrochemical process and some aspects about the proposed control system solution for the ethylene plant. Finally, an optimal operating point for the reactor is found, considering that the process has a nonlinear multi-variable structure. The results have been implemented on an assembly of pyrolysis reactors on a petrochemical platform from Romania.

  19. Generation III+ Reactor Portfolio

    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

  20. The fast breeder reactor

    Davis, D.A.; Baker, M.A.W.; Hall, R.S.

    1990-01-01

    Following submission of written evidence, the Energy Committee members asked questions of three witnesses from the Central Electricity Generating Board and Nuclear Electric (which will be the government owned company running nuclear power stations after privatisation). Both questions and answers are reported verbatim. The points raised include where the responsibility for the future fast reactor programme should lie, with government only or with private enterprise or both and the viability of fast breeder reactors in the future. The case for the fast reactor was stated as essentially strategic not economic. This raised the issue of nuclear cost which has both a construction and a decommissioning element. There was considerable discussion as to the cost of building a European Fast reactor and the cost of the electricity it would generate compared with PWR type reactors. The likely demand for fast reactors will not arrive for 20-30 years and the need to build a fast reactor now is questioned. (UK)

  1. One piece reactor removal

    Chia, Wei-Min; Wang, Song-Feng

    1993-01-01

    The strategy of Taiwan Research Reactor Renewal plan is to remove the old reactor block with One Piece Reactor Removal (OPRR) method for installing a new research reactor in original building. In this paper, the engineering design of each transportation works including the work method, the major equipments, the design policy and design criteria is described and discussed. In addition, to ensure the reactor block is safety transported for storage and to guarantee the integrity of reactor base mat is maintained for new reactor, operation safety is drawn special attention, particularly under seismic condition, to warrant safe operation of OPRR. ALARA principle and Below Regulatory Concern (BRC) practice were also incorporated in the planning to minimize the collective dose and the total amount of radioactive wastes. All these activities are introduced in this paper. (J.P.N.)

  2. Reactor power control device

    Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

    1998-01-01

    The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

  3. Reactor power monitoring device

    Dogen, Ayumi; Ozawa, Michihiro.

    1983-01-01

    Purpose: To significantly improve the working efficiency of a nuclear reactor by reflecting the control rod history effect on thermal variants required for the monitoring of the reactor operation. Constitution: An incore power distribution calculation section reads the incore neutron fluxes detected by neutron detectors disposed in the reactor to calculate the incore power distribution. A burnup degree distribution calculation section calculates the burnup degree distribution in the reactor based on the thus calculated incore power distribution. A control rod history date store device supplied with the burnup degree distribution renews the stored control rod history data based on the present control rod pattern and the burnup degree distribution. Then, thermal variants of the nuclear reactor are calculated based on the thus renewed control rod history data. Since the control rod history effect is reflected on the thermal variants required for the monitoring of the reactor operation, the working efficiency of the nuclear reactor can be improved significantly. (Seki, T.)

  4. The Maple reactor project

    Malkoske, G.R.; Labrie, J.-P.

    2003-01-01

    MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

  5. Fission reactors and materials

    Frost, B.R.T.

    1981-12-01

    The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

  6. 454 pyrosequencing analyses of bacterial and archaeal richness in 21 full-scale biogas digesters.

    Sundberg, Carina; Al-Soud, Waleed A; Larsson, Madeleine; Alm, Erik; Yekta, Sepehr S; Svensson, Bo H; Sørensen, Søren J; Karlsson, Anna

    2013-09-01

    The microbial community of 21 full-scale biogas reactors was examined using 454 pyrosequencing of 16S rRNA gene sequences. These reactors included seven (six mesophilic and one thermophilic) digesting sewage sludge (SS) and 14 (ten mesophilic and four thermophilic) codigesting (CD) various combinations of wastes from slaughterhouses, restaurants, households, etc. The pyrosequencing generated more than 160,000 sequences representing 11 phyla, 23 classes, and 95 genera of Bacteria and Archaea. The bacterial community was always both more abundant and more diverse than the archaeal community. At the phylum level, the foremost populations in the SS reactors included Actinobacteria, Proteobacteria, Chloroflexi, Spirochetes, and Euryarchaeota, while Firmicutes was the most prevalent in the CD reactors. The main bacterial class in all reactors was Clostridia. Acetoclastic methanogens were detected in the SS, but not in the CD reactors. Their absence suggests that methane formation from acetate takes place mainly via syntrophic acetate oxidation in the CD reactors. A principal component analysis of the communities at genus level revealed three clusters: SS reactors, mesophilic CD reactors (including one thermophilic CD and one SS), and thermophilic CD reactors. Thus, the microbial composition was mainly governed by the substrate differences and the process temperature. © 2013 Federation of European Microbiological Societies. Published by John Wiley & Sons Ltd. All rights reserved.

  7. A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

    Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

  8. ROSA full-core and DNBR capabilities

    Gibcus, H.P.M.; Verhagen, F.C.M.; Wakker, P.H.

    2012-01-01

    This paper presents the latest developments of the ROSA (Reloading Optimization by Simulated Annealing) code system with an emphasis on the first full-core version and the minimum DNBR (Departure from Nucleate Boiling Ratio) as a new optimization parameter. Designing the core loading pattern of nuclear power plants is becoming a more and more complex task. This task becomes even more complicated if asymmetries in the core loading pattern arise, for instance due to damaged fuel assemblies. For over almost two decades ROSA, NRG's (Nuclear Research and consultancy Group) loading pattern optimization code system for PWRs, has proven to be a valuable tool to reactor operators in accomplishing this task. To improve the use of ROSA for designing asymmetric loading patterns, NRG has developed a full-core version of ROSA besides the original quarter-core version which requires rotational symmetry in the computational domain. The extension of ROSA with DNBR as an optimization parameter is part of ROSA's continuous development. (orig.)

  9. ANALYTICAL SYNTHESIS OF CHEMICAL REACTOR CONTROL SYSTEM

    Alexander Labutin

    2017-02-01

    Full Text Available The problem of the analytical synthesis of the synergetic control system of chemical reactor for the realization of a complex series-parallel exothermal reaction has been solved. The synthesis of control principles is performed using the analytical design method of aggregated regulators. Synthesized nonlinear control system solves the problem of stabilization of the concentration of target component at the exit of reactor and also enables one to automatically transfer to new production using the equipment.

  10. Iser: an international inherently safe reactor concept

    Wakabayashi, Hiroaki

    1988-01-01

    Iser is a modular standardised 200-300 MWe power reactor based on the PIUS principle. It differs from PIUS in being simpler, and making full use of existing steel-vessel-based LWR technology. Iser is an inherently safe reactor concept under development in Japan. It is a generic concept, not a patented commodity, and it is expected that an international association to develop the concept will be formed. (U.K.)

  11. Petascale algorithms for reactor hydrodynamics

    Fischer, P.; Lottes, J.; Pointer, W.D.; Siegel, A.

    2008-01-01

    We describe recent algorithmic developments that have enabled large eddy simulations of reactor flows on up to P = 65, 000 processors on the IBM BG/P at the Argonne Leadership Computing Facility. Petascale computing is expected to play a pivotal role in the design and analysis of next-generation nuclear reactors. Argonne's SHARP project is focused on advanced reactor simulation, with a current emphasis on modeling coupled neutronics and thermal-hydraulics (TH). The TH modeling comprises a hierarchy of computational fluid dynamics approaches ranging from detailed turbulence computations, using DNS (direct numerical simulation) and LES (large eddy simulation), to full core analysis based on RANS (Reynolds-averaged Navier-Stokes) and subchannel models. Our initial study is focused on LES of sodium-cooled fast reactor cores. The aim is to leverage petascale platforms at DOE's Leadership Computing Facilities (LCFs) to provide detailed information about heat transfer within the core and to provide baseline data for less expensive RANS and subchannel models.

  12. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  13. Reactor Vessel Surveillance Program for Advanced Reactor

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  14. Research reactors in Argentina

    Carlos Ruben Calabrese

    1999-01-01

    Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

  15. Siting of research reactors

    1987-01-01

    The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

  16. Reactor core for FBR type reactor

    Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

    1996-01-01

    In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

  17. Thai research reactor

    Aramrattana, M.

    1987-01-01

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  18. Nuclear Reactor Physics

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  19. The magnetic form factor of the neutron, GMn, from the d(e,e'n)p reaction

    Markowitz, P.E.C.

    1992-01-01

    The author measured the d(e,e'n)p cross-section at three values of Q 2 : 0.255, 0.176 and 0.109 (GeV/c) 2 . The electrons were detected with the OHIPS magnetic spectrometer, and the neutrons were detected in a liquid mineral oil scintillator array. The measurement were made at a fixed neutron angle of θ n = 57 degrees; the Q 2 values were obtained by varying the incident electron energy and the scattering angle. These cross sections are sensitive primarily to the neutron magnetic form factor at these quasifree kinematics. The efficiency of the neutron detector was determined by the associated particle technique with the d(γ, pn) reaction for each of three neutron kinetic energies. The value of G n M extracted from the cross sections are consistent with the dipole parametrization at the two high momentum transfers; at the lowest momentum transfer the value of G n M is 10% higher than the dipole model. This enhancement at low momentum transfer is consistent with previous measurements

  20. Reactor containment and reactor safety in the United States

    Kouts, H.

    1986-01-01

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

  1. Developed Turbulence: From Full Simulations to Full Mode Reductions

    Grossmann, Siegfried; Lohse, Detlef; Reeh, Achim

    1996-01-01

    Developed Navier-Stokes turbulence is simulated with varying wave-vector mode reductions. The flatness and the skewness of the velocity derivative depend on the degree of mode reduction. They show a crossover towards the value of the full numerical simulation when the viscous subrange starts to be

  2. Performance of metallic fuels in liquid-metal fast reactors

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

    1984-01-01

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

  3. Nuclear reactor fuelling machine

    Peberdy, J.M.

    1976-01-01

    The refuelling machine described comprises a rotatable support structure having a guide tube attached to it by a parellel linkage mechanism, whereby the guide tube can be displaced sideways from the support structure. A gripper unit is housed within the guide tube for gripping the end of a fuel assembly or other reactor component and has means for maintenance in the engaging condition during travel of the unit along the guide tube, except for a small portion of the travel at one end of the guide tube, where the inner surface of the guide tube is shaped so as to maintain the gripper unit in a disengaging condition. The gripper unit has a rotatable head, means for moving it linearly within the guide tube so that a component carried by the unit can be housed in the guide tube, and means for rotating the head of the unit through 180 0 relative to its body, to effect rotation of a component carried by the unit. The means for rotating the head of the gripper unit comprises ring and pinion gearing, operable through a series of rotatable shafts interconnected by universal couplings. The reason for provision for 180 0 rotation is that due to the variation in the neutron flux across the reactor core the side of a fuel assembly towards the outside of the core will be subjected to a lower neutron flux and therefore will grow less than the side of the fuel assembly towards the inside of the core. This can lead to bowing and possible jamming of the fuel assemblies. Full constructional details are given. See also BP 1112384. (U.K.)

  4. E-beam heated linear solenoid reactors

    Benford, J.; Bailey, V.; Oliver, D.

    1976-01-01

    A conceptual design and system analysis shows that electron beam heated linear solenoidal reactors are attractive for near term applications which can use low gain fusion sources. Complete plant designs have been generated for fusion based breeders of fissile fuel over a wide range of component parameters (e.g., magnetic fields, reactor lengths, plasma densities) and design options (e.g., various radial and axial loss mechanisms). It appears possible that a reactor of 100 to 300 meters length operating at power levels of 1000 MWt can economically produce 2000 to 8000 kg/yr of 233 U to supply light water reactor fuel needs beyond 2000 A.D. Pure fusion reactors of 300 to 500 meter lengths are possible. Physics and operational features of reactors are described. Beam heating by classical and anomalous energy deposition is reviewed. The technology of the required beams has been developed to MJ/pulse levels, within a factor of 20 of that needed for full scale production reactors. The required repetitive pulsing appears practical

  5. Status report of Indonesian research reactor

    Arbie, B.; Supadi, S.

    1992-01-01

    A general description of three Indonesian research reactor, its irradiation facilities and its future prospect are described. Since 1965 Triga Mark II 250 KW Bandung, has been in operation and in 1972 the design powers were increased to 1000 KW. Using core grid form Triga 250 KW BATAN has designed and constructed Kartini Reactor in Yogyakarta which started its operation in 1979. Both of this Triga type reactors have served a wide spectrum of utilization such as training manpower in nuclear engineering, radiochemistry, isotope production and beam research in solid state physics. Each of this reactor have strong cooperation with Bandung Institute of Technology at Bandung and Gajah Mada University at Yogyakarta which has a faculty of Nuclear Engineering. Since 1976 the idea to have high flux reactor has been foreseen appropriate to Indonesian intention to prepare infrastructure for nuclear industry for both energy and non-energy related activities. The idea come to realization with the first criticality of RSG-GAS (Multipurpose Reactor G.A. Siwabessy) in July 1987 at PUSPIPTEK Serpong area. It is expected that by early 1992 the reactor will reached its full power of 30 MW and by end 1992 its expected that irradiation facilities will be utilized in the future for nuclear scientific and engineering work. (author)

  6. Multipurpose research reactors

    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

  7. TRIGA reactor main systems

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  8. Fast Spectrum Reactors

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  9. Upgradation of Apsara reactor

    Mammen, S.; Mukherjee, P.; Bhatnagar, A.; Sasidharan, K.; Raina, V.K.

    2009-01-01

    Apsara is a 1 MW swimming pool type research reactor using high enriched uranium as fuel with light water as coolant and moderator. The reactor is in operation for more than five decades and has been extensively used for basic research, radioisotope production, neutron radiography, detector testing, shielding experiments etc. In view of its long service period, it is planned to carry out refurbishment of the reactor to extend its useful life. During refurbishment, it is also planned to upgrade the reactor to a 2 MW reactor to improve its utilization and to upgrade the structure, system and components in line with the current safety standards. This paper gives a brief account of the design features and safety aspects of the upgraded Apsara reactor. (author)

  10. Confirmatory tests of full-scale condensers for SBWR

    Masoni, P.; Botti, S.; Fitzsimmons, G.W.

    1993-01-01

    A full-scale isolation condenser and a full-scale passive containment cooling condenser for the Simplified Boiling Water Reactor (SBWR) will be tested to confirm the thermal-hydraulic and structural design characteristics of these components. The condensers provide vital roles in removing heat from the reactor vessel and the containment during certain design basis events. This paper describes the condensers and the test facilities which are under construction and summarizes the test objectives, the planned instrumentation, and the conditions to be tested. The results of some pre-test performance predictions, calculated with the TRACG code are presented. The results of the testing program are expected to demonstrate that the condenser designs will provide the required heat removal capacity and will survive the design basis temperature/pressure cycles without structural damage

  11. Principle of human system interface (HSI) design for new reactor console of PUSPATI TRIGA Reactor (RTP)

    Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Idris Taib; Mohd Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha; Mohd Sabri Minhat; Izhar Abu Hussin

    2013-01-01

    Full-text: This paper will describe the principle of human system interface design for new reactor console in control room at TRIGA reactor facility. In order to support these human system interface challenges in digital reactor console. Software-based instrumentation and control (I and C) system for new reactor console could lead to new human machine integration. The proposed of Human System Interface (HSI) which included the large display panels which shows reactor status, compact and computer-based workstations for monitoring, control and protection function. The proposed Human System Interface (HIS) has been evaluated using various human factor engineering. It can be concluded that the Human System Interface (HIS) is designed as to address the safety related computer controlled system. (author)

  12. Reactor Engineering Division annual report

    1980-09-01

    Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

  13. Spectral shift reactor control method

    Impink, A.J. Jr.

    1981-01-01

    A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core wherein is established a reator coolant temperature set point at which it is desired to operate said reactor and first reactor coolant temperature band limits are provided within which said set point is located and it is desired to operate said reactor charactrized in that said reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in said core as said reactor coolant temperature approaches said first band limits thereby to maintain said reactor coolant temperature near said set point and within said first band limits

  14. Reactor Engineering Department annual report

    1985-08-01

    Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

  15. Reactor Engineering Division annual report

    Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

    1982-09-01

    Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

  16. Power reactors operational diagnosis

    Dach, K.; Pecinka, L.

    1976-01-01

    The definition of reactor operational diagnostics is presented and the fundamental trends of research are determined. The possible sources of power reactor malfunctions, the methods of defect detection, the data evaluation and the analysis of the results are discussed in detail. In view of scarcity of a theoretical basis and of insufficient in-core instrumentation, operational diagnostics cannot be as yet incorporated in a computer-aided reactor control system. (author)

  17. Fast reactors worldwide

    Hall, R.S.; Vignon, D.

    1985-01-01

    The paper concerns the evolution of fast reactors over the past 30 years, and their present status. Fast reactor development in different countries is described, and the present position, with emphasis on cost reduction and collaboration, is examined. The French development of the fast breeder type reactor is reviewed, and includes: the acquisition of technical skills, the search for competitive costs and the spx2 project, and more advanced designs. Future prospects are also discussed. (U.K.)

  18. Pressurised water reactor operation

    Birnie, S.; Lamonby, J.K.

    1987-01-01

    The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

  19. Reactor Safety Analysis

    Arien, B.

    1998-01-01

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

  20. Seals in nuclear reactors

    1979-01-01

    The aim of this invention is the provision of improved seals for reactor vessels in which fuel assemblies are located together with inlets and outlets for the circulation of a coolant. The object is to provide a seal arrangement for the rotatable plugs of nuclear reactor closure heads which has good sealing capacities over a wide gap during operation of the reactor but which also permits uninhibited rotation of the plugs for maintenance. (U.K.)

  1. Reactor BR2

    Gubel, P

    2000-07-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

  2. The Dragon reactor experiment

    Anon.

    1976-01-01

    The concept on which the Dragon Reactor Experiment was based was evolved at the Atomic Energy Research Establishment at Harwell in 1956, and in February of that year a High Temperature Gas- cooled Reactor Project Group was set up to study the feasibility of a helium-cooled reactor with a graphite or beryllium moderator, and with the emphasis on the thorium fuel cycle [af

  3. The replacement research reactor

    Cameron, R.

    1999-01-01

    As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described

  4. TRIGA reactor characteristics

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the general design, characteristics and parameters of TRIGA reactors and fuels. It is recommended that most of this information should be incorporated into any reactor operator training program and, in many cases, the facility Safety Analysis Report. It is oriented to teach the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and the differences between TRIGA fuels and others more traditional reactor fuels. (nevyjel)

  5. Integral fast reactor

    Chang, Y.I.

    1989-01-01

    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

  6. Reactor BR2

    Gubel, P.

    2000-01-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

  7. The Integral Fast Reactor

    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

  8. Mirror fusion reactors

    Anon.

    1978-01-01

    Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

  9. HOMOGENEOUS NUCLEAR POWER REACTOR

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  10. Australian research reactor studies

    McCulloch, D.B.

    1978-01-01

    The Australian AEC has two research reactors at the Lucas Heights Research Establishment, a 10 HW DIDO class materials testing reactor, HIFAR, and a smaller 100kW reactor MOATA, which was recently upgraded from 10kW power level. Because of the HIFAR being some 20 years old, major renewal and repair programmes are necessary to keep it operational. To enable meeting projected increases in demand for radioisotopes, plans for a new reactor to replace the HIFAR have been made and the design criteria are described in the paper. (author)

  11. Mirror fusion reactors

    Carlson, G.A.; Moir, R.W.

    1978-01-01

    We have carried out conceptual design studies of fusion reactors based on the three current mirror confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fission fuel for fission reactors. We have designed a large commercial hybrid based on standard mirror confinement, and also a small pilot plant hybrid. Tandem mirror designs include a commercial 1000 MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single cell pilot plant

  12. Reactor feedwater system

    Hikabe, Katsumi.

    1978-01-01

    Purpose: In order to prevent thermal stresses of a core of PWR type reactor, described has been a method for feeding heated recirculating water to the core in the case of the reactor start-up or shut-down. Constitution: A recirculating water is degassed, cleaned up and heated in the steam condensers, and then feeds the water to the reactor, characterized in that heaters are provided in the bypasses of the turbine, so that heated water is constantly supplied to the reactor. (Nakamura, S.)

  13. Remote Reactor Monitoring

    Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-10-21

    The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

  14. Fusion Reactor Materials

    Decreton, M.

    2002-01-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

  15. Trench reactor: an overview

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.; Sankoorikal, J.T.; Schmidt, R.S.; Lofshult, J.; Ramin, T.; Sokmen, N.; Lin, L.C.

    1988-01-01

    Recent fast, sodium-cooled reactor designs reflect new conditions. In nuclear energy these conditions are (a) emphasis on maintainability and operability, (b) design for more transparent safety, and (c) a surplus of uranium and enrichment availability that eases concerns about light water reactor fueling costs. In utility practice the demand is for less capital exposure, short construction time, smaller new unit sizes, and low capital cost. The PRISM, SAFR, and integral fast reactor (IFR) concepts are responses to these conditions. Fast reactors will not soon be deployed commercially, so more radical designs can be considered. The trench reactor is the product of such thinking. Its concepts are intended as contributions to the literature, which may be picked up by one of the existing programs or used in a new experimental project. The trench reactor is a thin-slab, pool-type reactor operated at very low power density and- for sodium-modest temperature. The thin slab is repeated in the sodium tank and the reactor core. The low power density permits a longer than conventional core height and a large-diameter fuel pin. Control is by borated steel slabs that can be lowered between the core and lateral sodium reflector. Shutdown is by semaphore slabs that can be swung into place just outside the control slabs. The paper presents major characteristics of the trench reactor that have been changed since the last report

  16. New reactor concepts

    Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A.

    1998-11-01

    The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

  17. Advanced CANDU reactors

    Dunn, J.T.; Finlay, R.B.; Olmstead, R.A.

    1988-12-01

    AECL has undertaken the design and development of a series of advanced CANDU reactors in the 700-1150 MW(e) size range. These advanced reactor designs are the product of ongoing generic research and development programs on CANDU technology and design studies for advanced CANDU reactors. The prime objective is to create a series of advanced CANDU reactors which are cost competitive with coal-fired plants in the market for large electricity generating stations. Specific plant designs in the advanced CANDU series will be ready for project commitment in the early 1990s and will be capable of further development to remain competitive well into the next century

  18. Mechanical spectral shift reactor

    Doshi, P.K.; George, R.A.; Dollard, W.J.

    1982-01-01

    A mechanical spectral shift arrangement for controlling a nuclear reactor includes a plurality of reactor coolant displacer members which are inserted into a reactor core at the beginning of the core life to reduce the volume of reactor coolant-moderator in the core at start-up. However, as the reactivity of the core declines with fuel depletion, selected displacer members are withdrawn from the core at selected time intervals to increase core moderation at a time when fuel reactivity is declining. (author)

  19. Nuclear reactor safety systems

    Ball, R.M.; Roberts, R.C.

    1980-01-01

    A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

  20. RA Reactor; Reaktor RA

    NONE

    1978-02-15

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation. [Serbo-Croat] Pored osnovnih karakeristika reaktora RA, organizacije rada i finansijskih pokazatelja, razmatra se stanje opreme reaktora nakon 18 godina rada, pitanja dozvole za rad sa 80% obogacenim gorivom, problem skladistenja isluzenog goriva u bazenu zgrade reaktora i potreba za obnavljanjem komponenti opreme, pre svega elektronske.