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Sample records for agr type reactors

  1. CO{sub 2} direct cycles suitable for AGR type reactors; Cycles directs de gaz carbonique applicables aux reacteurs du genre AGR

    Energy Technology Data Exchange (ETDEWEB)

    Maillet, E. [Commissariat a l' Energie Atomique. Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1967-10-01

    The perspectives given by the gas turbines under pressure, to build simple nuclear power plants and acieving significantly high yield, are specified. The CO{sub 2} is characterised by by good efficiency under moderate temperature (500 to 750 Celsius degrees), compactness and the simpleness of machines and the safe exploitation (supply, storage, relief cooling, thermosyphon). The revision of thermal properties of the CO{sub 2} and loss elements show that several direct cycles would fit in particular to the AGR type reactors. Cycles that would diverge a little from classical models and able to lead to power and heat generation can lead by simple means to the best results. Several satisfying solutions present for the starting up, the power regulation and the stopping. The nuclear power plant components and the functioning safety are equally considered in the present report. The conclusions stimulate the studies and realizations of carbon dioxide gas turbines in when approprite. [French] Les perspectives offertes par la turbine a gaz sous pression, pour construire des centrales nucleaires simples et de rendement progressivement eleve, se precisent actuellement. le CO{sub 2} se distingue par sa bonne efficacite a temperature moderee (500 a 750 degres celsius), la compacite et la simplicite des machines, et la surete qu'il apporte a l'exploitation ( approvisionnement, stockage, refroidissement de secours, thermosiphon). La revision des proprietes thermophysiques du CO{sub 2} et des elements de pertes montre que divers cycles directs conviendraient en particulier aux reacteurs agr ou derives. Des cycles s'ecartant peu des modeles classiques, et se pretant ulterieurement a la production simultanee d'electricite et de chaleur, peuvent conduire par des moyens simples aux meilleurs resultats d'ensemble. Plusieurs solutions satisfaisantes se presentent pour le demarrage, le reglage de la puissance et l'arret. Les composants de la centrale et la

  2. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  3. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

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    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  4. Irradiation performance of AGR-1 high temperature reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; John D. Hunn; Robert N. Morris; Charles A. Baldwin; Philip L. Winston; Jason M. Harp; Scott A. Ploger; Tyler Gerczak; Isabella J. van Rooyen; Fred C. Montgomery; Chinthaka M. Silva

    2014-10-01

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO-coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.5% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel–including the extent of fission product release and the evolution of kernel and coating microstructures–was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocrabon and compact matrix. The capsule-average fractional release from the compacts was 1×10 4 to 5×10 4 for 154Eu and 8×10 7 to 3×10 5 for 90Sr. The average 134Cs release from compacts was <3×10 6 when all particles maintained intact SiC. An estimated four particles out of 2.98×105 experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs release in two capsules to approximately 10 5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. Palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization

  5. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru

    1997-04-04

    An LMFBR type reactor comprises a plurality of reactor cores in a reactor container. Namely, a plurality of pot containing vessels are disposed in the reactor vessel and a plurality of reactor cores are formed in a state where an integrated-type fuel assembly is each inserted to a pot, and a coolant pipeline is connected to each of the pot containing-vessel to cool the reactor core respectively. When fuels are exchanged, the integrated-type fuel assembly is taken out together with the pot from the reactor vessel in a state where the integrated-type fuel assembly is immersed in the coolants in the pot as it is. Accordingly, coolants are supplied to each of the pot containing-vessel connected with the coolant pipeline and circulate while cooling the integrated-type fuel assembly for every pot. Then, when the fuels are exchanged, the integrated type fuel assembly is taken out to the outside of the reactor together with the pot by taking up the pot from the pot-containing vessel. Then, neutron economy is improved to thereby improve reactor power and the breeding ratio. (N.H.)

  6. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Takeshi; Iida, Masaaki; Moriki, Yasuyuki

    1994-10-18

    A reactor core is divided into a plurality of coolants flowrate regions, and electromagnetic pumps exclusively used for each of the flowrate regions are disposed to distribute coolants flowrates in the reactor core. Further, the flowrate of each of the electromagnetic pumps is automatically controlled depending on signals from a temperature detector disposed at the exit of the reactor core, so that the flowrate of the region can be controlled optimally depending on the burning of reactor core fuels. Then, the electromagnetic pumps disposed for every divided region are controlled respectively, so that the coolants flowrate distribution suitable to each of the regions can be attained. Margin for fuel design is decreased, fuels are used effectively, as well as an operation efficiency can be improved. Moreover, since the electromagnetic pump has less flow resistance compared with a mechanical type pump, and flow resistance of the reactor core flowrate control mechanism is eliminated, greater circulating flowrate can be ensured after occurrence of accident in a natural convection using a buoyancy of coolants utilizable for after-heat removal as a driving force. (N.H.).

  7. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kawakami, Hiroto

    1995-02-07

    A reactor container of the present invention has a structure that the reactor container is entirely at the same temperature as that at the inlet of the reactor and, a hot pool is incorporated therein, and the reactor container has is entirely at the same temperature and has substantially uniform temperature follow-up property transiently. Namely, if the temperature at the inlet of the reactor core changes, the temperature of the entire reactor container changes following this change, but no great temperature gradient is caused in the axial direction and no great heat stresses due to axial temperature distribution is caused. Occurrence of thermal stresses caused by the axial temperature distribution can be suppressed to improve the reliability of the reactor container. In addition, since the laying of the reactor inlet pipelines over the inside of the reactor is eliminated, the reactor container is made compact and the heat shielding structures above the reactor and a protection structure of container walls are simplified. Further, secondary coolants are filled to the outside of the reactor container to simplify the shieldings. The combined effects described above can improve economical property and reliability. (N.H.).

  8. Design and Status of the NGNP Fuel Experiment AGR-3/4 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in November 2013. Since the purpose of this experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  9. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Iwashige, Kengo

    1996-06-21

    In an LMFBR type reactor, partitions are disposed to a coolant channel at positions lower than the free liquid level, and the width of the partitions is adapted to have a predetermined condition. Namely, when low temperature fluid overflowing the wall of the coolant channel, flows down and collided against the free liquid surface in the coolant channel, since the dropping speed thereof is reduced abruptly, large pressure waves are caused by kinetic force of the low temperature fluid. However, if appropriate numbers of partitions having an appropriate shape are formed, the dropping speed of the low temperature fluid is moderated to reduce the pressure waves. In addition, since the pressure waves are dispersed to the circumferential and lateral directions of the coolant flow channel respectively, the propagation of the pressure waves can be prevented effectively. Further, when the flow of the low temperature fluid is changed to the circumferential direction, for example, by earthquakes, since the partitions act as members resisting against the circumferential change of the low temperature fluid, the change of the direction can be suppressed. (N.H.)

  10. New reactor type proposed

    CERN Multimedia

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  11. The Relationship Between Antibiotic Resistance and Agr Type in Methicillin-Resistant Staphylococcus aureus (MRSA Isolated From Burn Wound of Hospitalized Patient in Tehran

    Directory of Open Access Journals (Sweden)

    Mohammad Sadegh Vaziri

    2015-12-01

    Full Text Available Introduction: Staphylococcus aureus is the major cause of hospital and community-acquired infections. This bacterium possesses an accessory gene regulator (agr that plays role in colonization, expression of virulence factors and antibiotic resistance. It’s four major polypeptide with variable sequences lead to at least four agr type in S. aureus.The aim of this study was to determine the relationship between the antibiogram patterns with agr type of clinical S. aureus.Methods: Detection of methicillin-resistant Staphylococcus aureus (MRSA from burn wounds was performed by phenotypic and genotypic profiles. The antibiotics resistance pattern was determined by disk agar diffusion (Tigecycline (TGC, Ciprofloxacin(CIP, Erythromycin(E, Cloxacillin(CX, Clindamycin(CD, Imipenem(IMI, Co-trimoxazole(SXT, Kanamycin(K, Teicoplanin(TEC, Gentamicin(GM, Mupirocin(MUP, Ceftriaxone (CTR. The agr typing by PCR-RFLP method using the Restriction endonuclease ScaI was performed and spss19 was used for data analysis.Results: The total of 76 MRSA isolates was studied. The agr type distribution was 75.6% Type I, 8.2% Type II, 5.4% Type III, 10.8% type IV. The most antibiotics resistant agr type belongs to the type I. There was no significance relationship between every agr type and antibiotics but only a statistically significant association exist between CX, E, CTR, SXT, GM, CIP antibiotics and all agr types (P<0.05.Conclusion: There was no significance relationship between every agr type and antibiotics but significant relationship observed between resistance to some antibiotics with all agr types could be related to the number and source of isolated bacteria or extra use of these antibiotics. By considering that agr locus belongs to upstream genes so it may use the Quorum Sensing (QS system to induce the most drug resistance.

  12. Performance of AGR-1 High-Temperature Reactor Fuel During Post-Irradiation Heating Tests

    Energy Technology Data Exchange (ETDEWEB)

    Morris, Robert Noel [ORNL; Baldwin, Charles A [ORNL; Hunn, John D [ORNL; Demkowicz, Paul [Idaho National Laboratory (INL); Reber, Edward [Idaho National Laboratory (INL)

    2014-01-01

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide TRISO fuel compacts from the AGR-1 experiment has been evaluated at temperatures of 1600 1800 C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4 to 19.1% FIMA have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 10-6 after 300 h at 1600 C or 100 h at 1800 C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 C, and 85Kr release was very low during the tests (particles with breached SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 C in one compact. Post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.

  13. The DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification Program

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    David Petti; Hans Gougar; Gary Bell

    2005-05-01

    The Department of Energy has established the Advanced Gas Reactor Fuel Development and Qualification Program to address the following overall goals: Provide a baseline fuel qualification data set in support of the licensing and operation of the Next Generation Nuclear Plant (NGNP). Gas-reactor fuel performance demonstration and qualification comprise the longest duration research and development (R&D) task for the NGNP feasibility. The baseline fuel form is to be demonstrated and qualified for a peak fuel centerline temperature of 1250°C. Support near-term deployment of an NGNP by reducing market entry risks posed by technical uncertainties associated with fuel production and qualification. Utilize international collaboration mechanisms to extend the value of DOE resources. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, postirradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete fundamental understanding of the relationship between the fuel fabrication process, key fuel properties, the irradiation performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. Fuel performance modeling and analysis of the fission product behavior in the primary circuit are important aspects of this work. The performance models are considered essential for several reasons, including guidance for the plant designer in establishing the core design and operating limits, and demonstration to the licensing authority that the applicant has a thorough understanding of the in-service behavior of the fuel system. The fission product behavior task will also provide primary source term data needed for licensing. An overview of the program and recent progress will be presented.

  14. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  15. AGR-1 Data Qualification Report

    Energy Technology Data Exchange (ETDEWEB)

    Michael Abbott

    2010-03-01

    ABSTRACT Projects for the very high temperature reactor (VHTR) Technology Development Office (TDO) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the data streams associated with the first Advanced Gas Reactor experiment (AGR-1), the processing of these data within NDMAS, and reports the qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. They include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing, to confirm that the data are an accurate representation of the system or object being measured, and (3) documentation that the data were collected under an NQA-1 or equivalent quality assurance program. The NDMAS database processing and qualification status of the following five data streams is reported in this document: 1. Fuel fabrication data. All data have been processed into the NDMAS database and qualified (1,819 records). 2. Fuel irradiation data. Data from all 13 AGR-1 reactor cycles have been processed into the NDMAS database and tested. Of these, 85% have been qualified and 15% have failed NDMAS accuracy testing. 3. FPMS data. Reprocessed (January 2010) data from all 13 AGR-1 reactor cycles have been processed into the database and capture tested. Final qualification of these data will be recorded after QA approval of an Engineering Calculations and Analysis Report

  16. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V. [ed.; Feinberg, O.; Morozov, A. [Russian Research Centre `Kurchatov Institute`, Moscow (Russian Federation); Devell, L. [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1995-07-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

  17. A New Type of Deaggregators, the Bolaamphiphilic DeAgrs.Observations of the Dependence of Deaggregating Ability on Aggregator Concentration and Solvent Aggregating Power

    Institute of Scientific and Technical Information of China (English)

    ZHOU,Qi-Zhong(周其忠); SHI,Ji-Liang(史济良); JIANG,Xi-Kui(蒋锡夔); LI,Zhan-Ting(黎占亭)

    2002-01-01

    Four new amphiphilic molecules, i. e., the bolaamphipiles ω-trimethylamino n-hexadecanoate( deAgr 16 ), n-dodecanoate (deAgr 12), n-undecanoate(deAgr 11), and n-octanoate (deAgr 8)were synthesized. Their ability as deaggregators to break up the aggregating systems of hexadecyl β-naphthoate (A16) and dodecyl β-naphthoate (A12) in dioxanewater binary solvent systems was investigated in a mumber of solvent systems and at different A16 and A12 concentrations.The results reveal that this new type of bolaamphiphilic molecules appears to be one of the most efficient deaggregators reported, which, at very low concentrations, can effectively break up the aggregating systems of highly concentrated A16and A12. It is also found that bolaamphiphiles with longer aliphatic chains possess greater deaggregatibg ability. However, when the overall overall concentrations of A16 and A12 are very low but still above their critiical aggregating concentration,thee bolaamphiphilic molecules lose their deaggregating power and tend to coaggregate with the aggregated A16 or A12molecules. The quite efficient deaggregating ability of the longchain bolaamphiphiles (deAgr 16 and deAgr 12), as compared to the short-chain deAgr 8, has been ascribed to their dynamic switch-on/switch-off behavior involving coiled-up and uncoiled conformations of their long aliphatic chains.

  18. Pentose sugars inhibit metabolism and increase expression of an AgrD-type cyclic pentapeptide in Clostridium thermocellum

    Science.gov (United States)

    Verbeke, Tobin J.; Giannone, Richard J.; Klingeman, Dawn M.; Engle, Nancy L.; Rydzak, Thomas; Guss, Adam M.; Tschaplinski, Timothy J.; Brown, Steven D.; Hettich, Robert L.; Elkins, James G.

    2017-01-01

    Clostridium thermocellum could potentially be used as a microbial biocatalyst to produce renewable fuels directly from lignocellulosic biomass due to its ability to rapidly solubilize plant cell walls. While the organism readily ferments sugars derived from cellulose, pentose sugars from xylan are not metabolized. Here, we show that non-fermentable pentoses inhibit growth and end-product formation during fermentation of cellulose-derived sugars. Metabolomic experiments confirmed that xylose is transported intracellularly and reduced to the dead-end metabolite xylitol. Comparative RNA-seq analysis of xylose-inhibited cultures revealed several up-regulated genes potentially involved in pentose transport and metabolism, which were targeted for disruption. Deletion of the ATP-dependent transporter, CbpD partially alleviated xylose inhibition. A putative xylitol dehydrogenase, encoded by Clo1313_0076, was also deleted resulting in decreased total xylitol production and yield by 41% and 46%, respectively. Finally, xylose-induced inhibition corresponds with the up-regulation and biogenesis of a cyclical AgrD-type, pentapeptide. Medium supplementation with the mature cyclical pentapeptide also inhibits bacterial growth. Together, these findings provide new foundational insights needed for engineering improved pentose utilizing strains of C. thermocellum and reveal the first functional Agr-type cyclic peptide to be produced by a thermophilic member of the Firmicutes. PMID:28230109

  19. Moving hot cell for LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru

    1994-09-16

    A moving hot cell for an LMFBR type reactor is made movable on a reactor operation floor between a position just above the reactor container and a position retreated therefrom. Further, it comprises an overhung portion which can incorporate a spent fuel just thereunder, and a crane for moving a fuel assembly between a spent fuel cask and a reactor container. Further, an opening/closing means having a shielding structure is disposed to the bottom portion and the overhung portion thereof, to provide a sealing structure, in which only the receiving port for the spent fuel cask faces to the inner side, and the cask itself is disposed at the outside. Upon exchange of fuels, the movable hot cell is placed just above the reactor to take out the spent fuels, so that a region contaminated with primary sodium is limited within the hot cell. On the other hand, upon maintenance and repair for equipments, the hot cell is moved, thereby enabling to provide a not contaminated reactor operation floor. (N.H.).

  20. Sensitivity Evaluation of the Daily Thermal Predictions of the AGR-1 Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Grant Hawkes; James Sterbentz; John Maki

    2011-05-01

    A temperature sensitivity evaluation has been performed for the AGR-1 fuel experiment on an individual capsule. A series of cases were compared to a base case by varying different input parameters into the ABAQUS finite element thermal model. These input parameters were varied by ±10% to show the temperature sensitivity to each parameter. The most sensitive parameters are the outer control gap distance, heat rate in the fuel compacts, and neon gas fraction. Thermal conductivity of the compacts and graphite holder were in the middle of the list for sensitivity. The smallest effects were for the emissivities of the stainless steel, graphite, and thru tubes. Sensitivity calculations were also performed varying with fluence. These calculations showed a general temperature rise with an increase in fluence. This is a result of the thermal conductivity of the fuel compacts and graphite holder decreasing with fluence.

  1. Existence of two groups of Staphylococcus aureus strains isolated from bovine mastitis based on biofilm formation, intracellular survival, capsular profile and agr-typing.

    Science.gov (United States)

    Bardiau, Marjorie; Caplin, Jonathan; Detilleux, Johann; Graber, Hans; Moroni, Paolo; Taminiau, Bernard; Mainil, Jacques G

    2016-03-15

    Staphylococcus (S.) aureus is recognised worldwide as an important pathogen causing contagious acute and chronic bovine mastitis. Chronic mastitis account for a significant part of all bovine cases and represent an important economic problem for dairy producers. Several properties (biofilm formation, intracellular survival, capsular expression and group agr) are thought to be associated with this chronic status. In a previous study, we found the existence of two groups of strains based on the association of these features. The aim of the present work was to confirm on a large international and non-related collection of strains the existence of these clusters and to associate them with case history records. In addition, the genomes of eight strains were sequenced to study the genomic differences between strains of each cluster. The results confirmed the existence of both groups based on capsular typing, intracellular survival and agr-typing: strains cap8-positive, belonging to agr group II, showing a low invasion rate and strains cap5-positive, belonging to agr group I, showing a high invasion rate. None of the two clusters were associated with the chronic status of the cow. When comparing the genomes of strains belonging to both clusters, the genes specific to the group "cap5-agrI" would suggest that these strains are better adapted to live in hostile environment. The existence of these two groups is highly important as they may represent two clusters that are adapted differently to the host and/or the surrounding environment.

  2. Evidence that the Agr-like quorum sensing system regulates the toxin production, cytotoxicity and pathogenicity of Clostridium perfringens type C isolate CN3685.

    Science.gov (United States)

    Vidal, Jorge E; Ma, Menglin; Saputo, Julian; Garcia, Jorge; Uzal, Francisco A; McClane, Bruce A

    2012-01-01

    Clostridium perfringens possesses at least two functional quorum sensing (QS) systems, i.e. an Agr-like system and a LuxS-dependent AI-2 system. Both of those QS systems can reportedly control in vitro toxin production by C. perfringens but their importance for virulence has not been evaluated. Therefore, the current study assessed whether these QS systems might regulate the pathogenicity of CN3685, a C. perfringens type C strain. Since type C isolates cause both haemorrhagic necrotic enteritis and fatal enterotoxemias (where toxins produced in the intestines are absorbed into the circulation to target other internal organs), the ability of isogenic agrB or luxS mutants to cause necrotizing enteritis in rabbit small intestinal loops or enterotoxemic lethality in mice was evaluated. Results obtained strongly suggest that the Agr-like QS system, but not the LuxS-dependent AI-2 QS system, is required for CN3685 to cause haemorrhagic necrotizing enteritis, apparently because the Agr-like system regulates the production of beta toxin, which is essential for causing this pathology. The Agr-like system, but not the LuxS-mediated AI-2 system, was also important for CN3685 to cause fatal enterotoxemia. These results provide the first direct evidence supporting a role for any QS system in clostridial infections.

  3. AGR-1 Data Qualification Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    Machael Abbott

    2009-08-01

    Projects for the very-high-temperature reactor (VHTR) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR Program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the data streams associated with the first Advanced Gas Reactor (AGR-1) experiment, the processing of these data within NDMAS, and reports the interim FY09 qualification status of the AGR-1 data to date. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category, which is assigned by the data generator, and include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing, to confirm that the data are an accurate representation of the system or object being measured, and (3) documentation that the data were collected under an NQA-1 or equivalent QA program. The interim qualification status of the following four data streams is reported in this document: (1) fuel fabrication data, (2) fuel irradiation data, (3) fission product monitoring system (FPMS) data, and (4) Advanced Test Reactor (ATR) operating conditions data. A final report giving the NDMAS qualification status of all AGR-1 data (including cycle 145A) is planned for February 2010.

  4. AGR-1 Thermocouple Data Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jeff Einerson

    2012-05-01

    This report documents an effort to analyze measured and simulated data obtained in the Advanced Gas Reactor (AGR) fuel irradiation test program conducted in the INL's Advanced Test Reactor (ATR) to support the Next Generation Nuclear Plant (NGNP) R&D program. The work follows up on a previous study (Pham and Einerson, 2010), in which statistical analysis methods were applied for AGR-1 thermocouple data qualification. The present work exercises the idea that, while recognizing uncertainties inherent in physics and thermal simulations of the AGR-1 test, results of the numerical simulations can be used in combination with the statistical analysis methods to further improve qualification of measured data. Additionally, the combined analysis of measured and simulation data can generate insights about simulation model uncertainty that can be useful for model improvement. This report also describes an experimental control procedure to maintain fuel target temperature in the future AGR tests using regression relationships that include simulation results. The report is organized into four chapters. Chapter 1 introduces the AGR Fuel Development and Qualification program, AGR-1 test configuration and test procedure, overview of AGR-1 measured data, and overview of physics and thermal simulation, including modeling assumptions and uncertainties. A brief summary of statistical analysis methods developed in (Pham and Einerson 2010) for AGR-1 measured data qualification within NGNP Data Management and Analysis System (NDMAS) is also included for completeness. Chapters 2-3 describe and discuss cases, in which the combined use of experimental and simulation data is realized. A set of issues associated with measurement and modeling uncertainties resulted from the combined analysis are identified. This includes demonstration that such a combined analysis led to important insights for reducing uncertainty in presentation of AGR-1 measured data (Chapter 2) and interpretation of

  5. Study on the Adaptability of Etheriifcation Feedstock to Reactor Type

    Institute of Scientific and Technical Information of China (English)

    Mao Junyi; Yuan Qing; Wang Lei; Huang Tao

    2016-01-01

    A reactive C5 oleifns and methanol etheriifcation kinetic model based on E-R mechanism was established and three different types of reactors including the adiabatic ifxed-bed liquid reactor, the external loop reactor and the mixed-phase reactor were constructed by Aspen Plus. The adaptability of reactive C5 oleifns to these reactors was studied and simulated using various gasoline fractions with different oleifns content. After the theoretical model was validated by the experimental data of the etheriifcation of three C5 light cut fractions from different gasoline sources in different reactors, the simulated isoamylene conversion with reactive C5 olefin contents increasing from 10% to 60% was studied in the three different types of reactors for etheriifcation with methanol, respectively. Test results show that there is an obvious adaptability of the feedstock composition to the reactor type to achieve a high conversion.

  6. AGR-1, AGR-2 and AGR-3/4 Dimensional Change Data Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Herberger, Sarah E. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

    A series of Advanced Gas Reactor (AGR) experiments have been completed in the Advanced Test Reactor at Idaho National Laboratory in support of qualification and development of tristructural isotropic fuel. Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder. These capsules are instrumented with thermocouples embedded in the graphite, enabling temperature control. The fuel compacts are composed of fuel particles surrounded by a graphitic A3 matrix material. Dimensional change in AGR fuel compacts is vital because the swelling or shrinkage affects the size of the gas gaps that are used to control temperatures. Analysis of dimensional change in the AGR fuel compacts is needed to establish the variables directly relating to compact shrinkage. The variables initially identified for consideration were matrix density, compact density, fuel packing fraction, uranium loading, fuel particle diameter, cumulative fast neutron fluence, and volume average time average fuel temperature. In addition to the data available from the AGR experiments, the analysis included specimens formed from the same A3 matrix material used in Advanced Graphite Creep (AGC) experiments, which provide graphite creep data during irradiation for design and licensing purposes. The primary purpose of including the AGC specimens was to encompass dimensional behavior at zero packing fraction, zero uranium loading, and zero particle diameter. All possible combinations of first-order variable regressions were considered in the analysis. The study focused on identifying the best regression models for percent change in diameter, length, and volume. Bootstrap analysis was used to ensure the resulting regression models were robust and well-performing. The variables identified as very significant in predicting change in one or more dimensions (diameter, length, and volume) are volume average time average temperature, fast fluence

  7. Safety testing of AGR-2 UO2 compacts 3-3-2 and 3-4-2

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Morris, Robert Noel [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Baldwin, Charles A. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Montgomery, Fred C. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    Post-irradiation examination (PIE) is in progress on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2) [Collin 2014]. The AGR-2 PIE will build upon new information and understanding acquired throughout the recently-concluded six-year AGR-1 PIE campaign [Demkowicz et al. 2015] and establish a database for the different AGR-2 fuel designs.

  8. AGR-2 and AGR-3/4 Release-to-Birth Ratio Data Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh T. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Very High Temperature Reactor Technology Development Office; Einerson, Jeffrey J. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Very High Temperature Reactor Technology Development Office; Scates, Dawn M. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Very High Temperature Reactor Technology Development Office; Maki, John T. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Very High Temperature Reactor Technology Development Office; Petti, David A. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Very High Temperature Reactor Technology Development Office

    2014-09-01

    A series of Advanced Gas Reactor (AGR) irradiation tests is being conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) low enriched fuel used in the High Temperature Gas-cooled Reactor (HTGR). Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples embedded in the graphite enabling temperature control. AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through use of helium coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO coated particle fuel. Thus, these tests provide valuable irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, and support development and validation of fuel performance and fission product transport models and codes.

  9. Antibacterial Activity of Cold Atmospheric Pressure Argon Plasma against 78 Genetically Different (mecA, luk-P, agr or Capsular Polysaccharide Type) Staphylococcus aureus Strains.

    Science.gov (United States)

    Matthes, Rutger; Lührman, Anne; Holtfreter, Silva; Kolata, Julia; Radke, Dörte; Hübner, Nils-Olaf; Assadian, Ojan; Kramer, Axel

    2016-01-01

    Previous studies on the antimicrobial activity of cold atmospheric pressure argon plasma showed varying effects against mecA+ or mecA-Staphylococcus aureus strains. This observation may have important clinical and epidemiological implications. Here, the antibacterial activity of argon plasma was investigated against 78 genetically different S. aureus strains, stratified by mecA, luk-P, agr1-4, or the cell wall capsule polysaccharide types 5 and 8. kINPen09® served as the plasma source for all experiments. On agar plates, mecA+luk-P-S. aureus strains showed a decreased susceptibility against plasma compared to other S. aureus strains. This study underlines the high complexity of microbial defence against antimicrobial treatment and confirms a previously reported strain-dependent susceptibility of S. aureus to plasma treatment.

  10. Nox2 modification of LDL is essential for optimal apolipoprotein B-mediated control of agr type III Staphylococcus aureus quorum-sensing.

    Directory of Open Access Journals (Sweden)

    Pamela R Hall

    2013-02-01

    Full Text Available Staphylococcus aureus contains an autoinducing quorum-sensing system encoded within the agr operon that coordinates expression of virulence genes required for invasive infection. Allelic variation within agr has generated four agr specific groups, agr I-IV, each of which secretes a distinct autoinducing peptide pheromone (AIP1-4 that drives agr signaling. Because agr signaling mediates a phenotypic change in this pathogen from an adherent colonizing phenotype to one associated with considerable tissue injury and invasiveness, we postulated that a significant contribution to host defense against tissue damaging and invasive infections could be provided by innate immune mechanisms that antagonize agr signaling. We determined whether two host defense factors that inhibit AIP1-induced agrI signaling, Nox2 and apolipoprotein B (apoB, also contribute to innate control of AIP3-induced agrIII signaling. We hypothesized that apoB and Nox2 would function differently against AIP3, which differs from AIP1 in amino acid sequence and length. Here we show that unlike AIP1, AIP3 is resistant to direct oxidant inactivation by Nox2 characteristic ROS. Rather, the contribution of Nox2 to defense against agrIII signaling is through oxidation of LDL. ApoB in the context of oxLDL, and not LDL, provides optimal host defense against S. aureus agrIII infection by binding the secreted signaling peptide, AIP3, and preventing expression of the agr-driven virulence factors which mediate invasive infection. ApoB within the context of oxLDL also binds AIP 1-4 and oxLDL antagonizes agr signaling by all four agr alleles. Our results suggest that Nox2-mediated oxidation of LDL facilitates a conformational change in apoB to one sufficient for binding and sequestration of all four AIPs, demonstrating the interdependence of apoB and Nox2 in host defense against agr signaling. These data reveal a novel role for oxLDL in host defense against S. aureus quorum-sensing signaling.

  11. AGR-1 Irradiation Test Final As-Run Report

    Energy Technology Data Exchange (ETDEWEB)

    Blaise P. Collin

    2012-06-01

    This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 ?1025 n/m2 (E >0.18 MeV). We’ll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below 10-7 with only one

  12. Positron Annihilation Studies of VVER Type Reactor Steels

    OpenAIRE

    Brauer, G.

    1995-01-01

    A summary of recent positron annihilation work on Russian VVER type reactor steels is presented. Thereby, special attention is paid to the outline of basic processes that might help to understand the positron behaviour in this class of industrial material. The idea of positron trapping by irradiation-induced precipitates, which are probably carbides, is discussed in detail.

  13. Power plant performance - resurrection of the AGR

    Energy Technology Data Exchange (ETDEWEB)

    Kovan, Dick

    1994-04-01

    When the rest of the electricity supply industry was privatized in 1990, Nuclear Electric became responsible for the running of nuclear plants in England and Wales. These included five AGR stations, the output and revenue from which have increased dramatically. The absence of a fully representative prototype reactor led to deficiencies in the design and problems in the construction and commissioning of the AGR stations, particularly the first, Dungeness B. Subsequent plant and design improvements have led to the achievement of good performance, however, and Nuclear Electric's better understanding of AGR safety has been translated into higher output because it has been possible to run nearer to the limits on power ratings, temperatures and pressures. Increased efficiency in refuelling and greater productivity of boilers, gas circulators and turbines, together with improved servicing and maintenance procedures have all contributed to a reduction in reactor downtime. These and other actions which have been employed by Nuclear Electric to turn round the fortunes of the AGR stations are discussed. (UK).

  14. AGR-1 Safety Test Predictions using the PARFUME code

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2012-05-01

    The PARFUME modeling code was used to predict failure probability of TRISO-coated fuel particles and diffusion of fission products through these particles during safety tests following the first irradiation test of the Advanced Gas Reactor program (AGR-1). These calculations support the AGR-1 Safety Testing Experiment, which is part of the PIE effort on AGR-1. Modeling of the AGR-1 Safety Test Predictions includes a 620-day irradiation followed by a 300-hour heat-up phase of selected AGR-1 compacts. Results include fuel failure probability, palladium penetration, and fractional release of fission products. Results show that no particle failure is predicted during irradiation or heat-up, and that fractional release of fission products is limited during irradiation but that it significantly increases during heat-up.

  15. The Agr-like quorum-sensing system regulates sporulation and production of enterotoxin and beta2 toxin by Clostridium perfringens type A non-food-borne human gastrointestinal disease strain F5603.

    Science.gov (United States)

    Li, Jihong; Chen, Jianming; Vidal, Jorge E; McClane, Bruce A

    2011-06-01

    Clostridium perfringens type A strains producing enterotoxin (CPE) cause one of the most common bacterial food-borne illnesses, as well as many cases of non-food-borne human gastrointestinal disease. Recent studies have shown that an Agr-like quorum-sensing system controls production of chromosomally encoded alpha-toxin and perfringolysin O by C. perfringens, as well as sporulation by Clostridium botulinum and Clostridium sporogenes. The current study explored whether the Agr-like quorum-sensing system also regulates sporulation and production of two plasmid-encoded toxins (CPE and beta2 toxin) that may contribute to the pathogenesis of non-food-borne human gastrointestinal disease strain F5603. An isogenic agrB null mutant was inhibited for production of beta2 toxin during vegetative growth and in sporulating culture, providing the first evidence that, in C. perfringens, this system can control production of plasmid-encoded toxins as well as chromosomally encoded toxins. This mutant also showed reduced production of alpha-toxin and perfringolysin O during vegetative growth. Importantly, when cultured in sporulation medium, the mutant failed to efficiently form spores and was blocked for CPE production. Complementation partially or fully reversed all phenotypic changes in the mutant, confirming that they were specifically due to inactivation of the agr locus. Western blots suggest that this loss of sporulation and sporulation-specific CPE production for the agrB null mutant involves, at least in part, Agr-mediated regulation of production of Spo0A and alternative sigma factors, which are essential for C. perfringens sporulation.

  16. Fabrication and Comparison of Fuels for Advanced Gas Reactor Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey Phillips; Charles Barnes; John Hunn

    2010-10-01

    As part of the program to demonstrate TRISO-coated fuel for the Next Generation Nuclear Plant, a series of irradiation tests of Advanced Gas Reactor (AGR) fuel are being performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. In the first test, called “AGR-1,” graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 until November 2009. Development of AGR-1 fuel sought to replicate the properties of German TRISO-coated particles. No particle failures were seen in the nearly 3-year irradiation to a burn up of 19%. The AGR-1 particles were coated in a two-inch diameter coater. Following fabrication of AGR-1 fuel, process improvements and changes were made in each of the fabrication processes. Changes in the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a six-inch diameter coater using a change size about twenty-one times that of the two-inch diameter coater used to coat AGR-1 particles. Changes were also made in the compacting process, including increasing the temperature and pressure of pressing and using a different type of press. Irradiation of AGR-2 fuel began in late spring 2010. Properties of AGR-2 fuel compare favorably with AGR-1 and historic German fuel. Kernels are more homogeneous in shape, chemistry and density. TRISO-particle sphericity, layer thickness standard deviations, and defect fractions are also comparable. In a sample of 317,000 particles from deconsolidated AGR-2 compacts, 3 exposed kernels were found in a leach test. No SiC defects were found in a sample of 250,000 deconsolidated particles, and no IPyC defects in a sample of 64,000 particles. The primary difference in properties between AGR-1 and AGR-2 compacts is that AGR-2 compacts have a higher matrix density, 1.6 g/cm3 compared to about 1.3 g/cm3 for AGR-1 compacts. Based on

  17. Uncertainty Quantification of Calculated Temperatures for the AGR-1 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Binh T. Pham; Jeffrey J. Einerson; Grant L. Hawkes

    2012-04-01

    This report documents an effort to quantify the uncertainty of the calculated temperature data for the first Advanced Gas Reactor (AGR-1) fuel irradiation experiment conducted in the INL's Advanced Test Reactor (ATR) in support of the Next Generation Nuclear Plant (NGNP) R&D program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR-1 test, the results of the numerical simulations can be used in combination with the statistical analysis methods to improve qualification of measured data. Additionally, the temperature simulation data for AGR tests can be used for validation of the fuel transport and fuel performance simulation models. The crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. The report is organized into three chapters. Chapter 1 introduces the AGR Fuel Development and Qualification program and provides overviews of AGR-1 measured data, AGR-1 test configuration and test procedure, and thermal simulation. Chapters 2 describes the uncertainty quantification procedure for temperature simulation data of the AGR-1 experiment, namely, (i) identify and quantify uncertainty sources; (ii) perform sensitivity analysis for several thermal test conditions; (iii) use uncertainty propagation to quantify overall response temperature uncertainty. A set of issues associated with modeling uncertainties resulting from the expert assessments are identified. This also includes the experimental design to estimate the main effects and interactions of the important thermal model parameters. Chapter 3 presents the overall uncertainty results for the six AGR-1 capsules. This includes uncertainties for the daily volume-average and peak fuel temperatures, daily average temperatures at TC locations, and time-average volume-average and time-average peak fuel temperatures.

  18. Uncertainty Quantification of Calculated Temperatures for the AGR-1 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Binh T. Pham; Jeffrey J. Einerson; Grant L. Hawkes

    2013-03-01

    This report documents an effort to quantify the uncertainty of the calculated temperature data for the first Advanced Gas Reactor (AGR-1) fuel irradiation experiment conducted in the INL’s Advanced Test Reactor (ATR) in support of the Next Generation Nuclear Plant (NGNP) R&D program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR-1 test, the results of the numerical simulations can be used in combination with the statistical analysis methods to improve qualification of measured data. Additionally, the temperature simulation data for AGR tests can be used for validation of the fuel transport and fuel performance simulation models. The crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. The report is organized into three chapters. Chapter 1 introduces the AGR Fuel Development and Qualification program and provides overviews of AGR-1 measured data, AGR-1 test configuration and test procedure, and thermal simulation. Chapters 2 describes the uncertainty quantification procedure for temperature simulation data of the AGR-1 experiment, namely, (i) identify and quantify uncertainty sources; (ii) perform sensitivity analysis for several thermal test conditions; (iii) use uncertainty propagation to quantify overall response temperature uncertainty. A set of issues associated with modeling uncertainties resulting from the expert assessments are identified. This also includes the experimental design to estimate the main effects and interactions of the important thermal model parameters. Chapter 3 presents the overall uncertainty results for the six AGR-1 capsules. This includes uncertainties for the daily volume-average and peak fuel temperatures, daily average temperatures at TC locations, and time-average volume-average and time-average peak fuel temperatures.

  19. An innovative reactor-type biosensor for BOD rapid measurement.

    Science.gov (United States)

    Wang, Jianlong; Zhang, Yixin; Wang, Yeyao; Xu, Runhua; Sun, Zhonghua; Jie, Zhou

    2010-03-15

    Biochemical oxygen demand (BOD) is one of the most important and widely used parameters for characterizing the organic pollution of water and wastewater. In this paper, a novel reactor-type biosensor for rapid measurement of BOD was developed, based on using immobilized microbial cell (IMC) beads as recognition bio-element in a completely mixed reactor which was used as determining chamber, replacing the traditionally used membrane as recognition bio-element. The IMC beads were freely suspended in the aqueous solution, so the mass transfer resistance for dissolved oxygen and organic compounds significantly reduced, and the quantity of the microbial cells used as recognition element can be easily adjusted, in comparison with the traditional membrane-type BOD biosensor, in which exists a unadjustable contradiction between the quantity of biomass and the thickness of the bio-membrane, thus limiting the stability and the detection limit. This novel kind of BOD biosensor significantly increased the sensitivity of the response, the detecting precision and prolonged the life time of the recognition element. The experimental data showed that the most appropriate temperature for biochemical reaction in the reactor was 30 degrees C, and the IMC beads could keep the bioactivity for about 70d at the detecting frequency of 8 times every day. The standard deviation of repeatability and the reproducibility of responses were within +/-6.4% and +/-5.0%, respectively, which are within acceptable bias limits, and meet the requirement of BOD rapid measurement.

  20. Selection of Type I and Type II Methanotrophic Proteobacteria in a Fluidized Bed Reactor under Non-Sterile Conditions

    Science.gov (United States)

    2011-08-01

    00-00-2011 to 00-00-2011 4. TITLE AND SUBTITLE Selection of Type I and Type II methanotrophic proteobacteria in a fluidized bed reactor under...laboratory- scale fluidized bed reactor was initially inoculated with a Type II Methylocystis-like dominated culture. At elevated levels of dissolved...personal copy Selection of Type I and Type II methanotrophic proteobacteria in a fluidized bed reactor under non-sterile conditions Andrew R. Pfluger a, Wei

  1. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  2. Description of the magnox type of gas cooled reactor (MAGNOX)

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, S.E.; Nonboel, E

    1999-05-01

    The present report comprises a technical description of the MAGNOX type of reactor as it has been build in Great Britain. The Magnox reactor is gas cooled (CO{sub 2}) with graphite moderators. The fuels is natural uranium in metallic form, canned with a magnesium alloy called 'Magnox'. The Calder Hall Magnox plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other stations are given in tables with a summary of design data. Special design features are also shortly described. Where specific data for Calder Hall Magnox has not been available, corresponding data from other Magnox plants has been used. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 sub-project 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au)

  3. AGR-2 Data Qualification Report for ATR Cycle 154B

    Energy Technology Data Exchange (ETDEWEB)

    Binh Pham; Jeff Einerson

    2014-01-01

    This report provides the data qualification status of Advanced Gas Reactor-2 (AGR-2) fuel irradiation experimental data from Advanced Test Reactor (ATR) Cycle 154B as recorded in the Nuclear Data Management and Analysis System (NDMAS). This is the last cycle of AGR-2 irradiation, as the test train was pulled from the ATR core during the outage portion of ATR Cycle 155A. The AGR-2 data streams addressed in this report include thermocouple (TC) temperatures, sweep gas data (flow rates including new Fission Product Monitoring (FPM) downstream flows from Fission Product Monitoring System (FPMS) detectors, pressure, and moisture content), and FPMS data (release rates and release-to-birth rate ratios [R/Bs]) for each of the six capsules in the AGR-2 experiment. The final data qualification status for these data streams is determined by a Data Review Committee (DRC) comprised of AGR technical leads, Sitewide Quality Assurance (QA), and NDMAS analysts. The Data Review Committee reviewed the data acquisition process, considered whether the data met the requirements for data collection as specified in QA-approved Very High Temperature Reactor (VHTR) data collection plans, examined the results of NDMAS data testing and statistical analyses, and confirmed the qualification status of the data as given in this report.

  4. Development of toroid-type HTS DC reactor series for HVDC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwangmin, E-mail: kwangmin81@gmail.com [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of); Oh, Yunsang [Vector Fields Korea Inc., Pohang 790-834 (Korea, Republic of); Park, Minwon; Yu, In-Keun [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of)

    2015-11-15

    Highlights: • The authors developed the 400 mH, 400 A class toroid-type HTS DC reactor system. • The target temperature, inductance and operating current are under 20 K at magnet, 400 mH and 400 A, respectively. All target performances of the HTS DC reactor were achieved. • The HTS DC reactor was conducted through the interconnection operation with a LCC type HVDC system. • Now, the authors are studying the 400 mH, 1500 A class toroid-type HTS DC reactor for the next phase HTS DC reactor. - Abstract: This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

  5. Dimethylamine as a replacement for ammonia dosing in the secondary circuit of an advanced gas-cooled reactor (AGR) power station

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, Chris; Mitchell, Malcolm S. [EDF Energy, Hartlepool Power Station, Hartlepool (United Kingdom); Bull, Andrew E.A.; Quirk, Graham P.; Rudge, Andy [EDF Energy Nuclear Generation, Barnwood, Gloucester (United Kingdom). Central Technical Organisation; Woolsey, Ian S.

    2012-06-15

    Increasing flow resistance observed over recent years within the helical once-through boilers in the four advanced gas-cooled reactors at Hartlepool and Heysham 1 Power Stations have reduced boiler performance, resulting in reductions in feedwater flow, steam temperatures, and power output and in the need to carry out periodic chemical cleaning. The root cause is believed to be the development of magnetite deposits with high flow impedance in the 9Cr1Mo evaporator section of the boiler tubing. To prevent continued increases in boiler flow resistance, dimethylamine is being trialled, in one of the four affected units, as a replacement to the conventional ammonia dosing. Dimethylamine increases the pH at temperature around the secondary circuit and, based on full scale boiler rig simulations, is expected to reduce iron transport and prevent flow resistance increases within the evaporator section of the boiler. The dimethylamine plant trial commenced in January 2011 and is ongoing. The feedwater concentration of dimethylamine has been increased progressively towards a final target value of 900 {mu}g . kg{sup -1} and its effect on iron transport and boiler pressure loss is being closely monitored. The high steam temperature (> 500 C) of the secondary circuit leads to some decomposition of dimethylamine, which is being carefully monitored at various locations around the circuit. The decomposition products identified with dimethylamine dosing include ammonia, methylamine, formic acid, carbon dioxide and, as yet, unidentified neutral organic species. The effect of dimethylamine dosing on iron transport and boiler pressure drops and its decomposition behaviour around the secondary circuit during the plant trial will be presented in this paper. (orig.)

  6. Installation of a new type of nuclear reactor in Mexico: advantages and disadvantages; Instalacion de un nuevo tipo de reactor nuclear en Mexico: ventajas y desventajas

    Energy Technology Data Exchange (ETDEWEB)

    Jurado P, M.; Martin del Campo M, C. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: mjp_green@hotmail.com

    2005-07-01

    In this work the main advantages and disadvantages of the installation of a new type of nuclear reactor different to the BWR type reactor in Mexico are presented. A revision of the advanced reactors is made that are at the moment in operation and of the advanced reactors that are in construction or one has already planned its construction in the short term. Specifically the A BWR and EPR reactors are analyzed. (Author)

  7. Scale Effects on Magnet Systems of Heliotron-Type Reactors

    Institute of Scientific and Technical Information of China (English)

    S. Imagawa; A. Sagara

    2005-01-01

    For power plants heliotron-type reactors have attractive advantages, such as no current-disruptions, no current-drive, and wide space between helical coils for the maintenance of in-vessel components. However, one disadvantage is that a major radius has to be large enough to obtain large Q-value or to produce sufficient space for blankets. Although the larger radius is considered to increase the construction cost, the influence has not been understood clearly,yet. Scale effects on superconducting magnet systems have been estimated under the conditions of a constant energy confinement time and similar geometrical parameters. Since the necessary magnetic field with a larger radius becomes lower, the increase rate of the weight of the coil support to the major radius is less than the square root. The necessary major radius will be determined mainly by the blanket space. The appropriate major radius will be around 13 m for a reactor similar to the Large Helical Device (LHD).

  8. Development of toroid-type HTS DC reactor series for HVDC system

    Science.gov (United States)

    Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

    2015-11-01

    This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

  9. LMFBR type reactor and power generation system using the same

    Energy Technology Data Exchange (ETDEWEB)

    Otsubo, Akira.

    1994-02-25

    A reactor core void reactivity of a reactor main body is set to negative or zero. A heat insulation structure is disposed on the inner wall surface of a reactor container. Oxide fuels or nitride fuels are used. A fuel pin cladding tube has a double walled structure having an outer side of stainless steel and an inner side of niobium alloy. Upon imaginary event, boiling is allowed. Even if boiling of coolants should occur by temperature elevation of fuels upon imaginary event, since reactor core fuels comprises oxides or nitrides, they have a heat resistance, further, and since the fuel pin cladding tube has super heat resistance, it has a high temperature strength, so that it is not ruptured and durable to the coolant boiling temperature. Since the reactor core void reactivity is negative or zero, the reactor core is in a subcritical state by the boiling, and the reactor core power is reduced to several % of the rated power. Accordingly, boiling and non-boiling are repeated substantially permanently in the reactor core, during which safety can be kept with no operator's handling. Further, heat generated in the reactor core is gradually removed by an air cooling system for the reactor container. (N.H.).

  10. AGR-1 Post Irradiation Examination Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    The post-irradiation examination (PIE) of the Advanced Gas Reactor (AGR)-1 experiment was a multi-year, collaborative effort between Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) to study the performance of UCO (uranium carbide, uranium oxide) tristructural isotropic (TRISO) coated particle fuel fabricated in the U.S. and irradiated at the Advanced Test Reactor at INL to a peak burnup of 19.6% fissions per initial metal atom. This work involved a broad array of experiments and analyses to evaluate the level of fission product retention by the fuel particles and compacts (both during irradiation and during post-irradiation heating tests to simulate reactor accident conditions), investigate the kernel and coating layer morphology evolution and the causes of coating failure, and explore the migration of fission products through the coating layers. The results have generally confirmed the excellent performance of the AGR-1 fuel, first indicated during the irradiation by the observation of zero TRISO coated particle failures out of 298,000 particles in the experiment. Overall release of fission products was determined by PIE to have been relatively low during the irradiation. A significant finding was the extremely low levels of cesium released through intact coatings. This was true both during the irradiation and during post-irradiation heating tests to temperatures as high as 1800°C. Post-irradiation safety test fuel performance was generally excellent. Silver release from the particles and compacts during irradiation was often very high. Extensive microanalysis of fuel particles was performed after irradiation and after high-temperature safety testing. The results of particle microanalysis indicate that the UCO fuel is effective at controlling the oxygen partial pressure within the particle and limiting kernel migration. Post-irradiation examination has provided the final body of data that speaks to the quality of the AGR-1 fuel, building

  11. Design guide for Category III reactors: pool type reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-11-01

    The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems.

  12. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950`s are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  13. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950's are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  14. AGR-2 Irradiation Test Final As-Run Report

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Program

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel.

  15. AGR-2 Irradiation Test Final As-Run Report

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Program

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) program. The objectives of the AGR-2 experiment are to: 1. Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. 2. Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. 3. Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tristructural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S.-produced fuel.

  16. AGR-1 Irradiation Test Final As-Run Report, Rev. 3

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-01-01

    This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 x 1025 n/m2 (E >0.18 MeV). We’ll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below

  17. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  18. Steady Thermal Field Simulation of Forced Air-cooled Column-type Air-core Reactor

    Institute of Scientific and Technical Information of China (English)

    DENG Qiu; LI Zhenbiao; YIN Xiaogen; YUAN Zhao

    2013-01-01

    Modeling the steady thermal field of the column-type air-core reactor,and further analyzing its distribution regularity,will help optimizing reactor design as well as improving its quality.The operation mechanism and inner insulation structure of a novel current limiting column-type air-core reactor is introduced in this paper.The finite element model of five encapsulation forced air-cooled column type air-core reactor is constructed using Fluent.Most importantly,this paper present a new method that,the steady thermal field of reactor working under forced air-cooled condition is simulated without arbitrarily defining the convection heat transfer coefficient for the initial condition; The result of the thermal field distribution shows that,the maximum steady temperature rise of forced air-cooled columntype air-core reactor happens approximately 5% to its top.The law of temperature distribution indicates:In the 1/3part of the reactor to its bottom,the temperature will rise rapidly to the increasing of height,yet the gradient rate is gradually decreasing; In the 5 % part of the reactor to its top,the temperature will drop rapidly to the increasing of height; In the part between,the temperature will rise slowly to the increasing of height.The conclusion draws that more thermal withstand capacity should be considered at the 5 % part of the reactor to its top to achieve optimal design solution.

  19. Convective cooling in a pool-type research reactor

    Science.gov (United States)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

  20. Convective cooling in a pool-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sipaun, Susan, E-mail: susan@nm.gov.my [Malaysian Nuclear Agency, Industrial Technology Division, Blok 29T, Bangi 43200, Selangor (Malaysia); Usman, Shoaib, E-mail: usmans@mst.edu [Missouri University of Science and Technology, Nuclear Engineering, 222 Fulton Hall 301 W.14th St., Rolla 64509 MO (United States)

    2016-01-22

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

  1. Reactor pressure vessels as type B transport containment boundaries

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, R.E. [Applied Science and Technology, Inc., Poway, CA (United States); Griesbach, T.J. [ATI Consulting, Danville, CA (United States)

    1998-07-01

    Transportation risk and personnel exposure, as well as the cost of decommissioning nuclear power plants, can all be reduced significantly through the one-time use of the reactor pressure vessel as a containment boundary for shipping the activated internal components from the reactor site to a burial site. In order to help provide the technical basis for this end-use application, the ASME Board on Nuclear Codes and Standards, through its Subcommittee XI, has prepared a draft nuclear code case that contains requirements for any modifications to the vessel, including materials, design, fabrication, and examination. In particular, the requirements for evaluation of potential brittle fracture as the result of potentially low ambient shipping temperatures combined with hypothetical transportation accident loading are addressed. Existing ASME Code Section XI rules for linear elastic fracture mechanics evaluation of irradiated reactor pressure vessels have been adapted and included in the code case. (authors)

  2. EFFECT OF PARTICLE TYPE ON CYCLONE FORMATION INSIDE A SOLAR REACTOR

    Directory of Open Access Journals (Sweden)

    Min-Hsiu Chien

    2016-07-01

    Full Text Available Solar reactors featuring a circulating cyclone flow pattern provide enhanced heat transfer and longer residence time increasing conversion efficiency. Cyclone flow also works in reducing particle deposition on solar reactor walls and exit which is particularly important issue in solar cracking reactors to avoid clogging. This paper focuses on the physics of cyclone formation inside a solar cracking reactor and experimentally analyzes the effect of particle entrainment on the flow pattern via two dimensional Particle Image Velocimetry (PIV. The cyclone flow structure in the reactor is reconstructed by capturing images from orientations perpendicular or parallel to the geometrical axis of the reactor. In order to conduct PIV measurements and to reconstruct the cyclone structure inside the solar reactor, the experiment was operated at room temperature with the flow configuration matching that of a solar reactor operating at high temperatures. Two types of seeding particles were tested, namely tri-ethylene glycol (TEG and solid carbon. The effectiveness of the screening flow was evaluated by measuring the quantity of solid particles deposit on the reactor walls. The Stokes flow analysis of each particle species was performed and the cyclone vector fields generated by using different particles are compared.

  3. Study on Thermal-Hydraulic Behavior of an Integral Type Reactor under Heaving Condition

    OpenAIRE

    2014-01-01

    A self-developed program was used to study the thermal-hydraulic behavior of an integral type reactor under heaving condition. Comparison of calculated results with the data of experiments performed on a natural circulation loop designed with reference to an integral type reactor of Tsinghua University in inclination, heaving, and rolling motions was carried out. Characteristics of natural circulation in heaving motion and effect of motion parameters on natural circulation were investigated. ...

  4. AGR2 is induced in asthma and promotes allergen-induced mucin overproduction.

    Science.gov (United States)

    Schroeder, Bradley W; Verhaeghe, Catherine; Park, Sung-Woo; Nguyenvu, Louis T; Huang, Xiaozhu; Zhen, Guohua; Erle, David J

    2012-08-01

    Mucins are gel-forming proteins that are responsible for the characteristic viscoelastic properties of mucus. Mucin overproduction is a hallmark of asthma, but the cellular requirements for airway mucin production are poorly understood. The endoplasmic reticulum (ER) protein anterior gradient homolog 2 (AGR2) is required for production of the intestinal mucin MUC2, but its role in the production of the airway mucins MUC5AC and MUC5B is not established. Microarray data were analyzed to examine the relationship between AGR2 and MUC5AC expression in asthma. Immunofluorescence was used to localize AGR2 in airway cells. Coimmunoprecipitation was used to identify AGR2-immature MUC5AC complexes. Agr2(-/-) mice were used to determine the role of AGR2 in allergic airway disease. AGR2 localized to the ER of MUC5AC- and MUC5B-producing airway cells and formed a complex with immature MUC5AC. AGR2 expression increased together with MUC5AC expression in airway epithelium from "Th2-high" asthmatics. Allergen-challenged Agr2(-/-) mice had greater than 50% reductions in MUC5AC and MUC5B proteins compared with allergen-challenged wild-type mice. Impaired mucin production in Agr2(-/-) mice was accompanied by an increase in the proportion of mucins contained within the ER and by evidence of ER stress in airway epithelium. This study shows that AGR2 increases with mucin overproduction in individuals with asthma and in mouse models of allergic airway disease. AGR2 interacts with immature mucin in the ER and loss of AGR2 impairs allergen-induced MUC5AC and MUC5B overproduction.

  5. Concept of magnet systems for LHD-type reactor

    Science.gov (United States)

    Imagawa, S.; Takahata, K.; Tamura, H.; Yanagi, N.; Mito, T.; Obana, T.; Sagara, A.

    2009-07-01

    Heliotron reactors have attractive features for fusion power plants such as having no need for current drive and a wide space between the helical coils for the maintenance of in-vessel components. Their main disadvantage was considered to be the necessarily large size of their magnet systems. According to the recent reactor studies based on the experimental results in the Large Helical Device, a major radius of plasma of 14-17 m with a central toroidal field of 6-4 T is needed to attain the self-ignition condition with a blanket space thicker than 1.1 m. The stored magnetic energy is estimated at 120-140 GJ. Although both the major radius and the magnetic energy are about three times as large as ITER, the maximum magnetic field and mechanical stress are comparable. In the preliminary structural analysis, the maximum stress intensity including the peak stress is less than the 1000 MPa that is allowed for strengthened stainless steel. Although the length of the helical coil is more than 150 m, that is about five times as long as the ITER TF coil, cable-in-conduit conductors can be adopted with a parallel winding method of five-in-hand. The concept of the parallel winding is proposed. Consequently, the magnet systems for helical reactors can be realized with a small extension of the ITER technology.

  6. Sensitivity Analysis on LOCCW of Westinghouse typed Reactors Considering WOG2000 RCP Seal Leakage Model

    Energy Technology Data Exchange (ETDEWEB)

    Na, Jang-Hwan; Jeon, Ho-Jun; Hwang, Seok-Won [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this paper, we focus on risk insights of Westinghouse typed reactors. We identified that Reactor Coolant Pump (RCP) seal integrity is the most important contributor to Core Damage Frequency (CDF). As we reflected the latest technical report; WCAP-15603(Rev. 1-A), 'WOG2000 RCP Seal Leakage Model for Westinghouse PWRs' instead of the old version, RCP seal integrity became more important to Westinghouse typed reactors. After Fukushima accidents, Korea Hydro and Nuclear Power (KHNP) decided to develop Low Power and Shutdown (LPSD) Probabilistic Safety Assessment (PSA) models and upgrade full power PSA models of all operating Nuclear Power Plants (NPPs). As for upgrading full power PSA models, we have tried to standardize the methodology of CCF (Common Cause Failure) and HRA (Human Reliability Analysis), which are the most influential factors to risk measures of NPPs. Also, we have reviewed and reflected the latest operating experiences, reliability data sources and technical methods to improve the quality of PSA models. KHNP has operating various types of reactors; Optimized Pressurized Reactor (OPR) 1000, CANDU, Framatome and Westinghouse. So, one of the most challengeable missions is to keep the balance of risk contributors of all types of reactors. This paper presents the method of new RCP seal leakage model and the sensitivity analysis results from applying the detailed method to PSA models of Westinghouse typed reference reactors. To perform the sensitivity analysis on LOCCW of the reference Westinghouse typed reactors, we reviewed WOG2000 RCP seal leakage model and developed the detailed event tree of LOCCW considering all scenarios of RCP seal failures. Also, we performed HRA based on the T/H analysis by using the leakage rates for each scenario. We could recognize that HRA was the sensitive contributor to CDF, and the RCP seal failure scenario of 182gpm leakage rate was estimated as the most important scenario.

  7. Dependence of the characteristics of bubbles on types of sonochemical reactors.

    Science.gov (United States)

    Yasui, Kyuichi; Tuziuti, Toru; Iida, Yasuo

    2005-01-01

    Computer simulations of bubble oscillations in liquid water irradiated by an ultrasonic wave have revealed that the characteristic of bubbles depends on types of sonochemical reactors: a horn-type reactor and a standing-wave type reactor. When the acoustic amplitude is large at 20 kHz, the bubble content is mostly water vapor even at the end of the bubble collapse and the temperature inside a bubble at the collapse is relatively low. On the other hand, when the acoustic amplitude is relatively low, the bubble content is mostly noncondensable gas at the end of the bubble collapse and the bubble temperature is relatively high. In a horn-type sonochemical reactor, the former type of bubbles are dominant because many bubbles exist near the horn-tip where the acoustic amplitude is large, while in a standing-wave type reactor the latter type of bubbles are dominant because the Bjerknes force gathers bubbles at a region where acoustic amplitude is relatively low.

  8. Modification of Neutron Kinetic Code for Plate Type Fuel Nuclear Reactor

    Directory of Open Access Journals (Sweden)

    Salah Ud-Din Khan

    2013-01-01

    Full Text Available The research is conducted on the modification of neutron kinetic code for the plate type fuel nuclear reactor. REMARK is a neutron kinetic code that works only for the cylindrical type fuel nuclear reactor. In this research, our main emphasis is on the modification of this code in order to be applicable for the plate type fuel nuclear reactor. For this purpose, detailed mathematical studies have been performed and are subjected to write the program in Fortran language. Since REMARK code is written in Fortran language, so we have developed the program in Fortran and then inserted it into the source library of the code. The main emphasis is on the modification of subroutine in the source library of the code for hexagonal fuel assemblies with plate type fuel elements in it. The number of steps involved in the modification of the code has been included in the paper. The verification studies were performed by considering the small modular reactor with hexagonal assemblies and plate type fuel in it to find out the power distribution of the reactor core. The purpose of the research is to make the code work for the hexagonal fuel assemblies with plate type fuel element.

  9. LMFBR type reactor core and its fuel exchange method

    Energy Technology Data Exchange (ETDEWEB)

    Ishibashi, Yoko; Koyama, Jun-ichi; Aoyama, Motoo; Haikawa, Katsumasa; Yamanaka, Akihiro

    1996-08-20

    Upon initial loading, two kinds of fuel assemblies including first fuel assemblies having a highest enrichment degree and second fuel assemblies having a lowest enrichment degree are loaded. The average fuel enrichment degree of an upper region of the first fuel assembly is made greater than that of the lower region. The reactivity of the lower region of the first fuel assembly is made lower than that of the upper portion to reduce power peak. Upon transfer from a first cycle to a second cycle, at least one of the second fuel assemblies is exchanged by the same number of the third fuel assemblies. In this case, an average fuel enrichment degree of the upper region of the third fuel assembly is made greater than that of the lower region to suppress the reactivity in the lower region of the third fuel assembly lower than the reactivity in the upper region thereby reducing the power peak. Thus, the upper power peak over the entire reactor core is moderated thereby capable of ensuring the reactor shut down margin without deteriorating the same. (N.H.)

  10. AGR-2 IRRADIATION TEST FINAL AS-RUN REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Blaise, Collin

    2014-07-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test

  11. SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

    Science.gov (United States)

    Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; Allen, Todd R.

    2016-11-01

    Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy's Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representative of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.

  12. Fast Traveling-Wave Reactor of the Channel Type

    CERN Document Server

    Rusov, Vitaliy D; Vashchenko, Volodymyr N; Chernezhenko, Sergei A; Kakaev, Andrei A; Pantak, Oksana I

    2015-01-01

    The main aim of this paper is to solve the technological problems of the TWR based on the technical concept described in our priority of invention reference, which makes it impossible, in particular, for the fuel claddings damaging doses of fast neutrons to excess the ~200 dpa limit. Thus the essence of the technical concept is to provide a given neutron flux at the fuel claddings by setting the appropriate speed of the fuel motion relative to the nuclear burning wave. The basic design of the fast uranium-plutonium nuclear traveling-wave reactor with a softened neutron spectrum is developed, which solves the problem of the radiation resistance of the fuel claddings material.

  13. The cefazolin inoculum effect in methicillin-susceptible Staphylococcus aureus blood isolates: their association with dysfunctional accessory gene regulator (agr).

    Science.gov (United States)

    Wi, Yu Mi; Park, Young Kyoung; Moon, Chisook; Ryu, Seong Yeol; Lee, Hyuck; Ki, Hyun Kyun; Cheong, Hae Suk; Son, Jun Seong; Lee, Jin Seo; Kwon, Ki Tae; Kim, June Myong; Ha, Young Eun; Kang, Cheol In; Ko, Kwan Soo; Chung, Doo Ryeon; Peck, Kyong Ran; Song, Jae-Hoon

    2015-11-01

    We evaluated the clinical significance of the cefazolin inoculum effect (CIE) in methicillin-susceptible Staphylococcus aureus (MSSA) isolates. In total, 146 isolates were recovered from patients with MSSA bacteremia at 9 hospitals in Korea. The CIE was observed in 16 MSSA isolates, and while type A was the only detected β-lactamase in MSSA isolates exhibiting the CIE, no strains expressing type B, C, or D β-lactamases exhibited this effect. The CIE was only observed in agr group III and I isolates and was significantly more common in isolates with agr dysfunction than in those with functional agr (P<0.001). Even among isolates producing type A β-lactamase, the CIE was also prevalent in isolates with dysfunctional agr than in isolates with functional agr (P=0.025). This study demonstrates an association between the CIE of MSSA isolates and agr dysfunction, in addition to those between the CIE and type A β-lactamase.

  14. AGR-2 Safety Test Predictions Using the PARFUME Code

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2014-09-01

    This report documents calculations performed to predict failure probability of TRISO-coated fuel particles and diffusion of fission products through these particles during safety tests following the second irradiation test of the Advanced Gas Reactor program (AGR-2). The calculations include the modeling of the AGR-2 irradiation that occurred from June 2010 to October 2013 in the Advanced Test Reactor (ATR) and the modeling of a safety testing phase to support safety tests planned at Oak Ridge National Laboratory and at Idaho National Laboratory (INL) for a selection of AGR-2 compacts. The heat-up of AGR-2 compacts is a critical component of the AGR-2 fuel performance evaluation, and its objectives are to identify the effect of accident test temperature, burnup, and irradiation temperature on the performance of the fuel at elevated temperature. Safety testing of compacts will be followed by detailed examinations of the fuel particles to further evaluate fission product retention and behavior of the kernel and coatings. The modeling was performed using the particle fuel model computer code PARFUME developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact). PARFUME calculates the

  15. AGR-2 safety test predictions using the PARFUME code

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    This report documents calculations performed to predict failure probability of TRISO-coated fuel particles and diffusion of fission products through these particles during safety tests following the second irradiation test of the Advanced Gas Reactor program (AGR-2). The calculations include the modeling of the AGR-2 irradiation that occurred from June 2010 to October 2013 in the Advanced Test Reactor (ATR) and the modeling of a safety testing phase to support safety tests planned at Oak Ridge National Laboratory and at Idaho National Laboratory (INL) for a selection of AGR-2 compacts. The heat-up of AGR-2 compacts is a critical component of the AGR-2 fuel performance evaluation, and its objectives are to identify the effect of accident test temperature, burnup, and irradiation temperature on the performance of the fuel at elevated temperature. Safety testing of compacts will be followed by detailed examinations of the fuel particles to further evaluate fission product retention and behavior of the kernel and coatings. The modeling was performed using the particle fuel model computer code PARFUME developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact). PARFUME calculates the

  16. Study on the selection of nuclear fuel type for a hybrid power extraction reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Dong Han; Park, Won Suk [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    The development of a subcritical transmutation reactor concept is emerging for reducing the amounts of actinides and long-lived nuclides in the spent fuel from nuclear power plants. This technology may make contribution to reduce the human risks associated with constructing radio-waste disposal facilities. One of the important issues for the design of the reactor is the selection of a suitable nuclear fuel type. Choosing the best nuclear fuel type for the reactor may not be easy since there exist several criteria associated with neutronic aspects, thermal performance, safety problem, cost problem, radiation damage in the reactor, etc. The best option should be chosen based on the maximization of our needs in this situation. This study presents a logical decision model for this issue using an analytic hierarchy process (AHP). Hierarchy is a representation of a system to study the functional relations of its components and its impact on the entire system. The study shows first how to construct hierarchy representing their relations and then measure the individual element's impact to the entire system for a quantitative decision making. Current four fuel types; metal, oxide, molten salt, and nitride, were selected and analyzed based on several characteristics with respect to overall comparison. Based on the decision model developed, the study concludes that the metal fuel type is the best choice for the transmutation reactor. The proposed approach is intended to help people be rational and logical in making decisions such complex task. 13 refs., 16 figs., 16 tabs. (Author)

  17. Study on the selection of nuclear fuel type for a hybrid power extration reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, D. H.; Park, W. S. [KAERI, Taejon (Korea, Republic of)

    1999-05-01

    In order to solve the problem related to long-lived radioactive nuclides in spent fuel, development of a subcritical transmutation reactor concept is emerging. One of the important issues for the design of the reactor may be the selection of a suitable nuclear fuel type. This study presents a logical decision model for this issue using an analytic hierachy process (AHP). Hierarchy is a representation of a system to study the functional relations of its components and its impact on the entire system. The study shows first how to construct hierachy representing their relations and then measure the individual element's impact to the entire system for a quantitative decision making. Current four fuel types; metal, oxide, molten salt, and nitride, were selected and analyzed based on several characteristics with respect to overall comparison. Based on the decision model, the study concludes that the metal fuel type is the best choice for the transmutation reactor.

  18. Development of a Monolithic Research Reactor Fuel Type at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Clark, C.R.; Briggs, R.J.

    2004-10-06

    The Reduced Enrichment for Research and Test Reactors (RERTR) program has been tasked with the conversion of research reactors from highly enriched to low-enriched uranium (LEU). To convert several high power reactors, monolithic fuel, a new fuel type, is being developed. This fuel type replaces the standard fuel dispersion with a fuel alloy foil, which allows for fuel densities far in excess of that found in dispersion fuel. The single-piece fuel foil also contains a significantly lower interface area between the fuel and the aluminum in the plate than the standard fuel type, limiting the amount of detrimental fuel-aluminum interaction that can occur. Implementation of monolithic fuel is dependant on the development of a suitable fabrication method as traditional roll-bonding techniques are inadequate.

  19. Safety Testing of AGR-2 UCO Compacts 5-2-2, 2-2-2, and 5-4-1

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fusion and Materials for Nuclear Systems Division; Morris, Robert Noel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fusion and Materials for Nuclear Systems Division; Baldwin, Charles A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fusion and Materials for Nuclear Systems Division; Montgomery, Fred C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fusion and Materials for Nuclear Systems Division

    2016-08-01

    Post-irradiation examination (PIE) is being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2). This effort builds upon the understanding acquired throughout the AGR-1 PIE campaign, and is establishing a database for the different AGR-2 fuel designs. The AGR-2 irradiation experiment included TRISO fuel particles coated at BWX Technologies (BWXT) with a 150-mm-diameter engineering-scale coater. Two coating batches were tested in the AGR-2 irradiation experiment. Batch 93085 had 508-μm-diameter uranium dioxide (UO2) kernels. Batch 93073 had 427-μm-diameter UCO kernels, which is a kernel design where some of the uranium oxide is converted to uranium carbide during fabrication to provide a getter for oxygen liberated during fission and limit CO production. Fabrication and property data for the AGR-2 coating batches have been compiled and compared to those for AGR-1. The AGR-2 TRISO coatings were most like the AGR-1 Variant 3 TRISO deposited in the 50-mm-diameter ORNL lab-scale coater. In both cases argon-dilution of the hydrogen and methyltrichlorosilane coating gas mixture employed to deposit the SiC was used to produce a finer-grain, more equiaxed SiC microstructure. In addition to the fact that AGR-1 fuel had smaller, 350-μm-diameter UCO kernels, notable differences in the TRISO particle properties included the pyrocarbon anisotropy, which was slightly higher in the particles coated in the engineering-scale coater, and the exposed kernel defect fraction, which was higher for AGR-2 fuel due to the detected presence of particles with impact damage introduced during TRISO particle handling.

  20. Safety Testing of AGR-2 UCO Compacts 5-2-2, 2-2-2, and 5-4-1

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D [ORNL; Morris, Robert Noel [ORNL; Baldwin, Charles A [ORNL; Montgomery, Fred C [ORNL

    2016-10-01

    Post-irradiation examination (PIE) is being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2) [Collin 2014]. This effort builds upon the understanding acquired throughout the AGR-1 PIE campaign [Demkowicz et al. 2015a] and is establishing a database for the different AGR-2 fuel designs. The AGR-2 irradiation experiment included TRISO fuel particles coated at BWX Technologies (BWXT) with a 150-mm-diameter engineering-scale coater. Two coating batches were tested in the AGR-2 irradiation experiment. Batch 93085 had 508-μm-diameter uranium dioxide (UO2) kernels. Batch 93073 had 427-μm-diameter UCO kernels, which is a kernel design where some of the uranium oxide is converted to uranium carbide during fabrication to provide a getter for oxygen liberated during fission and limit CO production. Fabrication and property data for the AGR-2 coating batches have been compiled [Barnes and Marshall 2009] and compared to AGR-1 [Phillips, Barnes, and Hunn 2010]. The AGR-2 TRISO coatings were most like the AGR-1 Variant 3 TRISO deposited in the 50-mm-diameter ORNL lab-scale coater [Hunn and Lowden 2006]. In both cases argon-dilution of the hydrogen and methyltrichlorosilane coating gas mixture employed to deposit the SiC was used to produce a finer-grain, more equiaxed SiC microstructure [Lowden 2006; Gerczak et al. 2016]. In addition to the fact that AGR-1 fuel had smaller, 350-μm-diameter UCO kernels, notable differences in the TRISO particle properties included the pyrocarbon anisotropy, which was slightly higher in the particles coated in the engineeringscale coater, and the exposed kernel defect fraction, which was higher for AGR-2 fuel due to the detected presence of particles with impact damage introduced during TRISO particle handling [Hunn 2010].

  1. Operational characteristics analysis of a 8 mH class HTS DC reactor for an LCC type HVDC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. K.; Go, B. S.; Dinh, M. C.; Park, M.; Yu, I. K. [Changwon National University, Changwon (Korea, Republic of); Kim, J. H. [Daejeon University, Daejeon (Korea, Republic of)

    2015-03-15

    Many kinds of high temperature superconducting (HTS) devices are being developed due to its several advantages. In particular, the advantages of HTS devices are maximized under the DC condition. A line commutated converter (LCC) type high voltage direct current (HVDC) transmission system requires large capacity of DC reactors to protect the converters from faults. However, conventional DC reactor made of copper causes a lot of electrical losses. Thus, it is being attempted to apply the HTS DC reactor to an HVDC transmission system. The authors have developed a 8 mH class HTS DC reactor and a model-sized LCC type HVDC system. The HTS DC reactor was operated to analyze its operational characteristics in connection with the HVDC system. The voltage at both ends of the HTS DC reactor was measured to investigate the stability of the reactor. The voltages and currents at the AC and DC side of the system were measured to confirm the influence of the HTS DC reactor on the system. Two 5 mH copper DC reactors were connected to the HVDC system and investigated to compare the operational characteristics. In this paper, the operational characteristics of the HVDC system with the HTS DC reactor according to firing angle are described. The voltage and current characteristics of the system according to the types of DC reactors and harmonic characteristics are analyzed. Through the results, the applicability of an HTS DC reactor in an HVDC system is confirmed.

  2. Use of Stable Noble Gases as a Predictor of Reactor Fuel Type and Exposure

    Energy Technology Data Exchange (ETDEWEB)

    Fearey, B.L.; Charlton, W.S.; Perry, R.T.; Poths, J.; Wilson, W.B.; Hemberger, P.H.; Nakhleh, C.W.; Stanbro, W.D.

    1999-08-30

    Ensuring spent reactor fuel is not produced to provide weapons-grade plutonium is becoming a major concern as many countries resort to nuclear power as a solution to their energy problems. Proposed solutions range from the development of proliferation resistant fuel to continuous monitoring of the fuel. This paper discusses the use of the stable isotopes of the fissiogenic noble gases, xenon and krypton, for determining the burnup characteristics, fuel type, and the reactor type of the fuel from which the sample was obtained. The gases would be collected on-stack as the fuel is reprocessed, and thus confirm that the fuel is as declared.

  3. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  4. Cancer-secreted AGR2 induces programmed cell death in normal cells

    Science.gov (United States)

    Vitello, Elizabeth A.; Quek, Sue-Ing; Kincaid, Heather; Fuchs, Thomas; Crichton, Daniel J.; Troisch, Pamela; Liu, Alvin Y.

    2016-01-01

    Anterior Gradient 2 (AGR2) is a protein expressed in many solid tumor types including prostate, pancreatic, breast and lung. AGR2 functions as a protein disulfide isomerase in the endoplasmic reticulum. However, AGR2 is secreted by cancer cells that overexpress this molecule. Secretion of AGR2 was also found in salamander limb regeneration. Due to its ubiquity, tumor secretion of AGR2 must serve an important role in cancer, yet its molecular function is largely unknown. This study examined the effect of cancer-secreted AGR2 on normal cells. Prostate stromal cells were cultured, and tissue digestion media containing AGR2 prepared from prostate primary cancer 10-076 CP and adenocarcinoma LuCaP 70CR xenograft were added. The control were tissue digestion media containing no AGR2 prepared from benign prostate 10-076 NP and small cell carcinoma LuCaP 145.1 xenograft. In the presence of tumor-secreted AGR2, the stromal cells were found to undergo programmed cell death (PCD) characterized by formation of cellular blebs, cell shrinkage, and DNA fragmentation as seen when the stromal cells were UV irradiated or treated by a pro-apoptotic drug. PCD could be prevented with the addition of the monoclonal AGR2-neutralizing antibody P3A5. DNA microarray analysis of LuCaP 70CR media-treated vs. LuCaP 145.1 media-treated cells showed downregulation of the gene SAT1 as a major change in cells exposed to AGR2. RT-PCR analysis confirmed the array result. SAT1 encodes spermidine/spermine N1-acetyltransferase, which maintains intracellular polyamine levels. Abnormal polyamine metabolism as a result of altered SAT1 activity has an adverse effect on cells through the induction of PCD. PMID:27283903

  5. A comparative study of the attenuation of reactor thermal neutrons in different types of concrete

    Energy Technology Data Exchange (ETDEWEB)

    Bashiter, I.I. [Zagazig Univ. (Egypt). Dept. of Physics; El-Sayed Abdo, A.; Makarious, A.S. [Atomic Energy Authority, Cairo (Egypt). Nuclear Research Centre

    1996-05-20

    This study was carried out to assess the distribution of thermal neutrons emitted directly from the core of the ET-RR-1 reactor in ordinary concrete, ilmenite concrete and ilmenite-limonite concrete shields. Measurements were carried out by using a direct beam and a cadmium filtered beam of reactor neutrons. The neutron dose distributions were measured using Li{sub 2}B{sub 4}O{sub 7}:Mn thermoluminescent dosimeters. The data obtained show that ilmenite concrete is better for slow and thermal neutron attenuation than both ordinary and ilmenite-limonite concrete. Also it was concluded that thermal neutrons emitted directly from the reactor core are highly absorbed within the first few centimeters of each type of concrete. The thickness of ilmenite concrete required to attenuate the doses of neutrons to a certain value along the beam axis for a direct reactor beam estimated to be about 75 and 57% of the shield thickness made from ordinary and ilmenite-limonite concretes, respectively. Empirical formulae were derived to calculate the neutron dose distribution in ordinary, ilmenite and ilmenite-limonite concrete shields both along and perpendicular to the beam axis for both the direct reactor neutrons and the reactor thermal neutrons. (author).

  6. Theoretical study for ICRF sustained LHD type p-{sup 11}B reactor

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Tsuguhiro (ed.)

    2003-04-01

    This is a summary of the workshop on 'Theoretical Study for ICRF Sustained LHD Type p-{sup 11}B Reactor' held in National Institute for Fusion Science (NIFS) on July 25, 2002. In the workshop, study of LHD type D-{sup 3}He reactor is also reported. A review concerning the advanced nuclear fusion fuels is also attached. This review was reported at the workshop of last year. The development of the p-{sup 11}B reactor research which uses the LHD magnetic field configuration has been briefly summarized in section 1. In section 2, an integrated report on advanced nuclear fusion fuels is given. Ignition conditions in a D-{sup 3}He helical reactor are summarized in section 3. 0-dimensional particle and power balance equations are solved numerically assuming the ISS95 confinement law including a confinement factor ({gamma}{sub HH}). It is shown that high average beta plasma confinement, a large confinement factor ({gamma}{sub HH} > 3) and the hot ion mode (T{sub i}/T{sub e} > 1.4) are necessary to achieve the ignition of the D-{sup 3}He helical reactor. Characteristics of ICRF sustained p-{sup 11}B reactor are analyzed in section 4. The nuclear fusion reaction rate < {sigma}{upsilon} > is derived assuming a quasilinear plateau distribution function (QPDF) for protons, and an ignition condition of p-{sup 11}B reactor is shown to be possible. The 3 of the presented papers are indexed individually. (J.P.N.)

  7. NMR microimaging of fluid flow in model string-type reactors

    NARCIS (Netherlands)

    Koptyug, I.V.; Kovtunov, K.V.; Gerkema, E.; Kiwi-Minskerc, L.; Sagdeev, R.Z.

    2007-01-01

    Magnetic resonance microimaging (MRM) was employed to obtain quantitative velocity maps of water flowing in the channels possessing unconventional cross-section shapes formed by a bundle of parallel fibers within a tubular string-type reactor. The maps obtained demonstrate the presence of large amou

  8. Optimized core design and fuel management of a pebble-bed type nuclear reactor

    NARCIS (Netherlands)

    Boer, B.

    2009-01-01

    The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the si

  9. PIE on Safety-Tested AGR-1 Compact 5-1-1

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Morris, Robert Noel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Baldwin, Charles A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Fred C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gerczak, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

    Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High-Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energy's Office of Nuclear Energy as part of the Advanced Reactor Technologies (ART) initiative. AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for three years in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to examine fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing of irradiated AGR-1 Compact 5-1-1 in the ORNL Core Conduction Cooldown Test Facility (CCCTF) and post-safety testing PIE.

  10. Effects of reactor type and mass transfer on the morphology of CuS and ZnS crystals

    NARCIS (Netherlands)

    Al-Tarazi, Mousa; Heesink, A. Bert M.; Versteeg, Geert F.

    2005-01-01

    For the precipitation of CuS and ZnS, the effects of the reactor/precipitator type, mass transfer and process conditions on crystal morphology were studied. Either H2S gas or a S2- solution were applied. Three different types of reactors have been tested, namely a laminar jet, a bubble column and an

  11. Power regulating range broadening of the WWER-type reactor power units

    Energy Technology Data Exchange (ETDEWEB)

    Dement' ev, B.A.; Petrov, V.A.; Proskuryakov, A.G.; Puchkov, V.V. (Moskovskij Ehnergeticheskij Inst. (USSR))

    1984-02-01

    Calculational studies on the use of sliding pressure (SP) regime to expand the regulating range of the WWER-440 reactor power units are presented. Two operation regimes of a power unit have been considered: according to weekly and daily load swings in electrical grids. The conclusion is made that the use of SP regime in the secondary circuit improves manoeuvable characterstics of the power unit in the second half of operating cycle. T of the reactor (0.6 reactor. Besides, the use of SP regime during power unit operation with decreased loadings is the more efficient the smaller is the load. The range of operating cycle 0.8 <= T <= 1 makes the greatest contribution to regulating range broadening as a result of SP regime use. Conclusions of the calculational studies can be also applied to WWER reactors of other types as well as to RBMK reactors.

  12. SVBR-100 module-type fast reactor of the IV generation for regional power industry

    Science.gov (United States)

    Zrodnikov, A. V.; Toshinsky, G. I.; Komlev, O. G.; Stepanov, V. S.; Klimov, N. N.

    2011-08-01

    In the report the following is presented: basic conceptual provisions of the innovative nuclear power technology (NPT) based on modular fast reactors (FR) SVBR-100, summarized results of calculations of the reactor, analysis of the opportunities of multi-purpose application of such reactor facilities (RF) including export potentials with due account of nonproliferation requirements. The most important features of the proposed NPT analyzed in the report are as follows: (1) integral (monoblock) arrangement of the primary circuit equipment with entire elimination of the primary circuit pipelines and valves that considerably reduces the construction and assembly works period and coupling with high boiling point of lead-bismuth coolant (LBC) deterministically eliminates accidents of the LOCA type, (2) option for 100 MWe power and dimensions of the reactor provide: on the one hand, an opportunity to transport the reactor monoblock in factory-readiness by railway as well as other kinds of transport, on the other hand, core breeding ratio (CBR) exceeds 1 while MOX-fuel is used. The preferable area of application of RF SVBR-100 is regional and small power requiring power-units of electric power in a range of (100-600) MW, which could be used for cogeneration-based district heating while locating them nearby cities as well as for generation of electric power in a mode of load tracking in the regions with low network systems.

  13. Data Compilation for AGR-1 Variant 3 Compact Lot LEU01-49T-Z

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D [ORNL; Montgomery, Fred C [ORNL; Pappano, Peter J [ORNL

    2006-08-01

    This document is a compilation of characterization data for the AGR-1 vriant 3 fuel compact lot LEU01-49T-Z. The compacts were produced by ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program for the first AGR irradiation test train (AGR-1). This compact lot was fabricated using particle composite LEU01-49T, which was a composite of three batches of TRISO-coated 350 {micro}m diameter 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with an {approx} 50% dense carbon buffer layer (100 {micro}m nominal thickness), followed by a dense inner pyrocarbon layer (40 {micro}m nominal thickness), followed by a SiC layer (35 {micro}m nominal thickness), followed by another dense outer pyrocarbon layer (40 {micro}m nominal thickness). The kernels were obtained from BWXT and identified as composite G73D-20-69302. The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified as LEU01-?? (where ?? is a series of integers beginning with 01). A data compilation for the AGR-1 variant 3 coated particle composite LEU01-49t CAN BE FOUND IN ornl/tm-2006/022.

  14. Uncertainty Quantification of Calculated Temperatures for the AGR 3/4 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh Thi-Cam [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN 3636, “Technical Program Plan for INL Advanced Reactor Technologies Technology Development Office/Advanced Gas Reactor Fuel Development and Qualification Program”). The AGR 3/4 test was inserted in the Northeast Flux Trap position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in December 2011 and successfully completed irradiation in mid-April 2014, resulting in irradiation of the tristructural isotropic (TRISO) fuel for 369.1 effective full-power days (EFPDs) during approximately 2.4 calendar years. The AGR 3/4 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as run thermal analysis has been performed separately on each of twelve AGR 3/4 capsules for the entire irradiation as discussed in ECAR-2807, “AGR 3/4 Daily As Run Thermal Analyses”. The ABAQUS code’s finite element-based thermal model predicts the daily average volume average (VA) fuel temperature (FT), peak FT, and graphite matrix, sleeve, and sink temperature in each capsule. The JMOCUP simulation codes were also created to perform depletion calculations for the AGR 3/4 experiment (ECAR-2753, “JMOCUP As-Run Daily Physics Depletion Calculation for the AGR 3/4 TRISO Particle Experiment in ATR

  15. SPARC fast reactor design : Design of two passively safe metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control

    OpenAIRE

    Lindström, Tobias

    2015-01-01

    In this master thesis a small modular sodium-cooled metal-fuelled pool-type fast reactor design, called SPARC - Safe and Passive with Autonomous Reactivity control, has been designed. The long term reactivity changes in the SPARC are managed by implementation of the the Autonomous Reactivity Control (ARC) system, which is the novelty of the design. The overall design is mainly based on the Integral Fast Reactor project (IFR), which experimentally demonstrated the passive safety characteristic...

  16. AGR 3/4 Irradiation Test Final As Run Report

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-06-01

    Several fuel and material irradiation experiments have been planned for the Idaho National Laboratory Advanced Reactor Technologies Technology Development Office Advanced Gas Reactor Fuel Development and Qualification Program (referred to as the INL ART TDO/AGR fuel program hereafter), which supports the development and qualification of tristructural-isotropic (TRISO) coated particle fuel for use in HTGRs. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development and validation of fuel performance and fission product transport models and codes, and provide irradiated fuel and materials for post irradiation examination and safety testing (INL 05/2015). AGR-3/4 combined the third and fourth in this series of planned experiments to test TRISO coated low enriched uranium (LEU) oxycarbide fuel. This combined experiment was intended to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport by irradiating designed-to-fail fuel particles and by measuring subsequent fission metal transport in fuel-compact matrix material and fuel-element graphite. The AGR 3/4 fuel test was successful in irradiating the fuel compacts to the burnup and fast fluence target ranges, considering the experiment was terminated short of its initial 400 EFPD target (Collin 2015). Out of the 48 AGR-3/4 compacts, 42 achieved the specified burnup of at least 6% fissions per initial heavy-metal atom (FIMA). Three capsules had a maximum fuel compact average burnup < 10% FIMA, one more than originally specified, and the maximum fuel compact average burnup was <19% FIMA for the remaining capsules, as specified. Fast neutron fluence fell in the expected range of 1.0 to 5.5×1025 n/m2 (E >0.18 MeV) for all compacts. In addition, the AGR-3/4 experiment was globally successful in keeping the

  17. The use of experimental data in an MTR-type nuclear reactor safety analysis

    Science.gov (United States)

    Day, Simon E.

    Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

  18. Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tavera D, L.; Camacho L, M.E

    1991-01-15

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  19. Hydrogen energy recovery from high strength organic wastewater with ethanol type fermentation using acidogenic EGSB reactor

    Institute of Scientific and Technical Information of China (English)

    REN Nan-qi; GUO Wan-qian; WANG Xiang-jing; ZHANG Lu-si

    2005-01-01

    A lab-scale expanded granular sludge bed (EGSB) reactor was employed to evaluate the feasibility of the hydrogen energy recovery potential from high strength organic wastewater. The results showed that a maxioperation. At the acidogenic phase, COD removal rate was stable at about 15%. In the steady operation peri od, the main liquid end products were ethanol and acetic acid, which represented ethanol type fermentation. Among the liquid end products, the concentration percentage of ethanol and acetic acid amounted to 69.5% ~89.8% and the concentration percentage of ethanol took prominent about 51.7% ~ 59.1%, which is better than the utilization of substrate for the methanogenic bacteria. An ethanol type fermentation pathway was suggested in the operation of enlarged industrial continuous hydrogen bio-producing reactors.

  20. Study on Thermal-Hydraulic Behavior of an Integral Type Reactor under Heaving Condition

    Directory of Open Access Journals (Sweden)

    Beibei Feng

    2014-01-01

    Full Text Available A self-developed program was used to study the thermal-hydraulic behavior of an integral type reactor under heaving condition. Comparison of calculated results with the data of experiments performed on a natural circulation loop designed with reference to an integral type reactor of Tsinghua University in inclination, heaving, and rolling motions was carried out. Characteristics of natural circulation in heaving motion and effect of motion parameters on natural circulation were investigated. Results indicated that: (1 long-period heaving motion would lead to more significant influence than inclination and rolling motion; (2 it was an alternating force field which consisted of gravity and an additional force that decided the flow temperature and density difference of natural circulation; (3 effect of strength k and cycle T of heaving motion on flow fluctuation of natural circulation and condensate depression of heating section outlet was performed.

  1. AGR-2 Data Qualification Report for ATR Cycles 147A, 148A, 148B, and 149A

    Energy Technology Data Exchange (ETDEWEB)

    Michael L. Abbott; Keith A. Daum

    2011-08-01

    This report presents the data qualification status of fuel irradiation data from the first four reactor cycles (147A, 148A, 148B, and 149A) of the on-going second Advanced Gas Reactor (AGR-2) experiment as recorded in the NGNP Data Management and Analysis System (NDMAS). This includes data received by NDMAS from the period June 22, 2010 through May 21, 2011. AGR-2 is the second in a series of eight planned irradiation experiments for the AGR Fuel Development and Qualification Program, which supports development of the very high temperature gas-cooled reactor (VHTR) under the Next Generation Nuclear Plant (NGNP) Project. Irradiation of the AGR-2 test train is being performed at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) and is planned for 600 effective full power days (approximately 2.75 calendar years) (PLN-3798). The experiment is intended to demonstrate the performance of UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Data qualification status of the AGR-1 experiment was reported in INL/EXT-10-17943 (Abbott et al. 2010).

  2. A Study of the Energy Efficiency of Hadronic Reactors of Molecular Type

    OpenAIRE

    Aringazin, A. K.; Santilli, R. M.

    2001-01-01

    In this paper, we introduce an estimate of the "commercial efficiency" of Santilli's hadronic reactors of molecular type (Patented and International Patents Pending) which convert a liquid feedstock (such as automotive antifreeze and oil waste, city or farm liquid waste, crude oil, etc.) into the clean burning magnegas plus heat acquired by the liquid feedstock. The "commercial efficiency" is defined as the ratio between the total energy output (energy in magnegas plus heat) and the electric ...

  3. Performance of static var compensator control type thyristor controlled reactor and thyristor switched capacitor

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Josias M. de; Yung, Chou Shaw; Rose, Eber H.; Pantoja, Antonio L.A. [ELETRONORTE, Belem, PA (Brazil); Fouesnant, Thomas; Boissier, Luc

    1994-12-31

    This paper has the objective of presenting the philosophy of Static Var Compensator (SVC) Control as well the necessary adjustments in the project of control system to guarantee suitable performance under different operating conditions. The verification on the performance of the SVC control has been done by Transient Network Analyzer (TNA/CEPEL) studies, commissioning tests and a factory tests. The SVC is the type of Thyristor Controlled Reactor (TCR) and Thyristor Switched Capacitor (TSC). (author) 3 refs., 12 figs.

  4. Dismantling design for a reference research reactor of the WWR type

    Energy Technology Data Exchange (ETDEWEB)

    Lobach, Yu.N., E-mail: lobach@kinr.kiev.ua [Institute for Nuclear Research, Pr. Nauki, 47, Kiev 03680 (Ukraine); Cross, M.T., E-mail: Martin.Cross@nuvia.co.uk [Nuvia Ltd., Robinson House, Crow Park Way, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3HY (United Kingdom)

    2014-01-15

    Highlights: • Design features of WWRs relevant to decommissioning have been analysed. • The technical basis for the preparation and implementation of dismantling has been established for a reference WWR. • The applicability of existing proven dismantling technologies has been established. -- Abstract: A decommissioning study has been carried out for a reference research reactor of the WWR type. Many such reactors were constructed more than 50 years ago and most of them are still in operation. Decommissioning has now become an important consideration. This paper summarizes the main decommissioning steps and, on the basis of the reactor design features, technical aspects of the dismantling and removal of the contaminated/activated components have been analysed. The advisability of the removal of large components, such as the reactor vessel and the heat-exchangers, as one piece items has also been demonstrated. Additionally, a work schedule and an estimation of the collective dose for the preparation and implementation of dismantling have been established. The applicability of existing proven dismantling technologies has been identified together with some additional features for the dismantling.

  5. A High Operability Supervisory Digital System for TRIGA-Type Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Aronica, O.; Bove, R.; Cappelli, M.; Falconi, L.; Palomba, M.; Santoro, E.; Sepielli, M. [ENEA, UTFISST, Casaccia Research Center, Via Anguillarese, 301 Rome (Italy); Memmi, F. [University of Rome ' Roma Tre' , Department of Electrical Engineering, Via della Vasca Navale, 84 Rome (Italy)

    2011-07-01

    In this work, we propose an outline of a monitoring system to supervise variables coming from a fission nuclear reactor of TRIGA type (1-MW TRIGA reactor RC-1). The system can interface the control room instrumentation and can display the characteristic parameters (e.g. nuclear power, temperatures, flow rates, radiological parameters) in an intuitive, user-friendly way for plant operators. This aim is achieved using the Labview development environment. A front panel of a virtual instrument allows for a direct measure and a check that would not be possible by only reading the output data coming from the instruments of the control room, because of their standards and strict safety regulations. The acquisition system, for signals coming from the reactor, can process data and generate a detailed representation of the results. Statistics resulting from data analysis will be interpreted to optimize reactor management parameters. This system also includes a simulation tool to predict specific performances and investigate critical phenomena, or to optimize overall plant performances. In particular, it allows to have a feedback control and to perform predictive statistical surveys of all main process parameters. (author)

  6. Quantity of 135I released from the AGR-1, AGR-2, and AGR-3/4 experiments and discovery of 131I at the FPMS traps during the AGR-3/4 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Scates, Dawn M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    A series of three Advanced Gas Reactor (AGR) experiments have been conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). From 2006 through 2014, these experiments supported the development and qualification of the new U.S. tristructural isotropic (TRISO) particle fuel for Very High Temperature Reactors (VHTR). Each AGR experiment consisted of multiple fueled capsules, each plumbed for independent temperature control using a mix of helium and neon gases. The gas leaving a capsule was routed to individual Fission Product Monitor (FPM) detectors. For intact fuel particles, the TRISO particle coatings provide a substantial barrier to fission product release. However, particles with failed coatings, whether because of a minute percentage of initially defective particles, those which fail during irradiation, or those designed to fail (DTF) particles, can release fission products to the flowing gas stream. Because reactive fission product elements like iodine and cesium quickly deposit on cooler capsule components and piping structures as the effluent gas leaves the reactor core, only the noble fission gas isotopes of Kr and Xe tend to reach FPM detectors. The FPM system utilizes High Purity Germanium (HPGe) detectors coupled with a thallium activated sodium iodide NaI(Tl) scintillator. The HPGe detector provides individual isotopic information, while the NaI(Tl) scintillator is used as a gross count rate meter. During irradiation, the 135mXe concentration reaching the FPM detectors is from both direct fission and by decay of the accumulated 135I. About 2.5 hours after irradiation (ten 15.3 minute 135mXe half lives) the directly produced 135mXe has decayed and only the longer lived 135I remains as a source. Decay systematics dictate that 135mXe will be in secular equilibrium with its 135I parent, such that its production rate very nearly equals the decay rate of the

  7. Synthesis of layered double hydroxide nanosheets by coprecipitation using a T-type microchannel reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pang, Xiujiang; Sun, Meiyu; Ma, Xiuming [State Key Laboratory Base of Eco-chemical Engineering, College of Chemistry and Molecular Engineering, Qingdao University of Science and Technology, Qingdao 266042 (China); Hou, Wanguo, E-mail: wghou@sdu.edu.cn [Key Laboratory of Colloid and Interface Chemistry (Ministry of Education), Shandong University, Jinan 250100 (China)

    2014-02-15

    The synthesis of Mg{sub 2}Al–NO{sub 3} layered double hydroxide (LDH) nanosheets by coprecipitation using a T-type microchannel reactor is reported. Aqueous LDH nanosheet dispersions were obtained. The LDH nanosheets were characterized by X-ray diffraction, transmission electron microscopy, atomic force microscopy and particle size analysis, and the transmittance and viscosity of LDH nanosheet dispersions were examined. The two-dimensional LDH nanosheets consisted of 1–2 brucite-like layers and were stable for ca. 16 h at room temperature. In addition, the co-assembly between LDH nanosheets and dodecyl sulfate (DS) anions was carried out, and a DS intercalated LDH nanohybrid was obtained. To the best of our knowledge, this is the first report of LDH nanosheets being directly prepared in bulk aqueous solution. This simple, cheap method can provide naked LDH nanosheets in high quantities, which can be used as building blocks for functional materials. - Graphical abstract: Layered double hydroxide (LDH) nanosheets were synthesized by coprecipitation using a T-type microchannel reactor, and could be used as basic building blocks for LDH-based functional materials. Display Omitted - Highlights: • LDH nanosheets were synthesized by coprecipitation using a T-type microchannel reactor. • Naked LDH nanosheets were dispersed in aqueous media. • LDH nanosheets can be used as building blocks for functional materials.

  8. Deployable nuclear fleet based on available quantities of uranium and reactor types – the case of fast reactors started up with enriched uranium

    Directory of Open Access Journals (Sweden)

    Baschwitz Anne

    2016-01-01

    Full Text Available International organizations regularly produce global energy demand scenarios. To account for the increasing population and GDP trends, as well as to encompass evolving energy uses while satisfying constraints on greenhouse gas emissions, long-term installed nuclear power capacity scenarios tend to be more ambitious, even after the Fukushima accident. Thus, the amounts of uranium or plutonium needed to deploy such capacities could be limiting factors. This study first considers light-water reactors (LWR, GEN III using enriched uranium, like most of the current reactor technologies. It then examines the contribution of future fast reactors (FR, GEN IV operating with an initial fissile load and then using depleted uranium and recycling their own plutonium. However, as plutonium is only available in limited quantity since it is only produced in nuclear reactors, the possibility of starting up these Generation IV reactors with a fissile load of enriched uranium is also explored. In one of our previous studies, the uranium consumption of a third-generation reactor like an EPR™ was compared with that of a fast reactor started up with enriched uranium (U5-FR. For a reactor lifespan of 60 years, the U5-FR consumes three times less uranium than the EPR and represents a 60% reduction in terms of separative work units (SWU, though its requirements are concentrated over the first few years of operation. The purpose of this study is to investigate the relevance of U5-FRs in a nuclear fleet deployment configuration. Considering several power demand scenarios and assuming different finite quantities of available natural uranium, this paper examines what types of reactors must be deployed to meet the demand. The deployment of light-water reactors only is not sustainable in the long run. Generation IV reactors are therefore essential. Yet when started up with plutonium, the number of reactors that can be deployed is also limited. In a fleet deployment

  9. Development of core fuel management code system for WWER-type reactors

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    In this article, a core fuel management program for hexagonal pressurized water type WWER reactors (CFMHEX) has been developed, which is based on advanced three-dimensional nodal method and integrated with thermal hydraulic code to realize the coupling of neutronics and thermal-hydraulics. In CFMHEX, all these feedback effects such as burnup, power distribution, moderator density, and control rod insertion are considered. The verification and validation of the code system have been examined through the IAEA WWER-1000-type Kalinin NPP benchmark problem. The numerical results are in good agreement with measurements and are close to those of other international institutes.

  10. Computational simulation of fuel burnup estimation for research reactors plate type

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos, E-mail: nadiasam@gmail.com [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The aim of this study is to estimate the spatial fuel burnup, through computational simulation, in two research reactors plate type, loaded with dispersion fuel: the benchmark Material Test Research - International Atomic Energy Agency (MTR-IAEA) and a typical multipurpose reactor (MR). The first composed of plates with uranium oxide dispersed in aluminum (UAlx-Al) and a second composed with uranium silicide (U{sub 3}Si{sub 2}) dispersed in aluminum. To develop this work we used the deterministic code, WIMSD-5B, which performs the cell calculation solving the neutron transport equation, and the DF3DQ code, written in FORTRAN, which solves the three-dimensional neutron diffusion equation using the finite difference method. The methodology used was adequate to estimate the spatial fuel burnup , as the results was in accordance with chosen benchmark, given satisfactorily to the proposal presented in this work, even showing the possibility to be applied to other research reactors. For future work are suggested simulations with other WIMS libraries, other settings core and fuel types. Comparisons the WIMSD-5B results with programs often employed in fuel burnup calculations and also others commercial programs, are suggested too. Another proposal is to estimate the fuel burnup, taking into account the thermohydraulics parameters and the Xenon production. (author)

  11. Assessing optimal fermentation type for bio-hydrogen production in continuous-flow acidogenic reactors.

    Science.gov (United States)

    Ren, N Q; Chua, H; Chan, S Y; Tsang, Y F; Wang, Y J; Sin, N

    2007-07-01

    In this study, the optimal fermentation type and the operating conditions of anaerobic process in continuous-flow acidogenic reactors was investigated for the maximization of bio-hydrogen production using mixed cultures. Butyric acid type fermentation occurred at pH>6, propionic acid type fermentation occurred at pH about 5.5 with E(h) (redox potential) >-278mV, and ethanol-type fermentation occurred at pHhydrogen production capacities between the fermentation types, which remained stable when the organic loading rate (OLR) reached the highest OLR at 86.1kgCOD/m(3)d. The maximum hydrogen production reached up to 14.99L/d.

  12. TASS/SMR code improvement for small break LOCA applicability at an integral type reactor, SMART

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Young-Jong, E-mail: chung@kaeri.re.kr; Kim, Soo-Hyung; Lim, Sung-Won; Bae, Kyoo-Hwan

    2015-12-15

    Highlights: • SMART adopts a passive system to enhance its safety. • TASS/SMR code is developed to analyze thermal hydraulic phenomena of the SMART plant. • Improved TASS/SMR code predicts well the results of the OSU-MASLWR total-loss-of-feedwater test. - Abstract: Small reactors are a suitable option for nuclear system deployment in developing countries or non-electrical applications for various facilities. SMART is one of the small integral type reactors to apply flexibly local power demands or sea water desalination. A thermal hydraulic analysis code, TASS/SMR, having SMART specific models, was developed to simulate thermal hydraulic phenomena of the SMART plant. The improved TASS/SMR code predicts well the system behaviors under two-phase conditions compared with the OSU-MASLWR experimental results. A small break LOCA simulation of the SMART plant is improved a void distribution, a break flow, and a collapsed water level in the core.

  13. PWR type reactors. Normal and accidental operation; Reacteurs a eau sous pression. Fonctionnement normal et accidentel

    Energy Technology Data Exchange (ETDEWEB)

    Petetrot, J.F. [AREVA NP, Dept. Fonctionnement Reacteur et Etudes d' Accidents/Division, Tour AREVA, 92 - Paris La Defense (France)

    2009-07-15

    This article presents the general operation principles of PWR type reactors with the limits to be respected for the core and the steam supply system. Regulation systems controlling the main parameters are described as well: measurements used, functional structures, controlled actuators. The protection system which can lead to the automatic shutdown of the reactor (emergency rod drop) and to the start-up of safeguard functions is detailed. The interface for the conventional protection system is briefly described. The operation of the steam supply system with respect to the power grid needs is presented in relation with the regulation of the turbogenerator set: load follow-up, primary and secondary adjustment. Finally, the changes of the most important parameters during typical transients are commented. The main operations needed to move from the cold shutdown state to the nominal power are described as well. (J.S.)

  14. A Study of the Energy Efficiency of Hadronic Reactors of Molecular Type

    CERN Document Server

    Aringazin, A K

    2001-01-01

    In this paper, we introduce an estimate of the "commercial efficiency" of Santilli's hadronic reactors of molecular type (Patented and International Patents Pending) which convert a liquid feedstock (such as automotive antifreeze and oil waste, city or farm liquid waste, crude oil, etc.) into the clean burning magnegas plus heat acquired by the liquid feedstock. The "commercial efficiency" is defined as the ratio between the total energy output (energy in magnegas plus heat) and the electric energy used for its production, while the "scientific efficiency" is the usual ratio between the total energy output and the total energy input (the sum of the electric energy plus the energy in the liquid feedstock as well as that in the carbon electrodes). A primary purpose of this paper is to show that conventional thermochemistry does indeed predict a commercial efficiency bigger than one, although their values is considerably smaller than the actual efficiency measured in the reactors, thus indicating the applicabili...

  15. Data Compilation for AGR-1 Baseline Compact Lot LEU01-46T-Z

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D [ORNL; Montgomery, Fred C [ORNL; Pappano, Peter J [ORNL

    2006-08-01

    This document is a compilation of characterization data for the AGR-1 baseline compact lot LEU01-46T-Z. The compacts were produced by ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program for the first AGR irradiation test train (AGR-1). This compact lot was fabricated using particle composite LEU01-46T, which was a composite of four batches of TRISO-coated 350 {micro}m diameter 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with an {approx} 50% dense carbon buffer layer (100 {micro}m nominal thickness), followed by a dense inner pyrocarbon layer (40 {micro}m nominal thickness), followed by a SiC layer (35 {micro}m nominal thickness), followed by another dense outer pyrocarbon layer (40 {micro}m nominal thickness). The kernels were obtained from BWXT and identified as composite G73D-20-69302. The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified as LEU01-?? (where ?? is a series of integers beginning with 01). A data compilation for the AGR-1 baseline coated particle composite LEU01-46T can be found in ORNL/TM-2006/019. The AGR-1 Fuel product Specification and Characterization Guidance (INL EDF-4380) provides the requirements necessary for acceptance of the fuel manufactured for the AGR-1 irradiation test. Section 6.2 of EDF-4380 provides the property requirements for the heat treated compacts. The Statistical Sampling Plan for AGR Fuel materials (INL EDF-4542) provides additional guidance regarding statistical methods for product acceptance and recommended sample sizes. The procedures for characterizing and qualifying the compacts are outlined in ORNL product inspection plan AGR-CHAR-PIP-05. the inspection report forms generated by this product inspection plan document the product acceptance for the property requirements listed in section 6.2 of EDF-4380.

  16. Data Compilation for AGR-1 Variant 2 Compact Lot LEU01-48T-Z

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D [ORNL; Montgomery, Fred C [ORNL; Pappano, Peter J [ORNL

    2006-08-01

    This document is a compilation of characterization data for the AGR-1 variant 2 compact lot LEU01-48T-Z. The compacts were produced by ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program for the first AGR irradiation test train (AGR-1). This compact lot was fabricated using particle composite LEU01-48T, which was a composite of three batches of TRISO-coated 350 {micro}m diameter 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with an {approx} 50% dense carbon buffer layer (100 {micro}m nominal thickness), followed by a dense inner pyrocarbon layer (40 {micro}m nominal thickness), followed by a SiC layer (35 {micro}m nominal thickness), followed by another dense outer pyrocarbon layer (40 {micro}m nominal thickness). The kernels were obtained from BWXT and identified as composite G73D-20-69302. The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified as LEU01-?? (where ?? is a series of integers beginning with 01). A data compilation for the AGR-1 variant 2 coated particle composite LEU01-48T can be found in ORNL/TM-2006/021. The AGR-1 Fuel Product Specification and Characterization Guidance (INL EDF-4380) provides the requirements necessary for acceptance of the fuel manufactured for the AGR-1 irradiation test. Section 6.2 of EDF-4380 provides the property requirements for the heat treated compacts. The Statistical Sampling Plan for AGR Fuel materials (INL EDF-4542) provides additional guidance regarding statistical methods for product acceptance and recommended sample sizes. The procedures for characterizing and qualifying the compacts are outlined in ORNL product inspection plan AGR-CHAR-PIP-05. The inspection report forms generated by this product inspection plan document the product acceptance for the property requirements listed in section 6.2 of EDF-4380.

  17. Data Compilation for AGR-1 Variant 1 Compact Lot LEU01-47T-Z

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D [ORNL; Montgomery, Fred C [ORNL; Pappano, Peter J [ORNL

    2006-08-01

    This document is a compilation of characterization data for the AGR-1 variant 1 compact lot LEU01-47T-Z. The compacts were produced by ORNL for the ADvanced Gas Reactor Fuel Development and Qualification (AGR) program for the first AGR irradiation test train (AGR-1). This compact lot was fabricated using particle composite LEU01-47T, which was a composite of three batches of TRISO-coated 350 {micro}m diameter 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with an {approx} 50% dense carbon buffer layer (100 {micro}m nominal thickness), followed by a dense inner pyrocarbon layer (40 {micro}m nominal thickness), followed by a SiC layer (35 {micro}m nominal thickness), followed by another dense outer pyrcoarbon layer (40 {micro}m nominal thickness). The kernels were obtained from BWXT and identified as composite G73D-20-69302. The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified at LEU01-?? (where ?? is a series of integers beginning with 01). A data compilation for the AGR-1 variant 1 coated particle composite LEU01-47T can be found in ORNL/TM-2006/020. The AGR-1 Fuel Product Specification and Characterization Guidance (INL EDF-4380) provides the requirements necessary for acceptance of the fuel manufactured for the AGR-1 irradiation test. Section 6.2 of EDF-4380 provides the property requirements for the heat treated compacts. The Statistical Sampling Plan for AGR Fuel Materials (INL EDF-4542) provides additional guidance regarding statistical methods for product acceptance and recommended sample sizes. The procedures for characterizing and qualifying the compacts are outlined in ORNL product inspection plan AGR-CHAR-PIP-05. The inspection report forms generated by this product inspection plan document the product acceptance for the property requirements listed in section 6.2 of EDF-4380.

  18. Improvement of nuclear ship engineering simulation system. Hardware renewal and interface improvement of the integral type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Hiroki; Kyoya, Masahiko; Shimazaki, Junya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kano, Tadashi [KCS, Co., Mito, Ibaraki (Japan); Takahashi, Teruo [Energis, Co., Kobe, Hyogo (Japan)

    2001-10-01

    JAERI had carried out the design study about a lightweight and compact integral type reactor (an advanced marine reactor) with passive safety equipment as a power source for the future nuclear ships, and completed an engineering design. We have developed the simulator for the integral type reactor to confirm the design and operation performance and to utilize the study of automation of the reactor operation. The simulator can be used also for future research and development of a compact reactor. However, the improvement in a performance of hardware and a human machine interface of software of the simulator were needed for future research and development. Therefore, renewal of hardware and improvement of software have been conducted. The operability of the integral-reactor simulator has been improved. Furthermore, this improvement with the hardware and software on the market brought about better versatility, maintainability, extendibility and transfer of the system. This report mainly focuses on contents of the enhancement in a human machine interface, and describes hardware renewal and the interface improvement of the integral type reactor simulator. (author)

  19. MR-6 type fuel elements cooling in natural convection conditions after the reactor shut down

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Bykowski, W.; Moldysz, A. [Institute of Atomic Energy, Otwock Swierk (Poland)

    2002-07-01

    Natural cooling conditions of the nuclear fuel in the channel type reactor after its shut down are commonly determined with relatively high uncertainty. This is not only to he lack of adequate measurements of thermal parameters i.e. the residual power generation, the coolant flow and temperatures, but also due to indeterminate model of convection mechanism. The numerical simulation of natural convection in multitube fuel assembly in the fuel channel leads to various convection modes including evidently chaotic behaviour. To determine the real cooling conditions in the MARIA research reactor a series of experiments has been performed with fuel assembly equipped with a set of thermocouples. After some forced cooling period (the shortest was half an hour after the reactor shut down) the reactor was left with the only natural convection. Two completely different cooling modes have been observed. The MARIA core consists of series of individual fuel channel and so called bypasses, maintaining the hydraulic properties of the fuel channel, connected in parallel. Initially, the convection cells were established trough few so-called bypasses providing a very effective mode of cooling. In this mode the flow charts were identical to those existing in forced cooling mode. After certain period the system switched on the second cooling mode with natural circulation within the individual fuel cells. Higher temperatures and temperature fluctuations were characteristic for this mode approaching 30 deg in amplitude. In almost all the cases the system was switching few times between modes, but eventually remained in the second mode. The switching times were not regular and the process has a chaotic behaviour. (author)

  20. On the Scale-up of Gas-Hydrate-Forming Reactors: The Case of Gas-Dispersion-Type Reactors

    Directory of Open Access Journals (Sweden)

    Yasuhiko H. Mori

    2015-02-01

    Full Text Available For establishing hydrate-based technologies for natural-gas storage/transport, CO2 capture from industrial flue gases, etc., we need appropriate guidelines for the scale-up of hydrate production/processing equipment from laboratory scales to industrial scales. This paper aims to provide technical remarks on the scale-up of hydrate-forming reactors, the central components of hydrate production/processing equipment, particularly focusing on such a reactor design that hydrate-forming gas is dispersed in an aqueous phase which is either stirred in a tank or forced to flow through a tube. Based on the principles of classical fluid mechanics and heat-transfer analysis, the paper derives semi-empirical formulas that show how the capacity for heat discharge from each reactor and the power for operating the reactor are required to change with an increase in its size. Consequently, it is concluded that the stirred-tank design is unfavorable for significant scale-up and that the scale-up of tubular reactors should be made without significantly increasing the in-tube flow velocity.

  1. Thermal hydraulics characterization of the core and the reactor vessel type BWR; Caracterizacion termohidraulica del nucleo y de la vasija de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Zapata Y, M.; Lopez H, L.E. [CFE, Carretera Cardel-Nautla Km. 42.5, Municipio Alto Lucero, Veracruz (Mexico)]. e-mail: marxlenin.zapata@cfe.gob.mx

    2008-07-01

    The thermal hydraulics design of a reactor type BWR 5 as the employees in the nuclear power plant of Laguna Verde involves the coupling of at least six control volumes: Pumps jet region, Stratification region, Core region, Vapor dryer region, Humidity separator region and Reactor region. Except by the regions of the core and reactor, these control volumes only are used for design considerations and their importance as operative data source is limited. It is for that is fundamental to complement the thermal hydraulics relations to obtain major data that allow to determine the efficiency of internal components, such as pumps jet, humidity separator and vapor dryer. Like example of the previous thing, calculations are realized on the humidity of the principal vapor during starting, comparing it with the values at the moment incorporated in the data banks of the computers of process of both units. (Author)

  2. Artificial intelligence applied to fuel management in BWR type reactors; Inteligencia artificial aplicada a la administracion de combustible en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J

    1998-10-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  3. Radiological performance of hot water layer system in open pool type reactor

    Directory of Open Access Journals (Sweden)

    Amr Abdelhady

    2013-06-01

    Full Text Available The paper presents the calculated dose rate carried out by using MicroShield code to show the importance of hot water layer system (HWL in 22 MW open pool type reactor from the radiation protection safety point of view. The paper presents the dose rate profiles over the pool surface in normal and abnormal operations of HWL system. The results show that, in case of losing the hot water layer effect, the radiation dose rate profiles over the pool surface will increase from values lower than the worker permissible dose limits to values very higher than the permissible dose limits.

  4. Chemistry aspects of the source term formation for a severe accident in a CANDU type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Constantin, A.; Constantin, M. [Institute for Nuclear Research, Pitesti (Romania)

    2013-07-15

    The progression of a severe accident in a CANDU type reactor is slow because the core is surrounded by a large quantity of heavy and light water which acts as a heat sink to remove the decay heat. Therefore, the source term formation is a complex and long process involving fission products transport and releasing in the fuel matrix, thermal hydraulics of the transport fluid in the primary heat system and containment, deposition and transport of fission products, chemistry including the interaction with the dousing system, structural materials and paints, etc. The source term is strongly dependent on initial conditions and accident type. The paper presents chemistry aspects for a severe accident in a CANDU type reactor, in terms of the retention in the primary heat system. After releasing from the fuel elements, the fission products suffer a multitude of phenomena before they are partly transferred into the containment region. The most important species involved in the deposition were identified. At the same time, the influence of the break position in the transfer fractions from the primary heat system to the containment was investigated. (orig.)

  5. Proposal of rectifier type superconducting fault current limiter with non-inductive reactor (SFCL)

    Science.gov (United States)

    Mohammad Salim, Khosru; Muta, Itsuya; Hoshino, Tsutomu; Nakamura, Taketsune; Yamada, Masato

    2004-03-01

    A rectifier type superconducting fault current limiter (SFCL) with non-inductive reactor has been proposed. The concept behind this SFCL is the appearance of high impedance during non-superconducting state of the coil. In a hybrid bridge circuit, two superconducting coils connected in anti-parallel: a trigger coil and a limiting coil. Both the coils are magnetically coupled with each other and have same number of turns. There is almost zero flux inside the core and therefore the total inductance is small during normal operation. At fault time when the trigger coil current reaches to a certain level, the trigger coil changes from superconducting state to normal state. This super-to-normal transition of the trigger coil changes the current ratio of the coils and therefore the flux inside the reactor is no longer zero. So, the equivalent impedance of both the coils increased thus limits the fault current. We have carried out computer simulation using EMTDC and observed the results. A preliminary experiment has already been performed using copper wired reactor with simulated super-to-normal transition resistance and magnetic switches. Both the simulation and preliminary experiment shows good results. The advantage of using hybrid bridge circuit is that the SFCL can also be used as circuit breaker. Two separate bridge circuit can be used for both trigger coil and the limiter coil. In such a case, the trigger coil can be shutdown immediately after the fault to reduce heat and thus reduce the recovery time. Again, at the end of fault when the SFCL needs to re-enter to the grid, turning off the trigger circuit in the two-bridge configuration the inrush current can be reduced. This is because the current only flows through the limiting coil. Another advantage of this type of SFCL is that no voltage sag will appear during load increasing time as long as the load current stays below the trigger current level.

  6. Survival of Listeria monocytogenes in Soil Requires AgrA-Mediated Regulation.

    Science.gov (United States)

    Vivant, Anne-Laure; Garmyn, Dominique; Gal, Laurent; Hartmann, Alain; Piveteau, Pascal

    2015-08-01

    In a recent paper, we demonstrated that inactivation of the Agr system affects the patterns of survival of Listeria monocytogenes (A.-L. Vivant, D. Garmyn, L. Gal, and P. Piveteau, Front Cell Infect Microbiol 4:160, http://dx.doi.org/10.3389/fcimb.2014.00160). In this study, we investigated whether the Agr-mediated response is triggered during adaptation in soil, and we compared survival patterns in a set of 10 soils. The fate of the parental strain L. monocytogenes L9 (a rifampin-resistant mutant of L. monocytogenes EGD-e) and that of a ΔagrA deletion mutant were compared in a collection of 10 soil microcosms. The ΔagrA mutant displayed significantly reduced survival in these biotic soil microcosms, and differential transcriptome analyses showed large alterations of the transcriptome when AgrA was not functional, while the variations in the transcriptomes between the wild type and the ΔagrA deletion mutant were modest under abiotic conditions. Indeed, in biotic soil environments, 578 protein-coding genes and an extensive repertoire of noncoding RNAs (ncRNAs) were differentially transcribed. The transcription of genes coding for proteins involved in cell envelope and cellular processes, including the phosphotransferase system and ABC transporters, and proteins involved in resistance to antimicrobial peptides was affected. Under sterilized soil conditions, the differences were limited to 86 genes and 29 ncRNAs. These results suggest that the response regulator AgrA of the Agr communication system plays important roles during the saprophytic life of L. monocytogenes in soil.

  7. Optimizing the Design of Small Fast Spectrum Battery-Type Nuclear Reactors

    Directory of Open Access Journals (Sweden)

    Staffan Qvist

    2014-07-01

    Full Text Available This study is focused on defining and optimizing the design parameters of inherently safe “battery” type sodium-cooled metallic-fueled nuclear reactor cores that operate on a single stationary fuel loading at full power for 30 years. A total of 29 core designs were developed with varying power and flow conditions, including detailed thermal-hydraulic, structural-mechanical and neutronic analysis. Given set constraints for irradiation damage, primary cycle pressure drop and inherent safety considerations, the attainable power range and performance characteristics of the systems are defined. The optimum power level for a core with a coolant pressure drop limit of 100 kPa and an irradiation damage limit of 200 DPA (displacements per atom is found to be 100 MWt/40 MWe. Raising the power level of an optimized core gives significantly higher attainable power densities and burnup, but severely decreases safety margins and increases the irradiation damage. A fully optimized inherently safe battery-type fast reactor core with an active height and diameter of 150 cm (2.6 m3, a pressure drop limit of 100 kPa and an irradiation damage limit of 300 DPA can be designed to operate at 150 MWt/60 MWe for 30 years, reaching an average discharge burnup of 100 MWd/kg-actinide.

  8. Issues of intergranular embrittlement of VVER-type nuclear reactors pressure vessel materials

    Science.gov (United States)

    Zabusov, O.

    2016-04-01

    In light of worldwide tendency to extension of service life of operating nuclear power plants - VVER-type in the first place - recently a special attention is concentrated on phenomena taking place in reactor pressure vessel materials that are able to lead to increased level of mechanical characteristics degradation (resistibility to brittle fracture) during long term of operation. Formerly the hardening mechanism of degradation (increase in the yield strength under influence of irradiation) mainly had been taken into consideration to assess pressure vessel service life limitations, but when extending the service life up to 60 years and more the non-hardening mechanism (intergranular embrittlement of the steels) must be taken into account as well. In this connection NRC “Kurchatov Institute” has initiated a number of works on investigations of this mechanism contribution to the total embrittlement of reactor pressure vessel steels. The main results of these investigations are described in this article. Results of grain boundary phosphorus concentration measurements in specimens made of first generation of VVER-type pressure vessels materials as well as VVER-1000 surveillance specimens are presented. An assessment of non-hardening mechanism contribution to the total ductile-to- brittle transition temperature shift is given.

  9. Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwangmin, E-mail: kwangmin81@gmail.com [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Go, Byeong-Soo; Sung, Hae-Jin; Park, Hea-chul; Kim, Seokho [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of); Jin, Yoon-Su; Oh, Yunsang [Vector Fields Korea Inc., Pohang 790-834 (Korea, Republic of); Park, Minwon [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Yu, In-Keun, E-mail: yuik@changwon.ac.kr [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of)

    2014-09-15

    Highlights: • The authors designed and fabricated a D-shape coil based toroid-type HTS DC reactor using 2G GdBCO HTS wires. • The toroid-type magnet consisted of 30 D-shape double pancake coil (DDC)s. The total length of the wire was 2.32 km. • The conduction cooling method was adopted for reactor magnet cooling. • The maximum cooling temperature of reactor magnet is 5.5 K. • The inductance was 408 mH in the steady-state condition (300 A operating). - Abstract: This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

  10. ecotecnia agrícola

    Directory of Open Access Journals (Sweden)

    S. Orozco Cirilo

    2008-01-01

    Full Text Available Las Escuelas de Campo (EC son un método de extensión participativo que permite desarrollar capacidades analíticas, pensamiento crítico y, creatividad. La investigación se realizó en la región Mazateca de Oaxaca, la cual se caracteriza por su alto rezago tecnológico, porque resulta de relevancia científica, social y ecológica conocer, en este contexto, los alcances de metodologías alternativas de extensionismo sobre la adopción de ecotecniasagrícolas. Nuestro objetivo fue conocer la influencia del conocimiento otorgado en EC sobre el índice de adopción final de ecotecnia generada localmente, como alternativa a la enorme dependencia tecnológica. Las variables analizadas fueron: índice de adopción inicial, nivel inicial de conocimientos, número de cursos otorgados en EC, índice de adopción final y, nivel final de conocimientos. A una muestra de dos estratos participantes y un grupo testigo, fueron aplicados 135 cuestionarios. Para probar nuestras hipótesis los resultados de éstos se sometieron al análisis de varianza y correlación. Encontramos que solamente el índice de adopción final obtenido por los participantes experimentó incrementos estadísticamente significativos respecto a la condición inicial, debido a su participación en EC y al consecuente aumento en el conocimiento de la ecotecnia. Esto implica que la metodología propuesta incrementa la adopción de ecotecnias generadas localmente, consolidándose como un paradigmaalternativo en extensión agrícola.

  11. Anterior Gradient 2 (AGR2) Induced Epidermal Growth Factor Receptor (EGFR) Signaling Is Essential for Murine Pancreatitis-Associated Tissue Regeneration

    Science.gov (United States)

    Wodziak, Dariusz; Dong, Aiwen; Basin, Michael F.; Lowe, Anson W.

    2016-01-01

    A recently published study identified Anterior Gradient 2 (AGR2) as a regulator of EGFR signaling by promoting receptor presentation from the endoplasmic reticulum to the cell surface. AGR2 also promotes tissue regeneration in amphibians and fish. Whether AGR2-induced EGFR signaling is essential for tissue regeneration in higher vertebrates was evaluated using a well-characterized murine model for pancreatitis. The impact of AGR2 expression and EGFR signaling on tissue regeneration was evaluated using the caerulein-induced pancreatitis mouse model. EGFR signaling and cell proliferation were examined in the context of the AGR2-/- null mouse or with the EGFR-specific tyrosine kinase inhibitor, AG1478. In addition, the Hippo signaling coactivator YAP1 was evaluated in the context of AGR2 expression during pancreatitis. Pancreatitis-induced AGR2 expression enabled EGFR translocation to the plasma membrane, the initiation of cell signaling, and cell proliferation. EGFR signaling and tissue regeneration were partially inhibited by the tyrosine kinase inhibitor AG1478, but absent in the AGR2-/- null mouse. AG1478-treated and AGR2-/- null mice with pancreatitis died whereas all wild-type controls recovered. YAP1 activation was also dependent on pancreatitis-induced AGR2 expression. AGR2-induced EGFR signaling was essential for tissue regeneration and recovery from pancreatitis. The results establish tissue regeneration as a major function of AGR2-induced EGFR signaling in adult higher vertebrates. Enhanced AGR2 expression and EGFR signaling are also universally present in human pancreatic cancer, which support a linkage between tissue injury, regeneration, and cancer pathogenesis. PMID:27764193

  12. Analysis of Fission Products on the AGR-1 Capsule Components

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; Jason M. Harp; Philip L. Winston; Scott A. Ploger

    2013-03-01

    The components of the AGR-1 irradiation capsules were analyzed to determine the retained inventory of fission products in order to determine the extent of in-pile fission product release from the fuel compacts. This includes analysis of (i) the metal capsule components, (ii) the graphite fuel holders, (iii) the graphite spacers, and (iv) the gas exit lines. The fission products most prevalent in the components were Ag-110m, Cs 134, Cs 137, Eu-154, and Sr 90, and the most common location was the metal capsule components and the graphite fuel holders. Gamma scanning of the graphite fuel holders was also performed to determine spatial distribution of Ag-110m and radiocesium. Silver was released from the fuel components in significant fractions. The total Ag-110m inventory found in the capsules ranged from 1.2×10 2 (Capsule 3) to 3.8×10 1 (Capsule 6). Ag-110m was not distributed evenly in the graphite fuel holders, but tended to concentrate at the axial ends of the graphite holders in Capsules 1 and 6 (located at the top and bottom of the test train) and near the axial center in Capsules 2, 3, and 5 (in the center of the test train). The Ag-110m further tended to be concentrated around fuel stacks 1 and 3, the two stacks facing the ATR reactor core and location of higher burnup, neutron fluence, and temperatures compared with Stack 2. Detailed correlation of silver release with fuel type and irradiation temperatures is problematic at the capsule level due to the large range of temperatures experienced by individual fuel compacts in each capsule. A comprehensive Ag 110m mass balance for the capsules was performed using measured inventories of individual compacts and the inventory on the capsule components. For most capsules, the mass balance was within 11% of the predicted inventory. The Ag-110m release from individual compacts often exhibited a very large range within a particular capsule.

  13. Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

  14. AGR-5/6/7 LEUCO Kernel Fabrication Readiness Review

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Douglas W. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Design and Development; Bailey, Kirk W. [Idaho National Lab. (INL), Idaho Falls, ID (United States). ART Quality Assurance Engineer

    2015-02-01

    In preparation for forming low-enriched uranium carbide/oxide (LEUCO) fuel kernels for the Advanced Gas Reactor (AGR) fuel development and qualification program, Idaho National Laboratory conducted an operational readiness review of the Babcock & Wilcox Nuclear Operations Group – Lynchburg (B&W NOG-L) procedures, processes, and equipment from January 14 – January 16, 2015. The readiness review focused on requirements taken from the American Society Mechanical Engineers (ASME) Nuclear Quality Assurance Standard (NQA-1-2008, 1a-2009), a recent occurrence at the B&W NOG-L facility related to preparation of acid-deficient uranyl nitrate solution (ADUN), and a relook at concerns noted in a previous review. Topic areas open for the review were communicated to B&W NOG-L in advance of the on-site visit to facilitate the collection of objective evidences attesting to the state of readiness.

  15. Thermal-hydraulic Fortran program for steady-state calculations of plate-type fuel research reactors

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2008-01-01

    Full Text Available The safety assessment of research and power reactors is a continuous process covering their lifespan and requiring verified and validated codes. Power reactor codes all over the world are well established and qualified against real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume a lot of running time. On the other hand, most research reactor codes still require much more data for validation and qualification. It is, therefore, of benefit to any regulatory body to develop its own codes for the review and assessment of research reactors. The present paper introduces a simple, one-dimensional Fortran program called THDSN for steady-state thermal-hydraulic calculations of plate-type fuel research reactors. Besides calculating the fuel and coolant temperature distributions and pressure gradients in an average and hot channel, the program calculates the safety limits and margins against the critical phenomena encountered in research reactors, such as the onset of nucleate boiling, critical heat flux and flow instability. Well known thermal-hydraulic correlations for calculating the safety parameters and several formulas for the heat transfer coefficient have been used. The THDSN program was verified by comparing its results for 2 and 10 MW benchmark reactors with those published in IAEA publications and a good agreement was found. Also, the results of the program are compared with those published for other programs, such as the PARET and TERMIC.

  16. Propagation of cracks by stress corrosion in conditions of BWR type reactor; Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua en ebullicion (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, F.J. [ININ, 52045 Estado de Mexico (Mexico); Fuentes C, P. [ITT, Metepec, Estado de Mexico (Mexico)]. E-mail: fjmc@nuclear.inin.mx

    2004-07-01

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

  17. Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

    Energy Technology Data Exchange (ETDEWEB)

    Paul Demkowicz; Scott Ploger; John Hunn

    2012-05-01

    The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Five irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These five compacts also included all four TRISO coating variations irradiated in the AGR experiment. The five compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. Approximately 40 to 80 particles within each cross section were exposed near enough to mid-plane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 830 classified particles allowed other relationships among morphological types to be established.

  18. Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

    Energy Technology Data Exchange (ETDEWEB)

    Paul Demkowicz; Scott Ploger; John Hunn; Jay S. Kehn

    2012-09-01

    The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Six irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These six compacts also included all four TRISO coating variations irradiated in the AGR experiment. The six compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. From 36 to 79 particles within each cross section were exposed near enough to midplane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 931 classified particles allowed other relationships among morphological types to be established.

  19. Proposal for a novel type of small scale aneutronic fusion reactor

    Science.gov (United States)

    Gruenwald, J.

    2017-02-01

    The aim of this work is to propose a novel scheme for a small scale aneutronic fusion reactor. This new reactor type makes use of the advantages of combining laser driven plasma acceleration and electrostatic confinement fusion. An intense laser beam is used to create a lithium-proton plasma with high density, which is then collimated and focused into the centre of the fusion reaction chamber. The basic concept presented here is based on the 7Li-proton fusion reaction. However, the physical and technological fundamentals may generally as well be applied to 11B-proton fusion. The former fusion reaction path offers higher energy yields while the latter has larger fusion cross sections. Within this paper a technological realisation of such a fusion device, which allows a steady state operation with highly energetic, well collimated ion beam, is presented. It will be demonstrated that the energetic break even can be reached with this device by using a combination of already existing technologies.

  20. [Continuous operation of hydrogen bio-production reactor with ethanol-type fermentation].

    Science.gov (United States)

    Ren, Nan-qi; Gong, Man-li; Xing, De-feng

    2004-11-01

    The natural response of a continuous stirred tank reactor (CSTR) for hydrogen bio-production using molasses wastewater as substrate was investigated. Emphasis was placed on assessing the operational controlling strategy on the stable operation of CSTR with high efficiency. It was found that at an initial biomass of 15g/L, an equilibrial microbial community in the ethanol-type fermentation and efficient stable operation of CSTR could be established with following conditions: temperature of 35 degrees C +/- 1 degrees C, COD organic loading rate (OLR) of 40kg/(m3 x d), hydraulic retention time (HRT) of 4h, pH value of 4.6 - 4.9 and oxidation reduction potential (ORP) of -450 - -470mV. Following that, hydrogen production in the reactor was relatively stable. The observed maximal hydrogen bio-production rate was 7.63m3/(m3 x d). The content of hydrogen in the biogas was about 40% - 58%. COD removal rate was between 22% - 26%. The total content of ethanol and acetic acid in the fermentative end products was above 80%.

  1. Aging of reactor vessels in LWR type reactors; Envejecimiento de la vasija y de los internos del nuclear de los reactores tipo LWR

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Lapena, J.; Serrano, M.

    2004-07-01

    Most of the degradation mechanisms of nuclear components were not included on the design so they have to be treated a posteriori, and that imply a loss of capacity. In this paper the state of the art on the reactor pressure vessel neutron embrittlement and on the irradiation assisted stress corrosion cracking that affects internal components, are explained. Special attention is devoted on the influence of the neutron fluence on IASCC process, on the material alterations promoted by irradiation and their consequences on the susceptibility to this phenomenon. Regarding the reactor pressure vessel degradation, this paper discuss the application of the Master Curve on the structural integrity evaluation of the vessel. Other aspects related to further developments are also mentioned and the importance of a good materials ageing management on the operation of the plant is pointed out. (Author) 12 refs.

  2. E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

    Energy Technology Data Exchange (ETDEWEB)

    Van Tichelen, Katrien, E-mail: kvtichel@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Mirelli, Fabio, E-mail: fmirelli@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Greco, Matteo, E-mail: mgreco@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Viviani, Giorgia, E-mail: giorgiaviviani@gmail.com [University of Pisa, Lungarno Pacinotti 43, 56126 Pisa (Italy)

    2015-08-15

    Highlights: • The E-SCAPE facility is a thermal hydraulic scale model of the MYRRHA fast reactor. • The focus is on mixing and stratification in liquid-metal pool-type reactors. • Forced convection, natural convection and the transition are investigated. • Extensive instrumentation allows validation of computational models. • System thermal hydraulic and CFD models have been used for facility design. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant. The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing. The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE. The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

  3. Synthesis of layered double hydroxide nanosheets by coprecipitation using a T-type microchannel reactor

    Science.gov (United States)

    Pang, Xiujiang; Sun, Meiyu; Ma, Xiuming; Hou, Wanguo

    2014-02-01

    The synthesis of Mg2Al-NO3 layered double hydroxide (LDH) nanosheets by coprecipitation using a T-type microchannel reactor is reported. Aqueous LDH nanosheet dispersions were obtained. The LDH nanosheets were characterized by X-ray diffraction, transmission electron microscopy, atomic force microscopy and particle size analysis, and the transmittance and viscosity of LDH nanosheet dispersions were examined. The two-dimensional LDH nanosheets consisted of 1-2 brucite-like layers and were stable for ca. 16 h at room temperature. In addition, the co-assembly between LDH nanosheets and dodecyl sulfate (DS) anions was carried out, and a DS intercalated LDH nanohybrid was obtained. To the best of our knowledge, this is the first report of LDH nanosheets being directly prepared in bulk aqueous solution. This simple, cheap method can provide naked LDH nanosheets in high quantities, which can be used as building blocks for functional materials.

  4. AGR-3/4 Final Data Qualification Report for ATR Cycles 151A through 155B-1

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh T. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    This report provides the qualification status of experimental data for the entire Advanced Gas Reactor 3/4 (AGR 3/4) fuel irradiation. AGR-3/4 is the third in a series of planned irradiation experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for the AGR Fuel Development and Qualification Program, which supports development of the advanced reactor technology under the INL ART Technology Development Office (TDO). The main objective of AGR-3/4 irradiation is to provide a known source of fission products for subsequent transport through compact matrix and structural graphite materials due to the presence of designed-to-fail fuel particles. Full power irradiation of the AGR 3/4 test began on December 14, 2011 (ATR Cycle 151A), and was completed on April 12, 2014 (end of ATR Cycle 155B) after 369.1 effective full power days of irradiation. The AGR-3/4 test was in the reactor core for eight of the ten ATR cycles between 151A and 155B. During the unplanned outage cycle, 153A, the experiment was removed from the ATR northeast flux trap (NEFT) location and stored in the ATR canal. This was to prevent overheating of fuel compacts due to higher than normal ATR power during the subsequent Powered Axial Locator Mechanism cycle, 153B. The AGR 3/4 test was inserted back into the ATR NEFT location during the outage of ATR Cycle 154A on April 26, 2013. Therefore, the AGR-3/4 irradiation data received during these 2 cycles (153A and 153B) are irrelevant and their qualification status isnot included in this report. Additionally, during ATR Cycle 152A the ATR core ran at low power for a short enough duration that the irradiation data are not used for physics and thermal calculations. However, the qualification status of irradiation data for this cycle is still covered in this report. As a result, this report includes data from 8 ATR Cycles: 151A, 151B, 152A, 152B, 154A, 154B, 155A, and 155B, as recorded in the Nuclear Data Management and

  5. AGR-2 Irradiation Test Final As-Run Report, Rev 2

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a

  6. AGR-2 Irradiation Test Final As-Run Report, Rev 2

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test

  7. Microscopic analysis of irradiated AGR-1 coated particle fuel compacts

    Energy Technology Data Exchange (ETDEWEB)

    Scott Ploger; Paul Demkowicz; John Hunn; Robert Morris

    2012-10-01

    The AGR-1 experiment involved irradiation of 72 TRISO-coated particle fuel compacts to a peak burnup of 19.5% FIMA with no in-pile failures observed out of 3×105 total particles. Irradiated AGR-1 fuel compacts have been cross-sectioned and analyzed with optical microscopy to characterize kernel, buffer, and coating behavior. Five compacts have been examined so far, spanning a range of irradiation conditions (burnup, fast fluence, and irradiation temperature) and including all four TRISO coating variations irradiated in the AGR-1 experiment. The cylindrical specimens were sectioned both transversely and longitudinally, then polished to expose between approximately 40-80 individual particles on each mount. The analysis focused primarily on kernel swelling and porosity, buffer densification and fracturing, buffer-IPyC debonding, and fractures in the IPyC and SiC layers. Characteristic morphologies have been identified, over 800 particles have been classified, and spatial distributions of particle types have been mapped. No significant spatial patterns were discovered in these cross sections. However, some trends were found between morphological types and certain behavioral aspects. Buffer fractures were found in approximately 23% of the particles, and these fractures often resulted in unconstrained kernel swelling into the open cavities. Fractured buffers and buffers that stayed bonded to IPyC layers appear related to larger pore size in kernels. Buffer-IPyC interface integrity evidently factored into initiation of rare IPyC fractures. Fractures through part of the SiC layer were found in only three particles, all in conjunction with IPyC-SiC debonding. Compiled results suggest that the deliberate coating fabrication variations influenced the frequencies of IPyC fractures, IPyC-SiC debonds, and SiC fractures.

  8. Nave agrícola – Brasil

    Directory of Open Access Journals (Sweden)

    Levi, Rino

    1959-07-01

    Full Text Available Recientemente se ha realizado el proyecto de ejecución de una amplia nave dedicada a garaje, aparcamiento y conservación de maquinaria móvil agrícola en San José Dos Campos (Brasil.

  9. Research on Precaution and Detection Technology for Flow Blockage of Plate-type Fuel Element in Research Reactors

    Institute of Scientific and Technical Information of China (English)

    DING; Li; QIAO; Ya-xin; ZHANG; Nian-peng; LUO; Bei-bei; HUA; Xiao; JIA; Shu-jie; YAN; Hui-yang

    2013-01-01

    The main aim of this study is to offer the technical support for safety operation and management of research reactors using plate-type fuel assemblies in China,which is performed from analysis of precaution measures for flow blockage and detection methods of accidents.Study shows that most accidents were induced by in-core foreign objects and the swelling of fuel

  10. Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

    Science.gov (United States)

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

    2005-01-01

    Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment.

  11. Use of plutonium in PWR-type reactors; Utilisation du plutonium dans les REP

    Energy Technology Data Exchange (ETDEWEB)

    Berthet, A. [Electricite de France (EDF), 75 - Paris (France). Direction de l' Equipement

    1999-04-01

    The plutonium is used, as fuel, in the pressurized water reactors. It does not exist in nature; butit is fabricated in the reactor by neutrons capture. The MOX (Mixed Oxides) is its usual name. A part is consumed by the fission, the remainder is found in the used fuel released from the reactor. The paper deals with the plutonium specificities, the research and development programs about this fuel. The technical specifications of the PWR recycling the plutonium are also included (radiation protection, reactor fueling). (A.L.B.)

  12. Formation of a nuclear reactor's molten core bath in a crucible-type corium catcher for a nuclear power station equipped with VVER reactors

    Science.gov (United States)

    Beshta, S. V.; Vitol', S. A.; Granovskii, V. S.; Kalyago, E. K.; Kovtunova, S. V.; Krushinov, E. V.; Sulatskaya, M. B.; Sulatskii, A. A.; Khabenskii, V. B.; Al'Myashev, V. I.; Gusarov, V. V.

    2011-05-01

    Results from a calculation study on analyzing the formation of a melt bath in a crucible-type catcher for the conditions of a severe accident at a nuclear power station equipped with VVER-1000 reactors are presented. It is shown that the heat loads exerted on the water-cooled walls of the corium catcher shell are limited to a permissible level at which the necessary margins to nucleate boiling crisis and to destruction are ensured under the conditions of thermal and mechanical loading of the shell. An important role of sacrificial material in the efficient operation of the corium catcher is pointed out.

  13. Steam feed and effect of steam-thermal seal in thermolysis of tire shreds in a screw-type reactor

    Science.gov (United States)

    Kalitko, V. A.

    2010-05-01

    On the basis of experience in commercial operation, the effect of steam seal in tire-shred pyrolysis in a screw-type reactor with superheated steam has been considered and analytically substantiated; there, local steam feed produces the above effect at the total reduced pressure and keeps air from entering the reactor without sluices or valves used for hermetization of its loading and unloading. It has been shown that the increase in the production rate of pyrolysis due to the heating by steam amounts to 10-15% and is limited by the diffusion transfer in the reactor’s charge bed.

  14. Implications of reactor type and conditions on first-order hydrolysis rate assessment of maize silage.

    Science.gov (United States)

    Pabón Pereira, C P; Zeeman, G; Zhao, J; Ekmekci, B; van Lier, J B

    2009-01-01

    The biodegradability and first-order hydrolysis coefficient of maize silage have been assessed from batch experiments using different types of inoculum and substrate to inocula (S/I) ratios, and from CSTRs working at different hydraulic retention times (HRTs). In the batch experiments, the assessed maximum biodegradability of the maize silage was 68 (+/-2.7)% and 73(+/-2.9)% while the first order hydrolysis was 0.26 (+/-0.01) and 0.27(+/-0.02) d(-1), using granular and a mixture of granular and suspended inoculum, respectively. In the CSTR experiment biodegradability ranged from 41-65% depending on the HRT applied whereas the calculated first order hydrolysis coefficient was 0.32 d(-1). It is concluded that batch experiments can be used to assess first order hydrolysis constants and biodegradability provided that a well balanced inoculum is guaranteed. Further, it is shown that CSTR reactors digesting maize silage and operating at HRTs as low as 20 days can attain 88% of maximum biodegradability as long as pH fluctuations are minimized. 2 mmol NaHCO3 per gram maize silage was calculated to suffice for the purpose.

  15. Intercomparison of Different Types of Locally Prepared Concretes and Its Usability for Reactor Neutron Shielding

    Science.gov (United States)

    El-Kolaly, M. A.; Makarious, A. S.; Bashter, I. I.; Kansouh, W. A.

    Measurements have been carried out to study the attenuation of neutron from a horizontal channel of the ET-RR-1 reactor. The assessments of neutron distribution inside three different types of locally prepared concretes have been evaluated.Neutron intensities in ilmenite-limonite concrete shield show an exponential decrease with increasing concrete thickness. Ilmenite concrete is a good attenuator for thermal and intermediate neutrons. However, ordinary and ilmenite-limonite concretes show efficient shielding for fast neutrons.Translated AbstractVergleich verschiedener Zementarten hinsichtlich ihrer Brauchbarkeit zur Neutronenabschirmung von ReaktorenMessungen zur Untersuchung der Neutronenabschwächung in einem horizontalen Kanal eines ET-RR-1-Reaktors wurden durchgeführt. Die Charakteristika der Neutronenverteilung innerhalb dreier unterschiedlich zusammengesetzter Zemente wurden bestimmt. Die Neutronenintensität in einem Schild aus Ilmenite-Limonitezement zeigt einen exponentiellen Abfall mit wachsender Dicke. Ilmenitezement ist ein guter Schild für thermale und mittlere Neutronen. Normaler und Ilmenite-Limonitezement zeigen effektive Abschirmung bei schnellen Neutronen.

  16. Effect of conditions of air-lift type reactor work on cadmium adsorption

    Energy Technology Data Exchange (ETDEWEB)

    Filipkowska, Urszula; Szymczyk, Paula Szymczyk; Kuczajowska-Zadrozna, Malgorzata; Joezwiak, Tomasz [University of Warmia and Mazury in Olsztyn, Warszawska (Poland)

    2015-10-15

    We investigated cadmium sorption by activated sludge immobilized in 1.5% sodium alginate with 0.5% polyvinyl alcohol. Experiments were conducted in an air-lift type reactor at the constant concentration of biosorbent reaching 5 d.m./dm{sup 3}, at three flow rates: 0.1, 0.25 and 0.5 V/h, and at three concentrations of the inflowing cadmium solution: 10, 25 and 50mg/dm{sup 3}. Analyses determined adsorption capacity of activated sludge immobilized in alginate as well as reactor's work time depending on flow rate and initial concentration of the solution. Results achieved were described with the use of Thomas model. The highest adsorption capacity of the sorbent (determined from the Thomas model), i.e., 200.2mg/g d.m. was obtained at inflowing solution concentration of 50mg/dm{sup 3} and flow rate of 0.1V/h, whereas the lowest one reached 53.69mg/g d.m. at the respective values of 10mg/dm{sup 3} and 0.1 V/h. Analyses were also carried out to determine the degree of biosorbent adsorption capacity utilization at the assumed effectiveness of cadmium removal - at the breakthrough point (C=0.05*C{sub 0}) and at adsorption capacity depletion point (C−0.9*C0). The study demonstrated that the effectiveness of adsorption capacity utilization was influenced by both the concentration and flow rate of the inflowing solution. The highest degree of sorbent capacity utilization was noted at inflowing solution concentration of 50mg/dm{sup 3} and flow rate of 0.1 V/h, whereas the lowest one at the respective values of 10mg/dm{sup 3} and 0.1 V/h. The course of the process under dynamic conditions was evaluated using coefficients of tangent inclination - a, at point C/C{sub 0}=1/2. A distinct tendency was demonstrated in changes of tangent slope a as affected by the initial concentration of cadmium and flow rate of the solution. The highest values of a coefficient were achieved at the flow rate of 0.1 V/h and initial cadmium concentration of 50mg/dm{sup 3}.

  17. The Effect of Birthrate Granularity on the Release- to- Birth Ratio for the AGR-1 In-core Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Dawn Scates; John Walter

    2012-10-01

    The AGR-1 Advanced Gas Reactor (AGR) tristructural-isotropic-particle fuel experiment underwent 13 irradiation intervals from December 2006 until November 2009 within the Idaho National Laboratory Advanced Test Reactor in support of the Next Generation Nuclear Power Plant program. During this multi-year experiment, release-to-birth rate ratios were computed at the end of each operating interval to provide information about fuel performance. Fission products released during irradiation were tracked daily by the Fission Product Monitoring System using 8-hour measurements. Birth rates calculated by MCNP with ORIGEN for as-run conditions were computed at the end of each irradiation interval. Each time step in MCNP provided neutron flux, reaction rates and AGR-1 compact composition, which were used to determine birth rates using ORIGEN. The initial birth-rate data, consisting of four values for each irradiation interval at the beginning, end, and two intermediate times, were interpolated to obtain values for each 8-hour activity. The problem with this method is that any daily changes in heat rates or perturbations, such as shim control movement or core/lobe power fluctuations, would not be reflected in the interpolated data and a true picture of the system would not be presented. At the conclusion of the AGR-1 experiment, great efforts were put forth to compute daily birthrates, which were reprocessed with the 8-hour release activity. The results of this study are presented in this paper.

  18. Scaled Facility Design Approach for Pool-Type Lead-Bismuth Eutectic Cooled Small Modular Reactor Utilizing Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sangrok; Shin, Yong-Hoon; Lee, Jueun; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    2015-10-15

    In low carbon era, nuclear energy is the most prominent energy source of electricity. For steady ecofriendly nuclear energy supply, Generation IV reactors which are future nuclear reactor require safety, sustainability, economics and non-proliferation as four criteria. Lead cooled fast reactor (LFR) is one of these reactor type and Generation IV international forum (GIF) adapted three reference LFR systems which are a small and movable systems with long life without refueling, intermediate size and huge electricity generation system for power grid. NUTRECK (Nuclear Transmutation Energy Center of Korea) has been designed reactor called URANUS (Ubiquitous, Rugged, Accident-forgiving, Non-proliferating, and Ultra-lasting Sustainer) which is small modular reactor and using lead-bismuth eutectic coolant. To prove natural circulation capability of URANUS and analyze design based accidents, scaling mock-up experiment facility will be constructed. In this paper, simple specifications of URANUS will be presented. Then based on this feature, scaling law and scaled facility design results are presented. To validate safety feature and thermodynamics characteristic of URANUS, scaled mockup facility of URANUS is designed based on the scaling law. This mockup adapts two area scale factors, core and lower parts of mock-up are scaled for 3D flow experiment. Upper parts are scaled different size to reduce electricity power and LBE tonnage. This hybrid scaling method could distort some thermal-hydraulic parameters, however, key parameters for experiment will be matched for up-scaling. Detailed design of mock-up will be determined through iteration for design optimization.

  19. Large-scale surface dielectric barrier discharge type reactor : effect of the electric wind on the conversion effectiveness

    Energy Technology Data Exchange (ETDEWEB)

    Jolibois, J. [Univ. de Poitiers, Poitiers (France). Centre national de la recherche scientifique, Laboratoire de Catalyse en Chimie Organique; Poitiers Univ., Futuroscope Chasseneuil Cedex (France). Centre national de la recherche scientifique, Inst. Pprime; Zouzou, N.; Moreau, E. [Poitiers Univ., Futuroscope Chasseneuil Cedex (France). Centre national de la recherche scientifique, Inst. Pprime; Tatibouet, J.M. [Univ. de Poitiers, Poitiers (France). Centre national de la recherche scientifique, Laboratoire de Catalyse en Chimie Organique

    2010-07-01

    Non-thermal plasma (NTP) techniques offer an innovative approach for air pollution reduction. Most studies in NTP techniques use volumetric discharge reactors with small dimensions and low flow rates at laboratory scale. The objective of this study was to develop an air pollution control plasma reactor at industrial scale with surface discharge. Propene (C{sub 3}H{sub 6}) was oxidized at high flow rates in a large-scale plasma reactor based on surface dielectric barrier discharge (DBD). Three different configurations of surface discharges were tested with 15 ppm of C{sub 3}H{sub 6} in air at ambient temperature for a flow rate of 50 m{sup 3} per hour. The properties of these different surface discharges were analyzed using chemical measurements and 3 component particle image velocimetry (PIV) measurements. PIV measurements were used characterize the effect of the electric wind on the polluted gas airflow inside the reactor and to explain the differences of effectiveness of the three tested plasma generators. For the three plasma generators, a propene oxidation of up to 45 percent was obtained at one J per liter. The electric wind produced by the surface discharge resulted in the formation of vortices inside the plasma reactor. This electric wind can increase gas mixing inside the plasma reactor and therefore plays a key role in conversion efficiency. It was concluded that the electric wind produced by surface discharges enables the use of this type of discharge for VOC elimination at high flow rate, with the same effectiveness of volumetric discharges. 5 refs., 10 figs.

  20. Uncertainty Quantification of Calculated Temperatures for the U.S. Capsules in the AGR-2 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lybeck, Nancy [Idaho National Lab. (INL), Idaho Falls, ID (United States); Einerson, Jeffrey J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pham, Binh T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hawkes, Grant L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN-3636). The AGR-2 test was inserted in the B-12 position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in June 2010 and successfully completed irradiation in October 2013, resulting in irradiation of the TRISO fuel for 559.2 effective full power days (EFPDs) during approximately 3.3 calendar years. The AGR-2 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS) (Pham and Einerson 2014). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as-run thermal analysis has been performed separately on each of four AGR-2 U.S. capsules for the entire irradiation as discussed in (Hawkes 2014). The ABAQUS code’s finite element-based thermal model predicts the daily average volume-average fuel temperature and peak fuel temperature in each capsule. This thermal model involves complex physical mechanisms (e.g., graphite holder and fuel compact shrinkage) and properties (e.g., conductivity and density). Therefore, the thermal model predictions are affected by uncertainty in input parameters and by incomplete knowledge of the underlying physics leading to modeling assumptions. Therefore, alongside with the deterministic predictions from a set of input thermal conditions, information about prediction uncertainty is instrumental for the ART

  1. The upgrade and conversion of the ET-RR-1 research reactor using plate type fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Ashoub, N. [Reactor Physics Dept., Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt); Saleh, H.G. [Faculty of Girls for Arts and Education, Ain-Shams Univ., Cairo (Egypt)

    2001-11-01

    The ET-RR-1 research reactor has been operated at 2 MW since 1961 using EK-10 fuel elements with 10% enriched uranium. The reactor has been used for nuclear applied research and isotope production. In order to upgrade the reactor power to a reasonable limit facing up-to-date uses, core conversion by a new type of fuel element available is necessary. Two fuel elements in plate type are suggested in this study to be used in the ET-RR-1 reactor core rather than the utilized ones. The first element has a dimension of 8 x 8 x 50 cm and consists of 19.7% enriched uranium, which is typical for that utilized in the ET-RR-2 reactor, but with a different length. The other element is proposed with a dimension of 7 x 7 x 50 cm and has the same uranium enrichment. To accomplish safety requirements for these fuel elements, thermal-hydraulic evaluation has been carried out using the PARET code. To reach a core conversion of the ET-RR-1 reactor with the above two types of fuel elements, neutronic calculations have been performed using WIMSD4, DIXY2 and EREBUS codes. Some important nuclear parameters needed in the physical design of the reactor were calculated and included in this study. (orig.) [German] Der ET-RR-1 Forschungsreaktor wird seit 1961 unter Verwendung von EK-10 Brennelementen mit einer Leistung von 2 MW betrieben. Der Reaktor wird in der angewandten Forschung und zur Isotopenherstellung eingesetzt. Um die Reaktorleistung im Hinblick auf eine zeitgemaesse Nutzung der Anlage in einem vernuenftigen Mass zu erhoehen, ist eine Umwandlung des Kerns durch Verwendung neuartiger Brennelemente noetig. In der vorliegenden Untersuchung wird vorgeschlagen, anstelle der z. Z. verwendeten Elemente zwei neue, plattenfoermige Brennelemente zu verwenden. Das erste Element hat eine Groesse von 8 x 8 x 50 cm und besteht aus 19,7% angereichertem Uran, was den im ET-RR-2 Reaktor verwendeten Elementen entspricht, allerdings mit einer anderen Groesse. Das zweite Element hat die gleiche

  2. Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

    Science.gov (United States)

    Kim, Sung-Kyu; Kim, Kwangmin; Park, Minwon; Yu, In-Keun; Lee, Sangjin

    2015-11-01

    High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

  3. Numerical Analysis of Magnetic Force of Dry-Type Air-Core Reactor

    Institute of Scientific and Technical Information of China (English)

    LIUZhi-gang; GENGYing-san; WANGJian-hua

    2004-01-01

    This paper presents a coupled magnetic-circuit method for computing the magnetic force of air-core reactor under short-time current. The current and the magnetic flux density are computed first and then the magnetic force is obtained. Thus, the dynamic stability performance of air-core reactor can be analyzed at the design stage to reduce experimental cost and shorten the lead-time of product development.

  4. Mechanical, chemical and radiological characterization of the graphite of the UNGG reactors type; Caracterisation mecanique, chimique et radiologique du graphite des reacteurs de la filiere UNGG

    Energy Technology Data Exchange (ETDEWEB)

    Bresard, I.; Bonal, J.P

    2000-07-01

    In the framework of UNGG reactors type dismantling procedures, the characterization of the graphite, used as moderator, has to be realized. This paper presents the mechanical, chemical and radiological characterizations, the properties measured and gives some results in the case of the Bugey 1 reactor. (A.L.B.)

  5. Pyrolysis of aseptic packages (tetrapak) in a laboratory screw type reactor and secondary thermal/catalytic tar decomposition.

    Science.gov (United States)

    Haydary, J; Susa, D; Dudáš, J

    2013-05-01

    Pyrolysis of aseptic packages (tetrapak cartons) in a laboratory apparatus using a flow screw type reactor and a secondary catalytic reactor for tar cracking was studied. The pyrolysis experiments were realized at temperatures ranging from 650 °C to 850 °C aimed at maximizing of the amount of the gas product and reducing its tar content. Distribution of tetrapak into the product yields at different conditions was obtained. The presence of H2, CO, CH4, CO2 and light hydrocarbons, HCx, in the gas product was observed. The Aluminum foil was easily separated from the solid product. The rest part of char was characterized by proximate and elemental analysis and calorimetric measurements. The total organic carbon in the tar product was estimated by elemental analysis of tars. Two types of catalysts (dolomite and red clay marked AFRC) were used for catalytic thermal tar decomposition. Three series of experiments (without catalyst in a secondary cracking reactor, with dolomite and with AFRC) at temperatures of 650, 700, 750, 800 and 850 °C were carried out. Both types of catalysts have significantly affected the content of tars and other components in pyrolytic gases. The effect of catalyst on the tetrapack distribution into the product yield on the composition of gas and on the total organic carbon in the tar product is presented in this work.

  6. Properties of bio-oil generated by a pyrolysis of forest cedar residuals with the movable Auger-type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nishimura, Shun; Ebitani, Kohki, E-mail: ebitani@jaist.ac.jp [School of Materials Science, Japan Advanced Institute of Science and Technology, 1-1 Asahidai, Nomi, Ishikawa 923-1292 (Japan); Miyazato, Akio [Nanotechnology Center, Japan Advanced Institute of Science and Technology, 1-1 Asahidai, Nomi, Ishikawa 923-1292 (Japan)

    2016-02-01

    Our research project has developed the new movable reactor for bio-oil production in 2013 on the basis of Auger-type system. This package would be a great impact due to the concept of local production for local consumption in the hilly and mountainous area in not only Japan but also in the world. Herein, we would like to report the properties of the bio-oil generated by the developing Auger-type movable reactor. The synthesized bio-oil possessed C: 46.2 wt%, H: 6.5 wt%, N: wt%, S: <0.1 wt%, O: 46.8 wt% and H{sub 2}O: 18.4 wt%, and served a good calorific value of 18.1 MJ/kg. The spectroscopic and mass analyses such as FT-IR, GC-MS, {sup 13}C-NMR and FT-ICR MS supported that the bio-oil was composed by the fine mixtures of methoxy phenols and variety of alcohol or carboxylic acid functional groups. Thus, it is suggested that the bio-oil generated by the new movable Auger-type reactor has a significant potential as well as the existing bio-oil reported previously.

  7. Control of fermentation types in continuous-flow acidogenic reactors: effects of pH and redox potential

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The experiments were carried out in continuous-flow acidogenic reactors with molasses used as sub strate to study the effects of pH and redox potential on fermentation types. The conditions for each fermentation type were investigated at different experimental stages of start-up, pH-regulating and redox potential-regulating.The experiments confirmed that butyric acid-type fermentation would occur at pH > 6, the propionic acid-type fermentation at pH about 5.5 with Eh > - 278 mV, and the ethanol-type fermentation at pH < 4.5. A higher redox potential will lead to propionic acid-type fermentation because propionogens are facultative anaerobic bacteria.

  8. Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T [ORNL

    2014-05-01

    Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

  9. A note on the evaluation of the guest-gas uptake into a clathrate hydrate being formed in a semibatch- or batch-type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Yasuhiko H.; Komae, Naoya [Department of Mechanical Engineering, Keio University, 3-14-1 Hiyoshi, Kohoku-ku, Yokohama 223-8522 (Japan)

    2008-05-15

    This paper deals with the principle of determining the rate of guest-gas uptake into a clathrate hydrate being formed in a semibatch-type isobaric reactor or a batch-type closed reactor on the basis of experimental data for the guest-gas supply into the reactor or the pressure change inside the reactor. The specific issue considered here is the possible necessity of taking into account the effect of the change in the total volume of the condensed (liquid + hydrate) phases inside the reactor during each hydrate-forming operation. General schemes for evaluating this effect in semibatch and batch operations are formulated and applied to some specific hydrate-forming operations to evaluate the effect on estimating the guest-gas uptake into the hydrate. (author)

  10. The reactor core TRIGA Mark-III with fuels type 30/20; El nucleo del reactor TRIGA Mark-III con combustible tipo 30/20

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F., E-mail: fortunato.aguilar@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    This work describes the calculation series carried out with the program MCNP5 in order to define the configuration of the reactor core with fuels 30/20 (fuels with 30% of uranium content in the Or-Zr-H mixture and a nominal enrichment of 20%). To select the configuration of the reactor core more appropriate to the necessities and future uses of the reactor, the following criterions were taken into account: a) the excess in the reactor reactivity, b) the switch out margin and c) to have new irradiation facilities inside the reactor core. Taking into account these criterions is proceeded to know the characteristics of the components that form the reactor core (dimensions, geometry, materials, densities and positions), was elaborated a base model of the reactor core, for the MCNP5 code, with a configuration composed by 85 fuel elements, 4 control bars and the corresponding structural elements. The high reactivity excess obtained with this model, gave the rule to realize other models of the reactor core in which the reactivity excess and the switch out margin were approximate to the values established in the technical specifications of the reactor operation. Several models were realized until finding the satisfactory model; this is composite for 74 fuels, 4 control bars and 6 additional experimental positions inside the reactor core. (Author)

  11. Application of a novel type impinging streams reactor in solid-liquid enzyme reactions and modeling of residence time distribution using GDB model.

    Science.gov (United States)

    Fatourehchi, Niloufar; Sohrabi, Morteza; Dabir, Bahram; Royaee, Sayed Javid; Haji Malayeri, Adel

    2014-02-01

    Solid-liquid enzyme reactions constitute important processes in biochemical industries. The isomerization of d-glucose to d-fructose, using the immobilized glucose isomerase (Sweetzyme T), as a typical example of solid-liquid catalyzed reactions has been carried out in one stage and multi-stage novel type of impinging streams reactors. Response surface methodology was applied to determine the effects of certain pertinent parameters of the process namely axial velocity (A), feed concentration (B), nozzles' flow rates (C) and enzyme loading (D) on the performance of the apparatus. The results obtained from the conversion of glucose in this reactor were much higher than those expected in conventional reactors, while residence time was decreased dramatically. Residence time distribution (RTD) in a one-stage impinging streams reactor was investigated using colored solution as the tracer. The results showed that the flow pattern in the reactor was close to that in a continuous stirred tank reactor (CSTR). Based on the analysis of flow region in the reactor, gamma distribution model with bypass (GDB) was applied to study the RTD of the reactor. The results indicated that RTD in the impinging streams reactor could be described by the latter model.

  12. Design and analysis of 19 pin annular fuel rod cluster for pressure tube type boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Deokule, A.P., E-mail: abhijit.deokule1986@gmail.com [Homi Bhabha National Institute, Trombay 400 085, Mumbai (India); Vishnoi, A.K.; Dasgupta, A.; Umasankari, K.; Chandraker, D.K.; Vijayan, P.K. [Bhabha Atomic Research Centre, Trombay 400 085, Mumbai (India)

    2014-09-15

    Highlights: • Development of 19 pin annular fuel rod cluster. • Reactor physics study of designed annular fuel rod cluster. • Thermal hydraulic study of annular fuel rod cluster. - Abstract: An assessment of 33 pin annular fuel rod cluster has been carried out previously for possible use in a pressure tube type boiling water reactor. Despite the benefits such as negative coolant void reactivity and larger heat transfer area, the 33 pin annular fuel rod cluster is having lower discharge burn up as compared to solid fuel rod cluster when all other parameters are kept the same. The power rating of this design cannot be increased beyond 20% of the corresponding solid fuel rod cluster. The limitation on the power is not due to physics parameters rather it comes from the thermal hydraulics side. In order to increase power rating of the annular fuel cluster, keeping same pressure tube diameter, the pin diameter was increased, achieving larger inside flow area. However, this reduces the number of annular fuel rods. In spite of this, the power of the annular fuel cluster can be increased by 30% compared to the solid fuel rod cluster. This makes the nineteen pin annular fuel rod cluster a suitable option to extract more power without any major changes in the existing design of the fuel. In the present study reactor physics and thermal hydraulic analysis carried out with different annular fuel rod cluster geometry is reported in detail.

  13. Designing an epithermal neutron beam for boron neutron capture therapy for a DIDO type reactor using MCNP

    Science.gov (United States)

    Ross, D.; Constantine, G.; Weaver, D. R.; Beynon, T. D.

    1993-10-01

    This paper describes work undertaken to design an epithermal neutron beam for a DIDO type reactor for use in boron neutron capture therapy, a form of cancer treatment. It involved extensive use of MCNP, a Monte Carlo computer code. Initially, calculations were made with MCNP to simulate earlier experiments with an epithermal beam on the DIDO reactor. This comparison made it possible both to validate the Monte Carlo modelling of the reactor and to gain an insight into the important features of the simulation. Following this, MCNP was used to design a filtered epithermal neutron beam facility for DIDO's largest beam tube, a 13.7 cm radius horizontal tube which extends radially away from the core. First a selection was made of the optimum filter components for the beam. Then the research concentrated on combining these filter elements to construct a practical epithermal beam design. The results suggest that the optimum method of generating the epithermal neutron source is to employ a filter combination consisting principally of liquid argon with the addition of cadmium, aluminium, titanium and possibly tin. The calculations also show that the resultant neutron beam would have a flux greater than 1.0 × 10 9 n cm -2 s -1 and have sufficiently low fast-neutron and gamma-ray contamination.

  14. Development of a resonant-type microwave reactor and its application to the synthesis of positron emission tomography radiopharmaceuticals.

    Science.gov (United States)

    Kimura, Hiroyuki; Yagi, Yusuke; Ohneda, Noriyuki; Odajima, Hiro; Ono, Masahiro; Saji, Hideo

    2014-10-01

    Microwave technology has been successfully applied to enhance the effectiveness of radiolabeling reactions. The use of a microwave as a source of heat energy can allow chemical reactions to proceed over much shorter reaction times and in higher yields than they would do under conventional thermal conditions. A microwave reactor developed by Resonance Instrument Inc. (Model 520/521) and CEM (PETWave) has been used exclusively for the synthesis of radiolabeled agents for positron emission tomography by numerous groups throughout the world. In this study, we have developed a novel resonant-type microwave reactor powered by a solid-state device and confirmed that this system can focus microwave power on a small amount of reaction solution. Furthermore, we have demonstrated the rapid and facile radiosynthesis of 16α-[(18)F]fluoroestradiol, 4-[(18)F]fluoro-N-[2-(1-methoxyphenyl)-1-piperazinyl]ethyl-N-2-pyridinylbenzamide, and N-succinimidyl 4-[(18)F]fluorobenzoate using our newly developed microwave reactor.

  15. Heat Transfer Calculation on Plate-Type Fuel Assembly of High Flux Research Reactor

    Directory of Open Access Journals (Sweden)

    Daxin Gong

    2015-01-01

    Full Text Available Heat transfer characteristics of fuel assemblies for a high flux research reactor with a neutron trap are numerically investigated in this study. Single-phase turbulence flow is calculated by a commercial code, FLUENT, where the computational objective covers standard and control fuel assemblies. The simulation is carried out with an inlet coolant velocity varying from 4.5 m/s to 7.5 m/s in hot assemblies. The results indicate that the cladding temperature is always lower than the saturation temperature in the calculated ranges. The temperature rise in the control fuel assembly is smaller than that of the standard fuel assembly. Additionally, the assembly with a hot spot is specially studied, and the safety of the research reactor is also approved.

  16. CARBONACEOUS, NITROGENOUS AND PHOSPHORUS MATTERS REMOVAL FROM DOMESTIC WASTEWATER BY AN ACTIVATED SLUDGE REACTOR OF NITRIFICATION-DENITRIFICATION TYPE

    Directory of Open Access Journals (Sweden)

    MOHAMAD ALI FULAZZAKY

    2009-03-01

    Full Text Available This paper proposes an environmental engineering method based on biotechnology approach as one of the expected solutions that should be considered to implementing the activated sludge for improving the quality of water and living environment, especially to remove the major pollutant elements of domestic wastewater. Elimination of 3 major pollutant elements, i.e., carbon, nitrogen and phosphor containing the domestic wastewater is proposed to carry out biological method of an anoxic-aerobic reactor therein these types of pollutants should be consecutively processed in three steps. Firstly, eliminate the carbonaceous matter in the aerobic reactor. Secondly, to remove the carbonaceous and nitrogenous matters, it is necessary to modify the reactor’s nature from the aerobic condition to an anoxic-aerobic reactor. And finally, when the cycle of nitrification-denitrification is stable to achieve the target’s efficiency of reactor by adding the ferric iron into the activated sludge, it can be continued to remove the carbonaceous, nitrogenous and phosphorous matters simultaneously. The efficiency of carbonaceous and nitrogenous matters removal was confirmed with the effluent standard, COD is less than 100 mgO2/L and the value of global nitrogen is less than 10 mgN/L. The effectiveness of suspended matter removal is higher than 90% and the decantation of activated sludge is very good as identifying the Molhman’s index is below of 120 mL/L. The total phosphorus matter removal is more effective than the soluble phosphorus matter. By maintaining the reactor’s nature at the suitable condition, identifying the range of pH between 6.92 and 7.16 therefore the excellent abatement of phosphor of about 80% is achieving with the molar Fe/P ratio of 1.4.

  17. Determination of the Quantity of I-135 Released from the AGR Experiment Series

    Energy Technology Data Exchange (ETDEWEB)

    Scates, Dawn Marie [Idaho National Laboratory; Walter, John Bradley [Idaho National Laboratory; Reber, Edward Lawrence [Idaho National Laboratory; Sterbentz, James William [Idaho National Laboratory; Petti, David Andrew [Idaho National Laboratory

    2014-10-01

    A series of three Advanced Gas Reactor (AGR) experiments have been conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). From 2006 through 2014, these experiments supported the development and qualification of the new U.S. tri structural isotropic (TRISO) particle fuel for Very High Temperature Reactors (VHTR). Each AGR experiment consisted of multiple fueled capsules, each plumbed for independent temperature control using a mix of helium and neon gases. The gas leaving a capsule was routed to individual Fission Product Monitor (FPM) detectors. For intact fuel particles, the TRISO particle coatings provide a substantial barrier to fission product release. However, particles with failed coatings, whether because of a minute percentage of initially defective particles, those which fail during irradiation, or those designed to fail (DTF) particles, can release fission products to the flowing gas stream. Because reactive fission product elements like iodine and cesium quickly deposit on cooler capsule components and piping structures as the effluent gas leaves the reactor core, only the noble fission gas isotopes of Kr and Xe tend to reach FPM detectors. The FPM system utilizes High Purity Germanium (HPGe) detectors coupled with a thallium activated sodium iodide NaI(Tl) scintillator. The germanium detector provides individual isotopic information, while the NaI(Tl) scintillator is used as a gross count rate meter. During irradiation, the 135mXe concentration reaching the FPM detectors is from both direct fission and by decay of the accumulated 135I. About ~2.5 hours after irradiation (ten 15.3 minute 135mXe half lives) the directly produced 135mXe has decayed and only the longer lived 135I remains as a source. Decay systematics dictate that 135mXe will be in secular equilibrium with its 135I parent, such that it’s production rate very nearly equals the decay rate of the parent, and its concentration in the flowing gas stream will appear to

  18. Validation of the Physics Analysis used to Characterize the AGR-1 TRISO Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James W.; Harp, Jason M.; Demkowicz, Paul A.; Hawkes, Grant L.; Chang, Gray S.

    2015-05-01

    The results of a detailed physics depletion calculation used to characterize the AGR-1 TRISO-coated particle fuel test irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory are compared to measured data for the purpose of validation. The particle fuel was irradiated for 13 ATR power cycles over three calendar years. The physics analysis predicts compact burnups ranging from 11.30-19.56% FIMA and cumulative neutron fast fluence from 2.21?4.39E+25 n/m2 under simulated high-temperature gas-cooled reactor conditions in the ATR. The physics depletion calculation can provide a full characterization of all 72 irradiated TRISO-coated particle compacts during and post-irradiation, so validation of this physics calculation was a top priority. The validation of the physics analysis was done through comparisons with available measured experimental data which included: 1) high-resolution gamma scans for compact activity and burnup, 2) mass spectrometry for compact burnup, 3) flux wires for cumulative fast fluence, and 4) mass spectrometry for individual actinide and fission product concentrations. The measured data are generally in very good agreement with the calculated results, and therefore provide an adequate validation of the physics analysis and the results used to characterize the irradiated AGR-1 TRISO fuel.

  19. NUCLEAR REACTOR

    Science.gov (United States)

    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.

  20. Effect of application rates and media types on nitrogen and surfactant removal in trickling filters applied to the post-treatment of effluents from UASB reactors

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, P. G. S. de; Taveres, F. v. F.; Chernicharo, C. A. I.

    2009-07-01

    Tricking filters are a very promising alternative for the post treatment of effluents from UASB reactors treating domestic sewage,especially in developing countries. Although a fair amount of information is already available regarding organic mater removal in this combined system, very little is known in relation to nitrogen and surfactant removal in trickling filters post-UASB reactors. Therefore, the purpose of this study was to evaluate and compare the effect evaluate and compare the effect of different application rates and packing media types on trickling filters applied to the post-treatment of effluents from UASB reactors, regarding the removal of ammonia nitrogen and surfactants. (Author)

  1. Analysis on the `Thermite` reaction consequences in accidents involving research reactors using plate-type fuel; Analisis sobre las concequencias de la reaccion `Termita` en caso de accidentes en reactores de investigacion que utilizan combustible tipo placa

    Energy Technology Data Exchange (ETDEWEB)

    Boero, Norma L.; Bruno, Hernan R.; Camacho, Esteban F.; Cincotta, Daniel O.; Yorio, Daniel [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Constituyentes

    1999-11-01

    The mixture of Al-U{sub 3} O{sub 8} is not in a state of chemical equilibrium, and at temperatures of between 850 deg C and 1000 deg C, it reacts exo thermally. This is known, in corresponding bibliography as a `Thermite reaction. This mixture is used in the manufacturing of the plate-type fuel used in research reactors. It has been pointed out that the release of energy caused by this type of reactions might represent a risk in case of accidents in this type of reactor. Conclusions, in general, tend to indicate that no such risk exists, although no concrete assurance is given that this is the case, and this fact, therefore, leaves room for doubt. The objective of this paper is to provide an in-depth study of what happens to a fuel plate when it is subjected to thermite reaction. We will, furthermore, analyze the consequences of the release of energy generated by this type of reaction within the core of the reactor, clearly defining the problem for this type of fuel and this kind of reactor. (author) 3 refs., 9 figs., 1 tab.

  2. Bioregeneration of perchlorate-laden gel-type anion-exchange resin in a fluidized bed reactor.

    Science.gov (United States)

    Venkatesan, Arjun K; Sharbatmaleki, Mohamadali; Batista, Jacimaria R

    2010-05-15

    Selective ion-exchange resins are very effective to remove perchlorate from contaminated waters. However, these resins are currently incinerated after one time use, making the ion-exchange process incomplete and unsustainable for perchlorate removal. Resin bioregeneration is a new concept that combines ion-exchange with biological reduction by directly contacting perchlorate-laden resins with a perchlorate-reducing bacterial culture. In this research, feasibility of the bioregeneration of perchlorate-laden gel-type anion-exchange resin was investigated. Bench-scale bioregeneration experiments, using a fluidized bed reactor and a bioreactor, were performed to evaluate the feasibility of the process and to gain insight into potential mechanisms that control the process. The results of the bioregeneration tests suggested that the initial phase of the bioregeneration process might be controlled by kinetics, while the later phase seems to be controlled by diffusion. Feasibility study showed that direct bioregeneration of gel-type resin was effective in a fluidized-bed reactor, and that the resin could be defouled, reused, and repeatedly regenerated using the method applied in this research.

  3. Bioregeneration of perchlorate-laden gel-type anion-exchange resin in a fluidized bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Venkatesan, Arjun K.; Sharbatmaleki, Mohamadali [Department of Civil and Environmental Engineering, University of Nevada Las Vegas (UNLV), 4505 Maryland Parkway, Las Vegas, NV 89154-4015 (United States); Batista, Jacimaria R., E-mail: jaci@ce.unlv.edu [Department of Civil and Environmental Engineering, University of Nevada Las Vegas (UNLV), 4505 Maryland Parkway, Las Vegas, NV 89154-4015 (United States)

    2010-05-15

    Selective ion-exchange resins are very effective to remove perchlorate from contaminated waters. However, these resins are currently incinerated after one time use, making the ion-exchange process incomplete and unsustainable for perchlorate removal. Resin bioregeneration is a new concept that combines ion-exchange with biological reduction by directly contacting perchlorate-laden resins with a perchlorate-reducing bacterial culture. In this research, feasibility of the bioregeneration of perchlorate-laden gel-type anion-exchange resin was investigated. Bench-scale bioregeneration experiments, using a fluidized bed reactor and a bioreactor, were performed to evaluate the feasibility of the process and to gain insight into potential mechanisms that control the process. The results of the bioregeneration tests suggested that the initial phase of the bioregeneration process might be controlled by kinetics, while the later phase seems to be controlled by diffusion. Feasibility study showed that direct bioregeneration of gel-type resin was effective in a fluidized-bed reactor, and that the resin could be defouled, reused, and repeatedly regenerated using the method applied in this research.

  4. AGR-2 Final Data Qualification Report for U.S. Capsules - ATR Cycles 147A Through 154B

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh T. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Very High-Temperature Reactor Technology Development Office; Einerson, Jeffrey J. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Very High-Temperature Reactor Technology Development Office

    2014-07-01

    This report provides the data qualification status of AGR-2 fuel irradiation experimental data in four U.S. capsules from all 15 Advanced Test Reactor (ATR) Cycles 147A, 148A, 148B, 149A, 149B, 150A, 150B, 151A, 151B, 152A, 152B, 153A, 153B, 154A, and 154B, as recorded in the Nuclear Data Management and Analysis System (NDMAS). Thus, this report covers data qualification status for the entire AGR-2 irradiation and will replace four previously issued AGR-2 data qualification reports (e.g., INL/EXT-11-22798, INL/EXT-12-26184, INL/EXT-13-29701, and INL/EXT-13-30750). During AGR-2 irradiation, two cycles, 152A and 153A, occurred when the ATR core was briefly at low power, so AGR-2 irradiation data are not used for physics and thermal calculations. Also, two cycles, 150A and 153B, are Power Axial Locator Mechanism (PALM) cycles when the ATR power is higher than during normal cycles. During the first PALM cycle, 150A, the experiment was temporarily moved from the B-12 location to the ATR water canal and during the second PALM cycle, 153B, the experiment was temporarily moved from the B-12 location to the I-24 location to avoid being overheated. During the “Outage” cycle, 153A, seven flow meters were installed downstream from seven Fission Product Monitoring System (FPMS) monitors to measure flows from the monitors and these data are included in the NDMAS database.

  5. Cloning of an agr homologue of Staphylococcus saprophyticus.

    Science.gov (United States)

    Sakinc, Türkan; Kulczak, Pawel; Henne, Karsten; Gatermann, Sören G

    2004-08-01

    An agr homologue of Staphylococcus saprophyticus was identified, cloned and sequenced. The gene locus shows homologies to other staphylococcal agr systems, especially to those of S. epidermidis and S. lugdunensis. A putative RNAIII was identified and found to be differentially expressed during the growth phases. In contrast to the RNAIII molecules of S. epidermidis and S. aureus it does not contain an open reading frame that codes for a protein with homologies to the delta-toxin. Using PCR, the agr was found to be present in clinical isolates of S. saprophyticus.

  6. Analysis of Task Types and Error Types of the Human Actions Involved in the Human-related Unplanned Reactor Trip Events

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Whan; Park, Jin Kyun; Jung, Won Dea

    2008-02-15

    This report provides the task types and error types involved in the unplanned reactor trip events that have occurred during 1986 - 2006. The events that were caused by the secondary system of the nuclear power plants amount to 67 %, and the remaining 33 % was by the primary system. The contribution of the activities of the plant personnel was identified as the following order: corrective maintenance (25.7 %), planned maintenance (22.8 %), planned operation (19.8 %), periodic preventive maintenance (14.9 %), response to a transient (9.9 %), and design/manufacturing/installation (9.9%). According to the analysis of error modes, the error modes such as control failure (22.2 %), wrong object (18.5 %), omission (14.8 %), wrong action (11.1 %), and inadequate (8.3 %) take up about 75 % of all the unplanned trip events. The analysis of the cognitive functions involved showed that the planning function makes the highest contribution to the human actions leading to unplanned reactor trips, and it is followed by the observation function (23.4%), the execution function (17.8 %), and the interpretation function (10.3 %). The results of this report are to be used as important bases for development of the error reduction measures or development of the error mode prediction system for the test and maintenance tasks in nuclear power plants.

  7. Analysis of simulation results of damaged nuclear fuel accidents at NPPs with shell-type nuclear reactors

    Directory of Open Access Journals (Sweden)

    Igor L. Kozlov

    2015-03-01

    Full Text Available Lessons from the accident at the Fukushima Daiichi NPP made it necessary to reevaluate and intensificate the work on modeling and analyzing various scenarios of severe accidents with damage to the nuclear fuel in the reactor, containment and spent nuclear fuel storage pool with the expansion of the primary initiating event causes group listing. Further development of computational tools for modeling the explosion prevention criteria as to steam and gas mixtures, considering the specific thermal-hydrodynamic conditions and mechanisms of explosive situations arrival at different stages of a severe accident development, is substantiated. Based on the analysis of the known shell-type nuclear reactors accidents results the explosion safety thermodynamic criteria are presented, the parameters defining the steam and gas explosions conditions are found, the need to perform the further verification and validation of deterministic codes serving to simulate general accident processes behavior as well as phase-to-phase interaction calculated dependencies is established. The main parameters controlling and defining the criteria explosion safety effective regulation areas and their optimization conditions are found.

  8. Influences of Excess Oscillation of Voltage Pulse and Discharge Mode on NO Removal Using Barrier-Type Plasma Reactor

    Science.gov (United States)

    Kadowaki, Kazunori; Suzuki, Yoshiaki; Ihori, Haruo; Kitani, Isamu

    This paper presents experimental results of NO removal from a simulated exhausted-gas using a barrier type reactor with screw electrodes subjected to polarity-reversed voltage pulses. The polarity-reversed pulse was produced by direct grounding of a charged coaxial cable because a traveling wave voltage was negatively reflected at the grounding end with a change in its polarity and then it propagated to the plasma reactor at the opposite end. Influence of cable length on NO removal was studied for two kinds of cable connection, single-connected cable and parallel-connected cables. NO removal ratio for a 50m-long cable was lower than that for much shorter cables in both single and parallel connections when the applied voltage became high. Energy efficiency for NO removal also increased with decreasing the cable length. This was because excess discharges during the voltage oscillation caused by the large stored energy in the long cable resulted in reproduction of NO molecules. Energy efficiency was further improved by changing the discharge mode from dielectric barrier discharge (DBD) to surface discharge (SD). Energy efficiency was up to 110g/kWh with 55% NO removal ratio and 34g/kWh with 100% NO removal ratio by using a single 10m-long cable in SD mode.

  9. Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core

    Energy Technology Data Exchange (ETDEWEB)

    PARMA JR.,EDWARD J.

    2000-01-01

    Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public.

  10. Zoning Districts, ag-res, Published in 2008, Duchesne County.

    Data.gov (United States)

    NSGIC GIS Inventory (aka Ramona) — This Zoning Districts dataset, was produced all or in part from Other information as of 2008. It is described as 'ag-res'. Data by this publisher are often provided...

  11. Design analysis of the molten core confinement within the reactor vessel in the case of severe accidents at nuclear power plants equipped with a reactor of the VVER type

    Science.gov (United States)

    Zvonaryov, Yu. A.; Budaev, M. A.; Volchek, A. M.; Gorbaev, V. A.; Zagryazkin, V. N.; Kiselyov, N. P.; Kobzar', V. L.; Konobeev, A. V.; Tsurikov, D. F.

    2013-12-01

    The present paper reports the results of the preliminary design estimate of the behavior of the core melt in vessels of reactors of the VVER-600 and VVER-1300 types (a standard optimized and informative nuclear power unit based on VVER technology—VVER TOI) in the case of beyond-design-basis severe accidents. The basic processes determining the state of the core melt in the reactor vessel are analyzed. The concept of molten core confinement within the vessel based on the idea of outside cooling is discussed. Basic assumptions and models, as well as the results of calculation of the interaction between molten materials of the core and the wall of the reactor vessel performed by means of the SOCRAT severe accident code, are presented and discussed. On the basis of the data obtained, the requirements on the operation of the safety systems are determined, upon the fulfillment of which there will appear potential prerequisites for implementing the concept of the confinement of the core melt within the reactor in cases of severe accidents at nuclear power plants equipped with VVER reactors.

  12. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH; Desarrollo de modelos de radiacion de un reactor tipo BWR y su instalacion en el SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Barron A, I. [Facultad de Ingenieria, UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: isbarron@yahoo.com.mx

    2005-07-01

    This work about generation models, transport in processes and radioactive contamination of areas of a BWR central, is an amplification to the project developed in the UNAM to have a support tool in subjects or electric generation courses. It is planned about the implementation of models of radiation generation in a BWR type reactor for complement the functions developed in the University Simulator of Nucleo electric- Boiling water reactor (SUN-RAH) which it has been implemented in Simulink of MatLab and it has a model for the dynamics of one nucleo electric central that presents the main characteristics of the reactor vessel, the recirculation system, steam lines, turbines, generator, condensers and feeding water, defined by the main processes that intervene in the generation of energy of these plants. By this way the radiation monitoring systems for area and process, operate simultaneously with the processes of energy generation, with that is possible to observe the changes that present with respect to the operation conditions of the plant, and likewise to appreciate the radiation transport process through the components of the reactor, steam lines and turbines, for different operation conditions and possible faults that they could be presented during the reactor operation. (Author)

  13. Applied methods for mitigation of damage by stress corrosion in BWR type reactors; Metodos aplicados para la mitigacion del dano por corrosion bajo esfuerzo en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C. [Instituto Nacional de Investigaciones Nucleares, Gerencia de Ciencia de Materiales, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

  14. Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    This article presents a theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor. Through numerically solving the one-dimensional steady-state single-phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the steam generator, the natural circulation characteristics were studied. On the basis of the preliminary calculation analysis, it was found that natural circulation mass flow rate was proportional to the exponential function of the power and that the value of the exponent is related to the operating conditions of the secondary side of the steam generator. The higher the outlet pressure of the secondary side of the steam generator, the higher the primary natural circulation mass flow rate. The larger height difference between the core center and the steam generator center is favorable for the heat removal capacity of the natural circulation.

  15. Electrochemical incineration of vinasse in filter-press-type FM01-LC reactor using 3D BDD electrode.

    Science.gov (United States)

    Nava, J L; Recéndiz, A; Acosta, J C; González, I

    2008-01-01

    This work shows results obtained in the electrochemical incineration of a synthetic vinasse with initial chemical oxygen demand (COD) of 75.096 g L(-1) in aqueous media (which resembles vinasse industrial wastewater). Electrolyses in a filter-press-type FM01-LC electrochemical reactor equipped with a three-dimensional (3D) boron doped diamond electrode (BDD) were performed at Reynolds values between 22 unity and energy consumption of 168 KW-h m(-3), at Re =109. The mineralization of vinasse indicates that such degradation occurs via hydroxyl radicals formed by the oxidation of water on the BDD surface. Experimental data revealed that hydrodynamic conditions slightly influence the vinasse degradation rate and current efficiency, indicating that the oxidation involves a complex pathway.

  16. Neutron activation analysis at the Livermore pool-type reactor for the environmental research program. [Identification of trace element contaminants

    Energy Technology Data Exchange (ETDEWEB)

    Ragaini, R.C.; Heft, R.E.; Garvis, D.

    1976-07-02

    Instrumental neutron activation analysis is a technique of trace analysis using measurements of radioactivity induced in the sample by exposure to a source of neutrons. The induced activity is measured by the emitted gamma radiation. Each gamma emitter can then be identified by the energy of the photopeaks produced as the nuclide decays and by the half-life of the neutron-induced activity. A complex computer program GAMANAL has been used to accomplish the major tasks of nuclide identification and quantification. The nuclide data output from GAMANAL is processed by a second computer code NADAC, which develops elemental abundance data from disintegration rates observed. The methods are those employed at the Livermore Pool-Type Reactor in support of the environmental research trace element analysis program. Among the procedures described and discussed are sample preparation, irradiation, analysis, and application of the technique.

  17. Post-irradiation Examination of the AGR-1 Experiment: Plans and Preliminary Results

    Energy Technology Data Exchange (ETDEWEB)

    Paul Demkowicz

    2001-10-01

    Abstract – The AGR-1 irradiation experiment contains seventy-two individual cylindrical fuel compacts (25 mm long x 12.5 mm diameter) each containing approximately 4100 TRISO-coated uranium oxycarbide fuel particles. The experiment accumulated 620 effective full power days in the Advanced Test Reactor at the Idaho National Laboratory with peak burnups exceeding 19% FIMA. An extensive post-irradiation examination campaign will be performed on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature accident testing. PIE experiments will include dimensional measurements of fuel and irradiated graphite, burnup measurements, assessment of fission metals release during irradiation, evaluation of coating integrity using the leach-burn-leach technique, microscopic examination of kernel and coating microstructures, and accident testing of the fuel in helium at temperatures up to 1800°C. Activities completed to date include opening of the irradiated capsules, measurement of fuel dimensions, and gamma spectrometry of selected fuel compacts.

  18. AGR-3/4 Data Qualification Report for ATR Cycles 151A, 151B, 152A, 152B, 154A, and 154B

    Energy Technology Data Exchange (ETDEWEB)

    Binh T. Pham

    2014-02-01

    This data report provides the qualification status of Advanced Gas Reactor-3/4 (AGR-3/4) fuel irradiation experimental data from Advanced Test Reactor (ATR) Cycles 151A, 151B, 152A, 152B, 154A, and 154B, as recorded in the Nuclear Data Management and Analysis System (NDMAS). Of these cycles, ATR Cycle 152A is a low power cycle that occurred when the ATR core was briefly at low power. The irradiation data are not used for physics and thermal calculation, but the qualification status of these cycle data is still covered in this report. On the other hand, during ATR Cycles 153A (unplanned Outage cycle) and 153B (Power Axial Locator Mechanism [PALM] cycle), the AGR-3/4 was pulled out from the ATR core and stored in the canal to avoid being overheated. Therefore, qualification of the AGR-3/4 irradiation data from these 2 cycles was excluded in this report. By the end of ATR Cycle 154B, AGR-3/4 was irradiated for a total of 264.1 effective full power days. The AGR-3/4 data streams addressed in this report include thermocouple (TC) temperatures, sweep gas data (flow rates, pressure, and moisture content), and Fission Product Monitoring System (FPMS) data (release rates and release-to-birth rate ratios [R/Bs]) for each of the twelve capsules in the AGR-3/4 experiment. The final data qualification status for these data streams is determined by a Data Review Committee (DRC) composed of AGR technical leads, Sitewide Quality Assurance (QA), and NDMAS analysts. The DRC convened on February 12, 2014, reviewed the data acquisition process, and considered whether the data met the requirements for data collection as specified in QA-approved Very High Temperature Reactor (VHTR) Technology Development Office (TDO) data collection plans. The DRC also examined the results of NDMAS data testing and statistical analyses, and confirmed the qualification status of the data as given in this report.

  19. Evaluation of plate type fuel options for small power reactors; Avaliacao de alternativas de combustivel tipo placa para reatores de pequeno porte

    Energy Technology Data Exchange (ETDEWEB)

    Andrzejewski, Claudio de Sa

    2005-07-01

    Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO{sub 2} in stainless steel, of UO{sub 2} in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

  20. Development of Pool-type Sodium-cooled Fast Reactor System Analysis Code%池式钠冷快堆系统分析程序开发

    Institute of Scientific and Technical Information of China (English)

    王晋; 张东辉; 胡文军

    2016-01-01

    针对池式钠冷快堆的特点,在对快堆系统的水力模型、热工模型和中子动力学模型进行详细分类和建模的基础上,利用 FORTRAN95语言开发了可用于池式钠冷快堆事故分析的系统分析程序(FASYS程序)。以中国实验快堆为计算对象对FASYS程序模型进行了初步验证,所获得的结果和试验值与其他系统程序计算值符合良好,证明了所开发的系统分析程序的正确性。%According to the characteristics of pool‐type sodium‐cooled fast reactor ,and with the fast reactor hydraulic model , thermal model and neutron kinetics model thoroughly classified and developed ,a fast reactor system analysis code (FASYS code) was developed by FORTRAN95 language for pool‐type sodium‐cooled fast reactor acci‐dent analysis .Transient conditions in CEFR were calculated with FASYS code and the results were used for code validation .The calculation results are consistent with the test data and other fast reactor system analysis code results , and the correctness of the FASYS code is proved .

  1. Assessment of susceptibility of Type 304 stainless steel to intergranular stress corrosion cracking in simulated Savannah River Reactor environments

    Energy Technology Data Exchange (ETDEWEB)

    Ondrejcin, R.S.; Caskey, C.R. Jr.

    1989-12-01

    Intergranular stress corrosion cracking (IGSCC) of Type 304 stainless steel rate tests (CERT) of specimens machined was evaluated by constant extension from Savannah River Plant (SRP) decontaminated process water piping. Results from 12 preliminary CERT tests verified that IGSCC occurred over a wide range of simulated SRP envirorments. 73 specimens were tested in two statistical experimental designs of the central composite class. In one design, testing was done in environments containing hydrogen peroxide; in the other design, hydrogen peroxide was omitted but oxygen was added to the environment. Prediction equations relating IGSCC to temperature and environmental variables were formulated. Temperature was the most important independent variable. IGSCC was severe at 100 to 120C and a threshold temperature between 40C and 55C was identified below which IGSCC did not occur. In environments containing hydrogen peroxide, as in SRP operation, a reduction in chloride concentration from 30 to 2 ppB also significantly reduced IGSCC. Reduction in sulfate concentration from 50 to 7 ppB was effective in reducing IGSCC provided the chloride concentration was 30 ppB or less and temperature was 95C or higher. Presence of hydrogen peroxide in the environment increased IGSCC except when chloride concentration was 11 ppB or less. Actual concentrations of hydrogen peroxide, oxygen and carbon dioxide did not affect IGSCC. Large positive ECP values (+450 to +750 mV Standard Hydrogen Electrode (SHE)) in simulated SRP environments containing hydrogen peroxide and were good agreement with ECP measurements made in SRP reactors, indicating that the simulated environments are representative of SRP reactor environments. Overall CERT results suggest that the most effective method to reduce IGSCC is to reduce chloride and sulfate concentrations.

  2. Implications of reactor type and conditions on first-order hydrolysis rate assessment of maize

    NARCIS (Netherlands)

    Pabon Pereira, C.P.; Zeeman, G.; Zhao, R.; Ekmekci, B.; Lier, van J.B.

    2009-01-01

    The biodegradability and first-order hydrolysis coefficient of maize silage have been assessed from batch experiments using different types of inoculum and substrate to inocula (S/I) ratios, and from CSTRs working at different hydraulic retention times (HRTs). In the batch experiments, the assessed

  3. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Dawn M. Scates; John (Jack) K Hartwell; John B. Walter

    2008-09-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  4. Research on Three-Phase Magnetic Valve Type Controllable Reactor%三相磁阀式可控电抗器的研究

    Institute of Scientific and Technical Information of China (English)

    李海洋; 赵国生

    2011-01-01

    提出了一种新型的三相磁阀式可控电抗器,并介绍了其结构及原理,对其进行了电磁分析。%The new three-phase magnetic valve type controllable reactor is presented.Its structure and principle are introduced.Its electromagnetic problems are analyzed.

  5. Bellows-Type Accumulators for Liquid Metal Loops of Space Reactor Power Systems

    Science.gov (United States)

    Tournier, Jean-Michel; El-Genk, Mohamed S.

    2006-01-01

    In many space nuclear power systems, the primary and/or secondary loops use liquid metal working fluids, and require accumulators to accommodate the change in the liquid metal volume and maintain sufficient subcooling to avoid boiling. This paper developed redundant and light-weight bellows-type accumulators with and without a mechanical spring, and compared the operating condition and mass of the accumulators for different types of liquid metal working fluids and operating temperatures: potassium, NaK-78, sodium and lithium loops of a total capacity of 50 liters and nominal operating temperatures of 840 K, 860 K, 950 K and 1340 K, respectively. The effects of using a mechanical spring and different structural materials on the design, operation and mass of the accumulators are also investigated. The structure materials considered include SS-316, Hastelloy-X, C-103 and Mo-14Re. The accumulator without a mechanical spring weighs 23 kg and 40 kg for a coolant subcooling of 50 K and 100 K, respectively, following a loss of the fill gas. The addition of a mechanical spring comes with a mass penalty, in favor of higher redundancy and maintaining a higher liquid metal subcooling.

  6. Commercial-Scale Performance Predictions for High-Temperature Electrolysis Plants Coupled to Three Advanced Reactor Types

    Energy Technology Data Exchange (ETDEWEB)

    M. G. McKellar; J. E. O' Brien; J. S. Herring

    2007-09-01

    This report presents results of system analyses that have been developed to assess the hydrogen production performance of commercial-scale high-temperature electrolysis (HTE) plants driven by three different advanced reactor – power-cycle combinations: a high-temperature helium cooled reactor coupled to a direct Brayton power cycle, a supercritical CO2-cooled reactor coupled to a direct recompression cycle, and a sodium-cooled fast reactor coupled to a Rankine cycle. The system analyses were performed using UniSim software. The work described in this report represents a refinement of previous analyses in that the process flow diagrams include realistic representations of the three advanced reactors directly coupled to the power cycles and integrated with the high-temperature electrolysis process loops. In addition, this report includes parametric studies in which the performance of each HTE concept is determined over a wide range of operating conditions. Results of the study indicate that overall thermal-to- hydrogen production efficiencies (based on the low heating value of the produced hydrogen) in the 45 - 50% range can be achieved at reasonable production rates with the high-temperature helium cooled reactor concept, 42 - 44% with the supercritical CO2-cooled reactor and about 33 - 34% with the sodium-cooled reactor.

  7. The Performance Test for Reactor Coolant Pump (RCP) adopting Variable Restriction Orifice Type Control Valve

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.; Bae, B. U.; Cho, Y. J. and others

    2014-05-15

    The design values of the RCPTF are 17.2 MPa, 343 .deg. C, 11.7 m{sup 3}/s, and 13 MW in the maximum pressure, temperature, flow rate, and electrical power, respectively. In the RCPTF, various types of tests can be performed including a hydraulic performance test to acquire a H-Q curve as well seal transient tests, thrust bearing transient test, cost down test, NPSHR verification test, and so on. After a commissioning startup test was successfully perfomed, mechanical structures are improved including a flow stabilizer and variable restriction orifice. Two- branch pipe (Y-branch) was installed to regulate the flow rate in the range of performance tests. In the main pipe, a flow restrictor (RO: Restriction Orifice) for limiting the maximum flow rate was installed. In the branch pipe line, a globe valve and a butterfly valves for regulating the flow rate was located on the each branch line. When the pressure loss of the valve side is smaller than that of the RO side, the flow rate of valve side was increasing and the flow disturbance was occurred in the lower pipe line. Due to flow disturbnace, it is to cause an error when measuring RCP head and flow measurement of the venturi flow meter installed in the lower main pipe line, and thus leading to a decrease in measurement accuracy as a result. To increase the efficiency of the flow control availability of the test facility, the variable restriction orifice (VRO) type flow control valve was designed and manufactured. In the RCPTF in KAERI, the performance tests and various kinds of transient tests of the RCP were successfully performed. In this study, H-Q curve of the pump using the VRO revealed a similar trend to the result from two ROs. The VRO was confirmed to effectively cover the full test range of the flow rate.

  8. Sonochemical Reactors.

    Science.gov (United States)

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  9. Design of an equilibrium nucleus of a BWR type reactor based in a Thorium-Uranium fuel; Diseno de un nucleo de equilibrio de un reactor tipo BWR basado en un combustible de Torio-Uranio

    Energy Technology Data Exchange (ETDEWEB)

    Francois, J.L.; Nunez C, A. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Facultad de Ingenieria-UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)

    2003-07-01

    In this work the design of the reactor nucleus of boiling water using fuel of thorium-uranium is presented. Starting from an integral concept based in a type cover-seed assemble is carried out the design of an equilibrium reload for the nucleus of a reactor like that of the Laguna Verde Central and its are analyzed some of the main design variables like the cycle length, the reload fraction, the burnt fuel, the vacuum distribution, the generation of lineal heat, the margin of shutdown, as well as a first estimation of the fuel cost. The results show that it is feasible to obtain an equilibrium reload, comparable to those that are carried out in the Laguna Verde reactors, with a good behavior of those analyzed variables. The cost of the equilibrium reload designed with the thorium-uranium fuel is approximately 2% high that the uranium reload producing the same energy. It is concluded that it is convenient to include burnable poisons, type gadolinium, in the fuel with the end of improving the reload design, the fuel costs and the margin of shutdown. (Author)

  10. Trabajadores agrícolas y subjetividad en California

    OpenAIRE

    Florencio Posadas Segura

    2014-01-01

    En este trabajo se explora el problema de la subjetividad de los trabajadores agríco - las, partiendo de la hipótesis de que éstos consideran malas sus condiciones de vida y de trabajo. Mediante el trabajo de campo se demostraron sus conocimientos y opi - niones sobre su situación laboral y existencial en el Valle de San Joaquín California durante 2007 y 2008. Se descubrió que la visión del trabajador agrícola está basada en la subjetivación y reinterpretación de sus circunstancias objetivas ...

  11. Are fast explorers slow reactors? Linking personality type and anti-predator behaviour.

    Science.gov (United States)

    Jones, Katherine A; Godin, Jean-Guy J

    2010-02-22

    Response delays to predator attack may be adaptive, suggesting that latency to respond does not always reflect predator detection time, but can be a decision based on starvation-predation risk trade-offs. In birds, some anti-predator behaviours have been shown to be correlated with personality traits such as activity level and exploration. Here, we tested for a correlation between exploration behaviour and response latency time to a simulated fish predator attack in a fish species, juvenile convict cichlids (Amatitlania nigrofasciata). Individual focal fish were subjected to a standardized attack by a robotic fish predator while foraging, and separately given two repeated trials of exploration of a novel environment. We found a strong positive correlation between exploration and time taken to respond to the predator model. Fish that were fast to explore the novel environment were slower to respond to the predator. Our study therefore provides some of the first experimental evidence for a link between exploration behaviour and predator-escape behaviour. We suggest that different behavioural types may differ in how they partition their attention between foraging and anti-predator vigilance.

  12. Scaled-up bioconversion of fish waste to liquid fertilizer using a 5 L ribbon-type reactor.

    Science.gov (United States)

    Dao, Van Thingoc; Kim, Joong Kyun

    2011-10-01

    A scaled-up conversion process of fish waste to liquid fertilizer was performed in a 5 L ribbon-type reactor. Biodegradation was performed by inoculation of autoclaved fish waste with 5.84 × 10(5) CFU mL(-1) of mixed microorganisms for 96 h. As a result, the pH changed from 6.92 to 5.72, the cell number reached 7.28 × 10(5) CFU mL(-1), and approximately 430 g (28.3%) of fish waste was degraded. Analyses indicated that the 96 h culture of inoculated fish waste possessed comparable fertilizing ability to commercial fertilizers in hydroponic culture with amino acid contents of 6.91 g 100 g(-1). Therefore, the scaled-up production achieved a more satisfactory fish waste degradation rate (3.61 g h(-1)) than the flask-scale production (0.24 g h(-1)). The biodegraded broth of fish waste at room temperature did not undergo putrefaction for 6 months due to the addition of 1% lactate.

  13. Improvement of hydrogen production via ethanol-type fermentation in an anaerobic down-flow structured bed reactor.

    Science.gov (United States)

    Anzola-Rojas, Mélida del Pilar; Zaiat, Marcelo; De Wever, Heleen

    2016-02-01

    Although a novel anaerobic down-flow structured bed reactor has shown feasibility and stable performance for a long-term compared to other anaerobic fixed bed systems for continuous hydrogen production, the volumetric rates and yields have so far been too low. In order to improve the performance, an operation strategy was applied by organic loading rate (OLR) variation (12-96 g COD L(-1) d(-1)). Different volumetric hydrogen rates, and yields at the same OLR indicated that the system was mainly driven by the specific organic load (SOL). When SOL was kept between 3.8 and 6.2 g sucrose g(-1) VSS d(-1), the volumetric rates raised from 0.1 to 8.9 L H2 L(-1) d(-1), and the yields were stable around 2.0 mol H2 mol(-1) converted sucrose. Furthermore, hydrogen was produced mainly via ethanol-type fermentation, reaching a total energy conversion rate of 23.40 kJ h(-1) L(-1) based on both hydrogen and ethanol production.

  14. An Innovative Passive Residual Heat Removal System of an Open-Pool Type Research Reactor with Pump Flywheel and Gravity Core Cooling Tank

    Directory of Open Access Journals (Sweden)

    Kwon-Yeong Lee

    2015-01-01

    Full Text Available In an open-pool type research reactor, the primary cooling system can be designed to have a downward flow inside the core during normal operation because of the plate type fuel geometry. There is a flow inversion inside the core from the downward flow by the inertia force of the primary coolant to the upward flow by the natural circulation when the pump is turned off. To delay the flow inversion time, an innovative passive system with pump flywheel and GCCT is developed to remove the residual heat. Before the primary cooling pump starts up, the water level of the GCCT is the same as that of the reactor pool. During the primary cooling pump operation, the water in the GCCT is moved into the reactor pool because of the pump suction head. After the pump stops, the potential head generates a downward flow inside the core by moving the water from the reactor pool to the GCCT and removes the residual heat. When the water levels of the two pools are the same again, the core flow has an inversion of the flow direction, and natural circulation is developed through the flap valves.

  15. Accumulation of radioactive corrosion products on steel surfaces of VVER type nuclear reactors. I. 110mAg

    Science.gov (United States)

    Hirschberg, Gábor; Baradlai, Pál; Varga, Kálmán; Myburg, Gerrit; Schunk, János; Tilky, Péter; Stoddart, Paul

    Formation, presence and deposition of corrosion product radionuclides (such as 60Co, 51Cr, 54Mn, 59Fe and/or 110mAg) in the primary circuits of water-cooled nuclear reactors (PWRs) throw many obstacles in the way of normal operation. During the course of the work presented in this series, accumulations of such radionuclides have been studied at austenitic stainless steel type 08X18H10T (GOST 5632-61) surfaces (this austenitic stainless steel corresponds to AISI 321). Comparative experiments have been performed on magnetite-covered carbon steel (both materials are frequently used in some Soviet VVER type PWRs). For these laboratory-scale investigations a combination of the in situ radiotracer `thin gap' method and voltammetry is considered to be a powerful tool due to its high sensitivity towards the detection of the submonolayer coverages of corrosion product radionuclides. An independent technique (XPS) is also used to characterize the depth distribution and chemical state of various contaminants in the passive layer formed on austenitic stainless steel. In the first part of the series the accumulation of 110mAg has been investigated. Potential dependent sorption of Ag + ions (cementation) is found to be the predominant process on austenitic steel, while in the case of magnetite-covered carbon steel the silver species are mainly depleted in the form of Ag 2O. The XPS depth profile of Ag gives an evidence about the embedding of metallic silver into the entire passive layer of the austenitic stainless steel studied.

  16. Validation of deterministic and Monte Carlo codes for neutronics calculation of the IRT-type research reactor

    Science.gov (United States)

    Shchurovskaya, M. V.; Alferov, V. P.; Geraskin, N. I.; Radaev, A. I.

    2017-01-01

    The results of the validation of a research reactor calculation using Monte Carlo and deterministic codes against experimental data and based on code-to-code comparison are presented. The continuous energy Monte Carlo code MCU-PTR and the nodal diffusion-based deterministic code TIGRIS were used for full 3-D calculation of the IRT MEPhI research reactor. The validation included the investigations for the reactor with existing high enriched uranium (HEU, 90 w/o) fuel and low enriched uranium (LEU, 19.7 w/o, U-9%Mo) fuel.

  17. Importance of the global regulators Agr and SaeRS in the pathogenesis of CA-MRSA USA300 infection.

    Directory of Open Access Journals (Sweden)

    Christopher P Montgomery

    Full Text Available CA-MRSA infection, driven by the emergence of the USA300 genetic background, has become epidemic in the United States. USA300 isolates are hypervirulent, compared with other CA- and HA-MRSA strains, in experimental models of necrotizing pneumonia and skin infection. Interestingly, USA300 isolates also have increased expression of core genomic global regulatory and virulence factor genes, including agr and saeRS. To test the hypothesis that agr and saeRS promote the observed hypervirulent phenotype of USA300, isogenic deletion mutants of each were constructed in USA300. The effects of gene deletion on expression and protein abundance of selected downstream virulence genes were assessed by semiquantitative real-time reverse-transcriptase PCR (qRT-PCR and western blot, respectively. The effects of gene deletion were also assessed in mouse models of necrotizing pneumonia and skin infection. Deletion of saeRS, and, to a lesser extent, agr, resulted in attenuated expression of the genes encoding α-hemolysin (hla and the Panton-Valentine leukocidin (lukSF-PV. Despite the differences in hla transcription, the toxin was undetectable in culture supernatants of either of the deletion mutants. Deletion of agr, but not saeRS, markedly increased the expression of the gene encoding protein A (spa, which correlated with increased protein abundance. Each deletion mutant demonstrated significant attenuation of virulence, compared with wild-type USA300, in mouse models of necrotizing pneumonia and skin infection. We conclude that agr and saeRS each independently contribute to the remarkable virulence of USA300, likely by means of their effects on expression of secreted toxins.

  18. El Ingeniero Agrícola en la Agroindustria

    Directory of Open Access Journals (Sweden)

    Ramos Sarria Ramiro

    1985-12-01

    Full Text Available En este artículo se da a conocer la capacidad técnica y administrativa que posee el Ingeniero Agrícola, para organizar un sistema coherente que logre definir, planificar y promover una política de desarrollo científico y tecnológico en la agroindustria.

  19. A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Antonio Carlos Pinto Dias

    1993-12-31

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

  20. Discrimination of source reactor type by multivariate statistical analysis of uranium and plutonium isotopic concentrations in unknown irradiated nuclear fuel material.

    Science.gov (United States)

    Robel, Martin; Kristo, Michael J

    2008-11-01

    The problem of identifying the provenance of unknown nuclear material in the environment by multivariate statistical analysis of its uranium and/or plutonium isotopic composition is considered. Such material can be introduced into the environment as a result of nuclear accidents, inadvertent processing losses, illegal dumping of waste, or deliberate trafficking in nuclear materials. Various combinations of reactor type and fuel composition were analyzed using Principal Components Analysis (PCA) and Partial Least Squares Discriminant Analysis (PLSDA) of the concentrations of nine U and Pu isotopes in fuel as a function of burnup. Real-world variation in the concentrations of (234)U and (236)U in the fresh (unirradiated) fuel was incorporated. The U and Pu were also analyzed separately, with results that suggest that, even after reprocessing or environmental fractionation, Pu isotopes can be used to determine both the source reactor type and the initial fuel composition with good discrimination.

  1. H Reactor

    Data.gov (United States)

    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...

  2. Successful scaling-up of self-sustained pyrolysis of oil palm biomass under pool-type reactor.

    Science.gov (United States)

    Idris, Juferi; Shirai, Yoshihito; Andou, Yoshito; Mohd Ali, Ahmad Amiruddin; Othman, Mohd Ridzuan; Ibrahim, Izzudin; Yamamoto, Akio; Yasuda, Nobuhiko; Hassan, Mohd Ali

    2016-02-01

    An appropriate technology for waste utilisation, especially for a large amount of abundant pressed-shredded oil palm empty fruit bunch (OFEFB), is important for the oil palm industry. Self-sustained pyrolysis, whereby oil palm biomass was combusted by itself to provide the heat for pyrolysis without an electrical heater, is more preferable owing to its simplicity, ease of operation and low energy requirement. In this study, biochar production under self-sustained pyrolysis of oil palm biomass in the form of oil palm empty fruit bunch was tested in a 3-t large-scale pool-type reactor. During the pyrolysis process, the biomass was loaded layer by layer when the smoke appeared on the top, to minimise the entrance of oxygen. This method had significantly increased the yield of biochar. In our previous report, we have tested on a 30-kg pilot-scale capacity under self-sustained pyrolysis and found that the higher heating value (HHV) obtained was 22.6-24.7 MJ kg(-1) with a 23.5%-25.0% yield. In this scaled-up study, a 3-t large-scale procedure produced HHV of 22.0-24.3 MJ kg(-1) with a 30%-34% yield based on a wet-weight basis. The maximum self-sustained pyrolysis temperature for the large-scale procedure can reach between 600 °C and 700 °C. We concluded that large-scale biochar production under self-sustained pyrolysis was successfully conducted owing to the comparable biochar produced, compared with medium-scale and other studies with an electrical heating element, making it an appropriate technology for waste utilisation, particularly for the oil palm industry.

  3. Design of the VISTA-ITL Test Facility for an Integral Type Reactor of SMART and a Post-Test Simulation of a SBLOCA Test

    OpenAIRE

    2014-01-01

    To validate the performance and safety of an integral type reactor of SMART, a thermal-hydraulic integral effect test facility, VISTA-ITL, is introduced with a discussion of its scientific design characteristics. The VISTA-ITL was used extensively to assess the safety and performance of the SMART design, especially for its passive safety system such as a passive residual heat removal system, and to validate various thermal-hydraulic analysis codes. The VISTA-ITL program includes several tests...

  4. Daily Thermal Predictions of the AGR-1 Experiment with Gas Gaps Varying with Time

    Energy Technology Data Exchange (ETDEWEB)

    Grant Hawkes; James Sterbentz; John Maki; Binh Pham

    2012-06-01

    A new daily as-run thermal analysis was performed at the Idaho National Laboratory on the Advanced Gas Reactor (AGR) test experiment number one at the Advanced Test Reactor (ATR). This thermal analysis incorporates gas gaps changing with time during the irradiation experiment. The purpose of this analysis was to calculate the daily average temperatures of each compact to compare with experimental results. Post irradiation examination (PIE) measurements of the graphite holder and fuel compacts showed the gas gaps varying from the beginning of life. The control temperature gas gap and the fuel compact – graphite holder gas gaps were linearly changed from the original fabrication dimensions, to the end of irradiation measurements. A steady-state thermal analysis was performed for each daily calculation. These new thermal predictions more closely match the experimental data taken during the experiment than previous analyses. Results are presented comparing normalized compact average temperatures to normalized log(R/B) Kr-85m. The R/B term is the measured release rate divided by the predicted birth rate for the isotope Kr-85m. Correlations between these two normalized values are presented.

  5. NSGA-II Algorithm with a Local Search Strategy for Multiobjective Optimal Design of Dry-Type Air-Core Reactor

    Directory of Open Access Journals (Sweden)

    Chengfen Zhang

    2015-01-01

    Full Text Available Dry-type air-core reactor is now widely applied in electrical power distribution systems, for which the optimization design is a crucial issue. In the optimization design problem of dry-type air-core reactor, the objectives of minimizing the production cost and minimizing the operation cost are both important. In this paper, a multiobjective optimal model is established considering simultaneously the two objectives of minimizing the production cost and minimizing the operation cost. To solve the multi-objective optimization problem, a memetic evolutionary algorithm is proposed, which combines elitist nondominated sorting genetic algorithm version II (NSGA-II with a local search strategy based on the covariance matrix adaptation evolution strategy (CMA-ES. NSGA-II can provide decision maker with flexible choices among the different trade-off solutions, while the local-search strategy, which is applied to nondominated individuals randomly selected from the current population in a given generation and quantity, can accelerate the convergence speed. Furthermore, another modification is that an external archive is set in the proposed algorithm for increasing the evolutionary efficiency. The proposed algorithm is tested on a dry-type air-core reactor made of rectangular cross-section litz-wire. Simulation results show that the proposed algorithm has high efficiency and it converges to a better Pareto front.

  6. An Assessment of ORNL PIE Capabilities for the AGR Program Capsule Post Irradiation Examination

    Energy Technology Data Exchange (ETDEWEB)

    Morris, Robert Noel [ORNL

    2006-09-01

    ORNL has facilities and experienced staff that can execute +the Advanced Gas Reactor (AGR) Post Irradiation Examination (PIE) task. While the specific PIE breakdown needs to be more formally defined, the basic outline is clear and the existing capabilities can be assessed within the needs of the tasks defined in the program plan. A one-to-one correspondence between the program plan tasks and the current ORNL PIE status was conducted and while some shortcomings were identified, the general capability is available. Specific upgrade needs were identified and reviewed. A path forward was formulated. Building 3525 is available for this work and this building is currently receiving renewed attention from management so that it will be in good working order prior to the expected PIE start date. This building is equipped with the tools necessary for PIEs of this nature, but the long hiatus in coated particle fuel work has left it with aging analysis tools. This report identified several of these tools and rough estimates of what would be required to update and replace them. In addition, other ORNL buildings are available to support Building 3525 in specialized tasks along with the normal laboratory infrastructure. Before the AGR management embarks on any equipment development effort, the PIE tasks should be updated against current program (modeling and data) needs and better defined so that the items to be measured, their measurement uncertainties, and thru-put needs can be reviewed. A Data Task Matrix (DTM) should be prepared so that the program data needs can be compared against the identified PIE tasks and what is practical in the hot cell environment to make sure nothing is overlooked. Finally, thought should be given to the development of standardized equipment designs between sites to avoid redundant design efforts and different measurement techniques. This is a potentially cost saving effort that can also avoid data inconsistencies.

  7. Combining models of behaviour with operational data to provide enhanced condition monitoring of AGR cores

    Energy Technology Data Exchange (ETDEWEB)

    West, Graeme M., E-mail: graeme.west@strath.ac.uk; Wallace, Christopher J.; McArthur, Stephen D.J.

    2014-06-01

    Highlights: • Combining laboratory model outputs with operational data. • Isolation of single component from noisy data. • Better understanding of the health of graphite cores. • Extended plant operation through leveraging existing data sources. - Abstract: Installation of new monitoring equipment in Nuclear Power Plants (NPPs) is often difficult and expensive and therefore maximizing the information that can be extracted from existing monitoring equipment is highly desirable. This paper describes the process of combining models derived from laboratory experimentation with current operational plant data to infer an underlying measure of health. A demonstration of this process is provided where the fuel channel bore profile, a measure of core health, is inferred from data gathered during the refuelling process of an Advanced Gas-cooled Reactor (AGR) nuclear power plant core. Laboratory simulation was used to generate a model of an interaction between the fuel assembly and the core. This model is used to isolate a single frictional component from a noisy input signal and use this friction component as a measure of health to assess the current condition of the graphite bricks that comprise the core. In addition, the model is used to generate an expected refuelling response (the noisy input signal) for a given set of channel bore diameter measurements for either insertion of new fuel or removal of spent fuel, providing validation of the model. This benefit of this work is that it provides a greater understanding of the health of the graphite core, which is important for continued and extended operation of the AGR plants in the UK.

  8. Differential expression of anterior gradient gene AGR2 in prostate cancer

    Directory of Open Access Journals (Sweden)

    Knutzen Laura A

    2010-12-01

    Full Text Available Abstract Background The protein AGR2 is a putative member of the protein disulfide isomerase family and was first identified as a homolog of the Xenopus laevis gene XAG-2. AGR2 has been implicated in a number of human cancers. In particular, AGR2 has previously been found to be one of several genes that encode secreted proteins showing increased expression in prostate cancer cells compared to normal prostatic epithelium. Methods Gene expression levels of AGR2 were examined in prostate cancer cells by microarray analysis. We further examined the relationship of AGR2 protein expression to histopathology and prostate cancer outcome on a population basis using tissue microarray technology. Results At the RNA and protein level, there was an increase in AGR2 expression in adenocarcinoma of the prostate compared to morphologically normal prostatic glandular epithelium. Using a tissue microarray, this enhanced AGR2 expression was seen as early as premalignant PIN lesions. Interestingly, within adenocarcinoma samples, there was a slight trend toward lower levels of AGR2 with increasing Gleason score. Consistent with this, relatively lower levels of AGR2 were highly predictive of disease recurrence in patients who had originally presented with high-stage primary prostate cancer (P = 0.009. Conclusions We have shown for the first time that despite an increase in AGR2 expression in prostate cancer compared to non-malignant cells, relatively lower levels of AGR2 are highly predictive of disease recurrence following radical prostatectomy.

  9. Conceptual analysis of a preliminary model for instability study in normal operation of a natural circulation reactor type EBWR, using Relap5/Mod 3.2; Analisis conceptual de un modelo preliminar para el estudio de la inestabilidad en la operacion normal de un reactor de circulacion natural tipo ESBWR, usando Relap5/Mod 3.2

    Energy Technology Data Exchange (ETDEWEB)

    Ojeda S, J.; Morales S, J.; Chavez M, C. [UNAM, Facultad de Ingenieria, Circuito Exterior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)], e-mail: j.os.ojeda@hotmail.com

    2009-10-15

    This work intends a model using the code Relap5/Mod 3.2, for the instability study in normal operation of a natural circulation reactor type ESBWR. A conceptual analysis is considered because all the information was obtained of the open literature, and some of reactor operation or dimension (not available) parameters were approached. As starting point was took the pattern developed for reactor type BWR, denominated Browns Ferry and changes were focused in elimination of bonds of forced recirculation, in modification of operation parameters, dimensions and own control parameters, according to internal code structure. Additionally the nodalization outline is described analyzing for separate the four fundamental areas employees in peculiar geometry of natural circulation reactor. Comparative analysis of results of stability behavior obtained with those reported in the open literature were made, by part of commercial reactor designer ESBWR. (Author)

  10. Hybrid reactors. [Fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.

    1980-09-09

    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 /sup 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/sup -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.

  11. A new type of Neutrino Detector for Sterile Neutrino Search at Nuclear Reactors and Nuclear Nonproliferation Applications

    OpenAIRE

    Lane, C.; Usman, S. M.; Blackmon, J.; Rasco, C.; Mumm, H. P.; Markoff, D.; Jocher, G. R.; Dorrill, R.; Duvall, M.; J. G. Learned; Li, V; Maricic, J.; Matsuno, S.; Milincic, R.; Negrashov, S.

    2015-01-01

    We describe a new detector, called NuLat, to study electron anti-neutrinos a few meters from a nuclear reactor, and search for anomalous neutrino oscillations. Such oscillations could be caused by sterile neutrinos, and might explain the "Reactor Antineutrino Anomaly". NuLat, is made possible by a natural synergy between the miniTimeCube and mini-LENS programs described in this paper. It features a "Raghavan Optical Lattice" (ROL) consisting of 3375 boron or $^6$Li loaded plastic scintillator...

  12. Improving the performance of the Egyptian second testing nuclear research reactor using interval type-2 fuzzy logic controller tuned by modified biogeography-based optimization

    Energy Technology Data Exchange (ETDEWEB)

    Sayed, M.M., E-mail: M.M.Sayed@ieee.org; Saad, M.S.; Emara, H.M.; Abou El-Zahab, E.E.

    2013-09-15

    Highlights: • A modified version of the BBO was proposed. • A novel method for interval type-2 FLC design tuned by MBBO was proposed. • The performance of the ETRR-2 was improved by using IT2FLC tuned by MBBO. -- Abstract: Power stabilization is a critical issue in nuclear reactors. The conventional proportional derivative (PD) controller is currently used in the Egyptian second testing research reactor (ETRR-2). In this paper, we propose a modified biogeography-based optimization (MBBO) algorithm to design the interval type-2 fuzzy logic controller (IT2FLC) to improve the performance of the Egyptian second testing research reactor (ETRR-2). Biogeography-based optimization (BBO) is a novel evolutionary algorithm that is based on the mathematical models of biogeography. Biogeography is the study of the geographical distribution of biological organisms. In the BBO model, problem solutions are represented as islands, and the sharing of features between solutions is represented as immigration and emigration between the islands. A modified version of the BBO is applied to design the IT2FLC to get the optimal parameters of the membership functions of the controller. We test the optimal IT2FLC obtained by modified biogeography-based optimization (MBBO) using the integral square error (ISE) and is compared with the currently used PD controller.

  13. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    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. Influence of Sae-regulated and Agr-regulated factors on the escape of Staphylococcus aureus from human macrophages.

    Science.gov (United States)

    Münzenmayer, Lisa; Geiger, Tobias; Daiber, Ellen; Schulte, Berit; Autenrieth, Stella E; Fraunholz, Martin; Wolz, Christiane

    2016-08-01

    Although Staphylococcus aureus is not a classical intracellular pathogen, it can survive within phagocytes and many other cell types. However, the pathogen is also able to escape from cells by mechanisms that are only partially understood. We analysed a series of isogenic S. aureus mutants of the USA300 derivative JE2 for their capacity to destroy human macrophages from within. Intracellular S. aureus JE2 caused severe cell damage in human macrophages and could efficiently escape from within the cells. To obtain this full escape phenotype including an intermittent residency in the cytoplasm, the combined action of the regulatory systems Sae and Agr is required. Mutants in Sae or mutants deficient in the Sae target genes lukAB and pvl remained in high numbers within the macrophages causing reduced cell damage. Mutants in the regulatory system Agr or in the Agr target gene psmα were largely similar to wild-type bacteria concerning cell damage and escape efficiency. However, these strains were rarely detectable in the cytoplasm, emphasizing the role of phenol-soluble modulins (PSMs) for phagosomal escape. Thus, Sae-regulated toxins largely determine damage and escape from within macrophages, whereas PSMs are mainly responsible for the escape from the phagosome into the cytoplasm. Damage of macrophages induced by intracellular bacteria was linked neither to activation of apoptosis-related caspase 3, 7 or 8 nor to NLRP3-dependent inflammasome activation.

  15. La Misión del Agrónomo

    Directory of Open Access Journals (Sweden)

    Caicedo A.

    1942-08-01

    Full Text Available La agronomía, como todas las profesiones que surgen en el plano de las necesidades sociales, estará sujeta en sus primeros años de existencia a un proceso riguroso de adaptación. La primera fuerza opositora que halla el agrónomo en su labor es la impreparación ciega y el empirismo secular del campesino, quien cree abarcar todos los conocimientos y dominar todos los aspectos de su oficio. Aparte de esta misión exclusivamente social y por demás larga y difícil, caracteriza al agrónomo su calidad de creador de riqueza nacional.

  16. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Pal, Eshita, E-mail: eshi.pal@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Nayak, Arun K.; Vijayan, Pallipattu K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2015-09-15

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days.

  17. Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program; Simulacion en 3 dimensiones de un ciclo de 18 meses para un reactor BWR usando el programa Nod3D

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez, N.; Alonso, G. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: nhm@nuclear.inin.mx; Valle, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

  18. Trabajadores agrícolas y subjetividad en California

    Directory of Open Access Journals (Sweden)

    Florencio Posadas Segura

    2014-01-01

    Full Text Available En este trabajo se explora el problema de la subjetividad de los trabajadores agríco - las, partiendo de la hipótesis de que éstos consideran malas sus condiciones de vida y de trabajo. Mediante el trabajo de campo se demostraron sus conocimientos y opi - niones sobre su situación laboral y existencial en el Valle de San Joaquín California durante 2007 y 2008. Se descubrió que la visión del trabajador agrícola está basada en la subjetivación y reinterpretación de sus circunstancias objetivas de trabajo y de vida. Se concluyó que los trabajadores agrícolas opinan que sus condiciones son pé - simas porque sus derechos laborales, sociales y humanos no se reconocen o respetan, la crisis económica, social y política de Estados Unidos les perjudica y el Tratado de Libre Comercio entre México, Estados Unidos y Canadá ( TLCAN sólo podría beneficiarles con una reforma migratoria.

  19. Phenotypic characterization and virulence of a sae- agr- mutant of Staphylococcus aureus.

    Science.gov (United States)

    Giraudo, A T; Rampone, H; Calzolari, A; Nagel, R

    1996-02-01

    A sae::Tn551 agr::tetM double mutant was constructed and characterized. The production of several exoproteins (e.g., beta-hemolysin, DNase, and proteases) by this mutant was determined and found to be lower than the already diminished production of either isogenic single mutant sae- or agr-. The double mutant also showed, like the agr- mutant, null production of alpha- and delta-hemolysins and diminished levels of lipase. The reduced levels of many exoproteins in the double mutant as compared with their already diminished levels in either single mutant suggest that there is an additive or synergistic interaction between the two mutations involved, sae- and agr-. However, inactivation of both loci, sae and agr, had a different effect on the two exoproteins that are up regulated in the agr- mutant; thus, coagulase dropped to levels close to the null levels of the sae- parental strain, while extracellular protein A displayed the high levels characteristic of the agr- single mutant. The virulence of the sae- agr- double mutant, determined by intraperitoneal injection in mice, was found to be significantly diminished as compared with that of the sae+ agr+ parental strain or the sae- agr+ single mutant.

  20. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. XerC Contributes to Diverse Forms of Staphylococcus aureus Infection via agr-Dependent and agr-Independent Pathways.

    Science.gov (United States)

    Atwood, Danielle N; Beenken, Karen E; Loughran, Allister J; Meeker, Daniel G; Lantz, Tamara L; Graham, Justin W; Spencer, Horace J; Smeltzer, Mark S

    2016-04-01

    We demonstrate that mutation of xerC, which reportedly encodes a homologue of an Escherichia coli recombinase, limits biofilm formation in the methicillin-resistant Staphylococcus aureus strain LAC and the methicillin-sensitive strain UAMS-1. This was not due to the decreased production of the polysaccharide intracellular adhesin (PIA) in either strain because the amount of PIA was increased in a UAMS-1xerC mutant and undetectable in both LAC and its isogenic xerC mutant. Mutation of xerC also resulted in the increased production of extracellular proteases and nucleases in both LAC and UAMS-1, and limiting the production of either class of enzymes increased biofilm formation in the isogenic xerC mutants. More importantly, the limited capacity to form a biofilm was correlated with increased antibiotic susceptibility in both strains in the context of an established biofilm in vivo. Mutation of xerC also attenuated virulence in a murine bacteremia model, as assessed on the basis of the bacterial loads in internal organs and overall lethality. It also resulted in the decreased accumulation of alpha toxin and the increased accumulation of protein A. These findings suggest that xerC may impact the functional status of agr. This was confirmed by demonstrating the reduced accumulation of RNAIII and AgrA in LAC and UAMS-1xerC mutants. However, this cannot account for the biofilm-deficient phenotype of xerC mutants because mutation of agr did not limit biofilm formation in either strain. These results demonstrate that xerC contributes to biofilm-associated infections and acute bacteremia and that this is likely due to agr-independent and -dependent pathways, respectively.

  2. Reactor safeguards

    CERN Document Server

    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

  3. Reactor operation

    CERN Document Server

    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

  4. Análisis para la modelación y optimización geométrica de un reactor tipo tornillo sin-fin empleando el método de grafos dicromáticos//Analysis for geometric modeling and optimization of a worm type reactor using the method of dichromatic graph

    Directory of Open Access Journals (Sweden)

    Armando Díaz-Concepción

    2015-09-01

    Full Text Available En el presente trabajo se realiza la modelación, simulación y optimización de un reactor utilizado en las plantas para la obtención de un alimento animal, sobre la base de la predigestión del bagacillo de caña y el hidróxido de calcio en presencia de vapor denominado PREDICAL utilizando grafos dicromáticos. Se obtuvo el modelo matemático para el diseño del reactor, donde se vinculan las variables geométricas y tecnológicas. El modelo formulado permitió la optimización de la variable costo a partir de minimizar la variable geométrica diámetro exterior del reactor. Palabras claves: modelación reactor tipo tornillo sinfin, grafos dicromáticos, modelo matemático________________________________________________________________________________AbstractThe present work performs modeling, simulation and optimization of a reactor used in plants for the obtencion of animal feed. It's made on the basis of pre-digestion of cane bagasse and calcium hydroxide in the presence of steam called PREDICAL and using dichromatic graphs. It was achieved the mathematical model for the design of the reactor, where are linked geometric and technological variables. The model developed allowed cost optimization based on minimize the geometric variable outside diameter of the reactor. Key words: worm type reactor modeling, dichromatic graphs, mathematical model.

  5. Method for calculating coolant resonance frequencies under normal and accident conditions in nuclear power plants with WWER-type pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N. (Moskovskij Ehnergeticheskij Inst. (USSR))

    1983-03-01

    Mathematical models are proposed for calculating acoustic oscillation resonance frequencies in the coolant in various components of the WWER type primary circuit (core, steam generator, pressurizer, piping). Due to the correspondence between model calculations and experimental results obtained in operating nuclear power plants, the developed models can be used for practical calculations. The possibility of calculating the eigenfrequencies of the coolant oscillation under different operating conditions leads to the interpretation of operational data, to the analysis of operational conditions, to the detection of coolant boiling in the reactor, and to design changes in order to prevent resonance oscillations within the coolant.

  6. Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perneczky, L. [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1997-12-31

    Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

  7. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part); MCTP, un codigo para el analisis termo-mecanico de una barra combustible de reactores tipo BWR (Parte Neutronica)

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: hhl@nuclear.inin.mx

    2003-07-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  8. Life time of nuclear power plants and new types of reactors; La duree de vie des centrales nucleaires et les nouveaux types de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    This report, realized by the Evaluation Parliamentary Office of scientific and technological choices, aims to answer simple but fundamental questions for the french electric power production. What are the phenomena which may limit the exploitation time of nuclear power plants? How can we fight against the aging, at which cost and with which safety? The first chapter presents the management of the nuclear power plants life time, an essential element of the park optimization but not a sufficient element. The second chapter details the EPR and the other reactors for 2015 as a bond between the today and tomorrow parks. The last chapter deals with the necessity of efforts in the research and development to succeed in 2035 and presents other reactors in project. (A.L.B.)

  9. Chitinase expression in Listeria monocytogenes is positively regulated by the Agr system.

    Directory of Open Access Journals (Sweden)

    Dafni Katerina Paspaliari

    Full Text Available The food-borne pathogen Listeria monocytogenes encodes two chitinases, ChiA and ChiB, which allow the bacterium to hydrolyze chitin, the second most abundant polysaccharide in nature. Intriguingly, despite the absence of chitin in human and mammalian hosts, both of the chitinases have been deemed important for infection, through a mechanism that, at least in the case of ChiA, involves modulation of host immune responses. In this study, we show that the expression of the two chitinases is subject to regulation by the listerial agr system, a homologue of the agr quorum-sensing system of Staphylococcus aureus, that has so far been implicated in virulence and biofilm formation. We demonstrate that in addition to these roles, the listerial agr system is required for efficient chitin hydrolysis, as deletion of agrD, encoding the putative precursor of the agr autoinducer, dramatically decreased chitinolytic activity on agar plates. Agr was specifically induced in response to chitin addition in stationary phase and agrD was found to regulate the amount of chiA, but not chiB, transcripts. Although the transcript levels of chiB did not depend on agrD, the extracellular protein levels of both chitinases were reduced in the ΔagrD mutant. The regulatory effect of agr on chiA is potentially mediated through the small RNA LhrA, which we show here to be negatively regulated by agr. LhrA is in turn known to repress chiA translation by binding to the chiA transcript and interfering with ribosome recruitment. Our results highlight a previously unrecognized role of the agr system and suggest that autoinducer-based regulation of chitinolytic systems may be more commonplace than previously thought.

  10. Reactor Neutrinos

    OpenAIRE

    Soo-Bong Kim; Thierry Lasserre; Yifang Wang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  11. Levels of alpha-toxin correlate with distinct phenotypic response profiles of blood mononuclear cells and with agr background of community-associated Staphylococcus aureus isolates.

    Science.gov (United States)

    Mairpady Shambat, Srikanth; Haggar, Axana; Vandenesch, Francois; Lina, Gerard; van Wamel, Willem J B; Arakere, Gayathri; Svensson, Mattias; Norrby-Teglund, Anna

    2014-01-01

    Epidemiological studies of Staphylococcus aureus have shown a relation between certain clones and the presence of specific virulence genes, but how this translates into virulence-associated functional responses is not fully elucidated. Here we addressed this issue by analyses of community-acquired S. aureus strains characterized with respect to antibiotic resistance, ST types, agr types, and virulence gene profiles. Supernatants containing exotoxins were prepared from overnight bacterial cultures, and tested in proliferation assays using human peripheral blood mononuclear cells (PBMC). The strains displayed stable phenotypic response profiles, defined by either a proliferative or cytotoxic response. Although, virtually all strains elicited superantigen-mediated proliferative responses, the strains with a cytotoxic profile induced proliferation only in cultures with the most diluted supernatants. This indicated that the superantigen-response was masked by a cytotoxic effect which was also confirmed by flow cytometry analysis. The cytotoxic supernatants contained significantly higher levels of α-toxin than did the proliferative supernatants. Addition of α-toxin to supernatants characterized as proliferative switched the response into cytotoxic profiles. In contrast, no effect of Panton Valentine Leukocidin, δ-toxin or phenol soluble modulin α-3 was noted in the proliferative assay. Furthermore, a significant association between agr type and phenotypic profile was found, where agrII and agrIII strains had predominantly a proliferative profile whereas agrI and IV strains had a predominantly cytotoxic profile. The differential response profiles associated with specific S. aureus strains with varying toxin production could possibly have an impact on disease manifestations, and as such may reflect specific pathotypes.

  12. Verification Calculation Results to Validate the Procedures and Codes for Pin-by-Pin Power Computation in VVER Type Reactors with MOX Fuel Loading

    Energy Technology Data Exchange (ETDEWEB)

    Chizhikova, Z.N.; Kalashnikov, A.G.; Kapranova, E.N.; Korobitsyn, V.E.; Manturov, G.N.; Tsiboulia, A.A.

    1998-12-01

    One of the important problems for ensuring the VVER type reactor safety when the reactor is partially loaded with MOX fuel is the choice of appropriate physical zoning to achieve the maximum flattening of pin-by-pin power distribution. When uranium fuel is replaced by MOX one provided that the reactivity due to fuel assemblies is kept constant, the fuel enrichment slightly decreases. However, the average neutron spectrum fission microscopic cross-section for {sup 239}Pu is approximately twice that for {sup 235}U. Therefore power peaks occur in the peripheral fuel assemblies containing MOX fuel which are aggravated by the interassembly water. Physical zoning has to be applied to flatten the power peaks in fuel assemblies containing MOX fuel. Moreover, physical zoning cannot be confined to one row of fuel elements as is the case with a uniform lattice of uranium fuel assemblies. Both the water gap and the jump in neutron absorption macroscopic cross-sections which occurs at the interface of fuel assemblies with different fuels make the problem of calculating space-energy neutron flux distribution more complicated since it increases nondiffusibility effects. To solve this problem it is necessary to update the current codes, to develop new codes and to verify all the codes including nuclear-physical constants libraries employed. In so doing it is important to develop and validate codes of different levels--from design codes to benchmark ones. This paper presents the results of the burnup calculation for a multiassembly structure, consisting of MOX fuel assemblies surrounded by uranium dioxide fuel assemblies. The structure concerned can be assumed to model a fuel assembly lattice symmetry element of the VVER-1000 type reactor in which 1/4 of all fuel assemblies contains MOX fuel.

  13. State of art report for critical flow model to analyze a break flow in pressurizer of integral type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Yeon Moon; Lee, D. J.; Yoon, J. H.; Kim, J. P.; Kim, H. Y

    1999-03-01

    At a critical flow condition, the flow rate can't exceed a maximum value for given upstream conditions and the limited flow rate is called as a critical flow rate. The phenomena of critical flow occur at the discharge of a single phase gas or subcooled water through nozzles and pipes. Among the previous researches on critical flow, many accurate correlations on pressure, temperature and flow rate are represented for the single phase gas. However, for the two phase critical flow, the results of previous work showed that there was a large discrepancy between the analytical and experimental data and the data were in agreement for the limited thermodynamic conditions. Thus, further studies are required to enhance the two phase critical flow model. In the integral reactor, the critical flows of nitrogen gas and subcooled water are expected for the break of gas cylinder pipeline connected to the pressurizer. It requires that the inlet shape of the pipe and the nitrogen gas effect should be considered for the critical flow of integral reactor. The nitrogen gas exist in the pressurizer may affect the flow rate of primary coolant, which has been considered only for a few previous researches. Thus, the evaluation of the effect of the nitrogen on the critical flow gas should be preceded for the proper analysis of the critical flow in the integral reactor. In this report, not only the essences of previous work on critical flow were investigated and summarized but also the effect of nitrogen gas and the inlet shape of the pipe on the critical flow were also investigated. (author)

  14. Experimental results of acetone hydrogenation on a heat exchanger type reactor for solar chemical heat pump; Solar chemical heat pump ni okeru acetone suisoka hanno netsu kaishu jikken

    Energy Technology Data Exchange (ETDEWEB)

    Takashima, T.; Doi, T.; Tanaka, T.; Ando, Y. [Electrotechnical Laboratory, Tsukuba (Japan); Miyahara, R.; Kamoshida, J. [Shibaura Institute of Technology, Tokyo (Japan)

    1996-10-27

    With the purpose of converting solar heat energy to industrial heat energy, an experiment of acetone hydrogenation was carried out using a heat exchanger type reactor that recovers heat generated by acetone hydrogenation, an exothermic reaction, and supplies it to an outside load. In the experiment, a pellet-like activated carbon-supported ruthenium catalyst was used for the acetone hydrogenation with hydrogen and acetone supplied to the catalyst layer at a space velocity of 400-1,200 or so. In the external pipe of the double-pipe type reactor, a heating medium oil was circulated in parallel with the flow of the reactant, with the heat of reaction recovered that was generated from the acetone hydrogenation. In this experiment, an 1wt%Ru/C catalyst and a 5wt%Ru/C catalyst were used so as to examine the effects of variation in the space velocity. As a result, from the viewpoint of recovering the heat of reaction, it was found desirable to increase the reaction speed by raising catalytic density and also to supply the reactant downstream inside the reaction pipe by increasing the space velocity. 1 ref., 6 figs., 1 tab.

  15. A new type of Neutrino Detector for Sterile Neutrino Search at Nuclear Reactors and Nuclear Nonproliferation Applications

    CERN Document Server

    Lane, C; Blackmon, J; Rasco, C; Mumm, H P; Markoff, D; Jocher, G R; Dorrill, R; Duvall, M; Learned, J G; Li, V; Maricic, J; Matsuno, S; Milincic, R; Negrashov, S; Sakai, M; Rosen, M; Varner, G; Huber, P; Pitt, M L; Rountree, S D; Vogelaar, R B; Wright, T; Yokley, Z

    2015-01-01

    We describe a new detector, called NuLat, to study electron anti-neutrinos a few meters from a nuclear reactor, and search for anomalous neutrino oscillations. Such oscillations could be caused by sterile neutrinos, and might explain the "Reactor Antineutrino Anomaly". NuLat, is made possible by a natural synergy between the miniTimeCube and mini-LENS programs described in this paper. It features a "Raghavan Optical Lattice" (ROL) consisting of 3375 boron or $^6$Li loaded plastic scintillator cubical cells 6.3\\,cm (2.500") on a side. Cell boundaries have a 0.127\\,mm (0.005") air gap, resulting in total internal reflection guiding most of the light down the 3 cardinal directions. The ROL detector technology for NuLat gives excellent spatial and energy resolution and allows for in-depth event topology studies. These features allow us to discern inverse beta decay (IBD) signals and the putative oscillation pattern, even in the presence of other backgrounds. We discuss here test venues, efficiency, sensitivity an...

  16. Transmutation of actinides in power reactors.

    Science.gov (United States)

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

    2005-01-01

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

  17. The importance of carry out studies about the use of passive autocatalytic recombiners for hydrogen control in reactors type ESBWR; La importancia de realizar estudios sobre el uso de recombinadores autocataliticos pasivos para control de hidrogeno en reactores tipo ESBWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez J, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)], e-mail: jersonsanchez@gmail.com

    2009-10-15

    A way to satisfy and to guarantee the energy necessities in the future is increasing in a gradual way the creation of nuclear power plants, introducing advanced designs in its systems that contribute in way substantial in the security of the same nuclear plants. The tendency of new designs of these nuclear plants is the incorporation of systems more reliable and sure, and that the operation does not depend on external factors as the electric power, motors diesel or the action of the operator of nuclear plant, what is known as security passive systems. In this sense, the passive autocatalytic recombiners are a contribution toward the use of this type of systems. At the present time it is had studies of the incorporation of passive autocatalytic recombiners in nuclear plants in operation and that they have contributed to minimize the danger associated to hydrogen. The present work contains a first approach to the study of hydrogen recombiners incorporation in advanced nuclear plants, for this case in a nuclear power plant of ESBWR type. To achieve our objective it seeks to use specialized codes as RELAP/SCDAP to obtain simulations of passive autocatalytic recombiners behaviour and we can to estimate their operation inside the reactor contention, contemplating the possibility to use other codes like SCILAB and/or MATLAB for the simulation of a passive autocatalytic recombiner. (Author)

  18. The use of waveguide acoustic probes for void fraction measurement in the evaporator of BN-350-Type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Melnikov, V.I.; Nigmatulin, B.I.

    1995-09-01

    The present paper deals with some results of the experimental studies which have been carried out to investigate the steam generation dynamics in the Field tubes of sodium-water evaporators used in the BN-350 reactors. The void fraction measurements have been taken with the aid of waveguide acoustic transducers manufactured in accordance with a specially designed technology (waveguide acoustic transducers-WAT technology). Presented in this paper also the transducer design and calibration methods, as well as the diagram showing transducers arrengment in the evaporator. The transducers under test featured a waveguide of about 4 m in length and a 200-mm long sensitive element (probe). Besides, this paper specifies the void fraction data obtained through measurements in diverse points of the evaporator. The studies revealed that the period of observed fluctuations in the void fraction amounted to few seconds and was largely dependent on the level of water in the evaporator.

  19. de cultivos agrícolas, en la agricultura urbana

    Directory of Open Access Journals (Sweden)

    Marcial A. Méndez F.

    2006-01-01

    Full Text Available El uso de las aguas residuales se presenta como una de las fuentes alternativas para el riego en la agricultura urbana. Esto entraña un conjunto de interrogantes en cuanto a su manejo y las posibles afectaciones que ellas puedan ocasionar a los frutos cosechados, al suelo y al medio ambiente. Las informaciones que se presentan, son solo los primeros resultados de los estudios introductorios hasta ahora realizados. Las aguas residuales del afluente Luyanó en el municipio Arroyo Naranjo, de acuerdo con los análisis realizados, se clasifican como aguas de baja contaminación (según los valores DBO5 y DQO obtenidos. Estas aguas, además, toxicológicamente no constituyen un riesgo de contaminación ni para el suelo ni para el medio ambiente, aunque no deben utilizarse sin tratamiento previo en cultivos de consumo directo, dada la carga microbiana que presentan. Las parcelas experimentales que se establecieron con cultivos de zanahoria, rábano y flor del Marigold, mostraron en las primeras cosechas que es posible alcanzar en ellas altos rendimientos agrícolas, dado el valor fertilizante de estas aguas residuales. Los resultados de la cosecha del rábano, según análisis bacteriológicos realizados, no mostraron ningún tipo de contaminación. El suelo y el entorno agrícola, tampoco presentaron afectaciones.

  20. Assessment of the thorium fuel cycle in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled.

  1. A model for the analysis of loss of decay heat removal during loss of coolant accident in MTR pool type research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bousbia-salah, Anis [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2, 56126 Pisa (Italy)]. E-mail: b.salah@ing.unipi.it; Meftah, Brahim [Division Reacteur - Centre de Recherche Nucleaire Draria (CRND), BP 43 Sebala DRARIA - Algiers (Algeria); Hamidouche, Tewfik [Laboratoire des Analyses de Surete, Centre de Recherche Nucleaire d' Alger (CRNA), 02 Boulevard Frantz Fanon, B.P. 399, 16000 Algiers (Algeria)]. E-mail: thamidouche@comena-dz.org; Si-Ahmed, El Khider [Laboratoire des Ecoulements Polyhpasiques, Universite des Sciences et de la Technologie d' Alger, Algiers (Algeria)

    2006-03-15

    During a loss of coolant accident leading to total emptying of the reactor pool, the decay heat could be removed through air natural convection. However, under partial pool emptying the core is partially submerged and the coolant circulation inside the fuel element could no more be possible. Under such conditions, a core overheat takes place, and the thermal energy is essentially diffused from the core to its periphery by combined thermal radiation and conduction. In order to predict fuel element temperature evolution under such conditions a mathematical model is performed. The model is based on a 3D geometry and takes into account a variety of core configurations including fuel elements (standard and control), reflector elements and grid plates. The homogeneous flow model is used and the fluid conservation equations are solved using a semi-implicit finite difference method. Preliminary tests of the developed model were made by considering a series of hypothetical accidents. In the current framework a loss of decay heat removal accidents in the IAEA benchmark open pool MTR-type research reactor is considered. It is shown that in the case of a low core immersion height no water boiling is observed and the fuel surface temperature rise remains below the melting point of the aluminium cladding.

  2. AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

    Energy Technology Data Exchange (ETDEWEB)

    Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rice, Francine Joyce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip Lon [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of the rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite

  3. Proposal of novel method of continuous monitoring of possible fuel failure of a pool-type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sasaki, K. [Rikkyo University, Nishi-Ikebukuro, Toshima-ku, Tokyo (Japan). College of Science; Hayashi, S.A.; Matsura, T. [Rikkyo University, Nagasaka, Yokosuka (Japan). Institute for Atomic Energy

    1997-10-01

    During the course of studies on fuel failure detection, we have found that the bubbling of a gas such as nitrogen into a reactor coolant water effectively purges the dissolved fission rare gases ({sup 89}Kr, T{sub 1/2}=3.15 min, and {sup 138}Xe, T{sub 1/2}=14.08 min) and that the respective daughter nuclides ({sup 89}Rb, T{sub 1/2}=15.15 min and {sup 138}Cs, T{sub 1/2}=33.41 min) are detected in the washing water of the collected gas mixture. The detected activity depends on the time of standing between sampling and washing of the gas, and the dependence agreed well with the theoretical prediction from the consecutive radioactive decay for both pairs ({sup 89}Kr-{sup 89}Rb, and {sup 138}Xe-{sup 138}Cs). Based on these findings, we have recently constructed a semi-continuous fuel monitoring system, which consists of an automatic and intermittent gas sampler (1 litre bottles) and a bottle conveying unit. After standing for a definite time, bottled gas is shaken with a small amount of water, and the activity of the water is measured. This system operates satisfactorily, but the whole system involves several sophisticated steps so that is rather costly. Quite recently we have got an idea of a simpler, more economical, fully automated continuous system. The system consists in principle only of a large cylinder with packing materials just as in a fractional distiller. On the top of the cylinder there are an inlet of washing water and an outlet of the gas, and at the bottom there are an inlet of the collected gas from the coolant and an outlet of the washing water. The whole system can be operated fully automatically and continuously, with continuous feeding of bubbling gas into the reactor coolant. This has not yet been experimentally tested at present, and in this presentation, information about the setup parameters such as the flow rate of the bubbling gas, the volume of the cylinder and vacant space, the flow rate of the washing water, etc. are reported

  4. Comparison of fission product release predictions using PARFUME with results from the AGR-1 irradiation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2014-09-01

    This report documents comparisons between post-irradiation examination measurements and model predictions of silver (Ag), cesium (Cs), and strontium (Sr) release from selected tristructural isotropic (TRISO) fuel particles and compacts during the first irradiation test of the Advanced Gas Reactor program that occurred from December 2006 to November 2009 in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The modeling was performed using the particle fuel model computer code PARFUME (PARticle FUel ModEl) developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact) but it can be assessed at the particle level by adjusting the diffusivity in the fuel matrix to very high values. Furthermore, the diffusivity of each layer can be individually set to a high value (typically 10-6 m2/s) to simulate a failed layer with no capability of fission product retention. In this study, the comparison to PIE focused on fission product release and because of the lack of failure in the irradiation, the probability of particle failure was not calculated. During the AGR-1 irradiation campaign, the fuel kernel produced and released fission products, which migrated through the successive

  5. Fuel composition optimization in a 78-element fuel bundle for use in a pressure tube type supercritical water-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, D.W.; Novog, D.R. [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    A 78-element fuel bundle containing a plutonium-thorium fuel mixture has been proposed for a Generation IV pressure tube type supercritical water-cooled reactor. In this work, using a lattice cell model created with the code DRAGON,the lattice pitch, fuel composition (fraction of PuO{sub 2} in ThO{sub 2}) and radial enrichment profile of the 78-element bundle is optimized using a merit function and a metaheuristic search algorithm.The merit function is designed such that the optimal fuel maximizes fuel utilization while minimizing peak element ratings and coolant void reactivity. A radial enrichment profile of 10 wt%, 11 wt% and 20 wt% PuO{sub 2} (inner to outer ring) with a lattice pitch of 25.0 cm was found to provide the optimal merit score based on the aforementioned criteria. (author)

  6. Reduced-scale water test of natural circulation for decay heat removal in loop-type sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, T., E-mail: murakami@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Eguchi, Y., E-mail: eguchi@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Oyama, K., E-mail: kazuhiro_oyama@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan); Watanabe, O., E-mail: osamu4_watanabe@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan)

    2015-07-15

    Highlights: • The natural circulation characteristics of a loop-type SFR are examined by a water test. • The performance of decay heat removal system is evaluated using a similarity law. • The effects of flow deviation in the parallel piping of a primary loop are clarified. • The reproducibility of the natural circulation test is confirmed. - Abstract: Water tests of a loop-type sodium-cooled fast reactor have been conducted to physically evaluate the natural circulation characteristics. The water test apparatus was manufactured as a 1/10-scale mock-up of the Japan Sodium-Cooled Fast Reactor, which adopts a decay heat removal system (DHRS) utilizing natural circulation. Tests simulating a variety of events and operation conditions clarified the thermal hydraulic characteristics and core-cooling performance of the natural circulation in the primary loop. Operation conditions such as the duration of the pump flow coast-down and the activation time of the DHRS affect the natural circulation characteristics. A long pump flow coast-down cools the upper plenum of the reactor vessel (RV). This causes the loss of the buoyant force in the RV. The test result indicates that a long pump flow coast-down tends to result in a rapid increase in the core temperature because of the loss of the buoyant force. The delayed activation of the DHRS causes a decrease in the natural circulation flow rate and a temperature rise in the RV. Flow rate deviation and a reverse flow appear in the parallel cold-leg piping in some events, which cause thermal stratification in the cold-leg piping. The DHRS prevents the core temperature from fatally rise even for the most severe design-basis event, in which sodium leakage in a secondary loop of the DHRS and the opening failure of a single damper of the air cooler occur simultaneously. In the water test for the case of siphon break in the primary loop, which is one of the design extension conditions, a circulation flow consisting of ascendant

  7. Hydrogen Production in Fusion Reactors

    OpenAIRE

    Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Momota, H; Motojima, O.; Okamoto, M.; Ohnishi, M.; Onozuka, M; Uenosono, C.

    1993-01-01

    As one of methods of innovative energy production in fusion reactors without having a conventional turbine-type generator, an efficient use of radiation produced in a fusion reactor with utilizing semiconductor and supplying clean fuel in a form of hydrogen gas are studied. Taking the candidates of reactors such as a toroidal system and an open system for application of the new concepts, the expected efficiency and a concept of plant system are investigated.

  8. Acaros de importancia agrícola en Colombia.

    Directory of Open Access Journals (Sweden)

    Mesa Nora Cristina

    1999-06-01

    Full Text Available La acarología en Colombia ha contribuido de forma notoria al conocimiento de la diversidad de especies. Además los investigadores en este campo han desarrollado estudios sobre evaluación y manejo de los ácaros fitófagos. Los avances han sido importantes en los trabajos sobre aquellas especies depredadoras. Se puede afirmar que el país cuenta con información básica y aplicada sobre los ácaros, pero se necesita profundizar en algunos temas. Este documento adelanta el análisis de la retrospectiva de la acarología agrícola en el país y se detiene en consideraciones sobre las familias y especies de mayor importancia. Esta recopilación sobre los ácaros en Colombia, señala la trascendencia que tienen estos artrópodos.

  9. Nacionalismos e reforma agrária nos anos 50

    Directory of Open Access Journals (Sweden)

    Moreira Vânia Maria Losada

    1998-01-01

    Full Text Available O artigo visa recuperar os elementos fundamentais do ideário nacionalista durante a década de 1950, destacando a centralidade do conceito de desenvolvimento auto-sustentado. A diversidade política e ideológica do movimento nacionalista é problematizada através da análise de duas correntes preponderantes do período em foco: o nacional-desenvolvimentismo, de caráter liberal, e o nacionalismo econômico, isto é, a vertente defendida pelas esquerdas. Finalmente, é focalizado a importância da reforma agrária dentro do projeto social defendido sobretudo pelos nacionalistas econômicos, enquanto fator indispensável a uma industrialização com cidadania.

  10. Planificación de Maquinaria Agrícola

    Directory of Open Access Journals (Sweden)

    Alvarez Cardona Alberto

    1979-06-01

    Full Text Available Se entiende por planificación agrícola el conjunto de técnicas que permiten ver con anticipación suficiente la cantidad de maquinaria, tanto fuentes de potencia como implementos que serán requeridos para realizar un trabajo específico en un tiempo determinado; ó la cantidad de trabajo específico que podrá realizar un equipo con características ya definidas a través de un tiempo determinado; se analizan también bajo este concepto de planificación aquellos sistemas que permiten evaluar, controlar y ajustar la planificación propuesta a las circunstancias reales una vez iniciada ésta.

  11. Multifunctional reactors

    NARCIS (Netherlands)

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much

  12. Anaerobic digestion of solid waste in RAS: Effect of reactor type on the biochemical acidogenic potential (BAP) and assessment of the biochemical methane potential (BMP) by a batch assay

    DEFF Research Database (Denmark)

    Suhr, Karin Isabel; Letelier-Gordo, Carlos Octavio; Lund, Ivar

    2015-01-01

    additional 14 and 20 days) in continuously stirred tank reactors. Generally, the VFA yield increased with time and no effect of the reactor type used was found within the time frame of the experiment. At 10 days HT or 10 days HRT the VFA yield reached 222.3 ± 30.5 and 203.4 ± 11.2 mg VFA g-1 TVS0 (total...... volatile solids at day 0) in batch and fed-batch reactor, respectively. For the fedbatch reactor, increasing HRT from 5 to 10 days gained no significant additional VFA yield. Prolonging the batch reactor experiment to 20 days increased VFA production further (273.9 ± 1.6 mg VFA g-1 TVS0, n=2). After 10...... for the design of an acidogenic continuously stirred reactor tank in a RAS single-sludge denitrification set-up. The biochemical methane potential of the sludge was estimated to 318 ± 29 g CH4 g-1 TVS0 by a batch assay and represented a higher utility of the solid waste when comparing the methane yield...

  13. Increasing AIP Macrocycle Size Reveals Key Features of agr Activation in Staphylococcus aureus.

    Science.gov (United States)

    Johnson, Jeffrey G; Wang, Boyuan; Debelouchina, Galia T; Novick, Richard P; Muir, Tom W

    2015-05-04

    The agr locus in the commensal human pathogen, Staphylococcus aureus, is a two-promoter regulon with allelic variability that produces a quorum-sensing circuit involved in regulating virulence within the bacterium. Secretion of unique autoinducing peptides (AIPs) and detection of their concentrations by AgrC, a transmembrane receptor histidine kinase, coordinates local bacterial population density with global changes in gene expression. The finding that staphylococcal virulence can be inhibited through antagonism of this quorum-sensing pathway has fueled tremendous interest in understanding the structure-activity relationships underlying the AIP-AgrC interaction. The defining structural feature of the AIP is a 16-membered, thiolactone-containing macrocycle. Surprisingly, the importance of ring size on agr activation or inhibition has not been explored. In this study, we address this deficiency through the synthesis and functional analysis of AIP analogues featuring enlarged and reduced macrocycles. Notably, this study is the first to interrogate AIP function by using both established cell-based reporter gene assays and newly developed in vitro AgrC-I binding and autophosphorylation activity assays. Based on our data, we present a model for robust agr activation involving a cooperative, three-points-of-contact interaction between the AIP macrocycle and AgrC.

  14. Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2014-09-01

    Safety tests were conducted on fourteen fuel compacts from AGR-1, the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Development and Qualification program, at temperatures ranging from 1600 to 1800°C to determine fission product release at temperatures that bound reactor accident conditions. The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, strontium, and krypton from fuel compacts containing tristructural isotropic (TRISO) coated particles during the safety tests, and the predicted values were compared with experimental results. Preliminary comparisons between PARFUME predictions and post-irradiation examination (PIE) results of the safety tests show different trends in the prediction of the fractional release depending on the species, and it leads to different conclusions regarding the diffusivities used in the modeling of fission product transport in TRISO-coated particles: • For silver, the diffusivity in silicon carbide (SiC) might be over-estimated by a factor of at least 102 to 103 at 1600°C and 1700°C, and at least 10 to 102 at 1800°C. The diffusivity of silver in uranium oxy-carbide (UCO) might also be over-estimated, but the available data are insufficient to allow definitive conclusions to be drawn. • For cesium, the diffusivity in UCO might be over-estimated by a factor of at least 102 to 103 at 1600°C, 105 at 1700°C, and 103 at 1800°C. The diffusivity of cesium in SiC might also over-estimated, by a factor of 10 at 1600°C and 103 at 1700°C, based upon the comparisons between calculated and measured release fractions from intact particles. There is no available estimate at 1800°C since all the compacts heated up at 1800°C contain particles with failed SiC layers whose release dominates the release from intact particles. • For strontium, the diffusivity in SiC might be over-estimated by a factor of 10 to 102 at 1600 and 1700°C, and 102 to 103 at 1800°C. These

  15. HTR-2014 Paper Comparison of fission product release predictions using PARFUME with results from the AGR-1 irradiation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2001-10-01

    The PARFUME (PARticle FUel ModEl) code was used to predict fission product release from tristructural isotropic (TRISO) coated fuel particles and compacts during the first irradiation experiment (AGR-1) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-1 experiment used the fuel compact volume average temperature for each of the 620 days of irradiation to calculate the release of fission products silver, cesium, and strontium from a representative particle for a select number of AGR-1 compacts. Post-irradiation examination (PIE) measurements provided data on release of fission products from fuel compacts and fuel particles, and retention of fission products in the compacts outside of the SiC layer. PARFUME-predicted fractional release of these fission products was determined and compared to the PIE measurements. Results show an overall over-prediction of the fractional release of cesium by PARFUME. For particles with failed silicon carbide (SiC) layers, the over-prediction is by a factor of about two, corresponding to an over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of about 100. For intact particles, whose release is much lower, the over-prediction is by an average of about an order of magnitude, which could additionally be attributed to an over-estimated diffusivity in SiC by about 30%. The release of strontium from intact particles is also over-estimated by PARFUME, which also points towards an over-estimated diffusivity of strontium in either SiC or UCO, or possibly both. The measured strontium fractional release from intact particles varied considerably from compact to compact, making it difficult to assess the effective over-estimation of the diffusivities. Furthermore, the release of strontium from particles with failed SiC is difficult to observe experimentally due to the release from intact particles, preventing any conclusions to be made on the accuracy or validity of the

  16. Los suelos agrícolas frente al sistema legislativo Mexicano

    OpenAIRE

    Maycotte Luna, Eliot

    2012-01-01

    El presente estudio sobre legislación de los suelos agrícolas en México, se basa en la técnica que define el Derecho Parlamentario jugando un papel importante para explicar el desarrollo legislativo [retrospectivo y prospectivo] sobre el uso y protección de los recursos naturales y en específico el suelo con uso agrícola frente al crecimiento urbano. Los efectos muestran que la frontera urbana sigue destruyendo cualquier tipo de suelo y más los agrícolas. Ocasionando en ...

  17. Organismo de Certificación de Implementos y Maquinaria Agrícola (OCIMA)

    OpenAIRE

    Alma Velia Ayala Garay; Marco A. Audelo Benítez; Rocío Cervantes Osornio; Dora Ma. Sangerman-Jarquín; Miguel Sánchez Hernández; Martha Garay Hernández

    2012-01-01

    La certificación resulta trascendental puesto que da seguridad al usuario final en el funcionamiento y calidad de los tractores, la creación del Organismo de Certificación de Implementos y Maquinaria Agrícola (OCIMA) dependiente del INIFAP como Organismo de Certificación de Producto en Tractores surgió como respuesta a esta necesidad; convirtiéndose en el único a nivel nacional con el propósito de autenticar a los productores agrícolas que la compra de maquinaria agrícola cumple con las espec...

  18. Helias reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Beidler, C.D. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Grieger, G. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Harmeyer, E. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Kisslinger, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Karulin, N. [Nuclear Fusion Institute, Moscow (Russian Federation); Maurer, W. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany); Nuehrenberg, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Rau, F. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Sapper, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Wobig, H. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    1995-10-01

    The present status of Helias reactor studies is characterised by the identification and investigation of specific issues which result from the particular properties of this type of stellarator. On the technical side these are issues related to the coil system, while physics studies have concentrated on confinement, alpha-particle behaviour and ignition conditions. The usual assumptions have been made in those fields which are common to all toroidal fusion reactors: blanket and shield, refuelling and exhaust, safety and economic aspects. For blanket and shield sufficient space has been provided, a detailed concept will be developed in future. To date more emphasis has been placed on scoping and parameter studies as opposed to fixing a specific set of parameters and providing a detailed point study. One result of the Helias reactor studies is that physical dimensions are on the same order as those of tokamak reactors. However, it should be noticed that this comparison is difficult in view of the large spectrum of tokamak reactors ranging from a small reactor like Aries, to a large device such as SEAFP. The notion that the large aspect ratio of 10 or more in Helias configurations also leads to large reactors is misleading, since the large major radius of 22 m is compensated by the average plasma radius of 1.8 m and the average coil radius of 5 m. The plasma volume of 1400 m{sup 3} is about the same as the ITER reactor and the magnetic energy of the coil system is about the same or even slightly smaller than envisaged in ITER. (orig.)

  19. The improvement of control rod in experimental fast reactor JOYO. The development of a sodium bonded type control rod

    Energy Technology Data Exchange (ETDEWEB)

    Soga, T.; Miyakawa, S.; Mitsugi, T. [Japan Nuclear Cycle Development Inst., Oarai Engineering Center, Irradiation Center, Irradiation and Administration Section, Oarai, Ibaraki (Japan)

    1999-06-01

    Currently, the lifetime of control rods in JOYO is limited by Absorber-Cladding Mechanical Interaction (ACMI) due to swelling of B{sub 4}C(boron carbide) pellets accelerated by relocation of pellet fragments. A sodium bonded type control rod was developed which improves the thermal conductivity by means of charging sodium into the gap between B{sub 4}C and cladding and by utilizing a shroud which wraps the pellet fragments in a thin tube. This new design will be able to enlarge the gap between B{sub 4}C and cladding, without heating B{sub 4}C or fragment relocation, thus extending the life of the control rod. The sodium bonded type will be fabricated as the ninth reload control rods in JOYO. (1) The specification of a sodium bonded type control rod was determined with the wide gap between B{sub 4}C and cladding. In the design simulation, main component temperature were below the maximum limit. And the local heating by helium bubble generated from B{sub 4}C in the sodium gap, was not a serious problem in the analysis which was considered. (2) A structural design for the sodium entrance into the pin was determined. A formula was developed which the limit for sodium charging given physical dimension of the structure and sodium property. Result from sodium out-pile experiments validated the theoretical formula. (3) The analysis of ACMI indicated a lifetime extension of the sodium bonded type by 4.6% in comparison with lifetime of the helium bonded type of 1.6%. This is due to the boron10 burn-up rate being three times higher in the sodium bonded type than in the helium bonded type. To achieve a target burn-up 10% in the future, it will be necessary to modify design based on irradiation data which will be obtained by practical use of the sodium bonded control rods in JOYO. (4) The effects due to Absorber-Cladding Chemical Interaction (ACCI) were reduced by controlling the cladding temperature and chromium coating to the cladding's inner surface. It was confirmed

  20. Association between the agr locus and the presence of virulence genes and pathogenesis in Staphylococcus aureus using a Caenorhabditis elegans model

    Directory of Open Access Journals (Sweden)

    Terissa A. Thompson

    2017-01-01

    Conclusions: There was a strong association between the carriage of virulence genes and the presence of the agr operon in clinical strains of S. aureus. Further, agr-positive strains were more pathogenic than agr-negative strains, suggesting a correlation between the presence of agr, carriage of virulence determinants, and pathogenicity.

  1. Reactor vessel

    OpenAIRE

    Makkee, M.; Kapteijn, F.; Moulijn, J.A

    1999-01-01

    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and comprising connections for supplying (9, 10, 11) and discharging (13, 14, 15) via the channels (3) gases and/or liquids entering into a reaction with each other and substances formed upon this reactio...

  2. Mathematical model analysis on the enhancement of aeration efficiency using ladder-type flat membrane module forms in the Submerged Membrane Bio-reactor(SMBR)

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    The cross-flow shearing action produced from the inferior aeration in the Submerged Membrane Bio-reactor(SMBR) is an effective way to further improve anti-fouling effects of membrane modules.Based on the widely-applied vertical structure of flat membrane modules,improvements are made that ladder-type flat membrane structure is designed with a certain inclined angle θ so that the cross-flow velocity of bubble near the membrane surface can be held,and the intensity and times of elastic colli-sion between bubbles and membrane surface can be increased.This can improve scouring action of membrane surface on aeration and reduce energy consumption of strong aeration in SMBR.By de-ducing and improving the mathematics model of collision between bubble and vertical flat put forward by Vries,the relatively suitable incline angle θ under certain aeration place and in certain size rang of bubble can be obtained with the computer iterative calculation technology.Finally,for many groups of ladder-type flat membrane in parallel placement in the practical application of SMBR,some sugges-tions are offered:the interval distance of membrane modules is 8―15 mm,and aeration should be op-erated at 5―7 mm among membrane modules,and the optimal design angle of trapeziform membrane is 1.7°―2.5°.

  3. Mathematical model analysis on the enhancement of aeration efficiency using ladder-type flat membrane module forms in the Submerged Membrane Bio-reactor (SMBR)

    Institute of Scientific and Technical Information of China (English)

    LI Bo; YE MaoSheng; YANG FengLin; MA Hui

    2009-01-01

    The cross-flow shearing action produced from the inferior aeration in the Submerged Membrane Bio-reactor (SMBR) Is an effective way to further improve anti-fouling effects of membrane modules.Based on the widely-applied vertical structure of flat membrane modules, improvements are made that ladder-type flat membrane structure is designed with a certain inclined angle θ so that the cross-flow velocity of bubble near the membrane surface can be held, and the intensity and times of elastic colli-sion between bubbles and membrane surface can be increased. This can improve scouring action ofmembrane surface on aeration and reduce energy consumption of strong aeration in SMBR. By de-ducing and improving the mathematics model of collision between bubble and vertical flat put forward by Vries, the relatively suitable Incline angle θ under certain aeration place and in certain size rang ofbubble can be obtained with the computer iterative calculation technology. Finally, for many groups of ladder-type flat membrane in parallel placement in the practical application of SMBR, some sugges-tions are offered: the interval distance of membrane modules is 8--15 mm, and aeration should be op-erated at 5--7 mm among membrane modules, and the optimal design angle of trapeziform membrane is 1.7°--2.5°.

  4. Heat resistant reduced activation 12% Cr steel of 16Cr12W2VTaB type-advanced structural material for fusion and fast breeder power reactors

    Science.gov (United States)

    Ioltukhovskiy, A. G.; Leonteva-Smirnova, M. V.; Solonin, M. I.; Chernov, V. M.; Golovanov, V. N.; Shamardin, V. K.; Bulanova, T. M.; Povstyanko, A. V.; Fedoseev, A. E.

    2002-12-01

    Heat resistant 12% Cr steels of the 16Cr12W2VTaB type (12Cr-2W-V-Ta-B-0.16C) provide a reduced activation material that can be used as a structural material for fusion and fast breeder reactors. The composition under study meets scientific and engineering requirements and has an optimal base element composition to provide a δ-ferrite content of no more than 20%. It also has a minimum quantity of low melting impurity elements and non-metallic inclusions. Short-term tensile properties for the steel tested to 700 °C are provided after the standard heat treatment (normalization, temper). Rupture strength and creep properties for the steel depending on the initial heat treatment conditions are also given. The microstructural stability of the 16Cr12W2VTaB type steel at temperatures up to 650 °C is predicted to be good, and the properties of the steel after irradiation in BOR-60 are demonstrated.

  5. Heat resistant reduced activation 12% Cr steel of 16Cr12W2VTaB type-advanced structural material for fusion and fast breeder power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ioltukhovskiy, A.G. E-mail: iral@bochvar.ru; Leonteva-Smirnova, M.V.; Solonin, M.I.; Chernov, V.M.; Golovanov, V.N.; Shamardin, V.K.; Bulanova, T.M.; Povstyanko, A.V.; Fedoseev, A.E

    2002-12-01

    Heat resistant 12% Cr steels of the 16Cr12W2VTaB type (12Cr-2W-V-Ta-B-0.16C) provide a reduced activation material that can be used as a structural material for fusion and fast breeder reactors. The composition under study meets scientific and engineering requirements and has an optimal base element composition to provide a {delta}-ferrite content of no more than 20%. It also has a minimum quantity of low melting impurity elements and non-metallic inclusions. Short-term tensile properties for the steel tested to 700 deg. C are provided after the standard heat treatment (normalization, temper). Rupture strength and creep properties for the steel depending on the initial heat treatment conditions are also given. The microstructural stability of the 16Cr12W2VTaB type steel at temperatures up to 650 deg. C is predicted to be good, and the properties of the steel after irradiation in BOR-60 are demonstrated.

  6. Target-induced nano-enzyme reactor mediated hole-trapping for high-throughput immunoassay based on a split-type photoelectrochemical detection strategy.

    Science.gov (United States)

    Zhuang, Junyang; Tang, Dianyong; Lai, Wenqiang; Xu, Mingdi; Tang, Dianping

    2015-09-15

    Photoelectrochemical (PEC) detection is an emerging and promising analytical tool. However, its actual application still faces some challenges like potential damage of biomolecules (caused by itself system) and intrinsic low-throughput detection. To solve the problems, herein we design a novel split-type photoelectrochemical immunoassay (STPIA) for ultrasensitive detection of prostate specific antigen (PSA). Initially, the immunoreaction was performed on a microplate using a secondary antibody/primer-circular DNA-labeled gold nanoparticle as the detection tag. Then, numerously repeated oligonucleotide sequences with many biotin moieties were in situ synthesized on the nanogold tag via RCA reaction. The formed biotin concatamers acted as a powerful scaffold to bind with avidin-alkaline phosphatase (ALP) conjugates and construct a nanoenzyme reactor. By this means, enzymatic hydrolysate (ascorbic acid) was generated to capture the photogenerated holes in the CdS quantum dot-sensitized TiO2 nanotube arrays, resulting in amplification of the photocurrent signal. To elaborate, the microplate-based immunoassay and the high-throughput detection system, a semiautomatic detection cell (installed with a three-electrode system), was employed. Under optimal conditions, the photocurrent increased with the increasing PSA concentration in a dynamic working range from 0.001 to 3 ng mL(-1), with a low detection limit (LOD) of 0.32 pg mL(-1). Meanwhile, the developed split-type photoelectrochemical immunoassay exhibited high specificity and acceptable accuracy for analysis of human serum specimens in comparison with referenced electrochemiluminescence immunoassay method. Importantly, the system was not only suitable for the sandwich-type immunoassay mode, but also utilized for the detection of small molecules (e.g., aflatoxin B1) with a competitive-type assay format.

  7. Reactor Neutrinos

    Directory of Open Access Journals (Sweden)

    Soo-Bong Kim

    2013-01-01

    Full Text Available We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very recently the most precise determination of the neutrino mixing angle θ13. This paper provides an overview of the upcoming experiments and of the projects under development, including the determination of the neutrino mass hierarchy and the possible use of neutrinos for society, for nonproliferation of nuclear materials, and geophysics.

  8. Chemical Reactors.

    Science.gov (United States)

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  9. Heavy Water Reactor; Reacteurs a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

    Yu, St.; HOpwood, J.; Meneley, D. [Energie Atomique du Canada (Canada)

    2000-04-01

    This document deals with the Heavy Water Reactor (HWR) technology and especially the Candu (Canada Deuterium Uranium) reactor. This reactors type offers many advantages that promote them for the future. General concepts, a description of the Candu nuclear power plants, the safety systems, the fuel cycle and economical and environmental aspects are included. (A.L.B.)

  10. Reactor Neutrinos

    OpenAIRE

    Lasserre, T.; Sobel, H.W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrin...

  11. Crystal Structure of AGR_C_4470p from Agrobacterium tumefaciens

    Energy Technology Data Exchange (ETDEWEB)

    Vorobiev,S.; Neely, H.; Seetharaman, J.; Ma, L.; Xiao, R.; Acton, T.; Montelione, G.; Tong, L.

    2007-01-01

    We report here the crystal structure at 2.0 {angstrom} resolution of the AGR{_}C{_}4470p protein from the Gram-negative bacterium Agrobacterium tumefaciens. The protein is a tightly associated dimer, each subunit of which bears strong structural homology with the two domains of the heme utilization protein ChuS from Escherichia coli and HemS from Yersinia enterocolitica. Remarkably, the organization of the AGR{_}C{_}4470p dimer is the same as that of the two domains in ChuS and HemS, providing structural evidence that these two proteins evolved by gene duplication. However, the binding site for heme, while conserved in HemS and ChuS, is not conserved in AGR{_}C{_}4470p, suggesting that it probably has a different function. This is supported by the presence of two homologs of AGR{_}C{_}4470p in E. coli, in addition to the ChuS protein.

  12. Cross-Talk between Staphylococcus aureus and Other Staphylococcal Species via the agr Quorum Sensing System

    DEFF Research Database (Denmark)

    Canovas de la Nuez, Jaime; Baldry, Mara; Bojer, Martin S;

    2016-01-01

    between staphylococci and S. aureus, and show that this interaction may eventually lead to the identification of new anti-virulence candidates to target S. aureus infections. Here we show that culture supernatants of 37 out of 52 staphylococcal isolates representing 17 different species inhibit S. aureus......-inducing peptides (AIPs) sensed by AgrC, a two component histidine kinase. agr loci are found also in other staphylococcal species and for Staphylococcus epidermidis, the encoded AIP represses expression of agr regulated virulence genes in S. aureus. In this study we aimed to better understand the interaction....... To assess impact on S. aureus virulence, we co-inoculated S. aureus and S. schleiferi in vivo in the Galleria mellonella wax moth larva, and found that expression of key S. aureus virulence factors was abrogated. Our data show that the S. aureus agr locus is highly responsive to other staphylococcal species...

  13. Diseño, construcción y evaluación de un equipo tipo baúl para desinfección de sustratos agrícolas con calor Desing, construction and evaluation of trunk type equipment for agriculture sustrate disinfection with heat

    Directory of Open Access Journals (Sweden)

    Noel Chávez-Aguilera

    Full Text Available El objetivo de este estudio fue diseñar, construir y evaluar un equipo tipo baúl para desinfección en estático de sustratos agrícolas mediante aplicación de calor, a través de aire caliente y con vapor de agua aireado. La investigación se realizó en 2005 en los talleres y laboratorios de la Universidad Autónoma Chapingo, Texcoco, Estado de México, México. El diseño experimenetal utilizado fue factorial 2*2*3, el primer factor fue la fuente y medio de transferencia de calor: 1 resistencia eléctrica y aire caliente; y 2 combustión de diesel y vapor de agua aireado; el segundo, contenido de humedad del sustrato: 75 y 61%; el tercer factor fue tiempo de desinfección: 15, 30 y 45 min. Las variables respuesta fueron: tiempo para alcanzar la temperatura mínima deseada, rango de temperatura durante la desinfección, eficiencia de desinfección de Fusarium oxisporum culmorum y costo específico de energía. La desinfección con aire caliente mostró los mayores valores en tiempo para alcanzar la temperatura mínima deseada, rango de temperatura, eficiencia de desinfección y costo especifico de energía, en comparación con vapor de agua aireado.The objective of this study was the design, construction and evaluation of equipment for in static disinfection of agricultural substrates by applying heat from hot air and aerated steam sources. The research was conducted in 2005 at the workshops and laboratories of the Universidad Autonoma Chapingo, Texcoco, State of Mexico, Mexico. A factorial 2*2*3 experimental design was utilized, the first factor was the source and means of heat transfer: 1 electrical resistance and hot air and 2 combustion of diesel and aerated steam, the second factor was substrate moisture content:75 and 61% and disinfection time: 15, 30 and 45 min. The studied variables were: time to reach the desired temperature, temperature range during disinfection, disinfection efficiency of Fusarium oxisporum culmorum, specific

  14. AGR-2 Irradiated Test Train Preliminary Inspection and Disassembly First Look

    Energy Technology Data Exchange (ETDEWEB)

    Ploger, Scott [Idaho National Lab. (INL), Idaho Falls, ID (United States); Demkowciz, Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-05-01

    The AGR 2 irradiation experiment began in June 2010 and was completed in October 2013. The test train was shipped to the Materials and Fuels Complex in July 2014 for post-irradiation examination (PIE). The first PIE activities included nondestructive examination of the test train, followed by disassembly of the test train and individual capsules and detailed inspection of the capsule contents, including the fuel compacts and their graphite fuel holders. Dimensional metrology was then performed on the compacts, graphite holders, and steel capsule shells. AGR 2 disassembly and metrology were performed with the same equipment used successfully on AGR 1 test train components. Gamma spectrometry of the intact test train gave a preliminary look at the condition of the interior components. No evidence of damage to compacts or graphite components was evident from the isotopic and gross gamma scans. Disassembly of the AGR 2 test train and its capsules was conducted rapidly and efficiently by employing techniques refined during the AGR 1 disassembly campaign. Only one major difficulty was encountered while separating the test train into capsules when thermocouples (of larger diameter than used in AGR 1) and gas lines jammed inside the through tubes of the upper capsules, which required new tooling for extraction. Disassembly of individual capsules was straightforward with only a few minor complications. On the whole, AGR 2 capsule structural components appeared less embrittled than their AGR 1 counterparts. Compacts from AGR 2 Capsules 2, 3, 5, and 6 were in very good condition upon removal. Only relatively minor damage or markings were visible using high resolution photographic inspection. Compact dimensional measurements indicated radial shrinkage between 0.8 to 1.7%, with the greatest shrinkage observed on Capsule 2 compacts that were irradiated at higher temperature. Length shrinkage ranged from 0.1 to 0.9%, with by far the lowest axial shrinkage on Capsule 3 compacts

  15. Cross-talk between Staphylococcus aureus and other staphylococcal species via the agr quorum sensing system

    Directory of Open Access Journals (Sweden)

    Jaime Canovas de la Nuez

    2016-11-01

    Full Text Available Staphylococci are associated with both humans and animals. While most are non-pathogenic colonizers, Staphylococcus aureus is an opportunistic pathogen capable of causing severe infections. S. aureus virulence is controlled by the agr quorum sensing system responding to secreted auto-inducing peptides (AIPs sensed by AgrC, a two component histidine kinase. agr loci are found also in other staphylococcal species and for S. epidermidis, the encoded AIP represses expression of agr regulated virulence genes in S. aureus. In this study we aimed to determine how other staphylococci affect S. aureus agr, and if such interaction may point to new anti-virulence candidates to target S. aureus infections. Here we show that culture supernatants of 37 out of 52 staphylococcal isolates representing 17 different species inhibit S. aureus agr. The dog pathogen, S. schleiferi, expressed the most potent inhibitory activity and was active against all four agr classes found in S. aureus. By employing a S. aureus strain encoding a constitutively active AIP receptor we show that the activity is mediated via agr. Subsequent cloning and heterologous expression of the S. schleiferi AIP in S. aureus demonstrated that this molecule was likely responsible for the inhibitory activity, and further proof was provided when pure synthetic S. schleiferi AIP was able to completely abolish agr induction of an S. aureus reporter strain. To assess impact on S. aureus virulence, we co-inoculated S. aureus and S. schleiferi in vivo in the Galleria Mellonella wax moth larva, and found that expression of key S. aureus virulence factors was abrogated. Our data show that the S. aureus agr locus is highly responsive to other staphylococcal species suggesting that agr is an inter-species communication system. Based on these results we speculate that interactions between S. aureus and other colonizing staphylococci will significantly influence the ability of S. aureus to cause infection, and

  16. High temperature and pressure effects on the elastic properties of B2 intermetallics AgRE

    OpenAIRE

    Liu Lili; Wu Xiaozhi; Li Weiguo; Wang Rui; Liu Qing

    2015-01-01

    The high temperature and pressure effects on the elastic properties of the AgRE (RE=Sc, Tm, Er, Dy, Tb) intermetallic compounds with B2 structure have been performed from first principle calculations. For the temperature range 0-1000 K, the second order elastic constants for all the AgRE intermetallic compounds follow a normal behavior: they decrease with increasing temperature. The pressure dependence of the second order elastic constants has been investigated on the ...

  17. Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

    Energy Technology Data Exchange (ETDEWEB)

    A. J. Palmer; DC Haggard; J. W. Herter; M. Scervini; W. D. Swank; D. L. Knudson; R. S. Cherry

    2011-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina

  18. The Agr communication system provides a benefit to the populations of Listeria monocytogenes in soil.

    Science.gov (United States)

    Vivant, Anne-Laure; Garmyn, Dominique; Gal, Laurent; Piveteau, Pascal

    2014-01-01

    In this study, we investigated whether the Agr communication system of the pathogenic bacterium Listeria monocytogenes was involved in adaptation and competitiveness in soil. Alteration of the ability to communicate, either by deletion of the gene coding the response regulator AgrA (response-negative mutant) or the signal pro-peptide AgrD (signal-negative mutant), did not affect population dynamics in soil that had been sterilized but survival was altered in biotic soil suggesting that the Agr system of L. monocytogenes was involved to face the complex soil biotic environment. This was confirmed by a set of co-incubation experiments. The fitness of the response-negative mutant was lower either in the presence or absence of the parental strain but the fitness of the signal-negative mutant depended on the strain with which it was co-incubated. The survival of the signal-negative mutant was higher when co-cultured with the parental strain than when co-cultured with the response-negative mutant. These results showed that the ability to respond to Agr communication provided a benefit to listerial cells to compete. These results might also indicate that in soil, the Agr system controls private goods rather than public goods.

  19. Staphylococcus aureus interaction with phospholipid vesicles--a new method to accurately determine accessory gene regulator (agr activity.

    Directory of Open Access Journals (Sweden)

    Maisem Laabei

    Full Text Available The staphylococcal accessory gene regulatory (agr operon is a well-characterised global regulatory element that is important in the control of virulence gene expression for Staphylococcus aureus, a major human pathogen. Hence, accurate and sensitive measurement of Agr activity is central in understanding the virulence potential of Staphylococcus aureus, especially in the context of Agr dysfunction, which has been linked with persistent bacteraemia and reduced susceptibility to glycopeptide antibiotics. Agr function is typically measured using a synergistic haemolysis CAMP assay, which is believe to report on the level of expression of one of the translated products of the agr locus, delta toxin. In this study we develop a vesicle lysis test (VLT that is specific to small amphipathic peptides, most notably delta and Phenol Soluble Modulin (PSM toxins. To determine the accuracy of this VLT method in assaying Agr activity, we compared it to the CAMP assay using 89 clinical Staphylococcus aureus isolates. Of the 89 isolates, 16 were designated as having dysfunctional Agr systems by the CAMP assay, whereas only three were designated as such by VLT. Molecular analysis demonstrated that of these 16 isolates, the 13 designated as having a functional Agr system by VLT transcribed rnaIII and secreted delta toxin, demonstrating they have a functional Agr system despite the results of the CAMP assay. The agr locus of all 16 isolates was sequenced, and only the 3 designated as having a dysfunctional Agr system contained mutations, explaining their Agr dysfunction. Given the potentially important link between Agr dysfunction and clinical outcome, we have developed an assay that determines this more accurately than the conventional CAMP assay.

  20. Obreros agrícolas migrantes en Sinaloa

    Directory of Open Access Journals (Sweden)

    Florencio Posadas Segura

    2015-01-01

    Full Text Available el objetivo de este artículo es demostrar la presencia sig-nificativa de los obreros agrícolas migrantes, algunas carac-terísticas que la explican y la pauperización que experimen-tan. La causa principal de la migración es la desocupación;los empresarios controlan los métodos de contratación, tras-portación y explotación, y los jornaleros rurales migrantestrabajan más, ganan menos y su situación es mayormentemiserable. Esto se ilustra con el caso de Villa Benito Juárez,Sinaloa, que prueba la gran importancia de los asalariadosestacionales del campo, como auténticos productores de lariqueza, y la paradoja de su condición de pobreza aguda. Elpatrón migratorio interno y externo descubierto contradicelos estereotipos prevalecientes, en el marco de la reestruc-turación productiva y laboral de la agricultura mexicana yla incorporación de fuerza de trabajo asalariada pagada pordebajo de su valor, segmentada y dosificada conveniente-mente de acuerdo con la estrategia del capital, guiada por elespíritu de la ganancia máxima.

  1. Scientific-technical cooperation with Russia. Transient analyses for alternative types of water-cooled reactors. Final report; WTZ mit Russland. Transientenanalysen fuer wassergekuehlte Kernreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Rohde, Ulrich [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Kozmenkov, Yaroslav [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Institute of Physics and Power Engineering, Obninsk (Russian Federation); Pivovarov, Valeri; Matveev, Yurij [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    2010-12-15

    The recently developed multi-group version DYN3D-MG of the reactor dynamics code DYN3D has been qualified for applications to water-cooled reactor concepts different from industrial PWR and BWR. An extended DYN3D version was applied to the graphite-moderated pressure tube reactor EGP-6 (NPP Bilibino) and conceptual design studies of an advanced Boiling Water Reactor with reduced moderation (RMWR) as well as the RUTA-70 reactor for low temperature heat supply. Concerning the RUTA reactor, safe heat removal by natural circulation of the coolant at low pressure has to be shown. For the corresponding validation of thermo-hydraulic system codes like ATHLET and RELAP5, experiments on flashing-induced natural circulation instabilities performed at the CIRCUS test facility at the TU Delft were simulated using the RELAP5 code. For the application to alternative water-cooled reactors, DYN3D model extensions and modifications were implemented, in particular adaptations of heat conduction and heat transfer models. Performing code-to-code comparisons with the Russian fine-mesh neutron diffusion code ACADEM contributed to the verification of DYN3D-MG. Validation has been performed by calculating reactor dynamics experiments at the NPP Bilibino. For the reactors EGP-6, RMWR and RUTA, analyses of various protected and unprotected control rod withdrawal and ejection transients were performed. The beyond design basis accident (BDBA) scenario ''Coast-down of all main coolant pumps at nominal power without scram'' for the RUTA reactor was analyzed using the code complexes DYN3D/ATHLET and DYN3D/RELAP5. It was shown, that the reactor passes over to a save asymptotic state at reduced power with coolant natural circulation. Analyzing the BDBA ''Unprotected withdrawal of a control rod group'' for the RMWR, the safety against Departure from Nucleate Boiling (DNB) could not be shown with the necessary confidence. Finally, conclusions have been drawn

  2. Design of the VISTA-ITL Test Facility for an Integral Type Reactor of SMART and a Post-Test Simulation of a SBLOCA Test

    Directory of Open Access Journals (Sweden)

    Hyun-Sik Park

    2014-01-01

    Full Text Available To validate the performance and safety of an integral type reactor of SMART, a thermal-hydraulic integral effect test facility, VISTA-ITL, is introduced with a discussion of its scientific design characteristics. The VISTA-ITL was used extensively to assess the safety and performance of the SMART design, especially for its passive safety system such as a passive residual heat removal system, and to validate various thermal-hydraulic analysis codes. The VISTA-ITL program includes several tests on the SBLOCA, CLOF, and PRHRS performances to support a verification of the SMART design and contribute to the SMART design licensing by providing proper test data for validating the system analysis codes. A typical scenario of SBLOCA was analyzed using the MARS-KS code to assess the thermal-hydraulic similarity between the SMART design and the VISTA-ITL facility, and a posttest simulation on a SBLOCA test for the shutdown cooling system line break has been performed with the MARS-KS code to assess its simulation capability for the SBLOCA scenario of the SMART design. The SBLOCA scenario in the SMART design was well reproduced using the VISTA-ITL facility, and the measured thermal-hydraulic data were properly simulated with the MARS-KS code.

  3. Decay profiles of {beta} and {gamma} for a radionuclide inventory in equilibrium cycle of a BWR type reactor; Perfiles de decaimiento de radiacion {beta} y {gamma} para un inventario de radionuclidos en ciclo de equilibrio de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Salaices, M.; Sandoval, S.; Ovando, R. [Instituto de Investigaciones Electricas. Gerencia de Energia Nuclear, Av. Reforma 113 Col. Palmira. 62490 Cuernavaca, Morelos (Mexico)]. e-mail: sal@iie.org.mx

    2007-07-01

    Presently work the {beta} and {gamma} radiation decay profiles for a radionuclides inventory in equilibrium cycle of a BWR type reactor is presented. The profiles are presented in terms of decay in the activity of the total inventory as well as of the chemical groups that conform the inventory. In the obtaining of the radionuclides inventory in equilibrium cycle the ORIGEN2 code, version 1 was used, which simulates fuel burnup cycles and it calculates the evolution of the isotopic composition as a result of the burnt one, irradiation and decay of the nuclear fuel. It can be observed starting from the results that the decrease in the activity for the initial inventory and the different chemical groups that conform it is approximately proportional to the base 10 logarithm of the time for the first 24 hours of having concluded the burnt one. It can also be observed that the chemical groups that contribute in more proportion to the total activity of the inventory are the lanthanides-actinides and the transition metals, with 39% and 28%, respectively. The groups of alkaline earth metals, halogens, metalloids, noble gases and alkaline metals, contribute with percentages that go from the 8 to 5%. The groups that less they contribute to the total activity of the inventory they are the non metals and semi-metals with smaller proportions that 1%. The chemical groups that more contribute to the energy of {beta} and {gamma} radiation its are the transition metals and the lanthanides-actinides with a change in the order of importance at the end of the 24 hours period. The case of the halogens is of relevance for the case of the {gamma} radiation energy due that occupying the very near third site to the dimensions of the two previous groups. Additionally, the decay in the activity for the total inventory and the groups that conform it can be simulated by means of order 6 polynomials or smaller than describe its behavior appropriately. The results presented in this work, coupled

  4. Evolution of weld metals nanostructure and properties under irradiation and recovery annealing of VVER-type reactors

    Science.gov (United States)

    Gurovich, B.; Kuleshova, E.; Shtrombakh, Ya.; Fedotova, S.; Zabusov, O.; Prikhodko, K.; Zhurko, D.

    2013-03-01

    The results of VVER-440 steel Sv-10KhMFT and VVER-1000 steel SV-10KhGNMAA investigations by transmission electron microscopy, scanning electron microscopy, Auger-electron spectroscopy and mechanical tests are presented in this paper. The both types of weld metals with different content of impurities and alloying elements were studied after irradiations to fast neutron (E > 0.5 MeV) fluences in the wide range below and beyond the design values, after recovery annealing procedures and after re-irradiation following the annealing. The distinctive features of embrittlement kinetics of VVER-440 and VVER-1000 RPV weld metals conditioned by their chemical composition differences were investigated. It is shown that the main contribution into radiation strengthening within the design fluence can be attributed to radiation-induced precipitates, on reaching the design or beyond design values of fast neutron fluencies the main contribution into VVER-440 welds strengthening is made by radiation-induced dislocation loops, and in case of VVER-1000 welds - radiation-induced precipitates and grain-boundary phosphorous segregations. Recovery annealing of VVER-440 welds at 475 °C during 100 h causes irradiation-induced defects disappearance, transformation of copper enriched precipitates into bigger copper-rich precipitates with lower number density and leads to almost full recovery of mechanical properties followed by comparatively slow re-embrittlement rate. The recovery annealing temperature of VVER-1000 welds was higher - 565 °C during 100 h - to avoid temper brittleness. The annealing of VVER-1000 welds leads to almost full recovery of mechanical properties due to irradiation-induced defects disappearance and decrease in precipitates number density and grain-boundary segregation of phosphorus. The re-embrittlement rate of VVER-1000 weld during subsequent re-irradiation is at least not higher than the initial rate.

  5. Estimation of fast neutron fluence in steel specimens type Laguna Verde in TRIGA Mark III reactor; Estimacion de la fluencia de neutrones rapidos en probetas de acero tipo Laguna Verde en el reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    The main purpose of this work is to obtain the fluence of fast neutrons recorded within four specimens of carbon steel, similar to the material having the vessels of the BWR reactors of the nuclear power plant of Laguna Verde when subjected to neutron flux in a experimental facility of the TRIGA Mark III reactor, calculating an irradiation time to age the material so accelerated. For the calculation of the neutron flux in the specimens was used the Monte Carlo code MCNP5. In an initial stage, three sheets of natural molybdenum and molybdenum trioxide (MoO{sub 3}) were incorporated into a model developed of the TRIGA reactor operating at 1 M Wth, to calculate the resulting activity by setting a certain time of irradiation. The results obtained were compared with experimentally measured activities in these same materials to validate the calculated neutron flux in the model used. Subsequently, the fast neutron flux received by the steel specimens to incorporate them in the experimental facility E-16 of the reactor core model operating at nominal maximum power in steady-state was calculated, already from these calculations the irradiation time required was obtained for values of the neutron flux in the range of 10{sup 18} n/cm{sup 2}, which is estimated for the case of Laguna Verde after 32 years of effective operation at maximum power. (Author)

  6. The DOE advanced gas reactor fuel development and qualification program

    Science.gov (United States)

    Petti, David; Maki, John; Hunn, John; Pappano, Pete; Barnes, Charles; Saurwein, John; Nagley, Scott; Kendall, Jim; Hobbins, Richard

    2010-09-01

    The high outlet temperatures and high thermal-energy conversion efficiency of modular high-temperature gas-cooled reactors (HTGRs) enable an efficient and cost-effective integration of the reactor system with non-electricity-generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300°C and 900°C. The U.S. Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. An overview of the program and recent progress is presented.

  7. Evaluation of Erosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Brower, Jeffrey O. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Glazoff, Michael V. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Eiden, Thomas J. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Rezvoi, Aleksey V. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor

    2016-08-01

    Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady state conditions. However, after the cycle was over, when the fuel elements were removed from the core and inspected, several thousand flow-assisted erosion pits and “horseshoeing” defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum “dummy” plates that contain no fuel). In order to understand these erosion phenomena a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed scalloping

  8. Progress of China Experimental Fast Reactor in 2011

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    1 Background Fast reactor is the reactor which realized the chain fission with fast neutron.As an optional type of generation Ⅳ reactor,fast reactor has three characters:1) It can change 238U to 239Pu and raise the uranium resource utilization

  9. Antineutrino Monitoring of Thorium Reactors

    CERN Document Server

    Akindele, Oluwatomi A; Norman, Eric B

    2015-01-01

    Various groups have demonstrated that antineutrino monitoring can be successful in assessing the plutonium content in water-cooled nuclear reactors for nonproliferation applications. New reactor designs and concepts incorporate nontraditional fuels types and chemistry. Understanding how these properties affect the antineutrino emission from a reactor can extend the applicability of antineutrino monitoring.Thorium molten salt reactors (MSR) breed U-233, that if diverted constitute an IAEA direct use material. The antineutrino spectrum from the fission of U-233 has been determined, the feasibility of detecting the diversion of a significant quantity, 8 kg of U-233, within the IAEA timeliness goal of 30 days has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario at a 25 meter standoff by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos. It was found that the diversion of a signifi...

  10. La normalización y certificación de tractores agrícolas en México

    OpenAIRE

    Alma Velia Ayala Garay; Rocío Cervantes Osornio; Marco Antonio Audelo Benítez; Noé Velázquez López; José Manuel Vargas Sállago

    2013-01-01

    El tractor agrícola es la principal fuente de potencia dentro de una unidad de producción. Por lo tanto para los usuarios de maquinaria agrícola, resulta una prioridad contar con mecanismos que permitan dar seguridad al usuario final o productor agrícola en el funcionamiento y calidad de los tractores, para esto el objetivo del presente es dar a conocer la evolución de la normalización y certificación de la maquinaria agrícola en México, sus impactos y tendencias, por medio de una descripción...

  11. Construction of an external electrode for determination of electrochemical corrosion potential in normal operational conditions of an BWR type reactor for hot cells; Construccion de un electrodo externo para determinacion del potencial de corrosion electroquimico en condiciones normales de operacion de un reactor tipo BWR para celdas calientes

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar T, J.A.; Rivera M, H.; Hernandez C, R. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    The behavior of the corrosion processes at high temperature requires of external devices that being capable to resist a temperature of 288 Centigrade and a pressure of 80 Kg/cm{sup 2}, to give stable and reproducible results of some variable and resisting physically and chemically the radiation. The external electrode of Ag/AgCl fulfils all the requirements in the determination of the electrochemical corrosion potential under normal operational conditions of a BWR type reactor in hot cells. (Author)

  12. Oscillatory flow chemical reactors

    Directory of Open Access Journals (Sweden)

    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.

  13. AGR-1 Irradiated Test Train Preliminary Inspection and Disassembly First Look

    Energy Technology Data Exchange (ETDEWEB)

    Paul Demkowicz; Lance Cole; Scott Ploger; Philip Winston; Binh Pham; Michael Abbott

    2011-01-01

    The AGR-1 irradiation experiment ended on November 6, 2009, after 620 effective full power days in the Advanced Test Reactor, achieving a peak burnup of 19.6% FIMA. The test train was shipped to the Materials and Fuels Complex in March 2010 for post-irradiation examination. The first PIE activities included non-destructive examination of the test train, followed by disassembly of the test train and individual capsules and detailed inspection of the capsule contents, including the fuel compacts and the graphite fuel holders. Dimensional measurements of the compacts, graphite holders, and steel capsules shells were performed using a custom vision measurement system (for outer diameters and lengths) and conventional bore gauges (for inner diameters). Gamma spectrometry of the intact test train gave a preliminary look at the condition of the interior components. No evidence of damage to compacts or graphite components was evident from the isotopic and gross gamma scans. Neutron radiography of the intact Capsule 2 showed a high degree of detail of interior components and confirmed the observation that there was no major damage to the capsule. Disassembly of the capsules was initiated using procedures qualified during out-of-cell mockup testing. Difficulties were encountered during capsule disassembly due to irradiation-induced changes in some of the capsule components’ properties, including embrittled niobium and molybdenum parts that were susceptible to fracture and swelling of the graphite fuel holders that affected their removal from the capsule shells. This required various improvised modifications to the disassembly procedure to avoid damage to the fuel compacts. Ultimately the capsule disassembly was successful and only one compact from Capsule 4 (out of 72 total in the test train) sustained damage during the disassembly process, along with the associated graphite holder. The compacts were generally in very good condition upon removal. Only relatively minor

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

    Directory of Open Access Journals (Sweden)

    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

  15. Biotecnología agrícola

    Directory of Open Access Journals (Sweden)

    Melgarejo, Paloma

    2014-08-01

    Full Text Available The domestication of plants can be considered as the first genetic manipulation of living organisms and it is associated with the origin of agriculture, and represents the major contribution of biotechnology to the development of mankind. The challenges the agricultural sector faces in the coming decades are population growth and the increasing demand for raw materials for food, feed, as well as the consumer preferences, environmental and legal aspects, and the overall globalisation of the economy. The agricultural biotechnology applications will undoubtedly play a role in meeting these challenges. Spain has competent scientific groups devoted to basic and applied research, with a wide knowledge on agricultural biotechnology. The transfer of the business sector needs to the scientific field, coordinating the R&D companies and research groups from universities and public research organisms is an objective to be attained in the future.La domesticación de plantas, primera manipulación genética de los organismos vivos y origen de la Agricultura, representa la mayor contribución de la biotecnología al desarrollo de la humanidad. Los retos a superar en las próximas décadas para el sector agrario están determinados por el aumento demográfico y la demanda de materias primas para alimentación humana y animal, por las preferencias en el consumo, por aspectos medioambientales y legales, y por la globalización de la economía. Las aplicaciones de la biotecnología agrícola van a jugar un papel incuestionable en el logro de estos retos. Nuestro país posee unos buenos grupos en investigación básica y aplicada, con amplios conocimientos de la agrobiotecnología. El traslado de las necesidades del sector empresarial a la investigación, coordinando los departamentos de I+D+I de las empresas y los grupos de investigación de las universidades y organismos públicos de investigación es una finalidad a conseguir en el futuro.

  16. Brazilian multipurpose reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Brazilian Multipurpose Reactor (RMB) Project is an action of the Federal Government, through the Ministry of Science Technology and Innovation (MCTI) and has its execution under the responsibility of the Brazilian National Nuclear Energy Commission (CNEN). Within the CNEN, the project is coordinated by the Research and Development Directorate (DPD) and developed through research units of this board: Institute of Nuclear Energy Research (IPEN); Nuclear Engineering Institute (IEN); Centre for Development of Nuclear Technology (CDTN); Regional Center of Nuclear Sciences (CRCN-NE); and Institute of Radiation Protection and Dosimetry (IRD). The Navy Technological Center in Sao Paulo (CTMSP) and also the participation of other research centers, universities, laboratories and companies in the nuclear sector are important and strategic partnerships. The conceptual design and the safety analysis of the reactor and main facilities, related to nuclear and environmental licensing, are performed by technicians of the research units of DPD / CNEN. The basic design was contracted to engineering companies as INTERTHECNE from Brazil and INVAP from Argentine. The research units from DPD/CNEN are also responsible for the design verification on all engineering documents developed by the contracted companies. The construction and installation should be performed by specific national companies and international partnerships. The Nuclear Reactor RMB will be a open pool type reactor with maximum power of 30 MW and have the OPAL nuclear reactor of 20 MW, built in Australia and designed by INVAP, as reference. The RMB reactor core will have a 5x5 configuration, consisting of 23 elements fuels (EC) of U{sub 3}Si{sub 2} dispersion-type Al having a density of up to 3.5 gU/cm{sup 3} and enrichment of 19.75% by weight of {sup 23{sup 5}}U. Two positions will be available in the core for materials irradiation devices. The main objectives of the RMB Reactor and the other nuclear and radioactive

  17. Implications of Results from the Advanced Gas Reactor Fuel Development and Qualification Program on Licensing of Modular HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    David Petti

    2001-10-01

    The high level of safety of modular high temperature gas-cooled reactor (HTGR) designs is achieved by passively maintaining core temperatures below fission product release thresholds under all envisioned accident scenarios. This level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude relative to other reactor types but is predicated on exceptionally high coated-particle fuel fabrication quality and excellent fuel performance under normal operation and accident conditions. The Advanced Gas Reactor Fuel Development and Qualification (AGR) Program decided to qualify for uranium oxide/uranium carbide (UCO) TRISO coated-particle fuel in an operating envelope that would bound both pebble bed and prismatic modular HTGR options. By using a mixture of uranium oxide and uranium carbide, the kernel composition is engineered to minimize CO formation and fuel kernel migration, which is key to maintain to fuel integrity at the higher burnups, temperatures, and temperature gradients anticipated in prismatic HTGRs. Fuel fabrication conducted at both laboratory and engineering scale has demonstrated the ability to fabricate high quality UCO TRISO fuel with very low defects. The first irradiation (AGR 1) exposed about 300,000 TRISO fuel particles to a peak burnup of 19.6% FIMA, a peak fast-neutron fluence of about 4.3 × 1025 n/m2, and a maximum time-averaged fuel temperature of about 1,200°C without a single particle failure. The very low release of key metallic fission products (except silver) measured in post-irradiation examination (PIE) confirms the excellent performance measured under irradiation. Very low releases have been measured in accident simulation heatup testing (''safety testing'') after hundreds of hours at 1600 and 1700°C and no particle failures (no noble gas release measured) have been observed. Even after hundreds of hours at 1800°C, the releases are

  18. Modelos de vulnerabilidad agrícola ante los efectos del cambio climático

    Directory of Open Access Journals (Sweden)

    Claudia Hernández Ramírez

    2014-12-01

    Full Text Available La Vulnerabilidad Agrícola (VA es el resultado de la medición de la variabilidad climática dentro de un contexto de vulnerabilidad social y sectorial, que ha ido agregando conceptos e indicadores que enriquecen y complementan el análisis de la seguridad alimentaria, en el entendido que la disponibilidad de un pronóstico corresponde no solo al derecho primordial de cualquier ser humano, sino a la supervivencia del mismo. La metodología de medición de esta variable depende estrictamente del alcance de la investigación y de la disponibilidad de datos de la región y del producto agrícola seleccionado. Los modelos utilizados hasta la fecha son tres: el modelo Estructural, el modelo Espacial y la Vulnerabilidad Agrícola.

  19. Política Agrícola Canadiense. Algunas lecciones para Argentina

    Directory of Open Access Journals (Sweden)

    Rinaldo Antonio Colomé

    2010-09-01

    Full Text Available El objetivo general de este trabajo es analizar el diseño de la política agrícola canadiense, comparándola con las de los otros países altamente desarrollados y con la de Argentina.  Los objetivos específicos son: aAnalizar la política agrícola de Canadá referida a la producción y al comercio de granos y su efecto en los mercados internacionales; bAnalizar la política agrícola y comercial de Canadá referida al sector de oferta dirigida o regulada y sus efectos en el mercado doméstico y en los internacionales; c Realizar una comparación con Argentina. 

  20. Thermal-hydraulic interfacing code modules for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

    1997-07-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

  1. Optimally moderated nuclear fission reactor and fuel source therefor

    Science.gov (United States)

    Ougouag, Abderrafi M.; Terry, William K.; Gougar, Hans D.

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  2. Validation of WIMS-SNAP code systems for calculations in TRIGA-MARK II type reactors; Validacion del sistema de codigos WIMS-SNAP para calculos en reactores nucleares tipo TRIGA-MARK II

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez Valle, S.; Lopez Aldama, D. [Centro de Investigaciones Nucleares, Tecnologicas y Ambientales, La Habana (Cuba). E-mail: svalle@ctn.isctn.edu.cu

    2000-07-01

    The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d{sup 4} code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)

  3. High temperature and pressure effects on the elastic properties of B2 intermetallics AgRE

    Science.gov (United States)

    Liu, Lili; Wu, Xiaozhi; Li, Weiguo; Wang, Rui; Liu, Qing

    2015-02-01

    The high temperature and pressure effects on the elastic properties of the AgRE (RE=Sc, Tm, Er, Dy, Tb) intermetallic compounds with B2 structure have been performed from first principle calculations. For the temperature range 0-1000 K, the second order elastic constants for all the AgRE intermetallic compounds follow a normal behavior: they decrease with increasing temperature. The pressure dependence of the second order elastic constants has been investigated on the basis of the third order elastic constants. Temperature and pressure dependent elastic anisotropic parameters A have been calculated based on the temperature and pressure dependent elastic constants.

  4. Conceptos sobre Adecuación de Tierras Agrícolas

    Directory of Open Access Journals (Sweden)

    Bustamante Muñoz Hellodoro

    1986-06-01

    Full Text Available En este artículo se pretende de una manera muy resumida mostrar algunos conceptos y principios generales sobre la adecuación de tierras agrícolas Adecuación en términos simples significa acción o efecto de proporcionar o acomodar una cosa a otra; por extensión y a manera de ensayo de definición, adecuación de tierras agrícolas consiste en crear o proporcionar las condiciones necesarias para la eficiente explotación del campo.

  5. A eficiência nas caixas de Crédito Agrícola Mútuo

    OpenAIRE

    Pina, Licínio Manuel Prata

    2012-01-01

    Mestrado em Economia Portuguesa e Integração Internacional O Crédito Agrícola é o único banco cooperativo em actividade no mercado português. Constituído actualmente por oitenta e cinco caixas agrícolas e a caixa central, que sustentam a base do grupo com presença em quase todo o território nacional, caracteriza-se pela prática de banca de proximidade e de grande envolvimento com as comunidades locais, tentando contrariar a tendência para a centralidade territorial. O grupo ...

  6. La oferta agrícola, a corto y largo plazo

    OpenAIRE

    de Pablo, Juan Carlos

    2012-01-01

    Los productores agrícolas, y particularmente su dirigencia, cuentan con mi fraterna simpatía, a raíz de los sistemáticos ataques que sufren por parte de las autoridades nacionales, y la indiferencia que exhibe el resto de los dirigentes empresarios. Lo cual no implica compartir cualquier afirmación referida a lo que podría ocurrir con la producción agrícola de Argentina, en ausencia de trabas a las exportaciones e intervención directa en el mercado interno de dichos productos, tanto de índole...

  7. Diseño de un registrador de esfuerzos para maquinaria agrícola

    OpenAIRE

    2016-01-01

    La agricultura de precisión requiere de herramientas tecnológicas para la adquisición, registro y procesamiento de la información. La incorporación de los sistemas embebidos a la maquinaria agrícola permite realizar el control y monitoreo de forma constante y precisa de las variables que influyen en el rendimiento y afectan a la correcta utilización de la maquinaria. La eficiencia del tractor agrícola puede determinarse como la relación entre el consumo de combustible y la energía mecánica...

  8. CD147 and AGR2 expression promote cellular proliferation and metastasis of head and neck squamous cell carcinoma

    Energy Technology Data Exchange (ETDEWEB)

    Sweeny, Larissa, E-mail: larissasweeny@gmail.com [Department of Surgery, University of Alabama, Division of Otolaryngology-Head and Neck Surgery, 1670 University Boulevard, Volker Hall G082, Birmingham, Alabama (United States); Liu, Zhiyong; Bush, Benjamin D.; Hartman, Yolanda [Department of Surgery, University of Alabama, Division of Otolaryngology-Head and Neck Surgery, 1670 University Boulevard, Volker Hall G082, Birmingham, Alabama (United States); Zhou, Tong [Department of Medicine, Division of Immunology and Rheumatology, 1825 University Boulevard, Shelby Biomedical Research Building 302, Birmingham, Alabama (United States); Rosenthal, Eben L., E-mail: oto@uab.edu [Department of Surgery, University of Alabama, Division of Otolaryngology-Head and Neck Surgery, 1670 University Boulevard, Volker Hall G082, Birmingham, Alabama (United States)

    2012-08-15

    The signaling pathways facilitating metastasis of head and neck squamous cell carcinoma (HNSCC) cells are not fully understood. CD147 is a transmembrane glycoprotein known to induce cell migration and invasion. AGR2 is a secreted peptide also known to promote cell metastasis. Here we describe their importance in the migration and invasion of HNSCC cells (FADU and OSC-19) in vitro and in vivo. In vitro, knockdown of CD147 or AGR2 decreased cellular proliferation, migration and invasion. In vivo, knockdown of CD147 or AGR2 expression decreased primary tumor growth as well as regional and distant metastasis. -- Highlights: Black-Right-Pointing-Pointer We investigated AGR2 in head and neck squamous cell carcinoma for the first time. Black-Right-Pointing-Pointer We explored the relationship between AGR2 and CD147 for the first time. Black-Right-Pointing-Pointer AGR2 and CD147 appear to co-localize in head and squamous cell carcinoma samples. Black-Right-Pointing-Pointer Knockdown of both AGR2 and CD147 reduced migration and invasion in vitro. Black-Right-Pointing-Pointer Knockdown of both AGR2 and CD147 decreased metastasis in vivo.

  9. Reactor service life extension program

    Energy Technology Data Exchange (ETDEWEB)

    Caskey, G.R.; Sindelar, R.L.; Ondrejcin, R.S.; Baumann, E.W.

    1990-12-31

    A review of the Savannah River Site production reactor systems was initiated in 1980 and led to implementation of the Reactor Materials Program in 1984 to assess reactor safety and reactor service life. The program evaluated performance of the reactor tanks, primary coolant piping, and thermal shields, components of welded construction that were fabricated from Type 304 stainless steel. The structural integrity analysis of the primary coolant system has shown that the pressure boundary is not susceptible to gross rupture, including a double ended guillotine break or equivalent large area bank. Residual service life is potentially limited by two material degradation modes, irradiation damage and intergranular stress corrosion cracking. Analysis of the structural integrity of the tanks and piping has shown that continued safe operation of the reactors for several additional decades is not limited by the material performance of the primary coolant system. Although irradiation damage has not degraded material behavior to an unacceptable level, past experience has revealed serious difficulties with repair welding on irradiated stainless steel. Stress corrosion can be mitigated by newly identified limits on impurity concentrations in the coolant water and by stress mitigation of weld residual stresses. Work continues in several areas: the effects of helium on mechanical behavior of irradiated stainless steel; improved weld methods for piping and the reactor tanks; and a surveillance program to track irradiation effects on the tank walls.

  10. Reactor service life extension program

    Energy Technology Data Exchange (ETDEWEB)

    Caskey, G.R.; Sindelar, R.L.; Ondrejcin, R.S.; Baumann, E.W.

    1990-01-01

    A review of the Savannah River Site production reactor systems was initiated in 1980 and led to implementation of the Reactor Materials Program in 1984 to assess reactor safety and reactor service life. The program evaluated performance of the reactor tanks, primary coolant piping, and thermal shields, components of welded construction that were fabricated from Type 304 stainless steel. The structural integrity analysis of the primary coolant system has shown that the pressure boundary is not susceptible to gross rupture, including a double ended guillotine break or equivalent large area bank. Residual service life is potentially limited by two material degradation modes, irradiation damage and intergranular stress corrosion cracking. Analysis of the structural integrity of the tanks and piping has shown that continued safe operation of the reactors for several additional decades is not limited by the material performance of the primary coolant system. Although irradiation damage has not degraded material behavior to an unacceptable level, past experience has revealed serious difficulties with repair welding on irradiated stainless steel. Stress corrosion can be mitigated by newly identified limits on impurity concentrations in the coolant water and by stress mitigation of weld residual stresses. Work continues in several areas: the effects of helium on mechanical behavior of irradiated stainless steel; improved weld methods for piping and the reactor tanks; and a surveillance program to track irradiation effects on the tank walls.

  11. Use of standard spectra for the short life radionuclides and ratios for long life radionuclides in the wastes of EDF PWR type reactors; Utilisation de spectres types pour les radionucleides a vie courte et de ratios pour les radionucleides a vie longue dans les dechets de REP EDF

    Energy Technology Data Exchange (ETDEWEB)

    Lantes, B. [Electricite de France (EDF-DPN/Groupe Environnement), 31 - Toulouse (France); Bienvenu, Ph. [CEA Cadarache, Dept. d' Etudes des Dechets, DED, 13 - Saint-Paul-lez-Durance (France)

    2001-07-01

    This paper presents the type of declaration of radioactivity in the wastes of PWR type reactors park. Particularly, it insists on the justification of use of spectra for the declaration of short live radionuclides. It tackles the important developments of methods and measures of radiochemical analysis made by the Cea in order to determine the ratios to declare the long life radioisotopes. (N.C.)

  12. Experimental studies of local coolant hydrodynamics using a scaled model of cassette-type fuel assembly of a KLT-40S reactor

    Science.gov (United States)

    Dmitriev, S. M.; Barinov, A. A.; Varentsov, A. V.; Doronkov, D. V.; Solntsev, D. N.; Khrobostov, A. E.

    2016-08-01

    The results of experimental studies of local hydrodynamic and mass exchange characteristics of the coolant flow behind the spacer grid in the fuel assembly of a KLT-40S reactor are presented. The experiments were aimed at the investigation of representative domains of the fuel assembly with three tracer injection regions. The studies were performed at the aerodynamic test facility using the tracer gas diffusion method. According to the theory of hydrodynamic similarity, the obtained experimental results can be transferred to full-scale coolant flow conditions in standard fuel assemblies. The analysis of the tracer concentration propagation made it possible to determine in detail the flow pattern and find the main regularities and specific features of the coolant flow behind the plate spacer grid of KLT-40S fuel assembly. The hydraulic resistance coefficient of the spacer grid was experimentally determined. The coefficients of mass exchange between cells for representative cells of the displacer region in the KLT-40S fuel assembly were calculated for the first time; these results are presented in the form of the "mixing matrix." The results of studies of local coolant flow hydrodynamics in the KLT-40S fuel assembly are used at AO Afrikantov OKBM for estimation of thermotechnical reliability of active cores for reactors of floating nuclear power stations. The experimental data on hydrodynamic and mass exchange characteristics are included in the database for verification of CDF codes and detailed cell-wise calculation of the active core for KLT-40S reactor installation. The results of these studies can be used at FSUE RFNC-VNIIEF for testing and verification of domestic three-dimensional hydrodynamic CFD codes ("Logos") that are applied for substantiation of newly designed reactor installations. Practical recommendations on the application of the obtained results in thermohydraulic calculations of the active core for the KLT-40S reactor will be worked out. Proposals

  13. Analysis of Microbial Communities in Biofilms from CSTR-Type Hollow Fiber Membrane Biofilm Reactors for Autotrophic Nitrification and Hydrogenotrophic Denitrification.

    Science.gov (United States)

    Shin, Jung-Hun; Kim, Byung-Chun; Choi, Okkyoung; Kim, Hyunook; Sang, Byoung-In

    2015-10-01

    Two hollow fiber membrane biofilm reactors (HF-MBfRs) were operated for autotrophic nitrification and hydrogenotrophic denitrification for over 300 days. Oxygen and hydrogen were supplied through the hollow fiber membrane for nitrification and denitrification, respectively. During the period, the nitrogen was removed with the efficiency of 82-97% for ammonium and 87-97% for nitrate and with the nitrogen removal load of 0.09-0.26 kg NH4(+)-N/m(3)/d and 0.10-0.21 kg NO3(-)-N/m(3)/d, depending on hydraulic retention time variation by the two HF-MBfRs for autotrophic nitrification and hydrogenotrophic denitrification, respectively. Biofilms were collected from diverse topological positions in the reactors, each at different nitrogen loading rates, and the microbial communities were analyzed with partial 16S rRNA gene sequences in denaturing gradient gel electrophoresis (DGGE). Detected DGGE band sequences in the reactors were correlated with nitrification or denitrification. The profile of the DGGE bands depended on the NH4(+) or NO3(-) loading rate, but it was hard to find a major strain affecting the nitrogen removal efficiency. Nitrospira-related phylum was detected in all biofilm samples from the nitrification reactors. Paracoccus sp. and Aquaspirillum sp., which are an autohydrogenotrophic bacterium and an oligotrophic denitrifier, respectively, were observed in the denitrification reactors. The distribution of microbial communities was relatively stable at different nitrogen loading rates, and DGGE analysis based on 16S rRNA (341f /534r) could successfully detect nitrate-oxidizing and hydrogen-oxidizing bacteria but not ammonium-oxidizing bacteria in the HF-MBfRs.

  14. U.S. Nuclear Power Reactor Plant Status

    Data.gov (United States)

    Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

  15. Analysis of an accident type sbloca in reactor contention AP1000 with 8.0 Gothic code; Analisis de un accidente tipo Sbloca en la contencion del reactor AP1000 con el codigo Gothic 8.0

    Energy Technology Data Exchange (ETDEWEB)

    Goni, Z.; Jimenez Varas, G.; Fernandez, K.; Queral, C.; Montero, J.

    2016-08-01

    The analysis is based on the simulation of a Small Break Loss-of-Coolant-Accident in the AP1000 nuclear reactor using a Gothic 8.0 tri dimensional model created in the Science and Technology Group of Nuclear Fision Advanced Systems of the UPM. The SBLOCA has been simulated with TRACE 5.0 code. The main purpose of this work is the study of the thermo-hydraulic behaviour of the AP1000 containment during a SBLOCA. The transients simulated reveal close results to the realistic behaviour in case of an accident with similar characteristics. The pressure and temperature evolution enables the identification of the accident phases from the RCS point of view. Compared to the licensing calculations included in the AP1000 Safety Analysis, it has been proved that the average pressure and temperature evolution is similar, yet lower than the licensing calculations. However, the temperature and inventory distribution are significantly heterogeneous. (Author)

  16. Stability analysis of a recycling circuit of a BWR type reactor. Theoretical study; Analisis de estabilidad de un circuito de recirculacion de un reactor del tipo BWR. Estudio teorico

    Energy Technology Data Exchange (ETDEWEB)

    Salinas H, J.G.; Espinosa P, G. [Universidad Autonoma Metropolitana-Iztapalapa, 09000 Mexico D.F. (Mexico); Gonzalez M, V.M. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 04000 Mexico D.F. (Mexico)

    2000-07-01

    The Technology, Regulation and Services Management of the National Commission of Nuclear Safety and Safeguards financed and in coordinate form with the I.P.H. Department of the Metropolitan Autonomous-Iztapalapa University developed the present project with the purpose of studying the effect of the recycling system on the linear stability of a BWR reactor whose reference central is the Laguna Verde power station. The present project forms part of a work series focused to the linear stability of the nuclear reactor of the Unit 1 at Laguna Verde power station. The components of the recycling system considered for the study of stability are the recycling external circuit (recycling pumps, valves) and the internal circuit (downcomer, jet pumps, lower full, driers, separators). The mathematical model is obtained applying mass balances and movement quantity in each one of the mentioned circuits. With respect to the nucleus model two regions are considered, the first one is made of a flow in one phase and the second one of a flow in two phases. For modelling the biphasic region it is considered homogenous flow. Generally it is studied the system behavior in the frequency domain starting from the transfer function applied to four operational states which correspond to the lower stability zone in the map power-flow of the Unit 1 of Laguna Verde power station. The Nyquist diagrams corresponding to each state as well as their characteristic frequency were determined. The results show that exists a very clear dependence of the power-flow relation on the stability of the system. It was found that the boiling length is an important parameter for the linear stability of the system. The obtained results show that the characteristic frequencies in unstability zones are similar to the reported data of the Unit 1 of the Laguna Verde power station in the event of power oscillations carried out in January 1995. (Author)

  17. Development of a neutronic model for the fuel of a high temperature gas reactor type PBMR; Desarrollo de un modelo neutronico para el combustible de un reactor de gas de alta temperatura tipo PBMR

    Energy Technology Data Exchange (ETDEWEB)

    Oropeza C, I.; Carmona H, R.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: ivonucci@prodigy.net.mx

    2008-07-01

    In this work was developed the neutronic model of a fuel sphere of a nuclear reactor of gas of high temperature to modulate of bed of spheres (PBMR), using the Monte Carlo method with the MCNPx code. In order to be able to verify the fuel model constructed in this investigation, it is used a case of reference, based on an international exercise {sup b}enchmark{sup .} The benchmark report contains the results sent by different international participants for five phases with respect to the high temperature gas reactor (HTR), fed with uranium, plutonium and thorium. In particular, in first stage of benchmark an infinite adjustment of uranium compound fuel spheres is considered unique, with which our results were compared. This first stage considers two cases: cell calculations with spherical external frontier and cell calculations with cubic external frontier. The objective is to identify any increase in the uncertainty, related to the uranium fuel, that is associated with the plutonium and thorium fuels. In order to validate our results, the values of the neutron multiplication factor were taken in account, in cold and in the heat of the moment from the participants who sent their results obtained with Monte Carlo and deterministic calculations. The model of the fuel sphere developed in this work considers a regular distribution of 15000 Triso particles, in a cubic mesh centered within the sphere. For it was necessary to define the step firstly or {sup p}itch{sup o}f the cubic mesh. Generally, the results obtained by the participants of benchmark and those of this investigation present good agreement, nevertheless, appear some discrepancies, attributed to factors like different libraries of cross sections used, the nature of the solution: Monte Carlo or deterministic, and the difficulty of some participants to model the external frontier condition of reflection. (Author)

  18. Reactor Bolshoi Moshchnosti Kalani; Reacteurs RBMK

    Energy Technology Data Exchange (ETDEWEB)

    Bastien, D. [Conservatoire National des Arts et Metiers (CNAM), 75 - Paris (France)

    2000-01-01

    The Reactor Bolshoi Molshchnosti Kalani (RBMK) are pressure tubes reactor, boiling light water cooled. Exported since 1990 from the ex-USSR, they are today in three independent countries: Russian, Ukraine and Lithuania. Since this date, data exchange with the occident allowed the better knowledge of this reactor type. The design, the technical description (core, fuel, primary system), the safety and the improvement since Chernobyl are detailed. (A.L.B.)

  19. Nuevos neumáticos para tractores agrícolas

    OpenAIRE

    Arnó i Pujol, Judit; Masip Vilalta, Joan

    1997-01-01

    El neumático agrícola condiciona de forma determinante el rendimiento del tractor. Es necesario, por tanto, que el agricultar conozca las prestaciones de los neumáticos actuales. Ello le permitirá elegir el equipamiento más adecuado a sus necesidades.

  20. Colorado's AgrAbility Project's Effects on KASA and Practice Changes with Agricultural Producers and Professionals

    Science.gov (United States)

    Fetsch, Robert J.; Jackman, Danielle M.

    2015-01-01

    Disability rates resulting from work-related injuries remain steadily high among farmers and ranchers. To address the gap in services within this population, USDA implemented AgrAbility nationally. Using part of Bennett's hierarchical model, the current study evaluated the KASA and practice change levels of 401 farmers and ranchers and compared…

  1. Cross-Talk between Staphylococcus aureus and Other Staphylococcal Species via the agr Quorum Sensing System

    DEFF Research Database (Denmark)

    Canovas de la Nuez, Jaime; Baldry, Mara; Bojer, Martin S;

    2016-01-01

    Staphylococci are associated with both humans and animals. While most are non-pathogenic colonizers, Staphylococcus aureus is an opportunistic pathogen capable of causing severe infections. S. aureus virulence is controlled by the agr quorum sensing system responding to secreted auto-inducing pep...

  2. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    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.

  3. D and DR Reactors

    Data.gov (United States)

    Federal Laboratory Consortium — The world's second full-scale nuclear reactor was the D Reactor at Hanford which was built in the early 1940's and went operational in December of 1944.D Reactor ran...

  4. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    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.

  5. Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors

    Directory of Open Access Journals (Sweden)

    Peel Ross

    2016-01-01

    Full Text Available This paper discusses concepts for three-component fuel bundles containing plutonium, uranium and thorium for use in pressurised heavy water reactors, and cases for and against implementation of such a nuclear energy system in the United Kingdom. Heavy water reactors are used extensively in Canada, and are deploying within India and China, whilst the UK is considering the use of heavy water reactors to manage its plutonium inventory of 140 tonnes. The UK heavy water reactor proposal uses a mixed oxide (MOX fuel of plutonium in depleted uranium, within the enhanced CANDU-6 (EC-6 reactor. This work proposes an alternative heterogeneous fuel concept based on the same reactor and CANFLEX fuel bundle, with eight large-diameter fuel elements loaded with natural thorium oxide and 35 small-diameter fuel elements loaded with a MOX of plutonium and reprocessed uranium stocks from UK MAGNOX and AGR reactors. Indicative neutronic calculations suggest that such a fuel would be neutronically feasible. A similar MOX may alternatively be fabricated from reprocessed <5% enriched light water reactor fuel, such as the fuel of the AREVA EPR reactor, to consume newly produced plutonium from reprocessing, similar to the DUPIC (direct use of PWR fuel in CANDU process.

  6. Utilizing a one-dimensional multispecies model to simulate the nutrient reduction and biomass structure in two types of H2-based membrane-aeration biofilm reactors (H2-MBfR): model development and parametric analysis.

    Science.gov (United States)

    Wang, Zuowei; Xia, Siqing; Xu, Xiaoyin; Wang, Chenhui

    2016-02-01

    In this study, a one-dimensional multispecies model (ODMSM) was utilized to simulate NO3(-)-N and ClO4(-) reduction performances in two kinds of H2-based membrane-aeration biofilm reactors (H2-MBfR) within different operating conditions (e.g., NO3(-)-N/ClO4(-) loading rates, H2 partial pressure, etc.). Before the simulation process, we conducted the sensitivity analysis of some key parameters which would fluctuate in different environmental conditions, then we used the experimental data to calibrate the more sensitive parameters μ1 and μ2 (maximum specific growth rates of denitrification bacteria and perchlorate reduction bacteria) in two H2-MBfRs, and the diversity of the two key parameters' values in two types of reactors may be resulted from the different carbon source fed in the reactors. From the simulation results of six different operating conditions (four in H2-MBfR 1 and two in H2-MBfR 2), the applicability of the model was approved, and the variation of the removal tendency in different operating conditions could be well simulated. Besides, the rationality of operating parameters (H2 partial pressure, etc.) could be judged especially in condition of high nutrients' loading rates. To a certain degree, the model could provide theoretical guidance to determine the operating parameters on some specific conditions in practical application.

  7. Analysis of Discharge Fault in 35 kV Dry-type Air-Core Reactor%一起35 kV干式空心电抗器放电故障分析

    Institute of Scientific and Technical Information of China (English)

    张宁; 李洪伟

    2014-01-01

    本文介绍了一起500 kV变电站35 kV干式空心电抗器在运行当中发生放电烧损的故障情况,通过现场检查、试验,结合故障电抗器的解体检查结果,对故障原因进行了深入分析,发现故障的主要原因是由于在强磁场下涡流产生温升,破坏了电抗器本身绝缘,本文对防止同类故障的发生具有一定的借鉴意义。%In this paper, a fault of burn-out of 35 kV dry-type air-core reactor due to discharge in the operation at 500 kV substation is introduced. The reason of the fault is analyzed through site inspec-tion, test and disassembly inspection result of the faulty reactor. It is caused by high temperature rise caused by eddy current at strong magnetic field resulting to insulation damage of the reactor it-self. The analysis result has a definite reference to similar fault to be prevented.

  8. Study on the design of the new type of flocculation reactor and its performances%新型絮凝反应器的设计及应用性能研究

    Institute of Scientific and Technical Information of China (English)

    刘兴旺; 杨运泉

    2009-01-01

    A new type of flocculation reactor has been designed, whose performances have been evaluated by its residence time distribution. The new reactor's D/UL is 0.10, showing that it is close to plug flow reactor. Its flocculation efficiency is high. The new reactor is used for treating wastewater in iron and steel works. Dynamic tests are conducted according to the flocculant dosage ratio under static conditions. The hardness of the wastewater decreases from 143 mg/L to below 100 mg/L, its alkalinity decreases from 118 mg/L to below 100 mg/L, and its SS decreases from 213 mg/L to below 30 mg/L. The cost of coagulant can be controlled to 0.10 yuan/m3, which accords with the recycle standards of water quality and cost requirements of iron and steel works.%设计了一种新型结构的絮凝反应器,利用停留时间分布对其性能进行评价,测得它的分散数为0.10,接近于活塞流反应器,絮凝效率高.采用该絮凝器进行钢铁企业实际废水处理试验,结果表明:按静态条件下的药剂配比进行动态试验,废水的硬度、碱度和SS可以分别由进水时的143、118、213 mg/L控制在100、100、30 mg/L以内,药剂成本可控制在0.10元/m3,达到企业的回用水质和成本要求.

  9. DeveIopment of interactive safety anaIysis program for pooI type sodium cooIed fast reactor%池式钠冷快堆交互式安全分析软件开发

    Institute of Scientific and Technical Information of China (English)

    钱鸿涛; 李政昕; 胡文军; 宫宇

    2015-01-01

    为建立适用于池式钠冷快堆的仿真机,开发了基于法国快堆系统分析程序 OASIS 的交互式安全分析系统,实现了实时绘图、动态显示等可视化功能。利用该系统模拟了中国实验快堆的堆芯、主热传输系统、事故余热排出系统,以及控制调节系统和保护系统,分析了各个功率台阶的稳态及满功率下流量阶跃瞬态工况。分析结果与设计值符合度良好,表明该系统具有良好的适用性,可用于人员培训与安全审评等。%An interactive safety analysis program was developed and integrated into the simulation system for pool type sodium cooled fast reactor based on a French fast reactor system analysis code OASIS.The visualized functions of real-time plotting and dynamic display were provided.The core,main power transfer system,decay heat removal system,control and regulation system and reactor protection system of China Experimental Fast Reactor were simulated by the system.The various power level steady states and the transient of flow step at full power state were analyzed.The calculation results match well with the design data.It can be indicated that the program had a good applicability,and can be used for personnel training and safety review.

  10. AGR-2 Data Qualification Report for ATR Cycles 151B-2, 152A, 152B, 153A, 153B and 154A

    Energy Technology Data Exchange (ETDEWEB)

    Binh T. Pham; Jeffrey J. Einerson

    2013-09-01

    This report documents the data qualification status of AGR-2 fuel irradiation experimental data from Advanced Test Reactor (ATR) Cycles 152A, 152B, 153A, 153B, and 154A, as recorded in the Nuclear Data Management and Analysis System (NDMAS). The AGR-2 data streams addressed include thermocouple (TC) temperatures, sweep gas data (flow rate, pressure, and moisture content), and fission product monitoring system (FPMS) data for each of the six capsules in the experiment. A total of 13,400,520 every minute instantaneous TC and sweep gas data records were received and processed by NDMAS for this period. Of these data, 8,911,791 records (66.5% of the total) were determined to be Qualified based on NDMAS accuracy testing and data validity assessment. For temperature, there were 4,266,081 records (74% of the total TC data) that were Failed due to TC instrument failures. For sweep gas flows, there were 222,648 gas flow records (2.91% of the flow data) that were Failed. The inlet gas flow failures due to gas flow cross-talk and leakage problems that occurred after Cycle 150A were corrected by using the same gas mixture in all six capsules and the Leadout. For FPMS data, NDMAS received and processed preliminary release rate and release-to-birth rate ratio (R/B) data for three reactor cycles (Cycles 149B, 150B, and 151A) . This data consists of 45,983 release rate records and 45,235 R/B records for the 12 radionuclides reported. The qualification status of these FPMS data has been set to In Process until receipt of Quality Assurance-approved data generator reports. All of the above data have been processed and tested using a SAS®-based enterprise application software system, stored in a secure Structured Query Language database, made available on the NDMAS Web portal (http://ndmas.inl.gov), and approved by the INL STIM for release to both internal and appropriate external Very High Temperature Reactor Program participants.

  11. Test and Coupling Calculation of Temperature Field for UHV Dry-Type Air-Core Smoothing Reactor%特高压干式空心平波电抗器温度场耦合计算与试验

    Institute of Scientific and Technical Information of China (English)

    姜志鹏; 文习山; 王羽; 陈瑞珍; 曹继丰; 陈图腾

    2015-01-01

    为了研究特高压干式空心平波电抗器的温升分布特性,该文基于计算流体力学和传热学理论,建立了电抗器稳态流体与固体耦合温度场的数学计算模型.采用有限容积法对三维模型进行稳态流体场与温度场直接求解,获得其温度场分布特性,研究了包封轴向及径向温度分布规律.最后采用光纤测温法对自然对流下的电抗器进行温升测量.对比分析表明,计算与试验结果吻合较好,验证温度场数值计算的合理性和准确性,为特高压干式空心平波电抗器温升监测提供参考.%To research the distribution characteristics of temperature rise for UHV dry-type air-core smoothing reactor, according to computational fluid dynamics and heat transfer theory, this paper presented the mathematical model of temperature field coupling steady fluid and solid for the reactor. The finite volume method was employed to solve the steady flow and temperature fields of 3D model directly, and the temperature distribution characteristics of the reactor were obtained. Then the axial and radial temperature distributions of encapsulations were studied separately. Finally, optical fiber temperature measurement method was used to test temperature rise for the reactor under natural convection condition. Comparative analysis shows that the calculated results are in good agreement with the experiment, which verifies the rationality and accuracy of the temperature field numerical calculation. And it can provide references for the temperature rise monitoring of UHV dry-type air-core smoothing reactor.

  12. AGR3 in breast cancer: prognostic impact and suitable serum-based biomarker for early cancer detection.

    Science.gov (United States)

    Garczyk, Stefan; von Stillfried, Saskia; Antonopoulos, Wiebke; Hartmann, Arndt; Schrauder, Michael G; Fasching, Peter A; Anzeneder, Tobias; Tannapfel, Andrea; Ergönenc, Yavuz; Knüchel, Ruth; Rose, Michael; Dahl, Edgar

    2015-01-01

    Blood-based early detection of breast cancer has recently gained novel momentum, as liquid biopsy diagnostics is a fast emerging field. In this study, we aimed to identify secreted proteins which are up-regulated both in tumour tissue and serum samples of breast cancer patients compared to normal tissue and sera. Based on two independent tissue cohorts (n = 75 and n = 229) and one serum cohort (n = 80) of human breast cancer and healthy serum samples, we characterised AGR3 as a novel potential biomarker both for breast cancer prognosis and early breast cancer detection from blood. AGR3 expression in breast tumours is significantly associated with oestrogen receptor α (Pbreast tumours (multivariate hazard ratio: 2.186, 95% CI: 1.008-4.740, Pbreast cancer patients (n = 40, mainly low stage tumours) compared to healthy controls (n = 40). To develop a suitable biomarker panel for early breast cancer detection, we measured AGR2 protein in human serum samples in parallel. The combined AGR3/AGR2 biomarker panel achieved a sensitivity of 64.5% and a specificity of 89.5% as shown by receiver operating characteristic (ROC) curve statistics. Thus our data clearly show the potential usability of AGR3 and AGR2 as biomarkers for blood-based early detection of human breast cancer.

  13. NASA Reactor Facility Hazards Summary. Volume 1

    Science.gov (United States)

    1959-01-01

    The Lewis Research Center of the National Aeronautics and Space Administration proposes to build a nuclear research reactor which will be located in the Plum Brook Ordnance Works near Sandusky, Ohio. The purpose of this report is to inform the Advisory Committee on Reactor Safeguards of the U. S. Atomic Energy Commission in regard to the design Lq of the reactor facility, the characteristics of the site, and the hazards of operation at this location. The purpose of this research reactor is to make pumped loop studies of aircraft reactor fuel elements and other reactor components, radiation effects studies on aircraft reactor materials and equipment, shielding studies, and nuclear and solid state physics experiments. The reactor is light water cooled and moderated of the MTR-type with a primary beryllium reflector and a secondary water reflector. The core initially will be a 3 by 9 array of MTR-type fuel elements and is designed for operation up to a power of 60 megawatts. The reactor facility is described in general terms. This is followed by a discussion of the nuclear characteristics and performance of the reactor. Then details of the reactor control system are discussed. A summary of the site characteristics is then presented followed by a discussion of the larger type of experiments which may eventually be operated in this facility. The considerations for normal operation are concluded with a proposed method of handling fuel elements and radioactive wastes. The potential hazards involved with failures or malfunctions of this facility are considered in some detail. These are examined first from the standpoint of preventing them or minimizing their effects and second from the standpoint of what effect they might have on the reactor facility staff and the surrounding population. The most essential feature of the design for location at the proposed site is containment of the maximum credible accident.

  14. Experience feedback examination in PWR type reactors operating for the 1997-1999 period; Examen du retour d'experience en exploitation des reacteurs a eau sous pression pour la periode 1997-1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The present report is relative to the examination that the permanent group has made on the experience feedback in operation for PWR type reactors for the period 1997-1999 that was on eleven themes chosen by the Nuclear Safety and Radiation Protection Authority. It used analysis reports made by I.R.S.N. in support of four meetings of the permanent group devoted to this examination from April 2001 to June 2002. The different themes were operating uncertainties, machining to vibrations, analysis of incidents and gaseous releases, circuits, human factors, behaviour of electric batteries, risk of cold source loss. (N.C.)

  15. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    Energy Technology Data Exchange (ETDEWEB)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  16. 不同挡流板形式紫外线消毒仪杀菌效果模拟%Numerical simulation of sterilizing efficiency of ultraviolet disinfection reactors with different flow baffle types

    Institute of Scientific and Technical Information of China (English)

    牛培平; 丁日升; 宋卫堂; 王媛

    2015-01-01

    Nutrient solution recycling has become one of the essential techniques of soilless cultivation. But nutrient solutions are susceptible to be polluted by infectious diseases during the recycling process, so it is needed to disinfect nutrient solutions before recycling use. Compared to other disinfection methods of common nutrient solutions, ultraviolet (UV) disinfection has many advantages, such as high efficiency, low cost, not changing the physical and chemical properties of nutrient solutions, so UV disinfection is an environment-friendly technology of nutrient solution treatment.In general, experimental research and numerical simulation are the most common methods of UV disinfection. Although experimental results of the performance of UV disinfection reactor are credible, direct measurement is difficult and costly and thus seldom done. On the other hand, one can use numerical simulation techniques to model the UV disinfection. Computational fluid dynamics (CFD) has been widely used for simulating the UV disinfection. In previous studies, some researchers analyzed the performance of small-scale horizontal UV disinfection reactors with different flow baffle numbers and flow areas. They designed an alternately arranged flat-type flow baffle and analyzed the performance of the UV disinfection device for nutrient solutions using the CFD simulations and the measurements of biological bacterial disinfection. However, the effects of different flow baffle types on the performance of the UV disinfection reactor have seldom been investigated. In this study, we designed 5 different flow baffle types of the UV disinfection reactor. The main objectives were to increase disinfection efficiency by optimizing the flow baffle arrangement and to test the performance of the UV disinfection reactor with different flow baffle types. The 5 different flow baffle types were proposed: 2 circular channels and annulus alternation, 2 circular channels, 4 circular channels and annulus

  17. Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor; Aplicacion del algoritmo MOVE para la identificacion de modelos de orden reducido del nucleo de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Victoria R, M.A.; Morales S, J.B. [UNAM, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: angelvr@gmail.com

    2005-07-01

    Presently work is applied the modified algorithm of the ellipsoid of optimal volume (MOVE) to a reduced order model of 5 differential equations of the core of a boiling water reactor (BWR) with the purpose of estimating the parameters that model the dynamics. The viability is analyzed of carrying out an analysis that calculates the global dynamic parameters that determine the stability of the system and the uncertainty of the estimate. The modified algorithm of the ellipsoid of optimal volume (MOVE), is a method applied to the parametric identification of systems, in particular to the estimate of groups of parameters (PSE for their initials in English). It is looked for to obtain the ellipsoid of smaller volume that guarantees to contain the real value of the parameters of the model. The PSE MOVE is a recursive identification method that can manage the sign of noise and to ponder it, the ellipsoid represents an advantage due to its easy mathematical handling in the computer, the results that surrender are very useful for the design of Robust Control since to smaller volume of the ellipsoid, better is in general the performance of the system to control. The comparison with other methods presented in the literature to estimate the reason of decline (DR) of a BWR is presented. (Author)

  18. BOILER-SUPERHEATED REACTOR

    Science.gov (United States)

    Heckman, T.P.

    1961-05-01

    A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

  19. Reactor and method of operation

    Science.gov (United States)

    Wheeler, John A.

    1976-08-10

    A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

  20. Molecular ecology of anaerobic reactor systems

    DEFF Research Database (Denmark)

    Hofman-Bang, H. Jacob Peider; Zheng, D.; Westermann, Peter;

    2003-01-01

    Anaerobic reactor systems are essential for the treatment of solid and liquid wastes and constitute a core facility in many waste treatment plants. Although much is known about the basic metabolism in different types of anaerobic reactors, little is known about the microbes responsible...... to the abundance of each microbe in anaerobic reactor systems by rRNA probing. This chapter focuses on various molecular techniques employed and problems encountered when elucidating the microbial ecology of anaerobic reactor systems. Methods such as quantitative dot blot/fluorescence in-situ probing using various...

  1. Microchannel Reactors for ISRU Applications

    Science.gov (United States)

    Carranza, Susana; Makel, Darby B.; Blizman, Brandon; Ward, Benjamin J.

    2005-02-01

    Affordable planning and execution of prolonged manned space missions depend upon the utilization of local resources and the waste products which are formed in manned spacecraft and surface bases. Successful in-situ resources utilization (ISRU) will require component technologies which provide optimal size, weight, volume, and power efficiency. Microchannel reactors enable the efficient chemical processing of in situ resources. The reactors can be designed for the processes that generate the most benefit for each mission. For instance, propellants (methane) can be produced from carbon dioxide from the Mars atmosphere using the Sabatier reaction and ethylene can be produced from the partial oxidation of methane. A system that synthesizes ethylene could be the precursor for systems to synthesize ethanol and polyethylene. Ethanol can be used as a nutrient for Astrobiology experiments, as well as the production of nutrients for human crew (e.g. sugars). Polyethylene can be used in the construction of habitats, tools, and replacement parts. This paper will present recent developments in miniature chemical reactors using advanced Micro Electro Mechanical Systems (MEMS) and microchannel technology to support ISRU of Mars and lunar missions. Among other applications, the technology has been demonstrated for the Sabatier process and for the partial oxidation of methane. Microchannel reactors were developed based on ceramic substrates as well as metal substrates. In both types of reactors, multiple layers coated with catalytic material are bonded, forming a monolithic structure. Such reactors are readily scalable with the incorporation of extra layers. In addition, this reactor structure minimizes pressure drop and catalyst settling, which are common problems in conventional packed bed reactors.

  2. Decree n. 2007-534 of the 4. april 2007 allowing the creation of the base nuclear installation named Flamanville 3, including a EPR type reactor, on the site of Flamanville (Manche); Decret no 2007-534 du 10 avril 2007 autorisant la creation de l'installation nucleaire de base denommee Flamanville 3, comportant un reacteur nucleaire de type EPR, sur le site de Flamanville (Manche)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-04-15

    This decree gives the authorization to EDF to create on the site of Flamanville a nuclear installation including a PWR type reactor for a power of 4500 MW and devoted to the electric production. This reactor will can use uranium oxide or a mixture of uranium oxide and plutonium oxide. Considerations concerning the safety are given, as well as the control of the impact of this exploitation on the populations and the environment. (N.C.)

  3. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  4. State space modeling of reactor core in a pressurized water reactor

    Science.gov (United States)

    Ashaari, A.; Ahmad, T.; Shamsuddin, Mustaffa; M, Wan Munirah W.; Abdullah, M. Adib

    2014-07-01

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  5. Mecanização Agrícola. Bases de óleo-dinâmica. Transmissões hidráulicas em máquinas agrícolas

    OpenAIRE

    Peça, José

    2013-01-01

    Este trabalho destina-se a apoiar a aprendizagem de estudantes do ramo das ciências agrárias sobre aspectos relevantes das transmissões hidráulicas em máquinas agrícolas. A transmissão hidráulica efectua a transmissão de potência desde uma fonte (exemplo: o Serviço Externo do Sistema Hidráulico do tractor) para diversos actuadores hidráulicos nos equipamentos agrícolas. São apresentadas as transmissões hidráulicas de equipamentos agrícolas, com a preocupação focada na identificação dos co...

  6. Procedure of recovery of pin-by-pin fields of energy release in the core of VVER-type reactor for the BIPR-8 code

    Science.gov (United States)

    Gordienko, P. V.; Kotsarev, A. V.; Lizorkin, M. P.

    2014-12-01

    The procedure of recovery of pin-by-pin energy-release fields for the BIPR-8 code and the algorithm of the BIPR-8 code which is used in nodal computation of the reactor core and on which the recovery of pin-by-pin fields of energy release is based are briefly described. The description and results of the verification using the module of recovery of pin-by-pin energy-release fields and the TVS-M program are given.

  7. Some economic aspects of natural uranium graphite gas reactor types. Present status and trends of costs in France; Quelques aspects economiques de la filiere uranium naturel - Graphite - gaz. Etat actuel et tendance des couts en France

    Energy Technology Data Exchange (ETDEWEB)

    Gaussens, J.; Tanguy, P. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Leo, B. [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    The first part of this report defines the economic advantages of natural uranium fuels, which are as follows: the restricted number and relatively simple fabrication processes of the fuel elements, the low cost per kWh of the finished product and the reasonable capital investments involved in this type of fuel cycle as compared to that of enriched uranium. All these factors combine to reduce the arbitrary nature of cost estimates, which is particularly marked in the case of enriched uranium due to the complexity of its cycle and the uncertainties of plutonium prices). Finally, the wide availability of yellowcake, as opposed to the present day virtual monopoly of isotope separation, and the low cost of natural uranium stockpiling, offer appreciable guarantees in the way of security of supply and economic and political independence as compared with the use of enriched uranium. As far as overall capital investments are concerned, it is shown that, although graphite-gas reactor costs are higher than those of light water reactors in certain capacity ranges, the situation becomes far less clear when we start taking into account, in the interest of national independence, the cost of nuclear fuel production equipment in the case of each of these types of reactor. Finally, the marginal cost of the power capacity of a graphite-gas reactor is low and its technological limitations have receded (owing particularly to the use of prestressed concrete). It is a well known fact that the trend is now towards larger power station units, which means that the rentability of natural uranium graphite reactors as compared to other types of reactors will become more and more pronounced. The second section aims at presenting a realistic short and medium term view of the fuel, running, and investment costs of French natural uranium graphite gas, reactors. Finally, the economic goals which this type of reactor can reach in the very near future are given. It is thus shown that considerable

  8. Power Control Method for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baang, Dane; Suh, Yongsuk; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Considering safety-oriented design concept and other control environment, we developed a simple controller that provides limiting function of power change- rate as well as fine tracking performance. The design result has been well-proven via simulation and actual application to a TRIGA-II type research reactor. The proposed controller is designed to track the PDM(Power Demand) from operator input as long as maintaining the power change rate lower than a certain value for stable reactor operation. A power control method for a TRIGA-II type research reactor has been designed, simulated, and applied to actual reactor. The control performance during commissioning test shows that the proposed controller provides fine control performance for various changes in reference values (PDM), even though there is large measurement noise from neutron detectors. The overshoot at low power level is acceptable in a sense of reactor operation.

  9. A “questão agrária” em Mariátegui

    Directory of Open Access Journals (Sweden)

    Frederico Daia Firmiano

    2012-05-01

    Full Text Available Neste breve artigo temos como objetivo refletir sobre a “questão agrária” no pensamento de José Carlos Mariátegui, sobretudo a partir de sua principal obra, os “Sete ensaios de interpretação da realidade peruana”. Nosso argumento é que o socialista peruano não tratou os problemas da formação social peruana a partir da chave teórica dada pela “questão agrária”, mas a encontrou no curso de uma interpretação marxista, profundamente original e inventiva, sobre esta formação social (desenvolvida ao lado da intensa militância política, identificando o problema indígena ao problema da terra, aos que Amauta conferiu uma alternativa socialista.

  10. Desarrollo y perspectiva de la ingeniería agrícola

    OpenAIRE

    Ospina Mejía, Julio E.; Carlos A. González M.

    2011-01-01

    La Ingeniería Agrícola es la rama de la ingeniería orientada a dar respuesta a los problemas que plantea la agricultura moderna, mediante la aplicación de principios de las ciencias naturales y las matemáticas, procurando la conservación y el aumento del potencial del suelo, la producción de alimentos de calidad, el mantenimiento del medio ambiente y la dignificación del hombre rural. El objetivo de la ingeniería agrícola es la generación y aplicación de conocimientos científicos y tecnológic...

  11. Desempleo entre los jornaleros agrícolas, un fenómeno emergente

    Directory of Open Access Journals (Sweden)

    Antonieta Barrón

    2013-01-01

    Full Text Available Este artículo tiene como propósito presentar un fenómeno no planteado en el estudio de mercados de trabajo agrícolas y empleo rural, se cuestiona la definición de desem- pleo, y se plantea cómo esta definición, asumida por el inegi, se aplica en los mercados de trabajo formales pero no en los no formales, los agrícolas. Se hacen comparaciones entre los niveles nacional y rural para concluir con los estudios de caso que refuerzan la tesis inicial, el aumento del desempleo en los mercados de trabajo rurales. A lo largo del trabajo se resaltan las diferencias por sexo y se conlcuye que el desempleo rural afecta más a las mujeres que los hombres, fenómeno contrario al de nivel nacional.

  12. AN EVENT DRIVEN SIMULATION FOR ADAPTIVE GENTLE RANDOM EARLY DETECTION (AGRED ALGORITHM

    Directory of Open Access Journals (Sweden)

    Omid Seifaddini

    2014-01-01

    Full Text Available Simulations are used to find optimum answers for problems in wide areas. Active queue management algorithms such as RED, GRED, typically use simulators like ns2 which is an open source simulator or OPNET, OMNET which are commercial simulators. However, beside the benefits of using simulators like having defined modules, parameters. There are problems such as complexity, large integrated components and licensing cost. To have an ideal balance in mentioned benefits and problems and to further complement the repository of simulators, this study presents the description of a general-purpose programming language based discrete event simulation for active queue management. This research has focused at developing a discrete event simulator to implement one of active queue management algorithms which is called AGRED. The results showed that the developed simulator has successfully produced the same results with an average deviation of 1.5% as previous simulator in AGRED.

  13. La Ingeniería Agrícola en el País

    Directory of Open Access Journals (Sweden)

    Bustamante M. Hellodoro

    1985-12-01

    Full Text Available Es sumamente difícil situar en la historia, el origen de la Ingeniería Agrícola, aunque podría remontarse al propio origen de la tierra, de las plantas y de las aguas. La Ingeniería Agrícola está ocupando un campo de acción que se ha desarrollado a pesar de las limitaciones que le ofrece el mismo desarrollo del país y al escepticismo sin fundamento de algunos pregoneros del pesimismo nacional; le corresponde enfrentar además hoy al cáncer del desempleo y a la falta de oportunidad para demostrar al país que la inversión hecha por el estado en la formación de estos profesionales es productiva.

  14. Sector agrícola y política de competencia

    Directory of Open Access Journals (Sweden)

    Argüello C., Ricardo

    2006-12-01

    Full Text Available Los procesos de integración económica indican que la separación entre políticas comercial y de competencia no sólo es ficticia sino que contradice la realidad de la economía internacional. Este ensayo se discute la relación entre el sector agrícola y la política de competencia en un contexto de liberalización comercial. Sostiene que es necesario aplicar la política de competencia en este sector para incentivar los procesos de cambio tecnológico que requiere el desarrollo agrícola. Y muestra que es indispensable una aplicación rigurosa de la política de competencia que tenga en cuenta las particularidades de la estructura de mercado

  15. Transmissões presentes em tratores agrícolas no Brasil Transmissions present in agricultural tractors in Brazil

    Directory of Open Access Journals (Sweden)

    Rodrigo Lampert Ribas

    2010-10-01

    Full Text Available O sistema de transmissão de um trator agrícola tem por principal finalidade transmitir a potência gerada no motor ao sistema hidráulico, à tomada de potência e às rodas motrizes. A caixa de câmbio é formada por uma série de engrenagens que permitem a correta seleção da velocidade e do torque mais adequados a uma determinada operação. Assim, as transmissões podem ser classificadas, basicamente, em três categorias: mecânicas, hidrostáticas e hidrodinâmicas, sendo as transmissões mecânicas subdivididas em engrenagens deslizantes e sincronizadas. Este trabalho objetivou analisar os diferentes tipos de transmissões disponibilizados em 169 modelos de tratores fabricados no Brasil, com a finalidade de avaliar a real distribuição nas diferentes faixas de potência e estabelecer possíveis relações entre esses parâmetros. Como resultado, evidencia-se uma forte tendência de tratores agrícolas com potência menor que 111,8kW possuírem transmissão do tipo mecânica, e tratores acima de 111,8kW, do tipo hidrostática.The transmission system of an agricultural tractor has a main purpose to transmit the power generated in the hydraulic motor, the power take off and the drive wheels. The gearshift consists of a series of gears that allows the correct selection of speed and torque more appropriate to a particular operation. Then, the transmissions can be classified, basically, in three categories: mechanical, hydrostatic and hydrodynamic, being the mechanical transmissions subdivided in sliding gear and synchronized. Then, the objective of this research was to analyze the different types of transmissions available in 169 models of tractors made in Brazil in order to evaluate the real distribution of them in different power ranges and establish possible correlations. As a result, it was shown a strong tendency for agricultural tractors with power less than 111.8kW possessing mechanical transmission and tractors above 111.8k

  16. Redescubriendo El suelo: Su importancia ecológica y agrícola

    Directory of Open Access Journals (Sweden)

    Noé Manuel Montaño

    2016-10-01

    Full Text Available Redescubriendo el suelo: su importancia ecológica y agrícola es un libro muy completo, actualizado y sin duda será de consulta obligada para quienes hacen investigación edafológica y como material elemental para los cursos de pre- y posgrado de suelo y materias afines, tanto en México como en América Latina.

  17. A cidade, um foco de diversidade agrícola no Rio Negro (Amazonas, Brasil?

    Directory of Open Access Journals (Sweden)

    Laure Emperaire

    Full Text Available As regiões do Médio e Alto Rio Negro, apesar de serem isoladas da malha de comunicação terrestre e das frentes de colonização, caracterizam-se por uma articulação cada vez mais forte entre a área florestal, a das comunidades, e a urbana, ou seja, as pequenas cidades ribeirinhas. Esta complementaridade, temporária ou definitiva, se traduz por uma expansão da agricultura periurbana. Propomos uma abordagem comparativa da diversidade agrícola entre o urbano e o florestal. Analisamos as relações entre formas de manejo dos espaços cultivados (superfícies, ciclo de uso e práticas, plantas cultivadas e redes sociais envolvidas no acesso aos recursos fitogenéticos. A análise mostra uma recomposição dos sistemas agrícolas com a permanência de uma alta diversidade agrícola, porém marcada por uma maior vulnerabilidade do sistema em decorrência da diminuição do tempo de pousio e da diminuição da força de trabalho disponível. No contexto urbano, as estratégias tradicionais de manejo dos recursos agrícolas se combinam a um outro objetivo, o do acesso à terra. A análise aponta para a necessidade de uma reflexão sistémica sobre as possíveis formas de conservação deste patrimônio biocultural.

  18. Desarrollo y perspectiva de la Ingeniería Agrícola

    Directory of Open Access Journals (Sweden)

    Julio E. Ospina Mejía

    2011-02-01

    Full Text Available La Ingeniería Agrícola es la rama de la ingeniería orientada a dar respuesta a los problemas que plantea la agricultura moderna, mediante la aplicación de principios de las ciencias naturales y las matemáticas, procurando la conservación y el aumento del potencial del suelo, la producción de alimentos de calidad, el mantenimiento del medio ambiente y la dignificación del hombre rural. El objetivo de la ingeniería agrícola es la generación y aplicación de conocimientos científicos y tecnológicos de la ingeniería a la producción agropecuaria y otros biosistemas orientados a las áreas de adecuación de tierras, maquinaria agrícola, manejo de recursos hídricos, construcciones rurales, manejo poscosecha y procesamiento de producto agropecuarios, con criterios de sostenibilidad y competitividad.

  19. 螺旋升流塔式光催化反应器的设计研究%Design of Spiral Up-flow Tower-type Photocatalysis Reactor

    Institute of Scientific and Technical Information of China (English)

    徐璇; 吉芳英; 范子红

    2009-01-01

    Based on the cyclone separator model for granular pollutants, a spiral up-flow tower-type photocatalysis reactor was designed to increase the recovery rate of photocatalyst. In order to reach a high A/V ratio (the ratio of illumination area to reaction solution volume), a tower structure was used on this reactor, and the A/V ratio can reach 12.95 in this experimental condition. This reactor was used to treat the nitrobenzene simulated wastewater. When the initial concentration of nitrobenzene is 466 mg/L, the removal rate of nitrobenzene is about 60%, and the photocatalyst recovery rate reaches 92.80% after 12h.%采用颗粒污染物的旋流分离模型设计了螺旋升流塔式光催化反应器,能够在悬浮态光催化反应系统中提高光催化剂的回收率.反应器采用塔式结构布置,能有效提高反应器的光照面积与反应液体积之比(A/V值),在本试验条件下A/V值可达到12.95.采用该反应器处理硝基苯模拟废水,当硝基苯初始浓度为466 mg/L时,反应器对硝基苯的去除率稳定在60%左右,运行12 h后对光催化剂的回收率为92.80%.

  20. Study on an Automatic Knife Switch of New Type High-Voltage Small Current Reactor%新型高电压小电流电抗器自动刀闸的研究

    Institute of Scientific and Technical Information of China (English)

    许杰; 李世武; 孙伟

    2014-01-01

    针对当前电抗器投切刀闸存在的投切行程小、适用范围受限制、远程操控不方便等问题,研发了一种新型的高电压小电流电抗器自动刀闸,给出了自动刀闸结构图,分析了其工作原理。对该自动刀闸进行峰值耐受电流试验和机械操作试验,测试结果符合要求,有效通断率100%,提高了电抗器串并联自动化程度,易于远程控制且性能稳定,可广泛用于电力系统机构的各项试验。%Aiming at several problems such that the existing reactor knife switch is small in travel distance for knife making and breaking, limited in applicable range, not convenient in remote operating etc., this paper developed an automatic knife switch of new type high-voltage small current reactor and gave the structural diagram of the automatic knife switch, analyzing its working principle. The peak value withstand current and mechanical operating tests were carried out for the automatic knife switch. The test results are in conformity to the requirements, with 100% valid making-breaking rate, which raises the reactor serial and parallel connection automation degree, easy for remote control and stable in performance. The switch could be widely used in each test of electric power system organization.

  1. Sandia National Laboratories Medical Isotope Reactor concept.

    Energy Technology Data Exchange (ETDEWEB)

    Coats, Richard Lee; Dahl, James J.; Parma, Edward J., Jr.

    2010-04-01

    This report describes the Sandia National Laboratories Medical Isotope Reactor and hot cell facility concepts. The reactor proposed is designed to be capable of producing 100% of the U.S. demand for the medical isotope {sup 99}Mo. The concept is novel in that the fuel for the reactor and the targets for the {sup 99}Mo production are the same. There is no driver core required. The fuel pins that are in the reactor core are processed on a 7 to 21 day irradiation cycle. The fuel is low enriched uranium oxide enriched to less than 20% {sup 235}U. The fuel pins are approximately 1 cm in diameter and 30 to 40 cm in height, clad with Zircaloy (zirconium alloy). Approximately 90 to 150 fuel pins are arranged in the core in a water pool {approx}30 ft deep. The reactor power level is 1 to 2 MW. The reactor concept is a simple design that is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days. The fuel fabrication, reactor design and operation, and {sup 99}Mo production processing use well-developed technologies that minimize the technological and licensing risks. There are no impediments that prevent this type of reactor, along with its collocated hot cell facility, from being designed, fabricated, and licensed today.

  2. Neutron imaging on the VR-1 reactor

    Science.gov (United States)

    Crha, J.; Sklenka, L.; Soltes, J.

    2016-09-01

    Training reactor VR-1 is a low power research reactor with maximal thermal power of 1 kW. The reactor is operated by the Faculty of Nuclear Science and Physical Engineering of the Czech Technical University in Prague. Due to its low power it suits as a tool for education of university students and training of professionals. In 2015, as part of student research project, neutron imaging was introduced as another type of reactor utilization. The low available neutron flux and the limiting spatial and construction capabilities of the reactor's radial channel led to the development of a special filter/collimator insertion inside the channel and choosing a nonstandard approach by placing a neutron imaging plate inside the channel. The paper describes preliminary experiments carried out on the VR-1 reactor which led to first radiographic images. It seems, that due to the reactor construction and low reactor power, the neutron imaging technique on the VR-1 reactor is feasible mainly for demonstration or educational and training purposes.

  3. Human oral isolate Lactobacillus fermentum AGR1487 reduces intestinal barrier integrity by increasing the turnover of microtubules in Caco-2 cells.

    Directory of Open Access Journals (Sweden)

    Rachel C Anderson

    Full Text Available Lactobacillus fermentum is found in fermented foods and thought to be harmless. In vivo and clinical studies indicate that some L. fermentum strains have beneficial properties, particularly for gastrointestinal health. However, L. fermentum AGR1487 decreases trans-epithelial electrical resistance (TEER, a measure of intestinal barrier integrity. The hypothesis was that L. fermentum AGR1487 decreases the expression of intestinal cell tight junction genes and proteins, thereby reducing barrier integrity. Transcriptomic and proteomic analyses of Caco-2 cells (model of human intestinal epithelial cells treated with L. fermentum AGR1487 were used to obtain a global view of the effect of the bacterium on intestinal epithelial cells. Specific functional characteristics by which L. fermentum AGR1487 reduces intestinal barrier integrity were examined using confocal microscopy, cell cycle progression and adherence bioassays. The effects of TEER-enhancing L. fermentum AGR1485 were investigated for comparison. L. fermentum AGR1487 did not alter the expression of Caco-2 cell tight junction genes (compared to L. fermentum AGR1485 and tight junction proteins were not able to be detected. However, L. fermentum AGR1487 increased the expression levels of seven tubulin genes and the abundance of three microtubule-associated proteins, which have been linked to tight junction disassembly. Additionally, Caco-2 cells treated with L. fermentum AGR1487 did not have defined and uniform borders of zona occludens 2 around each cell, unlike control or AGR1485 treated cells. L. fermentum AGR1487 cells were required for the negative effect on barrier integrity (bacterial supernatant did not cause a decrease in TEER, suggesting that a physical interaction may be necessary. Increased adherence of L. fermentum AGR1487 to Caco-2 cells (compared to L. fermentum AGR1485 was likely to facilitate this cell-to-cell interaction. These findings illustrate that bacterial strains of the

  4. Estimation of damage by inmates of a PWR Reactor neutron irradiation; Medida de flujo adjunto en un reactor experimental

    Energy Technology Data Exchange (ETDEWEB)

    Blazquez, J.

    2013-07-01

    Flow measurement deputy in an experimental reactor This work focuses on the flow measurement attached with reactor subcritical, to be applied in fast, reactor type ADS (Accelerator Driven System). The role of the attached flow in perturbation theory of reactivity, as the theoretical basis for the design of the measurement technique is briefly reviewed. Used measures from the experimental fast reactor currently dismantled CORAL-I.

  5. Assessment of subcriticality during PWR-type reactor refueling; Evaluation de la sous-criticite lors des operations de chargement d'un reacteur nucleaire REP

    Energy Technology Data Exchange (ETDEWEB)

    Verdier, A

    2005-04-15

    During the core loading period of a PWR, any fuel assembly misplacements may significantly reduce the existing criticality margin. The Dampierre 4-18 event showed the present monitoring based on the variations of the outside-core detector counting rate cannot detect such misplacements. In order to circumvent that, a more detailed analysis of the available signal was done. We particularly focused on the neutronic noise analysis methods such as MSM (modified source multiplication), MSA (amplified source multiplication), Rossi-{alpha} and Feynman-{alpha} methods. The experimental part of our work was dedicated to the application of those methods to a research reactor. Finally, our results showed that those methods cannot be used with the present PWR instrumentation. Various detector positions were then studied using Monte Carlo calculations capable of following the neutron origin. Our results showed that the present technology does not allow us to use any solution based on neutron detection for monitoring core loading. (author)

  6. 新型罐式环氧丙烷反应器的结构设计优化%New tank type epoxy propane reactor structure design optimization

    Institute of Scientific and Technical Information of China (English)

    王海勇

    2015-01-01

    Propylene oxide is very important in the production of chemical industry production equipment.In the production process,using the new tank reactor,improve the response speed,reduces the consumption of chlorine gas,propylene.The design in propylene oxide production equipment,is the first.%环氧丙烷是化工生产中非常重要的生产设备。在环氧丙烷的生产过程中,采用新型罐式反应器,提高了反应速度,降低了氯气、丙烯的消耗量。该设计在环氧丙烷的生产设备中,属于首创。

  7. Status of the Combined Third and Fourth NGNP Fuel Irradiations In the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti; Michael E. Davenport

    2013-07-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in September 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. Since the purpose of this combined experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  8. Evaluation of integrated anaerobic/aerobic fixed-bed sequencing batch biofilm reactor for decolorization and biodegradation of azo dye acid red 18: comparison of using two types of packing media.

    Science.gov (United States)

    Hosseini Koupaie, E; Alavi Moghaddam, M R; Hashemi, S H

    2013-01-01

    Two integrated anaerobic/aerobic fixed-bed sequencing batch biofilm reactor (FB-SBBR) were operated to evaluate decolorization and biodegradation of azo dye Acid Red 18 (AR18). Volcanic pumice stones and a type of plastic media made of polyethylene were used as packing media in FB-SBBR1 and FB-SBBR2, respectively. Decolorization of AR18 in both reactors followed first-order kinetic with respect to dye concentration. More than 63.7% and 71.3% of anaerobically formed 1-naphthylamine-4-sulfonate (1N-4S), as one of the main sulfonated aromatic constituents of AR18 was removed during the aerobic reaction phase in FB-SBBR1 and FB-SBBR2, respectively. Based on statistical analysis, performance of FB-SBBR2 in terms of COD removal as well as biodegradation of 1N-4S was significantly higher than that of FB-SBBR1. Spherical and rod shaped bacteria were the dominant species of bacteria in the biofilm grown on the pumice stones surfaces, while, the biofilm grown on surfaces of the polyethylene media had a fluffy structure.

  9. Development of technology for next generation reactor - Development of next generation reactor in Korea -

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)] [and others

    1993-09-01

    The project, development of next generation reactor, aims overall related technology development and obtainment of related license in 2001. The development direction is to determine the reactor type and to build up the design concept in 1994. For development trend analysis of foreign next generation reactor, level-1 PSA, fuel cycle analysis and computer code development are performed on System 80+ and AP 600. Especially for design characteristics analysis and volume upgrade of AP 600, nuclear fuel and reactor core design analysis, coolant circuit design analysis, mechanical structure design analysis and safety analysis etc. are performed. (Author).

  10. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

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

  11. Reactivity control assembly for nuclear reactor. [LMFBR

    Science.gov (United States)

    Bollinger, L.R.

    1982-03-17

    This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

  12. Measurement of the Residual Stresses and Investigation of Their Effects on a Hardfaced Grid Plate due to Thermal Cycling in a Pool Type Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    S. Balaguru

    2016-01-01

    Full Text Available In sodium-cooled fast reactors (SFR, grid plate is a critical component which is made of 316 L(N SS. It is supported on core support structure. The grid plate supports the core subassemblies and maintains their verticality. Most of the components of SFR are made of 316 L(N/304 L(N SS and they are in contact with the liquid-metal sodium which acts as a coolant. The peak operating temperature in SFR is 550°C. However, the self-welding starts at 500°C. To avoid self-welding and galling, hardfacing of the grid plate has become necessary. Nickel based cobalt-free colmonoy 5 has been identified as the hardfacing material due to its lower dose rate by Plasma Transferred Arc Welding (PTAW. This paper is concerned with the measurement and investigations of the effects of the residual stress generated due to thermal cycling on a scale-down physical model of the grid plate. Finite element analysis of the hardfaced grid plate model is performed for obtaining residual stresses using elastoplastic analysis and hence the results are validated. The effects of the residual stresses due to thermal cycling on the hardfaced grid plate model are studied.

  13. Simulation of decay heat removal by natural convection in a pool type fast reactor model-ramona-with coupled 1D/2D thermal hydraulic code system

    Energy Technology Data Exchange (ETDEWEB)

    Kasinathan, N.; Rajakumar, A.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

    Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.

  14. Thermal-hydraulic behavior of physical quantities at critical velocities in a nuclear research reactor core channel using plate type fuel

    Directory of Open Access Journals (Sweden)

    Sidi Ali Kamel

    2012-01-01

    Full Text Available The thermal-hydraulic study presented here relates to a channel of a nuclear reactor core. This channel is defined as being the space between two fuel plates where a coolant fluid flows. The flow velocity of this coolant should not generate vibrations in fuel plates. The aim of this study is to know the distribution of the temperature in the fuel plates, in the cladding and in the coolant fluid at the critical velocities of Miller, of Wambsganss, and of Cekirge and Ural. The velocity expressions given by these authors are function of the geometry of the fuel plate, the mechanical characteristics of the fuel plate’s material and the thermal characteristics of the coolant fluid. The thermal-hydraulic study is made under steady-state; the equation set-up of the thermal problem is made according to El Wakil and to Delhaye. Once the equation set-up is validated, the three critical velocities are calculated and then used in the calculations of the different temperature profiles. The average heat flux and the critical heat flux are evaluated for each critical velocity and their ratio reported. The recommended critical velocity to be used in nuclear channel calculations is that of Wambsganss. The mathematical model used is more precise and all the physical quantities, when using this critical velocity, stay in safe margins.

  15. 发酵产氢的影响因素Ⅱ.反应器类型、营养条件和环境条件%Factors Influencing Fermentative Hydrogen Production Ⅱ.Reactor Type, Nutritional and Environmental Factors

    Institute of Scientific and Technical Information of China (English)

    沈李东; 金仁村

    2011-01-01

    Fermentative hydrogen production has captured extensive attention recently.But fermentative hydrogen production is a very complex process and influenced by many factors.Therefore,the researches of how to maintain stable and continuous hydrogen production and achieve a high yield have become the research focus as core role.This paper summarized several main factors (including inoculum,substrate,reactor type,nutritional and environmental factors) influencing fermentative hydrogen production,followed by some suggestions for the future work of fermentative hydrogen production based on both domestic literature and foreign journals.This part summarized the influence of reactor type, nutritional and environmental factors on fermentative hydrogen production.The effects of inoculum and substrate had been discussed in part Ⅰ.%发酵产氢技术在国内外受到了普遍关注,然该过程是一个极其复杂的生物过程,受诸多因素的干扰.如何优化产氢过程,确保稳定、持续和高效的产氢能力已成为发酵产氢研究的重点课题.本文在查阅了国内外大量文献的基础上,总结了主要影响因子(包括接种物、基质、反应器类型、营养条件和环境条件)对发酵产氢的影响,对发酵产氢朱朱的研发工作提出了建议.本部分主要针对反应器类型、营养条件和环境条件三类影响因素进行探讨,接种物和基质的影响已在1部分进行了论述.

  16. Development of New-type Connection Assembly in The Reactor Containment%反应堆安全壳内新型电缆连接装置研制

    Institute of Scientific and Technical Information of China (English)

    李伟

    2015-01-01

    针对田湾核电站反应堆控制保护系统安全壳内电缆绝缘性能降低现象,研制了一种新型电缆连接装置。该装置采用耐高温、耐辐照性能优良的密封填充材料,设计了以套筒附件为核心的复合密封结构,达到了在高温、高湿、高辐照等严酷环境下保持其绝缘性能的设计要求。该新型电缆连接装置对反应堆安全壳内电缆通道的设计和技术改造具有推广应用价值,也可为处理核工业其它领域同类问题提供借鉴和参考。%This paper focuses on the connection assembly in the reactor containment of Tianwan Nuclear Power Station, to develop a new-type connection assembly .After choosing excellent sealing material with high tem-perature resistant and high radiation resistant through the comparative analysis and experimental verification , designing a composite sealing structure with the impact socket as its core , the new-type connection assembly meet the design requirements of insulating property in harsh temperature , humidity and radiation conditions . This paper provides an example to solve cable channel insulation reduction problem in the nuclear reactor con -tainment unit and also provides reference for dealing with similar problems in other areas of nuclear industry .

  17. Thermal-hydraulic modeling of reactivity accidents in MTR reactors

    Directory of Open Access Journals (Sweden)

    Khater Hany

    2006-01-01

    Full Text Available This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactors during a reactivity insertion accident. The model is formulated for coupling reactor kinetics with feedback reactivity and reactor core thermal-hydraulics. To represent the reactor core, two types of channels are considered, average and hot channels. The developed computer program is compiled and executed on a personal computer, using the FORTRAN language. The model is validated by safety-related benchmark calculations for MTR-TYPE reactors of IAEA 10 MW generic reactor for both slow and fast reactivity insertion transients. A good agreement is shown between the present model and the benchmark calculations. Then, the model is used for simulating the uncontrolled withdrawal of a control rod of an ETRR-2 reactor in transient with over power scram trip. The model results for ETRR-2 are analyzed and discussed.

  18. The role of inoculum and reactor configuration for microbial community composition and dynamics in mainstream partial nitritation anammox reactors.

    Science.gov (United States)

    Agrawal, Shelesh; Karst, Søren M; Gilbert, Eva M; Horn, Harald; Nielsen, Per H; Lackner, Susanne

    2017-03-10

    Implementation of partial nitritation anammox (PNA) in the mainstream (municipal wastewater treatment) is still under investigation. Microbial community structure and reactor type can influence the performance of PNA reactor; yet, little is known about the role of the community composition of the inoculum and the reactor configuration under mainstream conditions. Therefore, this study investigated the community structure of inocula of different origin and their consecutive community dynamics in four different lab-scale PNA reactors with 16S rRNA gene amplicon sequencing. These reactors were operated for almost 1 year and subjected to realistic seasonal temperature fluctuations as in moderate climate regions, that is, from 20°C in summer to 10°C in winter. The sequencing analysis revealed that the bacterial community in the reactors comprised: (1) a nitrifying community (consisting of anaerobic ammonium-oxidizing bacteria (AnAOB), ammonia-oxidizing bacteria (AOB), and nitrite-oxidizing bacteria (NOB)); (2) different heterotrophic denitrifying bacteria and other putative heterotrophic bacteria (HB). The nitrifying community was the same in all four reactors at the genus level, although the biomasses were of different origin. Community dynamics revealed a stable community in the moving bed biofilm reactors (MBBR) in contrast to the sequencing batch reactors (SBR) at the genus level. Moreover, the reactor design seemed to influence the community dynamics, and reactor operation significantly influenced the overall community composition. The MBBR seems to be the reactor type of choice for mainstream wastewater treatment.

  19. A Revolução Agrária Cubana: conquistas e desafios

    Directory of Open Access Journals (Sweden)

    Juan Valdés Paz

    2011-01-01

    Full Text Available Após o triunfo da Revolução em 1959, iniciou-se uma profunda transformação da sociedade cubana, primeiro política, depois social e, desde muito cedo, agrária. A Revolução Cubana foi a superação da questão agrária originada na etapa republicana e uma opção socialista para o desenvolvimento socioeconômico do país, do qual o desenvolvimento agrário tem sido um eixo fundamental. A superação dessa "questão agrária" do socialismo cubano determina as estratégias em curso e suas metas imediatas. Como parte desse processo, examinamos seis aspectos que caracterizaram seu desenvolvimento; são eles: a estrutura da pose de terra; o uso do solo; a organização agrária; o modelo tecnológico; a economia agrária; e o desenvolvimento da sociedade rural.Con el triunfo de la Revolución en 1959, se inició una profunda transformación de la sociedad cubana, primero política, después social y desde muy tempranamente, agraria. La Revolución Cubana fue la superación de la cuestión agraria originada en la etapa republicana y una opción socialista para el desarrollo socioeconómico del país, del cual el desarrollo agrario ha sido un eje fundamental. La superación de esta "cuestión agraria" del socialismo cubano, determina las estrategias en curso y sus metas inmediatas. Como parte de ese proceso, se examinan seis de los aspectos que caracterizaron su desarrollo, a saber: la estructura de tenencia de la tierra; el uso del suelo; la organización agraria; el modelo tecnológico; la economía agraria; y el desarrollo de la sociedad rural.After the triumph of the revolution in 1959, a profound transformation of Cuban society began - first political, then social and, very early on, agrarian. The Cuban Revolution surmounted the agrarian issue that had originated during the republican phase and embodied a socialist option for socioeconomic development (of which agricultural development was a fundamental part. Cuban socialism's efforts to

  20. Comércio agrícola: o Brasil e a OMC

    Directory of Open Access Journals (Sweden)

    Danielle Annoni

    2004-03-01

    Full Text Available The agricultural and textile sectors, have been the greaters targets of the developed countries protectionism, that possess in common the fact that the amount of man-power used is sufficiently great, it means variation of the products final cost. Therefore, it is not difficult to understand why the most underdeveloped countries are more competitive in the dispute of the world-wide market’s control in these sectors. Thus the importance of the OMC elapses, in the search for the effectiveness of international trade regulation of agricultural goods is an activity that interests the developing countries, what is particularly applicable to Brazil, where the agricultural products answers almost for one third of the exportations.Los sectores agrícolas y textil, han sido los mayores objetivos del proteccionismo de los países desarrollados, que poseen en común la cantidad de mano-de-obra usado y esta energía es suficientemente grande y significa variación del costo final del producto. Por tanto, no es difícil comprender porque cuanto más subdesarrollados los países, más serán competitivos y amenazadores para el conflicto sobre el control del mercado mundial en estos sectores. Así la importancia del OMC transcurre, en la búsqueda para la eficacia de la regulación del comercio internacional de mercancías agrícolas es una, actividad que interesa a los países en vías de desarrollo en su totalidad, lo que es particularmente aplicable al Brasil, donde los productos agrícolas casi contestan para un tercio de las exportaciones.Os setores agrícola e têxtil, têm sido os maiores alvos do protecionismo dos países desenvolvidos, que possuem em comum o fato de que a quantidade de mão-de-obra utilizada é bastante grande o que significa margem de variação do custo final do produto. Portanto, não é difícil entender porque os países quão mais subdesenvolvidos forem mais competitivos e ameaçadores são na disputa pelo controle do mercado

  1. A History of Dosimetry for the Advanced Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    Shaw Simon

    2016-01-01

    Full Text Available This paper presents a summary of the methods used in the first ∼40 years of AGR neutron dosimetry and nuclear heating calculations, and the influence of the earlier Magnox reactor dosimetry programme. While the current state-of-the-art Monte Carlo methods are extremely powerful they still require very careful consideration of the quality of the input data, nuclear data validation and variance reduction techniques; in particular, this paper examines the difficulties in assuring the adequate convergence of calculations when Monte Carlo acceleration is applied in the presence of significant streaming paths through attenuating or scattering media.

  2. A History of Dosimetry for the Advanced Gas-cooled Reactors

    Science.gov (United States)

    Shaw, Simon; Thornton, Dean

    2016-02-01

    This paper presents a summary of the methods used in the first ˜40 years of AGR neutron dosimetry and nuclear heating calculations, and the influence of the earlier Magnox reactor dosimetry programme. While the current state-of-the-art Monte Carlo methods are extremely powerful they still require very careful consideration of the quality of the input data, nuclear data validation and variance reduction techniques; in particular, this paper examines the difficulties in assuring the adequate convergence of calculations when Monte Carlo acceleration is applied in the presence of significant streaming paths through attenuating or scattering media.

  3. Influence of filling ratio and carrier type on organic matter removal in a moving bed biofilm reactor with pretreatment of electrocoagulation in wastewater treatment.

    Science.gov (United States)

    Lopez-Lopez, C; Martín-Pascual, J; González-Martínez, A; Calderón, K; González-López, J; Hontoria, E; Poyatos, J M

    2012-01-01

    At present, there is great concern about limited water resources and water quality, which require a more advanced technology. The Moving Bed Biofilm Reactor (MBBR) has been shown to be an efficient technology for removal of organic matter and nutrients in industrial and urban wastewater treatment. However, there are some pollutants which are more difficult to remove by biological processes, so this process can be improved with additional physical and chemical treatments such as electrocoagulation, which appears to be a promising technology in electrochemical treatments. In this research, urban wastewater was treated in an MBBR plant with an electrocoagulation pre-treatment. K1 from AnoxKaldnes and AQWISE ABC5 from Aqwise were the carriers studied under three different filling ratios (20, 35, and 50%). The experimental pilot plant had four bioreactors with 20 L of operation volume and a common feed tank with 100 L of operation volume. The movement of the carriers was generated by aeration and stirrer systems. Organic matter removal was studied by analysis of soluble chemical oxygen demand (sCOD). The maximum organic matter removal in this MBBR system was 65.8% ± 1.4% and 78.4% ± 0.1% for K1 and Aqwise ABC5 carriers, respectively. Moreover, the bacterial diversity of the biofilm was studied by temperature-gradient gel electrophoresis (TGGE) of PCR-amplified partial 16S rRNA genes. 20 prominent TGGE bands were successfully reamplified and sequenced, being the predominant population: β-Proteobacteria, α-Proteobacteria, and Actinobacteria.

  4. Catalyst dynamics: consequences for classical kinetic descriptions of reactors

    DEFF Research Database (Denmark)

    Johannessen, Tue; Larsen, Jane Hvolbæk; Chorkendorff, Ib

    2001-01-01

    The modelling of catalytic reactions/reactors has undergone great improvements since the introduction of empirical power-law kinetics in chemical reaction engineering and micro-kinetic models based on insight into the nature of elementary steps have appeared for many reactions. However, recent...... of the dynamical behaviour of some catalytic systems and discuss the corresponding Limitations in existing models for catalytic reactions and reactors. Catalytic reactors operated in non-steady-state are becoming more frequent in industry. The additional efforts needed to accurately simulate these types...... of reactors are discussed. Finally, we discuss the role of computational fluid dynamics (CFD) as a tool for detailed simulation of catalytic reactors....

  5. Single Pellet String Reactor for Intensification of Catalyst Testing in Gas/Liquid/Solid Configuration Réacteur catalytique de type “filaire” pour l’intensification de tests catalytiques en configuration gaz/liquide/solide

    Directory of Open Access Journals (Sweden)

    Hipolito A.I.

    2010-09-01

    Full Text Available Catalyst improvement is a key route toward process improvement in terms of yield, energy efficiency and selectivity optimization. The catalyst development strategy includes catalyst testing on a model or real feedstock. This key step has been the focus of many studies during the last decades concerning reactor design, analytical tool development and operating procedures. Most studies aim to determine catalytic grain activity in isothermal conditions so as to be able to understand and predict the kinetics. With catalyst improvement, in the lab-scale reactors available, the mass transfer rate can become the limiting step compared with the reaction rate, especially for fast exothermic reactions. A new reactor geometry is proposed to intensify the mass transfer and to accelerate the fluid superficial velocities: the single pellet string reactor. To characterize this new geometry, a hydrodynamic study was carried out in a horizontal single pellet string reactor with a 4.0 × 4.0 mm2 square section, filled with spherical particles of diameter varying between 2.0 and 4.0 mm. In this hydrodynamic study, visual observations of the flow patterns were performed, as well as pressure drop measurements and residence time distribution analysis in single liquid phase flow and two-phase flows. In every configuration tested, two main regimes were identified: the “isolated bubbles” regime and the “stratified” regime. Peclet number and liquid hold-up were deduced from the residence time distribution analysis. The measured liquid hold-ups are always higher than 0.6, which indicates, in addition to the visual observations and colorimetric tests, that the catalyst is always fully wetted by the liquid film. The axial dispersion measurements showed that the single liquid phase flow cannot be interpreted by a classical axial dispersion model. However, when a gas phase is added, the flow becomes closer to plug flow, with Peclet numbers always higher than 40. It

  6. Programa computacional para calcular a potência requerida de máquinas e implementos agrícolas

    Directory of Open Access Journals (Sweden)

    Pablo Pereira Corrêa Klaver

    2013-12-01

    Full Text Available O uso de programas computacionais no setor agrícola permite atingir objetivos específicos na área. Dentre esses, um dos mais complexos é a seleção adequada de máquinas e implementos agrícolas visando à otimização de operações agrícolas, devido, principalmente, à grande variedade de equipamentos existentes no mercado e a gama de tarefas e situações de trabalho que estas são submetidas no campo. O objetivo deste trabalho foi desenvolver um programa computacional para calcular a potência requerida de máquinas e implementos agrícolas normalmente utilizados na condução de operações de campo, desde o preparo do solo até as operações de implantação de culturas. Desenvolvido em linguagem PHP, o programa computacional baseia-se na norma ASAE D497.4 - Agricultural Machinery Management Data como referência para desenvolvimento de cálculos. A partir do programa desenvolvido, tornou-se possível a execução de tarefas para cálculos de avaliação da demanda de potência de máquinas e implementos agrícolas de forma simplificada pela internet.

  7. Listeria monocytogenes differential transcriptome analysis reveals temperature-dependent Agr regulation and suggests overlaps with other regulons.

    Science.gov (United States)

    Garmyn, Dominique; Augagneur, Yoann; Gal, Laurent; Vivant, Anne-Laure; Piveteau, Pascal

    2012-01-01

    Listeria monocytogenes is a ubiquitous, opportunistic pathogenic organism. Environmental adaptation requires constant regulation of gene expression. Among transcriptional regulators, AgrA is part of an auto-induction system. Temperature is an environmental cue critical for in vivo adaptation. In order to investigate how temperature may affect AgrA-dependent transcription, we compared the transcriptomes of the parental strain L. monocytogenes EGD-e and its ΔagrA mutant at the saprophytic temperature of 25°C and in vivo temperature of 37°C. Variations of transcriptome were higher at 37°C than at 25°C. Results suggested that AgrA may be involved in the regulation of nitrogen transport, amino acids, purine and pyrimidine biosynthetic pathways and phage-related functions. Deregulations resulted in a growth advantage at 37°C, but affected salt tolerance. Finally, our results suggest overlaps with PrfA, σB, σH and CodY regulons. These overlaps may suggest that through AgrA, Listeria monocytogenes integrates information on its biotic environment.

  8. Light water reactor safety

    CERN Document Server

    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

  9. Nuclear reactor physics

    CERN Document Server

    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

  10. Sistemas de producción agrícola sostenible

    Directory of Open Access Journals (Sweden)

    Róger Martínez Castillo

    2009-01-01

    Full Text Available El desarrollo sostenible se fundamenta en principios éticos, como el respeto y armonía con la naturaleza; valores políticos, como la democracia participativa y equidad social; y normas morales, como racionalidad ambiental. El desarrollo sostenible es igualitario, descentralizado y autogestionario, capaz de satisfacer las necesidades básicas de la población, respetando la diversidad cultural y mejorando la calidad de vida. La agricultura y el desarrollo sostenible se refieren a la necesidad de minimizar la degradación de la tierra agrícola, maximizando a su vez la producción. Este considera el conjunto de las actividades agrícolas, como el manejo de suelos y aguas,el manejo de cultivos y la conservación de la biodiversidad; considerando a su vez el suministro de alimentos y materias primas. La sostenibilidad de los sistemas de producción agrícola se refiere a la capacidad del sistema para mantener su productividad a pesar de las perturbaciones económicas y naturales, externas o internas. La sostenibilidad es función de las características naturales del sistema y las presiones e intervenciones que sufre, así como aquellas intervenciones sociales, económicas y técnicas que se hacen para contrarrestar presiones negativas; destacándose la resiliencia del sistema.

  11. Correlation of radiation-induced changes in microstructure/microchemistry, density and thermo-electric power of type 304L and 316 stainless steels irradiated in the Phénix reactor

    Energy Technology Data Exchange (ETDEWEB)

    Renault Laborne, Alexandra, E-mail: alexandra.renault@cea.fr [CEA, DEN, SRMA, F-91191 Gif-sur-Yvette (France); Gavoille, Pierre [CEA, DEN, SEMI, F-91191 Gif-sur-Yvette (France); Malaplate, Joël [CEA, DEN, SRMA, F-91191 Gif-sur-Yvette (France); Pokor, Cédric [EDF R& D, MMC, Site des Renardières, F-77818 Morêt-sur-Loing cedex (France); Tanguy, Benoît [CEA, DEN, SEMI, F-91191 Gif-sur-Yvette (France)

    2015-05-15

    Annealed specimens of type 304L and 316 stainless steel and cold-worked 316 specimens were irradiated in the Phénix reactor in the temperature range 381–394 °C and to different damage doses up to 39 dpa. The microstructure and microchemistry of both 304L and 316 have been examined using the combination of the different techniques of TEM to establish the void swelling and precipitation behavior under neutron irradiation. TEM observations are compared with results of measurements of immersion density and thermo-electric power obtained on the same irradiated stainless steels. The similarities and differences in their behavior on different scales are used to understand the factors in terms of the chemical composition and metallurgical state of steels, affecting the precipitation under irradiation and the swelling behavior. Irradiation induces the formation of some precipitate phases (e.g., M{sub 6}C and M{sub 23}C{sub 6}-type carbides, and γ’- and G-phases), Frank loops and cavities. According to the metallurgical state and chemical composition of the steel, the amount of each type of radiation-induced defects is not the same, affecting their density and thermo-electric power.

  12. Pyrometric fuel particle measurements in pressurised reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hernberg, R.; Joutsenoja, T. [Tampere Univ. of Technology (Finland)

    1996-12-01

    A fiberoptic two-colour pyrometric technique for fuel particle temperature and size measurement is modified and applied to three pressurized reactors of different type in Finland, Germany and France. A modification of the pyrometric method for simultaneous in situ measurement of the temperature and size of individual pulverized coal particles at the pressurized entrained flow reactor in Jyvaeskylae was developed and several series of measurements were made. In Orleans a fiberoptic pyrometric device was installed to a pressurised thermogravimetric reactor and the two-colour temperatures of fuel samples were measured. Some results of these measurements are presented. The project belongs to EU`s Joule 2 extension research programme. (author)

  13. Monitoring and control of anaerobic reactors

    DEFF Research Database (Denmark)

    Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær;

    2003-01-01

    measurements are reviewed in detail. In the sequel, possible manipulated variables, such as the hydraulic retention time, the organic loading rate, the sludge retention time, temperature, pH and alkalinity are evaluated with respect to the two main reactor types: high-rate and low-rate. Finally, the different......The current status in monitoring and control of anaerobic reactors is reviewed. The influence of reactor design and waste composition on the possible monitoring and control schemes is examined. After defining the overall control structure, and possible control objectives, the possible process...

  14. Manejo y uso de biosólidos en suelos agrícolas

    OpenAIRE

    Robledo Santoyo, Edmundo

    2012-01-01

    Como subproducto del tratamiento de las aguas residuales de la ciudad de Aguascalientes, Ags., en la Planta de Tratamiento “El Niágara”, se generan diariamente de 150 a 170 m3 de biosólidos. Estos materiales son aplicados a suelos agrícolas de áreas cercanas a la planta, sin regulación alguna, por lo que se han obtenido buenos resultados o causado quemaduras en plántulas o endurecimiento de suelos. El objetivo de este proyecto fue generar estrategias para la disposición y aprovechamiento de l...

  15. A crise de identidade das cooperativas agrárias na França

    OpenAIRE

    Vargas-Prieto, Amanda

    2014-01-01

    El contexto evolutivo del sector agrícola en Francia llevó a las cooperativas agrarias a crear grupos de empresas que combinan las cooperativas tradicionales —encargadas del proceso de producción— y las filiales, encargadas de agregar valor a la producción. En este artículo se presenta la especificidad de los grupos cooperativos agrarios en Francia. La nueva estructura cooperativa lleva a asociar el modo de gobernanza asociativo de la estructura tradicional y el modo de gobernanza accionarial...

  16. Biorremediación de un suelo agrícola impactado con aceite residual automotriz

    OpenAIRE

    Pasaye Anaya, Lizbeth

    2011-01-01

    El aceite residual automotriz (ARA) derivado del uso de automotores es una mezcla de hidrocarburos (HC) alifáticos, aromáticos y otros que en el suelo impide la actividad microbiana para el reciclaje de elementos esenciales para la vida. El objetivo de este trabajo fue analizar la biorremediación (BR) por bioestimulación (BS) de un suelo agrícola impactado con ARA. Para ello, el suelo contaminado con la mezcla de HC se biorremedió por BS con una solución mineral (SM), lombricomposta (LC) y ab...

  17. Comportamiento adsortivo-desortivo del lindano en un suelo agrícola

    OpenAIRE

    Ireri V. Robles-González; Elvira Ríos-Leal; Juvencio Galíndez-Mayer; Sergio Caffarel-Méndez; Josefina Barrera-Cortés; Fernando J. Esparza-García; Poggi-Varaldo, Héctor M

    2006-01-01

    El objetivo de este trabajo fue estudiar el comportamiento adsortivo-desortivo del pesticida lindano en un suelo agrícola con importante contenido de materia orgánica, utilizando el coeficiente de histéresis diferencial (CHPoggi). Además se obtuvo una expresión analítica simple del CHPoggi para el caso mixto, cuando la isoterma de adsorción sigue el modelo lineal y la isoterma de desorción sigue el modelo de Langmuir. Se obtuvo una isoterma de adsorción lineal, q= 3,91·C y una isoterma de des...

  18. Escuelas de campo y adopción de ecotecnia agrícola

    OpenAIRE

    2008-01-01

    Las Escuelas de Campo (EC) son un método de extensión participativo que permite desarrollar capacidades analíticas, pensamiento crítico y, creatividad. La investigación se realizó en la región Mazateca de Oaxaca, la cual se caracteriza por su alto rezago tecnológico, porque resulta de relevancia científica, social y ecológica conocer, en este contexto, los alcances de metodologías alternativas de extensionismo sobre la adopción de ecotecnias agrícolas. Nuestro objetivo fue conocer la...

  19. Readiness Review of BWXT for Fabrication of AGR-5/6/7 TRISO Particles

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Douglas William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sharp, Michelle Tracy [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

    INL readiness review assessment of BWXT readiness to commence fabrication of low-enriched TRISO coated fuel particles for the AGR-5/6/7 irradiation experiments. BWXT self-identified equipment issues preventing operation. INL identified two findings. The first was that disposition codes had not been assigned and documented on BWXT forms to ensure that off-specification materials could not be used in the fabrication of TRISO particles. The second was that chemical purity specifications were not reliably passed on to chemical suppliers, which resulted in the receipt of one acetylene cylinder with suspect impurity levels.

  20. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report; Qualifizierung des Kernmodells DYN3D im Komplex mit dem Stoerfallcode ATHLET als fortgeschrittenes Werkzeug fuer die Stoerfallanalyse von WWER-Reaktoren. T. 1. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [Deutsch] Das Reaktorkernmodell DYN3D mit 3D Neutronenkinetik wurde an den Thermohydraulik-Systemcode ATHLET angekoppelt. Im vorliegenden Bericht werden Arbeiten zur Qualifizierung des gekoppelten Codekomplexes zu einem validierten Hilfsmittel fuer Stoerfallablaufanalysen zu Reaktoren des russischen Typs WWER dargestellt. Diese umfassten im einzelnen: - Beitraege zur Validierung der Einzelcodes ATHLET und DYN3D anhand der Nachrechnung von Experimenten zum

  1. Spinning fluids reactor

    Science.gov (United States)

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  2. Influence of operation factors on brittle fracture initiation and critical local normal stress in SE(B) type specimens of VVER reactor pressure vessel steels

    Science.gov (United States)

    Kuleshova, E. A.; Erak, A. D.; Kiselev, A. S.; Bubyakin, S. A.; Bandura, A. P.

    2015-12-01

    A complex of mechanical tests and fractographic studies of VVER-1000 RPV SE(B) type surveillance specimens was carried out: the brittle fracture origins were revealed (non-metallic inclusions and structural boundaries) and the correlation between fracture toughness parameters (CTOD) and fracture surface parameters (CID) was established. A computational and experimental method of the critical local normal stress determination for different origin types was developed. The values of the critical local normal stress for the structural boundary origin type both for base and weld metal after thermal exposure and neutron irradiation are lower than that for initial state due to the lower cohesive strength of grain boundaries as a result of phosphorus segregation.

  3. Experimental study on the treatment of wastewater from food waste by a new type of internal circulation reactor%新型IC反应器处理餐厨垃圾废水的实验研究

    Institute of Scientific and Technical Information of China (English)

    王罕; 蒋文化; 顾礼炜; 马三剑

    2014-01-01

    采用内循环厌氧反应器(IC)处理餐厨垃圾废水。结果表明:采用快速提升负荷至5 kg/(m3·d)并稳定运行19 d这一启动方式有利于提高污泥的活性。负荷提升中后期,出水pH高于进水pH。IC处理餐厨垃圾废水的最大容积负荷为25.2 kg/(m3·d),此时COD去除率下降到86%。稳定运行期,当进水COD达到22.4 mg/L,出水COD稳定在1650~1950 mg/L,COD去除率高达91.8%。%The new type of internal circulation (IC ) reactor has been used for treating the wastewater from food waste water. The results show that in the start-up period,the start-up form of raising the load rapidly to 5 kg/(m3·d) and running the system steadily for 19 d,is good for improving the sludge activity. In the mid late period of load lifting,the pH of effluent is higher than that of influent. The maximum volume load of food-waste wastewater treated by IC reactor is 25.2 kg/(m3·d). At this time,the COD removing rate declines to 86%. In the steadily running period,when COD concentration of influent reaches 22.4 mg/L,the COD concentration of effluent stabilizes between 1 650-1 950 mg/L,and the COD removing rate reaches 91.8%.

  4. Development of System Analysis Code for Pool-Type Fast Reactor Under Transient Operation%池式快堆系统瞬态分析软件开发

    Institute of Scientific and Technical Information of China (English)

    陆道纲; 隋丹婷

    2012-01-01

    为实现快堆系统分析软件国产化,在已开发的适用于稳态计算的池式快堆系统分析软件SAC-CFR的基础上,进一步开发了系统各部件的瞬态模型、控制系统和保护系统模型、瞬态工况热工水力学的求解逻辑,完成瞬态计算功能的开发.通过对日本文殊快堆45%功率汽机跳闸工况进行建模分析,验证了SAC-CFR用于系统瞬态分析的有效性,为进一步开发非能动余热排出系统分析模型打下了基础.%Aiming at developing system analysis code independently, the system analysis code for pool-type fast reactor in China (SAC-CFR) under transient operation was developed with further development of component transient model, plant control and protection system model, calculation logic for system transient thermal-hydraulic analysis based on the former SAC-CFR version applicable to steady state analysis. The transient started from turbine trip test at 45 % thermal output in the Monju Plant was analyzed with the developed SAC-CFR. A good agreement between the calculated results and the test data was obtained. SAC-CFR is now ready to incorporate passive residual heat removal model for China Experimental Fast Reactor.

  5. Trabajadores agrícolas migrantes en Baja California. Vinculación con la migración internacional

    Directory of Open Access Journals (Sweden)

    Ma. Eugenia Anguiano Téllez

    1989-09-01

    Full Text Available En este ensayo los jornaleros agrícolas migrantes son clasificados en dos estratos: el primero, considera la migración internacional, es decir, aquéllos que se internan en Estados Unidos con el propósito de conseguir empleo; y el segundo, corresponde a la migración interna de los trabajadores agrícolas que permanecen en Baja California para laborar en los valles de Mexicali y San Quintín.En este trabajo se señalan las diferencias entre estos dos estratos, tanto de carácter económico como en los niveles de educación; asimismo, se comparan las características de los jornaleros agrícolas migrantes que cruzan a Estados Unidos por Mexicali y por Tijuana, según su actividad económica de procedencia y su expectativa de empleo en los Estados Unidos.

  6. Analysis of the MEX-15 multipurpose reactor using SRAC code system

    Energy Technology Data Exchange (ETDEWEB)

    Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1992-12-15

    The MEX-15 is a conceptual design of a Multipurpose Reactor with thermal power of 15 MW and this reactor is pool type with fuel plates U{sub 3}0{sub 8}-Al of low enrichment uranium. This report presents the static calculation for the MEX-15 reactor using SRAC code system and was developed under the collaboration agreement between ININ-JAERI in Research Reactor Technology Development Division of Department of Research Reactor in Tokai Research Establishment. (Author)

  7. An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

    Science.gov (United States)

    Koontz, Steven L.; Davis, Dennis D.; Hansen, Merrill

    1988-01-01

    A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

  8. Uncertainty quantification approaches for advanced reactor analyses.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Nuclear Engineering Division

    2009-03-24

    The original approach to nuclear reactor design or safety analyses was to make very conservative modeling assumptions so as to ensure meeting the required safety margins. Traditional regulation, as established by the U. S. Nuclear Regulatory Commission required conservatisms which have subsequently been shown to be excessive. The commission has therefore moved away from excessively conservative evaluations and has determined best-estimate calculations to be an acceptable alternative to conservative models, provided the best-estimate results are accompanied by an uncertainty evaluation which can demonstrate that, when a set of analysis cases which statistically account for uncertainties of all types are generated, there is a 95% probability that at least 95% of the cases meet the safety margins. To date, nearly all published work addressing uncertainty evaluations of nuclear power plant calculations has focused on light water reactors and on large-break loss-of-coolant accident (LBLOCA) analyses. However, there is nothing in the uncertainty evaluation methodologies that is limited to a specific type of reactor or to specific types of plant scenarios. These same methodologies can be equally well applied to analyses for high-temperature gas-cooled reactors and to liquid metal reactors, and they can be applied to steady-state calculations, operational transients, or severe accident scenarios. This report reviews and compares both statistical and deterministic uncertainty evaluation approaches. Recommendations are given for selection of an uncertainty methodology and for considerations to be factored into the process of evaluating uncertainties for advanced reactor best-estimate analyses.

  9. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    Energy Technology Data Exchange (ETDEWEB)

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  10. Analysis of the horizontal flow in the advanced gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Duan, Y. [Department of Mechanical Engineering, University of Sheffield, Sheffield S1 3JD (United Kingdom); He, S., E-mail: s.he@sheffield.ac.uk [Department of Mechanical Engineering, University of Sheffield, Sheffield S1 3JD (United Kingdom); Ganesan, P. [Department of Mechanical Engineering, University of Malaya, Kuala Lumpur 50603 (Malaysia); Gotts, J. [EDF Energy, Barnwood, Gloucester GL4 3RS (United Kingdom)

    2014-06-01

    Highlights: • CFD is used to assess the effect of horizontal flows in AGRs. • The horizontal flows can reduce the graphite brick temperature significantly. • Such effects are not taken into consideration in current engineering calculations. • There might be flow instabilities when the fuel channel flow is very low but horizontal flows reduce its possibility. - Abstract: The purpose of the paper is to report a computational investigation of horizontal flows in the UK advanced-gas-cooled reactor (AGR) by using computational fluid dynamics with ANSYS FLUENT. The study is relevant to practical issues encountered in some AGR stations currently in operation in the UK. It is carried out using a comparative approach based on the results of two contrasting models: one simulating the full effect of the cross flow, the other simulating the simplified approach currently employed by the industry which neglects the momentum of the horizontal cross flow. The study reveals that the horizontal cross flow plays a significant role in the cooling of the moderator brick, while the axial variation of the brick geometry also significantly changes the distribution of the temperature within the brick. It is also found that under some circumstances the so-called horizontal inter-brick leakage (HIBL) flow could influence the cooling performance in the narrow gaps, resulting in a local hot spot. Furthermore, there may be flow instabilities in the flows in AGR fuel channels due to the interactions between the flow in the main arrowhead flow passages and that in some narrow passages connected to it, but the influence on the brick temperature is negligible. Horizontal cross flow has an effect of reducing such instabilities.

  11. Post-processor for simulations of the ORIGEN program and calculation of the composition of the activity of a burnt fuel core by a BWR type reactor; Post-procesador para simulaciones del programa ORIGEN y calculo de la composicion de la actividad de un nucleo de combustible quemado por un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sandoval V, S. [IIE, Av. Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico)]. e-mail: sandoval@iie.org.mx

    2006-07-01

    The composition calculation and the activity of nuclear materials subject to processes of burnt, irradiation and decay periods are of utility for diverse activities inside the nuclear industry, as they are it: the processes design and operations that manage radioactive material, the calculation of the inventory and activity of a core of burnt nuclear fuel, for studies of type Probabilistic Safety Analysis (APS), as well as for regulation processes and licensing of nuclear facilities. ORIGEN is a program for computer that calculates the composition and the activity of nuclear materials subject to periods of burnt, irradiation and decay. ORIGEN generates a great quantity of information whose processing and analysis are laborious, and it requires thoroughness to avoid errors. The automation of the extraction, conditioning and classification of that information is of great utility for the analyst. By means of the use of the post-processor presented in this work it is facilitated, it speeds up and wide the capacity of analysis of results, since diverse consultations with several classification options and filtrate of results can be made. As illustration of the utility of the post-processor, and as an analysis of interest for itself, it is also presented in this work the composition of the activity of a burned core in a BWR type reactor according to the following classification criteria: by type of radioisotope (fission products, activation products and actinides), by specie type (gassy, volatile, semi-volatile and not volatile), by element and by chemical group. The results show that the total activity of the studied core is dominated by the fission products and for the actinides, in proportion four to one, and that the gassy and volatile species conform a fifth part of the total activity of the core. (Author)

  12. Cocurrent downflow circulating fluidized bed (downer) reactors - a state of the art review

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, J.-X.; Yu, Z.-Q.; Jin, Y.; Grace, J.R.; Issangya, A. [University of Western Ontario, London, ON (Canada). Department of Chemical and Biochemical Engineering

    1995-10-01

    A new type of chemical reactor known as the cocurrent downflow fluidized bed reactor (or reversed riser reactor or downer reactor), that overcomes some of the disadvantages of the riser reactor, is described. Since both the gas and solids flow directions are downwards in the cocurrent downflow fluidized bed reactor, particle residence times are uniform, and there is no backmixing. The literature on downer studies is reviewed. Laboratory results on axial voidage profiles, pressure profiles, radial flow, mixing and residence time distribution, heat transfer, and particle velocities are summarized. Suggestions are made both for possible industrial applications of downer reactors and for suitable research directions. 56 refs., 18 figs., 1 tab.

  13. Unidades integrales de servicios técnicos de maquinaria agrícola

    Directory of Open Access Journals (Sweden)

    José Suárez

    2011-01-01

    Full Text Available En cada municipio cubano se inició la creación de centros encargados de prestar servicios de mecanización agrícola y asistencia técnica a los productores. Para esto deben disponer de tractores, cosechadoras y otras máquinas e implementos propios, así como de talleres para la reparación y mantenimiento. Cada uno de estos centros será una Unidad Integral de Servicios Técnicos (UIST con el carácter de Unidad Empresarial de Base (UEB subordinada a la Empresa Agropecuaria Municipal. Para constituir estas Unidades fue necesario realizar un estudio de la disponibilidad de talleres, equipamiento, maquinaria y personal de mecanización en las empresas agrícolas y otras entidades productivas de todos los municipios del país, pues ellos constituirán el núcleo fundamental del nuevo sistema. Se ha conformado la propuesta del esquema de funcionamiento de las Unidades, especialmente en lo que respecta a áreas, funciones, equipamiento, modelo de gestión, contratación de servicios, planificación, organización, control, relaciones con otras entidades productivas y de dirección, etc. Estas Unidades se están creando en municipios seleccionados, con vistas a evaluar los resultados, y, teniendo en cuenta las experiencias obtenidas, conformar el programa para extender el sistema a todo el país.

  14. Existe convergência espacial da produtividade agrícola no Brasil?

    Directory of Open Access Journals (Sweden)

    Eduardo Simões de Almeida

    2008-03-01

    Full Text Available Este trabalho analisa a convergência espacial da produtividade agrícola da terra para as microrregiões brasileiras entre o período de 1991 a 2003. Tal análise avalia se há evidências para a existência de convergência beta da produtividade agrícola, controlando-se explicitamente para efeitos espaciais. Para tanto, conduziu-se preliminarmente uma análise exploratória de dados espaciais, que detectou a presença de autocorrelação espacial para a produtividade da terra. Numa etapa posterior, desenvolveram-se modelos econométricos espaciais para a análise da convergência.This work analyzes the convergence of land productivity for Brazilian regions from 1991 to 2003. It evaluates if there is or not beta convergence of the agricultural productivity, controlling explicitly for spatial effects. To do so, an exploratory spatial data analysis was previously implemented and the presence of spatial autocorrelation for the land productivity was detected. In addition, spatial econometric models were developed for the convergence analysis.

  15. Evaluación de residuos agrícolas para la producción del hongo Pleurotus ostreatus

    OpenAIRE

    María Bernarda Ruilova Cueva; Aldo Hernández Monzón

    2014-01-01

    Se caracterizaron los residuos agrícolas de la provincia de Bolívar, Ecuador, desde el punto de vista de su composición físico-química para su empleo en el cultivo del hongo Pleurotus ostreatus. Los residuos agrícolas utilizados fueron: maíz, trigo, cebada, lenteja, bagazo de caña de azúcar y cascarilla de arroz. Se realizaron análisis de: humedad, cenizas, extracto etéreo, proteína, fibra, extracto libre de nitrógeno, fibra detergente neutra, fibra detergente ácida, lignina, celulosa, hemice...

  16. Mercado de derechos de agua para uso agrícola en el noroeste de México

    OpenAIRE

    Lorenzo Fidel Cota-Verdugo; Alfredo Beltrán-Morales; Enrique Troyo-Diéguez; José Luís García-Hernández; Luis Felipe Beltrán-Morales; Víctor Ángel Hernández-Trejo; Alfredo Ortega-Rubio; Agustín Navarro Bravo

    2013-01-01

    El estado de Baja California sur, es una zona árida y la provisión de agua para la agricultura y demás usos, se sustenta en acuíferos subterráneos. La legislación en materia de agua se ha modificado acorde al problema evidente del crecimiento en el consumo, especialmente para uso agrícola, incluyendo la figura de mercados de derechos de agua. El objetivo de este trabajo, es buscar mecanismos de mercado que legalicen las transacciones de agua entre usuarios agrícolas, proponer la implementaci...

  17. Oxidation efficiency of elemental mercury in two DBD plasma reactors

    Science.gov (United States)

    Jiang, Yuze; An, Jiutao; Shang, Kefeng; Jiang, Diwen; Li, Jie; Lu, Na; Wu, Yan

    2013-03-01

    Configuration of plasma reactors influences the generation of active species including the energized electrons, active radicals and the distribution of active species in reactor, and thus influences the removal efficiency of pollutants. Oxidation efficiency of elemental mercury (Hg0) in two different DBD plasma reactors was studied in this paper. One plasma reactor is a surface discharge reactor (SDR) with a spiral stainless steel thread as the high voltage electrode, and the other plasma reactor is a concentric cylinder type DBD reactor (CCDR) with a copper screw rod as the high voltage electrode. The oxidation efficiencies of Hg0 under different specific energy density (SED), oxygen content, flue gas residence time and the temperature of flue gas indicate that SDR had a better performance than CCDR in oxidation of Hg0, which can be attributed to the higher generation efficiency of ozone in SDR than in CCDR.

  18. System assessment of helical reactors in comparison with tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Yamazaki, K.; Imagawa, S.; Muroga, T.; Sagara, A.; Okamura, S.

    2002-10-01

    A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-{beta}{sub N} tokamak reactors. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  20. An agr-like two-component regulatory system in Lactobacillus plantarum is involved in production of a novel cyclic peptide and regulation of adherence

    NARCIS (Netherlands)

    Sturme, M.H.J.; Nakayama, J.; Molenaar, D.; Murakami, Y.; Kunugi, R.; Fujii, T.; Vaughan, E.E.; Kleerebezem, M.; Vos, de W.M.

    2005-01-01

    We have analyzed a locus on the annotated Lactobacillus plantarum WCFS1 genome that showed homology to the staphylococcal agr quorum-sensing system and designated it lam for Lactobacillus agr-like module. Production of the lamBDCA transcript was shown to be growth phase dependent. Analysis of a resp

  1. Staphylococcal superantigen-like genes, ssl5 and ssl8, are positively regulated by Sae and negatively by Agr in the Newman strain.

    Science.gov (United States)

    Pantrangi, Madhulatha; Singh, Vineet K; Wolz, Christiane; Shukla, Sanjay K

    2010-07-01

    Some of the staphylococcal superantigen-like (SSL) proteins SSL5, SSL7, SSL9, and SSL11 act as immunomodulatory proteins in Staphylococcus aureus. However, little is known about their regulatory mechanisms. We determined the expression levels of ssl5 and ssl8 in seven clinically important S. aureus strains and their regulatory mechanisms in the Newman strain, which had the highest ssl5 and ssl8 expression. Independent comparisons of ssl5 or ssl8 coding and upstream sequences in these strains identified multiple haplotypes that did not correlate with the differential expression of ssl5 and ssl8, suggesting the role of additional regulatory elements. Using knockout mutant strains of known S. aureus global regulators such as Agr, Sae, and SigB in the Newman strain, we showed that both ssl5 and ssl8 were induced by Sae and repressed by Agr, suggesting that Sae and Agr are the positive and the negative regulators, respectively, of these two ssl genes. Moreover, we observed upregulation of sae in the agr mutant and upregulation of agr in the sae mutant compared with the isogenic Newman strain, suggesting that the Agr and Sae may be inhibiting each other. The SigB mutation did not affect ssl5 and ssl8 expression, but they were downregulated in the agr/sigB double mutant, indicating that SigB probably acts synergistically with Agr in their upregulation.

  2. Green Fluorescent Protein (GFP-Based Overexpression Screening and Characterization of AgrC, a Receptor Protein of Quorum Sensing in Staphylococcus aureus

    Directory of Open Access Journals (Sweden)

    Shengdi Fan

    2013-09-01

    Full Text Available Staphylococcus aureus AgrC is an important component of the agr quorum-sensing system. AgrC is a membrane-embedded histidine kinase that is thought to act as a sensor for the recognition of environmental signals and the transduction of signals into the cytoplasm. However, the difficulty of expressing and purifying functional membrane proteins has drastically hindered in-depth understanding of the molecular structures and physiological functions of these proteins. Here, we describe the high-yield expression and purification of AgrC, and analyze its kinase activity. A C-terminal green fluorescent protein (GFP fusion to AgrC served as a reporter for monitoring protein expression levels in real time. Protein expression levels were analyzed by the microscopic assessment of the whole-cell fluorescence. The expressed AgrC-GFP protein with a C-terminal His-tagged was purified using immobilized metal affinity chromatography (IMAC and size exclusion chromatography (SEC at yields of ≥10 mg/L, following optimization. We also assessed the effects of different detergents on membrane solubilization and AgrC kinase activity, and polyoxyethylene-(23-lauryl-ether (Brij-35 was identified as the most suitable detergent. Furthermore, the secondary structural stability of purified AgrC was analyzed using circular dichroism (CD spectroscopy. This study may serve as a general guide for improving the yields of other membrane protein preparations and selecting the appropriate detergent to stabilize membrane proteins for biophysical and biochemical analyses.

  3. Green fluorescent protein (GFP)-based overexpression screening and characterization of AgrC, a Receptor protein of quorum sensing in Staphylococcus aureus.

    Science.gov (United States)

    Wang, Lina; Quan, Chunshan; Liu, Baoquan; Xu, Yongbin; Zhao, Pengchao; Xiong, Wen; Fan, Shengdi

    2013-09-06

    Staphylococcus aureus AgrC is an important component of the agr quorum-sensing system. AgrC is a membrane-embedded histidine kinase that is thought to act as a sensor for the recognition of environmental signals and the transduction of signals into the cytoplasm. However, the difficulty of expressing and purifying functional membrane proteins has drastically hindered in-depth understanding of the molecular structures and physiological functions of these proteins. Here, we describe the high-yield expression and purification of AgrC, and analyze its kinase activity. A C-terminal green fluorescent protein (GFP) fusion to AgrC served as a reporter for monitoring protein expression levels in real time. Protein expression levels were analyzed by the microscopic assessment of the whole-cell fluorescence. The expressed AgrC-GFP protein with a C-terminal His-tagged was purified using immobilized metal affinity chromatography (IMAC) and size exclusion chromatography (SEC) at yields of ≥ 10 mg/L, following optimization. We also assessed the effects of different detergents on membrane solubilization and AgrC kinase activity, and polyoxyethylene-(23)-lauryl-ether (Brij-35) was identified as the most suitable detergent. Furthermore, the secondary structural stability of purified AgrC was analyzed using circular dichroism (CD) spectroscopy. This study may serve as a general guide for improving the yields of other membrane protein preparations and selecting the appropriate detergent to stabilize membrane proteins for biophysical and biochemical analyses.

  4. Reactor core design and characteristics of the Fugen

    Energy Technology Data Exchange (ETDEWEB)

    Matsumoto, Mitsuo; Kowata, Yasuki; Sugawara, Satoru; Deshimaru, Takehide

    1988-03-01

    The heavy water moderated, boiling light water cooled pressure tube type reactor Fugen uses plutonium-uranium mixed oxide as a fuel. Heavy water as the moderator and the light water of coolant are separated by the pressure tubes and calandria tubes. Thereby, the reactor core is heterogenes compared with that of LWRs. This paper describes the development of reactor core design procedure based on the feature of the Fugen type reactor, the feasibility test and the validity of nuclear and thermalhydraulic design based on the operating experience.

  5. Reactor pulse repeatability studies at the annular core research reactor

    Energy Technology Data Exchange (ETDEWEB)

    DePriest, K.R. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Trinh, T.Q. [Nuclear Facility Operations, Sandia National Laboratories, Mail Stop 0614, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Luker, S. M. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States)

    2011-07-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories is a water-moderated pool-type reactor designed for testing many types of objects in the pulse and steady-state mode of operations. Personnel at Sandia began working to improve the repeatability of pulse operations for experimenters in the facility. The ACRR has a unique UO{sub 2}-BeO fuel that makes the task of producing repeatable pulses difficult with the current operating procedure. The ACRR produces a significant quantity of photoneutrons through the {sup 9}Be({gamma}, n){sup 8}Be reaction in the fuel elements. The photoneutrons are the result of the gammas produced during fission and in fission product decay, so their production is very much dependent on the reactor power history and changes throughout the day/week of experiments in the facility. Because the photoneutrons interfere with the delayed-critical measurements required for accurate pulse reactivity prediction, a new operating procedure was created. The photoneutron effects at delayed critical are minimized when using the modified procedure. In addition, the pulse element removal time is standardized for all pulse operations with the modified procedure, and this produces less variation in reactivity removal times. (authors)

  6. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  7. SNTP program reactor design

    Science.gov (United States)

    Walton, Lewis A.; Sapyta, Joseph J.

    1993-06-01

    The Space Nuclear Thermal Propulsion (SNTP) program is evaluating the feasibility of a particle bed reactor for a high-performance nuclear thermal rocket engine. Reactors operating between 500 MW and 2,000 MW will produce engine thrusts ranging from 20,000 pounds to 80,000 pounds. The optimum reactor arrangement depends on the power level desired and the intended application. The key components of the reactor have been developed and are being tested. Flow-to-power matching considerations dominate the thermal-hydraulic design of the reactor. Optimal propellant management during decay heat cooling requires a three-pronged approach. Adequate computational methods exist to perform the neutronics analysis of the reactor core. These methods have been benchmarked to critical experiment data.

  8. Fast Spectrum Reactors

    CERN Document Server

    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. Using thermal balance model to determine optimal reactor volume and insulation material needed in a laboratory-scale composting reactor.

    Science.gov (United States)

    Wang, Yongjiang; Pang, Li; Liu, Xinyu; Wang, Yuansheng; Zhou, Kexun; Luo, Fei

    2016-04-01

    A comprehensive model of thermal balance and degradation kinetics was developed to determine the optimal reactor volume and insulation material. Biological heat production and five channels of heat loss were considered in the thermal balance model for a representative reactor. Degradation kinetics was developed to make the model applicable to different types of substrates. Simulation of the model showed that the internal energy accumulation of compost was the significant heat loss channel, following by heat loss through reactor wall, and latent heat of water evaporation. Lower proportion of heat loss occurred through the reactor wall when the reactor volume was larger. Insulating materials with low densities and low conductive coefficients were more desirable for building small reactor systems. Model developed could be used to determine the optimal reactor volume and insulation material needed before the fabrication of a lab-scale composting system.

  10. Una aproximación a las características actuales de la mano de obra agrícola en el valle de Mexicali

    Directory of Open Access Journals (Sweden)

    José Ascención Moreno Mena

    1994-07-01

    Full Text Available Este trabajo aborda, de manera sintética, algunas características de la estructura agraria, así como de productores agrícolas del valle de Mexicali. Se presenta un acercamiento a las condiciones en que se desarrolla la mano de obra actual en el sector agrícola del municipio de Mexicali. Para ello se toma como base información primaria obtenida de dos encuestas realizadas durante el periodo 1991-1994; laprimera producto de un muestreo no aleatorio por cuotas aplicada a 150 trabajadores agrícolas del algodón; y la segunda de carácter aleatoria aplicada en 499 viviendas de trabajadores agrícolas en el valle de Mexicali por el Instituto de investigaciones Sociales de la Universidad Autónoma de Baja California y el Programa Nacional de Solidaridad con Jornaleros Agrícolas.

  11. Multi purpose research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Raina, V.K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)]. E-mail: vkrain@magnum.barc.ernet.in; Sasidharan, K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sengupta, Samiran [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Singh, Tej [Research Reactor Services Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2006-04-15

    At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor.

  12. INVAP's Research Reactor Designs

    Directory of Open Access Journals (Sweden)

    Eduardo Villarino

    2011-01-01

    Full Text Available INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors.

  13. A unified theory of autism revisited: linkage evidence points to chromosome X using a high-risk subset of AGRE families.

    Science.gov (United States)

    Allen-Brady, Kristina; Cannon, Dale; Robison, Reid; McMahon, William M; Coon, Hilary

    2010-04-01

    Zhao et al. [2007] in their "Unified Theory of Autism" hypothesized that incidence of autism in males could be explained by essentially two types of family structures: majority of autism cases are from low-risk autism families with de novo mutations, and a minority of cases are from high-risk multiplex families, where risk to male offspring approximates 50% consistent with a dominant model and high penetrance. Using the Autism Genetic Resource Exchange (AGRE) data set, Zhao et al. identified 86 high-risk families with likely dominant transmission. As genotype data are now available for many members of the AGRE resource, the objective of this manuscript was to determine if dominant linkage evidence for an autism predisposition gene exists in these 86 high-risk families. HumanHap550K Illumina SNP data were available for 92% of 455 total family members in these 86 high-risk families. We performed a linkage analysis using a pruned subset of markers where markers in high linkage disequilibrium were removed. We observed a single suggestive peak (maximum LOD 2.01, maximum HLOD 2.08) under a dominant model on chromosome Xp22.11-p21.2 that encompasses the IL1RAPL1 gene. Mutations or deletions in IL1RAPL1 have been previously reported in three families with autism. In our study, 11 families contributed nominally (P0.588) to the chromosome X peak. These results demonstrate that identification of a more homogeneous subset of autism cases, which was based on family structure in this study, may help to identify, localize and further our understanding of autism predisposition genes.

  14. The counter-rotating twin screw extruder as a polymerization reactor

    NARCIS (Netherlands)

    Ganzeveld, Klaassien Jakoba

    1992-01-01

    The goal of the research was to examine the possibilities of this type of extruder as a polymerization reactor, and to develop models of the extruder reactor which accurately describe the reaction progress in the extruder. See summary

  15. Antineutrino reactor safeguards - a case study

    CERN Document Server

    Christensen, Eric; Jaffke, Patrick

    2013-01-01

    Antineutrinos have been proposed as a means of reactor safeguards for more than 30 years and there has been impressive experimental progress in neutrino detection. In this paper we conduct, for the first time, a case study of the application of antineutrino safeguards to a real-world scenario - the North Korean nuclear crisis in 1994. We derive detection limits to a partial or full core discharge in 1989 based on actual IAEA safeguards access and find that two independent methods would have yielded positive evidence for a second core with very high confidence. To generalize our results, we provide detailed estimates for the sensitivity to the plutonium content of various types of reactors, including most types of plutonium production reactors, based on detailed reactor simulations. A key finding of this study is that a wide class of reactors with a thermal power of less than 0.1-1 GWth can be safeguarded achieving IAEA goals for quantitative sensitivity and timeliness with detectors right outside the reactor ...

  16. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. A brief history of design studies on innovative nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com [Emeritus Professor, Tokyo Institute of Technology (Japan)

    2014-09-30

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  18. A brief history of design studies on innovative nuclear reactors

    Science.gov (United States)

    Sekimoto, Hiroshi

    2014-09-01

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970's the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980's the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  19. Modular Lead-Bismuth Fast Reactors in Nuclear Power

    OpenAIRE

    Vladimir Petrochenko; Georgy Toshinsky

    2012-01-01

    On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral) design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing...

  20. Biofilm reactors for industrial bioconversion processes: employing potential of enhanced reaction rates

    Directory of Open Access Journals (Sweden)

    Karcher Patrick

    2005-08-01

    Full Text Available Abstract This article describes the use of biofilm reactors for the production of various chemicals by fermentation and wastewater treatment. Biofilm formation is a natural process where microbial cells attach to the support (adsorbent or form flocs/aggregates (also called granules without use of chemicals and form thick layers of cells known as "biofilms." As a result of biofilm formation, cell densities in the reactor increase and cell concentrations as high as 74 gL-1 can be achieved. The reactor configurations can be as simple as a batch reactor, continuous stirred tank reactor (CSTR, packed bed reactor (PBR, fluidized bed reactor (FBR, airlift reactor (ALR, upflow anaerobic sludge blanket (UASB reactor, or any other suitable configuration. In UASB granular biofilm particles are used. This article demonstrates that reactor productivities in these reactors have been superior to any other reactor types. This article describes production of ethanol, butanol, lactic acid, acetic acid/vinegar, succinic acid, and fumaric acid in addition to wastewater treatment in the biofilm reactors. As the title suggests, biofilm reactors have high potential to be employed in biotechnology/bioconversion industry for viable economic reasons. In this article, various reactor types have been compared for the above bioconversion processes.

  1. The Anterior Gradient Homolog 3 (AGR3) Gene Is Associated with Differentiation and Survival in Ovarian Cancer

    Science.gov (United States)

    King, Erin R.; Tung, Celestine S.; Tsang, Yvonne T.M.; Zu, Zhifei; Lok, Gabriel T.M.; Deavers, Michael T.; Malpica, Anais; Wolf, Judith K.; Lu, Karen H.; Birrer, Michael J.; Mok, Samuel C.; Gershenson, David M.; Wong, Kwong-Kwok

    2011-01-01

    Low-grade serous ovarian carcinoma is believed to arise from serous borderline ovarian tumors, yet the progression from serous borderline tumors to low-grade serous ovarian carcinoma remains poorly understood. The purpose of this study was to identify differentially expressed genes between the two groups. Expression profiles were generated from 6 human ovarian surface epithelia (HOSE), 8 serous borderline ovarian tumors (SBOT), 13 low-grade serous ovarian carcinomas (LG), and 24 high-grade serous ovarian carcinomas (HG). The anterior gradient homolog 3 (AGR3) gene was found to be highly upregulated in serous borderline ovarian tumors; this finding was validated by real-time quantitative RT-PCR, Western blotting, and immunohistochemistry. Anti-AGR3 immunohistochemistry was performed on an additional 56 LG and 103 HG tissues and the results were correlated with clinical data. Expression profiling determined that 1254 genes were differentially expressed (P 10%) of AGR3 positively stained tumor cells were associated with improved longer median survival in both the LG (P = 0.013) and HG (P = 0.008) serous ovarian carcinoma groups. The progression of serous borderline ovarian tumors to low-grade serous ovarian carcinoma may involve the de-differentiation of ciliated cells. AGR3 could serve as a prognostic marker for survival in patients with low-grade and high-grade serous ovarian carcinomas. PMID:21451362

  2. PEMBUATAN BAHAN BAKU SPREADS KAYA KAROTEN DARI MINYAK SAWIT MERAH MELALUI INTERESTERIFIKASI ENZIMATIK MENGGUNAKAN REAKTOR BATCH [Preparation of Red Palm Oil Based-Spreads Stock Rich in Carotene Through Enzymatic Interesterification in Batch-type Reactor

    Directory of Open Access Journals (Sweden)

    Nur Wulandari1,2

    2012-12-01

    Full Text Available Enzymatic interesterification of red palm oil (a mixture of red palm olein/RPO and red palm stearin/RPS in 1:1 weight ratio and coconut oil (CNO blends of varying proportions using a non-specific immobilized Candida antartica lipase (Novozyme 435 was studied for the preparation of spread stock. The interesterification reaction was held in a batch-type reactor. Two substrate blends were chosen for the production of spread stock i.e. 77.5:22,5 and 82.5:17.5 (RPO/RPS:CNO, by weight through enzymatic interesterification in three different reaction times (2, 4, and 6 hours. The interesterification reactions were conducted at 60°C, 200 rpm agitation speed and 10% of Novozyme 435. The interesterified products were evaluated for their physical characteristics (slip melting point or SMP and solid fat content or SFC and chemical characteristics (carotene retention, moisture content, and free fatty acid/FFA content. All of the interesterified products had lower SFC and SMP as compared to the initial blends. The SMP and SFC increased in longer reaction times. The SMP ranged from 30.8°C to 34.9°C. The carotene retention ranged from 74.80% to 81.08%, while the moisture content and FFA content increased in longer reaction times. The interesterified products had desirable physical properties for possible use as a spread stock rich in carotene.

  3. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  4. Space Nuclear Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-06

    We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

  5. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

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

  6. Ingeniería agrícola, una alternativa para las necesidades del agro

    OpenAIRE

    2011-01-01

    La ingeniería agrícola como profesión que articula los conceptos de la ingeniería y los procesos biológicos, está llamada a desempeñar un papel protagónico en el desarrollo del campo colombiano. Por ello requerimos espacio político para poder hacer parte de los estamentos de decisión y difundir estrategias en: - Control de erosión. - Desarrollo de tecnología apropiada a través de la investigación en las universidades colombianas. - Racionalización de las pérdidas en frutas, hortalizas y cerea...

  7. Plano de negócios Monsaluz, Sociedade Agrícola Lda.

    OpenAIRE

    Rente, José Eduardo Lopes

    2014-01-01

    Este trabalho de projeto tem como finalidade a elaboração de um plano de negócios, para reconversão de uma empresa agrícola no Alentejo, com um investimento na produção de ovinos de aptidão leiteira. O presente trabalho de projeto é apresentado metodologicamente através de um plano de negócio, que se pretende constitua um guia de apoio á tomada de decisão na implantação e controlo desse projeto. Com base na caracterização e análise da situação atual da empresa, na análise do...

  8. Modelo organizacional da Caixa Agrícola do Litoral Centro

    OpenAIRE

    Trindade, Tiago João Pereira Cachulo da

    2009-01-01

    Relatório de Projecto / Mestrado em Gestão Este projecto estuda a implementação de um novo Modelo Organizacional para a Caixa Agrícola do Litoral Centro que deverá constituir um referencial organizacional que permita o crescimento da Instituição, que contribua para a focalização na vertente do negócio, que cumpra as exigências da Supervisão e que potencie a descentralização da tomada de decisões da Direcção para as estruturas funcionais. Como enquadramento teórico deste trab...

  9. América Latina y las exportaciones de recursos naturales agrícolas

    OpenAIRE

    Piñeiro, Martín; Bianchi, Eduardo

    2012-01-01

    La discusión sobre las estrategias de desarrollo en América Latina ha estado fuertemente influida por dos conceptos desarrollados en la literatura económica: (a) la hipótesis de que una buena dotación de recursos naturales puede tener un efecto negativo sobre el desarrollo o "la maldición de los recursos naturales'' y (b) la evidencia empírica que sugería que los términos de intercambio de la producción agrícola sufren un permanente deterioro. Una revisión de la literatura indica que la hipót...

  10. Ingeniería agrícola, una alternativa para las necesidades del agro

    Directory of Open Access Journals (Sweden)

    Julio E. Ospina M.

    2011-02-01

    Full Text Available La ingeniería agrícola como profesión que articula los conceptos de la ingeniería y los procesos biológicos, está llamada a desempeñar un papel protagónico en el desarrollo del campo colombiano. Por ello requerimos espacio político para poder hacer parte de los estamentos de decisión y difundir estrategias en: - Control de erosión. - Desarrollo de tecnología apropiada a través de la investigación en las universidades colombianas. - Racionalización de las pérdidas en frutas, hortalizas y cereales, entre otros".

  11. Modelo de Precificação de Capital: Segmento de Commodities Agrícola

    Directory of Open Access Journals (Sweden)

    TÁCITO AUGUSTO FARIAS

    2014-12-01

    Full Text Available O objetivo do artigo é a aplicação do modelo de precificação de ativos conhecido na literatura de economia financeira como CAPITAL ASSET PRICING MODEL – MODELO DE PRECIFICAÇÃO DE ATIVOS DE CAPITAL, utilizando como ferramentas os balanços patrimoniais das empresas selecionadas. Brasil Ecodiesel indústria e comercio de biocombustíveis e óleos vegetais s.a. (v-agro e SLC Agrícola s.a. no período relativo ao ano de 2009 e 2010. Resultado fundamental: ambas as empresas no período em estudo apresentaram um beta inferior a 1, ou seja, risco menor que o risco do mercado. Descrevemos alguns resultados estatísticos que mostram o comportamento de ambas as empresas do setor de frigorifico. 

  12. AGR-1 Compact 4-1-1 Post-Irradiation Examination Results

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ploger, Scott A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); van Rooyen, Isabella J. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

    Destructive post-irradiation examination was performed on AGR-1 fuel Compact 4-1-1, which was irradiated to a final compact-average burnup of 19.4% FIMA (fissions per initial metal atom) and a time-average, volume-average temperature of 1072°C. The analysis of this compact focused on characterizing the extent of fission product release from the particles and examining particles to determine the condition of the kernels and coating layers. The work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, metallurgical preparation of selected particles, and examination of particle cross-sections with optical microscopy, electron microscopy, and elemental analysis. Deconsolidation-leach-burn-leach (DLBL) analysis revealed no particles with failed TRISO or failed SiC layers (as indicated by very low uranium inventory in all of the leach solutions). The total fractions of the predicted compact inventories of fission products Ce-144, Cs-134, Cs-137, and Sr-90 that were present in the compact outside of the SiC layers were <2×10-6, based on DLBL data. The Ag-110m fraction in the compact outside the SiC layers was 3.3×10-2, indicating appreciable release of silver through the intact coatings and subsequent retention in the OPyC layers or matrix. The Eu-154 fraction was 2.4×10-4, which is equivalent to the inventory in one average particle, and indicates a small but measurable level of release from the intact coatings. Gamma counting of 61 individual particles indicated no particles with anomalously low fission product retention. The average ratio of measured inventory to calculated inventory was close to a value of 1.0 for several fission product isotopes (Ce-144, Cs-134, and Cs-137), indicating good retention and reasonably good agreement with the predicted inventories. Measured-to-calculated (M/C) activity ratios for fission products Eu-154, Eu-155, Ru-106, Sb

  13. Transients in reactors for power systems compensation

    Science.gov (United States)

    Abdul Hamid, Haziah

    This thesis describes new models and investigations into switching transient phenomena related to the shunt reactors and the Mechanically Switched Capacitor with Damping Network (MSCDN) operations used for reactive power control in the transmission system. Shunt reactors and MSCDN are similar in that they have reactors. A shunt reactor is connected parallel to the compensated lines to absorb the leading current, whereas the MSCDN is a version of a capacitor bank designed as a C-type filter for use in the harmonic-rich environment. In this work, models have been developed and transient overvoltages due to shunt reactor deenergisation were estimated analytically using MathCad, a mathematical program. Computer simulations used the ATP/EMTP program to reproduce both single-phase and three-phase shunt reactor switching at 275 kV operational substations. The effect of the reactor switching on the circuit breaker grading capacitor was also examined by considering various switching conditions.. The main original achievement of this thesis is the clarification of failure mechanisms occurring in the air-core filter reactor due to MSCDN switching operations. The simulation of the MSCDN energisation was conducted using the ATP/EMTP program in the presence of surge arresters. The outcome of this simulation shows that extremely fast transients were established across the air-core filter reactor. This identified transient event has led to the development of a detailed air-core reactor model, which accounts for the inter-turn RLC parameters as well as the stray capacitances-to-ground. These parameters are incorporated into the transient simulation circuit, from which the current and voltage distribution across the winding were derived using electric field and equivalent circuit modelling. Analysis of the results has revealed that there are substantial dielectric stresses imposed on the winding insulation that can be attributed to a combination of three factors. (i) First, the

  14. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  15. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

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

  16. Effects of Types and Dosages of Accelerator on the Properties of Impregnating Resin for Reactor%促进剂种类和用量对电抗器用浸渍树脂性能的影响

    Institute of Scientific and Technical Information of China (English)

    徐旭; 黄孙息; 王轶

    2012-01-01

    To solve the product quality problem caused by outflow of adhesive from large scale core reactor in the coil winding and curing process, the viscosity, gel time, impact strength and heat distortion temperature of impregnating resin with different curing systems were tested, and the effects of different accelerators and dosages on the properties of impregnating resin were studied. By making coil models of three types of curing systems, the resin loss amount in curing process and distribution uniformity of resin in coil after curing were tested. The results show that the three curing systems, 2, 4-EMI, DMP-30 and BDMA can meet the property requirements for the impregnating resin for reactor coil. The distribution uniformity of resin with BDMA curing system is better than that with 2, 4-EMI and DMP-30 curing systems, and the loss amount of resin with BDMA curing system is 0.47%, which is obviously lower than that with 2, 4-EMI and DMP-30 curing systems.%为解决大型空心电抗器线圈在绕制和固化过程中发生流胶而影响产品质量的问题,通过测试浸渍树脂不同固化体系的粘度、凝胶时间、冲击强度和热变形温度,研究了不同种类促进剂及其用量对浸渍树脂性能的影响,并通过制作线圈模型,测试了3种固化体系线圈模型固化过程的树脂流失量和固化后线圈内部的树脂分布均匀性.结果表明:2,4-EMI、DMP-30、BDMA 3种固化体系均能满足电抗器线圈用浸渍树脂的性能要求;BDMA固化体系对树脂分布均匀性的改善效果明显优于2,4-EMI、DMP-30固化体系,其树脂流失量为0.47%,明显小于2,4-EMI、DMP-30固化体系的树脂流失量.

  17. Data Compilation for AGR-3/4 Designed-to-Fail (DTF) Fuel Particle Batch LEU04-02DTF

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D [ORNL; Miller, James Henry [ORNL

    2008-10-01

    This document is a compilation of coating and characterization data for the AGR-3/4 designed-to-fail (DTF) particles. The DTF coating is a high density, high anisotropy pyrocarbon coating of nominal 20 {micro}m thickness that is deposited directly on the kernel. The purpose of this coating is to fail early in the irradiation, resulting in a controlled release of fission products which can be analyzed to provide data on fission product transport. A small number of DTF particles will be included with standard TRISO driver fuel particles in the AGR-3 and AGR-4 compacts. The ORNL Coated Particle Fuel Development Laboratory 50-mm diameter fluidized bed coater was used to coat the DTF particles. The coatings were produced using procedures and process parameters that were developed in an earlier phase of the project as documented in 'Summary Report on the Development of Procedures for the Fabrication of AGR-3/4 Design-to-Fail Particles', ORNL/TM-2008/161. Two coating runs were conducted using the approved coating parameters. NUCO425-06DTF was a final process qualification batch using natural enrichment uranium carbide/uranium oxide (UCO) kernels. After the qualification run, LEU04-02DTF was produced using low enriched UCO kernels. Both runs were inspected and determined to meet the specifications for DTF particles in section 5 of the AGR-3 & 4 Fuel Product Specification (EDF-6638, Rev.1). Table 1 provides a summary of key properties of the DTF layer. For comparison purposes, an archive sample of DTF particles produced by General Atomics was characterized using identical methods. This data is also summarized in Table 1.

  18. La normalización y certificación de tractores agrícolas en México

    Directory of Open Access Journals (Sweden)

    Alma Velia Ayala Garay

    2013-01-01

    Full Text Available El tractor agrícola es la principal fuente de potencia dentro de una unidad de producción. Por lo tanto para los usuarios de maquinaria agrícola, resulta una prioridad contar con mecanismos que permitan dar seguridad al usuario final o productor agrícola en el funcionamiento y calidad de los tractores, para esto el objetivo del presente es dar a conocer la evolución de la normalización y certificación de la maquinaria agrícola en México, sus impactos y tendencias, por medio de una descripción de conceptos y organismos internacionales que han sido referencia para México, así como la descripción de las actividades del CENEMA y OCIMA. Para esto se tomaron definiciones y datos estadísticos de cantidad de modelos de tractores certificados, cantidad de certificados otorgados sobre estructuras de protección por OCIMA, cantidad de motocultores e implementos certificados, dictaminaciones de tractores no favorables, de múltiples sectores. Como resultado de la certificación de la ma- quinaria agrícola se obtuvo hasta junio de 2013, existen 76 tractores certificados con una potencia a la “toma de fuerza” que varía de los 18,3 a los 130 hp. Por ultimo cabe destacar que la finalidad de OCIMA es respaldar al agricultor y avalar que se cumple con las normas mexicanas (NMX y las especificaciones necesarias para garantizar el buen desempeño de su equipo.

  19. Productividad del agua en algunos cultivos agrícolas en Cuba

    Directory of Open Access Journals (Sweden)

    Felicita González Robaina

    2014-01-01

    Full Text Available Para el mejor uso económico y social del agua se requieren de métodos para evaluar su productividad, a fin de tomar mejores decisiones en cuanto a políticas y estrategias de utilización de manera sostenible. Los administradores del agua para el riego necesitan iden- tificar tendencias en los patrones de uso y niveles de eficiencia con vistas a fijar metas y mejorar la productividad por unidad de volumen utilizado y/o consumido. El objetivo de este trabajo fue presentar los indicadores de productividad agronómica del agua de 17 cultivos a partir de la información obtenida en más de 100 experimentos de campo disponibles en la base de datos sobre necesidades hídricas del Instituto de Investigaciones de Ingeniería Agrícola. Se determinaron los rendimientos máximo y mínimo y sus respectivos valores de consumo, agua aplicada por riego y la productividad agronómica del agua aplicada para los 17 cultivos estudiados. Cultivos como los frijoles, maíz, boniato y los plátanos obtienen incrementos de rendimiento superiores al 100% cuando se aplica el agua necesaria para satisfacer su demanda hídrica, mientras que la yuca obtuvo valores menores del 50%. Los altos valores obtenidos de productividad en base al agua utilizada como riego no son un buen indicador del uso eficiente del agua por las plantas para las condiciones de la zona de estudio, donde se informan para los años analizados, porcentajes de lluvia aprovechable de hasta el 41%, supliendo ésta una parte importante de las necesidades de los cultivos agrícolas.

  20. CONTINUOUS PRODUCTION OF HYDROXYPROPYL STARCH IN A STATIC MIXER REACTOR

    NARCIS (Netherlands)

    LAMMERS, G; STAMHUIS, EJ; BEENACKERS, AACM

    1993-01-01

    A novel type of reactor for the chemical derivatization of starch pastes is presented. The design is based on the application of static mixers. The reactor shows excellent plug flow behaviour with a Peclet number of about 100. The viscosity behaviour of concentrated starch pastes in the static mixer

  1. A Fixed Bed Barrier Reactor with Separate Feed of Reactants

    NARCIS (Netherlands)

    Neomagus, H.W.J.P.; Saracco, G.; Versteeg, G.F.

    2001-01-01

    A new type of gas-solid reactor was developed and characterised in the series of reactor configurations with separate feed of reactants studied by our group. The novelty in the proposed design lies in the use of a fixed bed of small catalytic particles instead of a porous catalytic membrane. The maj

  2. Core Physics of Pebble Bed High Temperature Nuclear Reactors

    NARCIS (Netherlands)

    Auwerda, G.J.

    2014-01-01

    To more accurately predict the temperature distribution inside the reactor core of pebble bed type high temperature reactors, in this thesis we investigated the stochastic properties of randomly stacked beds and the effects of the non-homogeneity of these beds on the neutronics and thermal-hydraulic

  3. Analytical model of plasma-chemical etching in planar reactor

    Science.gov (United States)

    Veselov, D. S.; Bakun, A. D.; Voronov, Yu A.; Kireev, V. Yu; Vasileva, O. V.

    2016-09-01

    The paper discusses an analytical model of plasma-chemical etching in planar diode- type reactor. Analytical expressions of etch rate and etch anisotropy were obtained. It is shown that etch anisotropy increases with increasing the ion current and ion energy. At the same time, etch selectivity of processed material decreases as compared with the mask. Etch rate decreases with the distance from the centre axis of the reactor. To decrease the loading effect, it is necessary to reduce the wafer temperature and pressure in the reactor, as well as increase the gas flow rate through the reactor.

  4. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  5. Related activities on management of ageing of Dalat Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pham Van Lam [Reactor Dept., Nuclear Research Institute, Dalat (Viet Nam)

    1998-10-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the previous 250 kW TRIGA-MARK II reactor. The reactor core, the control and instrumentation system, the primary and secondary cooling systems as well as other associated systems were newly designed and installed. The renovated reactor reached its initial criticality in November 1983 and attained its nominal power of 500 kW in February 1984. Since then DNRR has been operated safely. Retained structures of the former reactor such as the reactor aluminum tank, the graphite reflector, the thermal column, the horizontal beam tubes and the radiation concrete shielding are 35 years old. During the recent years, in-service inspection has been carried out, the reactor control and instrumentation system were renovated due to ageing and obsolescence of its components, reactor general inspection and refurbishment were performed. Efforts are being made to cope with ageing of old reactor components to maintain safe operation of the DNRR. (author)

  6. Strengthening IAEA Safeguards for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reid, Bruce D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anzelon, George A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Budlong-Sylvester, Kory [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-01

    During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half a dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors, for example, such characteristics include rapid on-line refueling, and a core design with room for such a large number of assemblies or targets that it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors, such as hot cells, where plutonium could be separated, could pose a safeguards challenge because, in some cases, they are not declared (because they are not located in the facility or because nuclear materials are not foreseen to be processed inside) and may not be accessible to inspectors in States without an Additional Protocol in force.

  7. Valoración físico-química y fitotóxica de las aguas residuales depuradas/regeneradas destinadas al riego agrícola

    Directory of Open Access Journals (Sweden)

    J.A. Pérez Donato

    2015-01-01

    Full Text Available La aplicación más extendida de la reutilización de aguas residuales domésticas depuradas/regeneradas es el riego agrícol a. Las normativas comunes sobre calidad de aguas regeneradas para reutilización contemplan paráme tros físico-químicos y biológico s para minimizar riesgos sanitarios, pero pocas veces tienen en c uenta parámetros fitotóxicos que puede n ser limitantes para aplicacio nes agrícolas. Los parámetros físico-químicos y fitotóxicos cambian al avanzar en el grado de depuración/regeneración de las aguas residuales, por lo que resulta d e interés conocer que etapa de la instalación proporciona el agua a decuada para el riego. Este tr abajo presenta un estudio de correlaci ón de fitotoxicidad con parámet ros físico-químicos en agua de muestras tomadas al final de cada un a de las etapas de tratamiento conve ncional de aguas residuales domésticas, que incluye reactor b iológico de lodos activados y tratamiento terciario con filtración de arena, cloración, filtr ación en cartucho y desalinización por electrodiálisis reversible.En cad a muestra se midió: demanda química de oxígeno, pH, conductividad , carbono orgánico, fósforo total, proteínas solubles, y los ione s de Na, K, Ca, Mg, B, F y Cl. Los criter ios de fitotoxicidad se evaluar on mediante el biomarcador Lactuca sativa. Los resultados mostraron reducciones importantes de los pa rámetros físico-químicos en el proceso de electrodiálisis, mien tras que los efectos de fitotox icidad mostraron reducciones en los pro cesos anteriores a la etapa fin al de electrodiálisis. Se concluye que l os parámetros físico-químicos , conjuntamente con el marcador biológico, constituyen aspectos d e interés práctico que relaciona e l tipo de tratamiento con la ca lidad físico-química y la fitotoxicida d del agua depurada/regenerada obtenida y su idoneidad para aplicaciones de riego agrícola.

  8. Necesidades educacionales sobre riesgo de plaguicidas en el contexto socio-ambiental de las comunidades agrícolas de Sonora

    OpenAIRE

    M.I. Silveira Gramont; L. Aldana Madrid; A.I. Valenzuela Quintanar; C.B. Ochoa Nogales; G. Jasa-Silveira; Beatriz Camarena Gómez

    2016-01-01

    Para proponer estrategias de educación y comunicación sobre plaguicidas en las comunidades agrícolas de Sonora, se requiere situar la problemática de la exposición a plaguicidas en el contexto socio-cultural y ambiental de las comunidades afectadas, tomando en cuenta los factores que contribuyen al mayor riesgo de exposición de los trabajadores agrícolas y de sus familias. Este estudio examina los factores po