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Sample records for avr reactor juelich

  1. 10 years of current generation in the AVR reactor

    International Nuclear Information System (INIS)

    Schenk, P.; Nehrling, H.; Daeunert, U.; Schulten, R.

    1978-01-01

    On 17th December 1967, the experimental nuclear power plant (AVR) in Juelich supplied RWE network with power for the first time. With the start of the power operation of the first German high-temperature reactor (HTR), a milestone was reached in the development of this new and progressive line of construction. On the same day exactly 10 years later, the successful work with the hottest nuclear reactor in the world was reviewed in the presence of 15 associates of the Arbeitsgemeinschaft Versuchsreaktor (AVR) Ltd. and the personnel of the experimental nuclear power plant at a festival event in the main auditorium of the nuclear power plant at Juelich before some 300 guests from central and local government, the board of control, representatives of the population of the Dueren area and the town of Juelich, as well as bodies of the power producing industry. (orig.) [de

  2. 10 years of current generation in the AVR reactor. The high-temperature pebble-bed reactor - a hot tip for our future

    Energy Technology Data Exchange (ETDEWEB)

    Schenk, P [Arbeitsgemeinschaft Versuchs-Reaktor G.m.b.H., Juelich (Germany, F.R.); A G, Stadtwerke Duesseldorf [Germany, F.R.; Vereinigung Deutscher Elektrizitaetswerke e.V. (VDEW), Frankfurt am Main (Germany, F.R.)); Nehrling, H [Ministerium fuer Wirtschaft, Mittelstand und Verkehr des Landes Nordrhein-Westfalen, Duesseldorf (Germany, F.R.); Daeunert, U [Bundesministerium fuer Forschung und Technologie, Bonn-Bad Godesberg (Germany, F.R.); Schulten, R [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung; Mattick, W [Brown, Boveri und Cie A.G., Mannheim (Germany, F.R.)

    1978-02-01

    On 17th December 1967, the experimental nuclear power plant (AVR) in Juelich supplied RWE network with power for the first time. With the start of the power operation of the first German high-temperature reactor (HTR), a milestone was reached in the development of this new and progressive line of construction. On the same day exactly 10 years later, the successful work with the hottest nuclear reactor in the world was reviewed in the presence of 15 associates of the Arbeitsgemeinschaft Versuchsreaktor (AVR) Ltd. and the personnel of the experimental nuclear power plant at a festival event in the main auditorium of the nuclear power plant at Juelich before some 300 guests from central and local government, the board of control, representatives of the population of the Dueren area and the town of Juelich, as well as bodies of the power producing industry.

  3. The importance of the AVR pebble-bed reactor for the future of nuclear power

    International Nuclear Information System (INIS)

    Pohl, P.

    2006-01-01

    The AVR pebble-bed high temperature gas-cooled reactor (HTGR) at Juelich (Germany)) operated from 1967 to 1988 and was certainly the most important HTGR project of the past. The reactor was the mass test bed for all development steps of HTGR pebble fuel. Some early fuel charges failed under high temperature conditions and contaminated the reactor. An accurate pebble measurement (Cs 137) allowed to clean the core from unwanted pebbles after 1981. The coolant activity went down and remained very low for the remaining reactor operation. A melt-wire experiment in 1986 revealed max. coolant temperatures of >1280 deg. C and fuel temperatures of >1350 deg. C, explained by under-estimated bypasses. The fuel still in the core achieved high burn-ups and showed under the extreme temperature conditions excellent fission product retention. Thus, the AVR operation qualified the HTGR fuel, and an average discharge burn-up of 112% fifa revealed an excellent fuel economy of the pebble-bed reactor. Furthermore, the AVR operation offers many meaningful data for code-to-experiment comparisons. (authors)

  4. Combined conditioning in the high-temperature experimental nuclear reactor (AVR) at Juelich

    International Nuclear Information System (INIS)

    Nieder, R.; Vey, K.; Ivens, G.

    1984-01-01

    The high temperature experimental nuclear reactor (AVR) is the first nuclear power plant in which combined cycle operation has been introduced. The water-steam cycle has been operated for about 15 years according to the alkali method of working with ammonia and hydrazine. The VGB-guidelines have been adhered to througout. Since January 1983 cobined cycle operation has been employed, and in this process a pH-value of about 8.5 and an oxygen concentration of about 200 μg/kg in the feedwater have been used. A distinct reduction of tritium concentration in the water-steam cycle was the outstanding new result. (orig.) [de

  5. Compilation of reactor physics data of the year 1984, AVR reactor

    International Nuclear Information System (INIS)

    Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

    1985-12-01

    The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1984 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

  6. Compilation of reactor physics data of the year 1983, AVR reactor

    International Nuclear Information System (INIS)

    Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

    1985-06-01

    The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1983 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

  7. A safeguards concept for the AVR fuel element storage areas at the KFA-Juelich

    International Nuclear Information System (INIS)

    Canty, M.J.; Buttler, R.

    1980-11-01

    The storage of spent AVR fuel in the KFA-Juelich has been discussed in relation to the obligations of the FRG under NPT. The present system of material accountancy and the associated procedures for physical inventory taking, while adequate from the operational standpoint, fall short of providing sufficient safequards for the fissile material involved. It is essential to complement existing controls by providing the safeguards authorities with the means of verifying the nuclear materials accountancy data of the storage facility operators. Due to the difficulties associated with the assay of irradiated fuel, the verification measurements must be carried out with the close cooperation of the operators. It was demonstrated that, given appropriate measuring devices, a high assurance for the non-diversion of a significant quantity can be obtained with an acceptable manpower effort. In this regard, the highly diluted form of the fissile material plays a favourable role. (orig.) [de

  8. The HTR 500 concept based on pratical THTR and AVR experience

    International Nuclear Information System (INIS)

    Wachholz, W.; Weicht, U.

    1988-01-01

    This paper discusses progress during the past ten years in the development of a specific HTR safety concept. This has been mainly characterized by the abandonment of the LWR specific safety principles and making use of the safety characteristics typical of the high-temperature reactor (HTR). In the design, construction and operation of high-temperature reactors - especially AVR (15 MWe plant in Juelich, FRG) and THTR (300 MWe plant in Hamm-Uentrop, FRG) - experience has been gained in the field of accident topology and plant risk of HTRs in recent years. This experience, based on detailed accident analyses performed by manufacturers and experts, is relevant for the entire HTR line independent of specific projects. The authors focus on the HTR 500, the first commercial high temperature reactor with a pebble bed core. Its design principles and the design of its systems are based on the earlier AVR and THTR projects

  9. Compilation of reactor-physical data of the AVR experimental reactor for 1982

    International Nuclear Information System (INIS)

    Werner, H.; Wawrzik, U.; Grotkamp, T.; Buettgen, I.

    1983-12-01

    Since the end of 1981 the calculation model AVR-80 has been taken as a basis for compiling reactor-physical data of the AVR experimental reactor. A brief outline of the operation history of 1982 is given, including the beginning of a large-scale experiment dealing with change-over from high enriched uranium to low enriched uranium. Calculations relative to spectral shift, diffusion, temperature, burnup, and recirculation of the fuel elements are described in brief. The essential results of neutron-physical and thermodynamic calculations and the characteristical data of the various types of fuel used are shown in tables and illustrations. (RF) [de

  10. The present state of the HTR concept based on experience gained from AVR and THTR

    International Nuclear Information System (INIS)

    Wachholz, W.

    1989-01-01

    During the past ten years the development of a specific HTR concept has made remarkable progress. This has been mainly characterized by making use of the safety characteristics typical of the High-Temperature Reactor (HTR). In the design, construction and operation of High-Temperature Reactors - especially AVR (15 MWe plant in Juelich, FRG) and THTR (300 MWe plant in Hamm-Uentrop, FRG) - comprehensive experience has been gained in the field of operational availability and safety, accident topology and plant risk of HTRs in recent years. This experience is relevant for the entire HTR line independent of specific projects. (author). 3 refs, 5 figs, 1 tab

  11. The Siemens-Argonaut reactor as a driver zone for a high-temperature reactor cell. Der Siemens-Argonaut-Reaktor als Treiberzone fuer eine Hochtemperaturreaktorzelle

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, H; Schuerrer, F; Ninaus, W; Oswald, K; Rabitsch, H; Kreiner, H [Technische Univ., Graz (Austria). Inst. fuer Theoretische Physik und Reaktorphysik; Neef, R D [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung

    1984-12-15

    To enable a validation of neutron physics calculation methods for pebble bed reactors the inner reflector of an Argonaut research reactor was substituted by a full of about 1200 fuel elements of the AVR-Juelich type. The report describes the measuring instruments and the reactor physical layout of the arrangement by the code packages GAMTEREX, ZUT-D.G.L. and MUPO. Comparison of calculated reaction rates with measurements show good agreement. Application of the codes to high-temperature reactors in abnormal states is envisaged. (Author, translated by G.Q.)

  12. Leakage in the Juelich research reactor

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    On August 17, 1978, a leakage occurred in the DIDO research reactor. Early in the afternoon of this day, the valves of the coolant loop had been checked with the reactor shut off. When the mechanics wanted to oil a tight valve and drilled a hole in the valve cover for this, heavy water started to leak (leakage in the valve membrane). The mechanics left the shielded valve space at once; directly after having a shower, they underwent a radiation protection examination. It was found that none of the mechanics had been exposed to an excessive dose. When other mechanics in protective suits had closed the leak in the valve, a total of 150 liters had leaked into the sump pump at the valve entrance. They were pumped back into the cooling system. About 5 liters of water were evaporated and, via the stack, escaped into the environment. The activity released was about 40 curie; this is less than the permissible amount of 60 curie per week during normal operation. Neither the KFA personnel nor the inhabitants of Juelich and its surroundings were in danger at any moment. Calculations so far yield a maximum radiation exposure below 1mrem at the point of maximum exposure. The cooling circuit could be entered again only one day after the incident. The present shut-off phase of the reactor is not unduly prolonged by this accident. (orig./HP) [de

  13. Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    Rainer Moormann

    2008-01-01

    Full Text Available Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA. The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1  m, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 hours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.

  14. A safety re-evaluation of the AVR pebble bed reactor operation and its consequences for future HTR concepts

    Energy Technology Data Exchange (ETDEWEB)

    Moormann, R.

    2008-06-15

    The AVR pebble bed reactor (46 MW{sub th}) was operated 1967-88 at coolant outlet temperatures up to 990 C. A principle difference of pebble bed HTRs as AVR to conventional reactors is the continuous movement of fuel element pebbles through the core which complicates thermohydraulic, nuclear and safety estimations. Also because of a lack of other experience AVR operation is still a relevant basis for future pebble bed HTRs and thus requires careful examination. This paper deals mainly with some insufficiently published unresolved safety problems of AVR operation and of pebble bed HTRs but skips the widely known advantageous features of pebble bed HTRs. The AVR primary circuit is heavily contaminated with metallic fission products (Sr-90, Cs-137) which create problems in current dismantling. The amount of this contamination is not exactly known, but the evaluation of fission product deposition experiments indicates that the end of life contamination reached several percent of a single core inventory, which is some orders of magnitude more than precalculated and far more than in large LWRs. A major fraction of this contamination is bound on graphitic dust and thus partly mobile in depressurization accidents, which has to be considered in safety analyses of future reactors. A re-evaluation of the AVR contamination is performed here in order to quantify consequences for future HTRs (400 MW{sub th}). It leads to the conclusion that the AVR contamination was mainly caused by inadmissible high core temperatures, increasing fission product release rates, and not - as presumed in the past - by inadequate fuel quality only. The high AVR core temperatures were detected not earlier than one year before final AVR shut-down, because a pebble bed core cannot yet be equipped with instruments. The maximum core temperatures are still unknown but were more than 200 K higher than calculated. Further, azimuthal temperature differences at the active core margin of up to 200 K were

  15. A safety re-evaluation of the AVR pebble bed reactor operation and its consequences for future HTR concepts

    International Nuclear Information System (INIS)

    Moormann, R.

    2008-06-01

    The AVR pebble bed reactor (46 MW th ) was operated 1967-88 at coolant outlet temperatures up to 990 C. A principle difference of pebble bed HTRs as AVR to conventional reactors is the continuous movement of fuel element pebbles through the core which complicates thermohydraulic, nuclear and safety estimations. Also because of a lack of other experience AVR operation is still a relevant basis for future pebble bed HTRs and thus requires careful examination. This paper deals mainly with some insufficiently published unresolved safety problems of AVR operation and of pebble bed HTRs but skips the widely known advantageous features of pebble bed HTRs. The AVR primary circuit is heavily contaminated with metallic fission products (Sr-90, Cs-137) which create problems in current dismantling. The amount of this contamination is not exactly known, but the evaluation of fission product deposition experiments indicates that the end of life contamination reached several percent of a single core inventory, which is some orders of magnitude more than precalculated and far more than in large LWRs. A major fraction of this contamination is bound on graphitic dust and thus partly mobile in depressurization accidents, which has to be considered in safety analyses of future reactors. A re-evaluation of the AVR contamination is performed here in order to quantify consequences for future HTRs (400 MW th ). It leads to the conclusion that the AVR contamination was mainly caused by inadmissible high core temperatures, increasing fission product release rates, and not - as presumed in the past - by inadequate fuel quality only. The high AVR core temperatures were detected not earlier than one year before final AVR shut-down, because a pebble bed core cannot yet be equipped with instruments. The maximum core temperatures are still unknown but were more than 200 K higher than calculated. Further, azimuthal temperature differences at the active core margin of up to 200 K were observed

  16. CONDITIONING OF INTERMEDIATE-LEVEL WASTE AT FORSCHUNGSZENTRUM JUELICH GMBH

    International Nuclear Information System (INIS)

    Krumbach, H.

    2003-01-01

    This contribution to the group of low-level, intermediate, mixed and hazardous waste describes the conditioning of intermediate-level mixed waste (dose rate above 10 mSv/h at the surface) from Research Centre Juelich (FZJ). Conditioning of the waste by supercompaction is performed at Research Centre Karlsruhe (FZK). The waste described is radioactive waste arising from research at Juelich. This waste includes specimens and objects from irradiation experiments in the research reactors Merlin (FRJ-1) and Dido (FRJ-2) at FZJ. In principle, radioactive waste at Forschungszentrum Juelich GmbH is differentiated by the surface dose rate at the waste package. Up to a surface dose rate of 10 mSv/h, the waste is regarded as low-level. The radioactive waste described here has a surface dose rate above 10 mSv/h. Waste up to 10 mSv/h is conditioned at the Juelich site according to different conditioning methods. The intermediate-level waste can only be conditioned by supercompaction in the processing facility for intermediate-level waste from plant operation at Research Centre Karlsruhe. Research Centre Juelich also uses this waste cell to condition its intermediate-level waste from plant operation

  17. Numerical investigation of the 3-dimensional steady-state temperature- and flow distribution in the core of a pebble bed high temperature reactor

    International Nuclear Information System (INIS)

    Verfondern, K.

    1983-01-01

    This work presents a computer model determining the steady-state temperature- and flow field in 3 dimensions in the core of a pebble bed high temperature reactor. The numerical sprinkler method, basind on the Thermix-model, allows to describe the thermo-hydraulics of a non-rotational-symmetric core-geometry. The AVR-reactor in Juelich, in operation since 1967, represents a suitable investigation-object for the computer model of Thermix-3D. It is in a 3D-mesh-structure to reproduce very precisely the so called ''graphite noses'', in which the shut-down rods are conducted as well as the filling cones in the inner and outer area. The results of the final calculation of the normal operation condition for the AVR-reactor unambiguously show, that within the core reproduced in 3 dimensions there are evident deviations in the flow profile and in the temperatures of the cooling gas in contrast to a 2D-handling. (orig.) [de

  18. Conditioning of intermediate-level waste at Forschungszentrum Juelich GmbH

    International Nuclear Information System (INIS)

    Krumbach, H.

    2003-01-01

    This contribution to the group of low-level, intermediate, mixed and hazardous waste describes the conditioning of intermediate-level mixed waste (dose rate above 10 mSv/h at the surface) from Research Centre Juelich (FZJ). Conditioning of the waste by supercompaction is performed at Research Centre Karlsruhe (FZK). The waste described is radioactive waste arising from research at Juelich. This waste includes specimens and objects from irradiation experiments in the research reactors Merlin (FRJ-1) and Dido (FRJ-2) at FZJ. In principle, radioactive waste at Forschungszentrum Juelich GmbH is differentiated by the surface dose rate at the waste package. Up to a surface dose rate of 10 mSv/h, the waste is regarded as low-level. The radioactive waste described here has a surface dose rate above 10 mSv/h. Waste up to 10 mSv/h is conditioned at the Juelich site according to different conditioning methods. The intermediate-level waste can only be conditioned by supercompaction in the processing facility for intermediate-level waste from plant operation at Research Centre Karlsruhe. Research Centre Juelich also uses this waste cell to condition its intermediate-level waste from plant operation. (orig.)

  19. Decommissioning of the AVR reactor, concept for the total dismantling

    International Nuclear Information System (INIS)

    Marnet, C.; Wimmers, M.; Birkhold, U.

    1998-01-01

    After more than 21 years of operation, the 15 MWe AVR experimental nuclear power plant with pebble bed high temperature gas-cooled reactor was shout down in 1988. Safestore decommissioning began in 1994. In order to completely dismantle the plant, a concept for Continued dismantling was developed according to which the plant could be dismantled in a step-wise procedure. After each step, there is the possibility to transform the plant into a new state of safe enclosure. The continued dismantling comprises three further steps following Safestore decommissioning: 1. Dismantling the reactor vessels with internals; 2. Dismantling the containment and the auxiliary units; 3. Gauging the buildings to radiation limit, release from the validity range of the AtG (Nuclear Act), and demolition of the buildings. For these steps, various technical procedures and concepts were developed, resulting in a reference concept in which the containment will essentially remain intact (in-situ concept). Over the top of the outer reactor vessel a disassembling area for remotely controlled tools will be erected that tightens on that vessel and can move down on the vessel according to the dismantling progress. (author)

  20. Testing of HTR UO{sub 2} TRISO fuels in AVR and in material test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kania, Michael J., E-mail: MichaelJKania@googlemail.com [Retired from Lockheed Martin Corp, 20 Beach Road, Averill Park, NY 12018 (United States); Nabielek, Heinz, E-mail: heinznabielek@me.com [Retired from Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Verfondern, Karl [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Allelein, Hans-Josef [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, 52072 Aachen (Germany)

    2013-10-15

    The German High Temperature Reactor Fuel Development Program successfully developed, licensed and manufactured many thousands of spherical fuel elements that were used to power the experimental AVR reactor and the commercial THTR reactor. In the 1970s, this program extended the performance envelope of HTR fuels by developing and qualifying the TRISO-coated particle system. Irradiation testing in real-time AVR tests and accelerated MTR tests demonstrated the superior manufacturing process of this fuel and its irradiation performance. In the 1980s, another program direction change was made to a low enriched UO{sub 2} TRISO-coated particle system coupled with high-quality manufacturing specifications designed to meet new HTR plant design needs. These needs included requirements for inherent safety under normal operation and accident conditions. Again, the German fuel development program met and exceeded these challenges by manufacturing and qualifying the low-enriched UO{sub 2} TRISO-fuel system for HTR systems with steam generation, gas-turbine systems and very high temperature process heat applications. Fuel elements were manufactured in production scale facilities that contained near defect free UO{sub 2} TRISO coated particles, homogeneously distributed within a graphite matrix with very low levels of uranium contamination. Good irradiation performance for these elements was demonstrated under normal operating conditions to 12% FIMA and under accident conditions not exceeding 1600 °C.

  1. Radioactivity monitoring by the official monitoring stations in North-Rhine Westphalia and the Juelich Nuclear Research Centre after the Chernobyl reactor accident

    International Nuclear Information System (INIS)

    1986-01-01

    This official report presents a governmental declaration of the prime minister of NRW, Mr. Rau, concerning the reactor accident at Chernobyl, and a joint declaration of ministers of NRW, concerning the impact of the accident on the Land NRW. These statements are completed by six official reports on radioactivity measurements carried out by the official monitoring stations of the Land and by the KFA Juelich. These reports inform about methods, scope, and results of the measuring campaigns accomplished by the Zentralstelle fuer Sicherheitstechnik (ZFS), the public materials testing office (MPA), the Chemisches Untersuchungsamt, the Landesamt fuer Wasser und Abfall, and the KFA Juelich. (DG) [de

  2. The programme 'fission product deposition' at the IRB of Juelich nuclear research centre

    International Nuclear Information System (INIS)

    Gottaut, H.; Iniotakis, N.; Malinowski, J.; Muenchow, K.H.; Sackmann, B.

    1976-01-01

    The transport and deposition behaviour of the non-gaseous fission and activation products in the primary circuit of HTR-type reactors determines the possibility of inspection and maintenance of single components of the primary circuit as well as the safety of the reactor in normal operation and during accidents. For the investigation of these problems, the programme 'fission product deposition' was started at Juelich nuclear research centre in 1969 in cooperation with a number of industrial firms. The programme covers in-pile and out-of-pile experiments, in which the HTR conditions are simulated as realistically as possible, as well as various laboratory experiments and extensive theoretical studies. It is the objective of this work to establish a realistic physical model and computer programme with which the transport and deposition of nuclides in the primary circuit of HTR reactors can be calculated in advance. A report is given on the experimental and theoretical studies carried out at the IRB of Juelich nuclear research centre. (orig./AK) [de

  3. Twenty-fifth anniversary of the Juelich Nuclear Research Center

    International Nuclear Information System (INIS)

    Haefele, W.

    1982-01-01

    On December 10, 1981, KFA Juelich celebrated its 25th year of existence; on December 11, 1956, the land parliament of North Rhine Westphalia had decided in favour of the erection of a joint nuclear research facility of the land of North Rhine Westphalia. In contrast to other nuclear research centers, the Juelich centre was to develop and operate large-scale research equipment and infrastructure for joint use by the universities of the land. This cooperation has remained an important characteristic in spite of the independent scientific work of KFA institutes, Federal government majorities, and changes in research fields and tasks. KFA does fundamental research in nuclear and plasma physics, solid state research, medicine, life sciences, and environmental research; other activities are R + D tasks for the HTR reactor and its specific applications as well as energy research in general. (orig.) [de

  4. Safety concept of high-temperature reactors based on the experience with AVR and THTR

    International Nuclear Information System (INIS)

    Wachholz, Winfried; Kroeger, Wolfgang

    1990-01-01

    In the Federal Republic of Germany a reactor is considered safe if verification has been furnished that the requirements contained in paragraph 7 of the Federal German Atomic Energy Act are met for this reactor: demonstration of sufficient precautions against damage required according to the actual state of the art, and especially compliance with the dose rate limits for normal operation and accidental conditions. These requirements result in a deterministic multi-stage safety concept with specified requirements for the engineered safety systems. In recent years, proposals for enhanced safety of nuclear power reactors or a radical change in safety philosophy have been made. This is characterised by 'inherently safe', 'super safe' and similar slogans. A quantitative definition of these requirements has not yet been established, but it is clear as a common objective that the event of beyond design basis accidents evacuation, relocation, and large scale contamination of ground should not occur. As a consequence of the Chernobyl accident the safety of all the NPPs in Germany has been reviewed. This analysis was completed for the THTR reactor in 1988. The same has been done for AVR reactor. The final evaluation of the HTR specific safety features have been fully confirmed. The HTR concepts under development are based on this experience. The HTR-Modul unit is currently being designed

  5. Forschungszentrum Juelich. Annual report 2013; Forschungszentrum Juelich. Jahresbericht 2013

    Energy Technology Data Exchange (ETDEWEB)

    Frick, Frank; Roegener, Wiebke

    2014-07-15

    The annual report 2013 of the Forschungszentrum Juelich covers research activities, including high-lights of brain science, electrically controllable quantum bits, climate science and atmosphere research, knowledge management, including education and international cooperation, and an economic survey.

  6. Scientific and technical publications of Juelich Research Centre, January 1988-June 1992

    International Nuclear Information System (INIS)

    1992-01-01

    The scientific and technical publications of Juelich Research Centre from January 1988 through June 1992 are listed under the following headings: General publications; Mathematics, computers, cybernetics; General physics; Atomic and nuclear physics; Solid state physics; Materials; Analytical chemistry; Engineering; Reactor Engineering; Metrology; Biology; Biotechnology; Agriculture; Energy; Medicine; Ecology; Plasma physics and fusion reactor technology; Physico-chemistry; Nuclear chemistry and radiochemistry; Chemical engineering; Electrical engineering, electronics; Geosciences. There is an author's index with report numbers (JUEL-, JUEL-BIBL-, JUEL-CONF, JUEL-SPEZ). (orig./BBR) [de

  7. List of reports of the Kernforschungsanlage Juelich published between October 1972 and December 1975

    International Nuclear Information System (INIS)

    1976-03-01

    This is a compilation of the Juel reports published by Kernforschungsanlage Juelich between October, 1972, and December, 1975. The compilation was made by automatic data processing. It applies to the fields mathematics, computers, kybernetics; general and solid state physics; plasma physics and fusion reactor technology; atomic and nuclear physics; reactor engineering, reactor physics, applied nuclear physics; thermionics and technical physics, organic, physical, radio-radiation and nuclear chemistry, analytical chemistry; chemical engineering, materials, general engineering; instruments and methods of measurement; biology and agriculture; medicine; environment, radiation protection; energy problems. (HK) [de

  8. Forschungszentrum Juelich. Annual report 2016; Forschungszentrum Juelich. Jahresbericht 2016

    Energy Technology Data Exchange (ETDEWEB)

    Frick, Frank; Lueers, Katja; Roegener, Wiebke; Stahl-Busse, Brigitte

    2017-07-15

    The annual report 2016 of the Forschungszentrum Juelich covers research activities, including high-lights of structural biochemistry (Alzheimer research), material research (skyrmions), computer simulation (e.g. of flexible blood cells), quantum physics (100 qubit era), photovoltaics, battery research, environmental research, climate research, biotechnology and community codes, including education and international cooperation.

  9. Identification of proteins similar to AvrE type III effector proteins from ...

    African Journals Online (AJOL)

    Type III effector proteins are injected into host cells through type III secretion systems. Some effectors are similar to host proteins to promote pathogenicity, while others lead to the activation of disease resistance. We used partial least squares alignment-free bioinformatics methods to identify proteins similar to AvrE proteins ...

  10. Installation of a three-dimensional simulation method for core-physical description of pebble bed reactors with multiple recycling process at the example of the AVR

    International Nuclear Information System (INIS)

    Grotkamp, T.

    1984-01-01

    To describe the core-physical behaviour of pebble bed reactors simulation models are used, which reproduce the burn-up/recycling and - resulting - calculate criticality and neutron spectrum as well as neutron flux and temperature distribution. Modelling the AVR-reactor requires a three-dimensional treating for detailed considerations because of the graphite noses extending into the core. Such a system is built up in the present work and compared with the results of the two-dimensional model standardizated from operational side. The agreement is so good that the latter one is sufficient for the calculations accompanying the operation. The comparison with results of measurement is very satisfying in regard to fuel element distribution and temperature coefficient. As in all theoretical investigations there stays a discrepance of a little more than 1 nile against the measurement at the reactivity equivalence of the AVR rod-bank. On the other hand it is possible to reproduce the rod-bank curve resulting of the calibration very exactly with the present model. (orig.) [de

  11. Analysis of AVR4 promoter by sequential response-element deletion ...

    African Journals Online (AJOL)

    An Avr4 promoter region ligated to chloramphenicol acetyltransferase plasmid vector (pBLCAT2) to produce recombinant plasmid Avr4pBLCAT2 was sequentially deleted to produce five distinct mutants: Avr4pBLCAT2907-176, Avr4pBLCAT2809-176, Avr4pBLCAT2789-176, Avr4pBLCAT2429-176 and Avr4pBLCAT2 ...

  12. The Activation of Phytophthora Effector Avr3b by Plant Cyclophilin is Required for the Nudix Hydrolase Activity of Avr3b.

    Science.gov (United States)

    Kong, Guanghui; Zhao, Yao; Jing, Maofeng; Huang, Jie; Yang, Jin; Xia, Yeqiang; Kong, Liang; Ye, Wenwu; Xiong, Qin; Qiao, Yongli; Dong, Suomeng; Ma, Wenbo; Wang, Yuanchao

    2015-08-01

    Plant pathogens secrete an arsenal of effector proteins to impair host immunity. Some effectors possess enzymatic activities that can modify their host targets. Previously, we demonstrated that a Phytophthora sojae RXLR effector Avr3b acts as a Nudix hydrolase when expressed in planta; and this enzymatic activity is required for full virulence of P. sojae strain P6497 in soybean (Glycine max). Interestingly, recombinant Avr3b produced by E. coli does not have the hydrolase activity unless it was incubated with plant protein extracts. Here, we report the activation of Avr3b by a prolyl-peptidyl isomerase (PPIase), cyclophilin, in plant cells. Avr3b directly interacts with soybean cyclophilin GmCYP1, which activates the hydrolase activity of Avr3b in a PPIase activity-dependent manner. Avr3b contains a putative Glycine-Proline (GP) motif; which is known to confer cyclophilin-binding in other protein substrates. Substitution of the Proline (P132) in the putative GP motif impaired the interaction of Avr3b with GmCYP1; as a result, the mutant Avr3bP132A can no longer be activated by GmCYP1, and is also unable to promote Phytophthora infection. Avr3b elicits hypersensitive response (HR) in soybean cultivars producing the resistance protein Rps3b, but Avr3bP132A lost its ability to trigger HR. Furthermore, silencing of GmCYP1 rendered reduced cell death triggered by Avr3b, suggesting that GmCYP1-mediated Avr3b maturation is also required for Rps3b recognition. Finally, cyclophilins of Nicotiana benthamiana can also interact with Avr3b and activate its enzymatic activity. Overall, our results demonstrate that cyclophilin is a "helper" that activates the enzymatic activity of Avr3b after it is delivered into plant cells; as such, cyclophilin is required for the avirulence and virulence functions of Avr3b.

  13. Activities at Forschungszentrum Juelich in Safeguards Analytical Techniques and Measurements

    International Nuclear Information System (INIS)

    Duerr, M.; Knott, A.; Middendorp, R.; Niemeyer, I.; Kueppers, S.; Zoriy, M.; Froning, M.; Bosbach, D.

    2015-01-01

    The application of safeguards by the IAEA involves analytical measurements of samples taken during inspections. The development and advancement of analytical techniques with support from the Member States contributes to strengthened and more efficient verification of compliance with non-proliferation obligations. Since recently, a cooperation agreement has been established between Forschungszentrum Juelich and the IAEA in the field of analytical services. The current working areas of Forschungszentrum Juelich are: (i) Production of synthetic micro-particles as calibration standard and reference material for particle analysis, (ii) qualification of the Forschungszentrum Juelich as a member of the IAEA network of analytical laboratories for safeguards (NWAL), and (iii) analysis of impurities in nuclear material samples. With respect to the synthesis of particles, a dedicated setup for the production of uranium particles is being developed, which addresses the urgent need for material tailored for its use in quality assurance and quality control measures for particle analysis of environmental swipe samples. Furthermore, Forschungszentrum Juelich has been nominated as a candidate laboratory for membership in the NWAL network. To this end, analytical capabilities at Forschungszentrum Juelich have been joined to form an analytical service within a dedicated quality management system. Another activity is the establishment of analytical techniques for impurity analysis of uranium-oxide, mainly focusing on inductively coupled mass spectrometry. This contribution will present the activities at Forschungszentrum Juelich in the area of analytical measurements and techniques for nuclear verification. (author)

  14. Forschungszentrum Juelich. Annual report 2015

    International Nuclear Information System (INIS)

    Frick, Frank; Lueers, Katja; Roegener, Wiebke; Stettien, Annette; Trautwein, Ilse; Stahl-Busse, Brigitte

    2016-07-01

    The annual report 2015 of the Forschungszentrum Juelich covers research activities, including high-lights of brain science, electrically controllable quantum bits, climate science and atmosphere research, knowledge management, including education and international cooperation, and an economic survey.

  15. Forschungszentrum Juelich. Annual report 2013

    International Nuclear Information System (INIS)

    Frick, Frank; Roegener, Wiebke

    2014-07-01

    The annual report 2013 of the Forschungszentrum Juelich covers research activities, including high-lights of brain science, electrically controllable quantum bits, climate science and atmosphere research, knowledge management, including education and international cooperation, and an economic survey.

  16. Program status of the high temperature reactor development in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    1984-01-01

    The status of the HTR development program in the Federal Republic of Germany in 1984 is characterized by the beginning of a transition phase from a national program to a commercial program. In the last 20 years the HTR technology program was strongly, nearly completely supported by the Federal Government and the State Government of North-Rhine-Westfalia. Funding of the program up to now exceeded 5 billion DM. Within this framework it was possible to establish competent-reactor-system companies, to enable industries to supply HTR- specific components including fuel elements and nuclear graphites, to maintain the strong engagement of the national centre KFA Juelich in general R and D activities, to build and operate the AVR-plant for more than 16 years, to erect the demonstration plant THTR-300 now approaching completion and to build and operate many efficient test facilities. Thereby the HTR technology development achieved a stage of maturity which is not only considered to be most advanced, but is also ready now for commerical deployment. The assessment report which comprised both the fast breeder and the HTR development included all major impacts, such as history, status, prospects, benefits, industrial aspects and international developments of the technology. The program description is facilitated by distinguishing the five major program elements: AVR, THTR-300, THTR follow-up plant, nuclear process heat program, fuel cycle activities

  17. Radiochemical analysis in the nuclear research establishment (KFA) Juelich, FRG

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    KFA Juelich is one of the two great nuclear research centres of the Federal Republic of Germany. About 3700 employees including about 700 scientists are engaged in a great number of programs and projects belonging to six main fields of research and development: high temperature reactor and energy techniques; nuclear fusion; properties of materials; materials research; life and environment; methods. In the article the radiochemical analysis work of the former Central Institute of Analytical Chemistry and its two successors is described: activation analysis, application of tracer techniques, fission product analysis. Further on the irradiation facilities are described, a short survey is given on the instrumentation, and the future work is outlined. (T.G.)

  18. Gas chromatographic measurement in water-steam circuits

    International Nuclear Information System (INIS)

    Zschetke, J.; Nieder, R.

    1984-01-01

    A gas chromatographic technique for measurements in water-steam circuits, which has been well known for many years, has been improved by design modifications. A new type of equipment developed for special measuring tasks on nuclear engineering plant also has a general application. To date measurements have been carried out on the ''Otto Hahn'' nuclear powered ship, on the KNK and AVR experimental nuclear power plants at Karlsruhe and Juelich respectively and on experimental boiler circuits. The measurements at the power plants were carried out under different operating conditions. In addition measurements during the alkali operating mode and during combined cycle operation were carried out on the AVR reactor. It has been possible to draw new conclusion from the many measurements undertaken. (orig.) [de

  19. The new polarized neutron reflectometer in Juelich

    International Nuclear Information System (INIS)

    Ruecker, U.; Alefeld, B.; Bergs, W.; Kentzinger, E.; Brueckel, T.

    1999-01-01

    On the basis of the HADAS spectrometer in the guide hall of the Juelich research reactor FRJ-2 a polarized neutron reflectometer is build with a 2D-position sensitive detector system. The new spectrometer is optimized for reflectivity and diffuse magnetic scattering measurements with small incident angles on thin magnetic films with thicknesses in the nm range. The polarization analyzer covers the whole detector area, so that a range of 2.5 deg in the scattering angle can be measured simultaneously. The analyzer consists of a stack of supermirrors tilted against the scattering plane. In this reflection geometry, the momentum transfer resolution of the instrument is not reduced, but the sample height is limited to 17 mm. For the monochromator, polarizer and collimation different setups have been compared on the basis of Monte-Carlo calculations: a focusing elliptical supermirror monochromator, a cylindrical mirror, a focusing pyrolytic graphite double monochromator and a double monochromator with bent perfect Si crystals. (author)

  20. Forschungszentrum Juelich. Annual report 2016

    International Nuclear Information System (INIS)

    Frick, Frank; Lueers, Katja; Roegener, Wiebke; Stahl-Busse, Brigitte

    2017-07-01

    The annual report 2016 of the Forschungszentrum Juelich covers research activities, including high-lights of structural biochemistry (Alzheimer research), material research (skyrmions), computer simulation (e.g. of flexible blood cells), quantum physics (100 qubit era), photovoltaics, battery research, environmental research, climate research, biotechnology and community codes, including education and international cooperation.

  1. International HTR activities

    International Nuclear Information System (INIS)

    Baust, E.; Weisbrodt, I.

    1989-01-01

    Asea Brown Boveri AG (ABB) and their subsidiary High Temperature Reactor Construction GmbH (HRB) have brought the pebble bed high temperature reactor to the edge of being ready for the market with the construction and operation of the AVR reactor at Juelich and the THTR 300 at Hamm-Uentrop. Siemens/Interatom have developed the HTR modular concept and, together with their partners HRB, KFA, Rheinbraun Bergbauforschung have taken the nuclear process heat project to its present advanced state of development. The further introduction of the HTR to the market is a long-term objective, due to the present market situation. ABB and Siemens AG have therefore agreed to collaborate by forming a joint company. (orig.)

  2. Operational monitoring of the release behaviour of the AVR core

    International Nuclear Information System (INIS)

    Wawrzik, U.; Ivens, G.

    1985-01-01

    The AVR reactor has been used for the mass testing of spherical HTR fuel elements for more than 17 years. To date 14 fuel element types have been used, some of which differ considerably with reference to the heavy metal content, fuel coating and chemical fuel composition. The aim of the measurements at the reactor and of a comprehensive post-irradiation examination programme for fuel elements is to check and evaluate the behaviour of these fuel elements under real reactor conditions. This paper considers only those measurements which are of interest for reactor operation. The integral release behaviour of the fuel elements is continuously monitored by measuring the noble fission gas activity in the primary system. This value is directly determined by the heavy metal contamination. If any significant particle defects occur during operation, these are immediately indicated by a considerable increase in the noble gas activity. The integral release of the solid fission products is monitored by means of filter samples both in the hot and in the cold gas area; this examination, however, is performed intermittently and with a time delay. As these integral measurements only allow one to draw limited conclusions about the behaviour of single fuel element charges or types, they are supplemented by the systematic extraction of fuel elements. These elements are then subjected to standardized annealing tests (KFA) to determine the individual noble gas release, and to examinations of the fuel-free shell to establish the distribution of the solid fission products in it (AVR). The latter method, in particular, has proved to be practicable, as particle defects are detected at an early stadium. During operation to date, only one fuel element charge exhibited incipient particle defects shortly before reaching its final burnup. It was possible to limit the activity release by altering the charging strategy, which resulted in lower fuel element temperatures, and by systematically

  3. Computer codes for the operational control of the research reactors

    International Nuclear Information System (INIS)

    Kalker, K.J.; Nabbi, R.; Bormann, H.J.

    1986-01-01

    Four small computer codes developed by ZFR are presented, which have been used for several years during operation of the research reactors FRJ-1, FRJ-2, AVR (all in Juelich) and DR-2 (Riso, Denmark). Because of interest coming from the other reactor stations the codes are documented within the frame work of the IAEA Research Contract No. 3634/FG. The zero-dimensional burnup program CREMAT is used for reactor cores in which flux measurements at each individual fuel element are carried out during operation. The program yields burnup data for each fuel element and for the whole core. On the basis of these data, fuel reloading is prepared for the next operational period under consideration of the permitted minimum shut down reactivity of the system. The program BURNY calculates burnup for fuel elements inaccessible for flux measurements, but for which 'position weighting factors' have been measured/calculated during zero power operation of the core, and which are assumed to be constant in all operational situations. The code CURIAX calculates post-irradiation data for discharged fuel elements needed in their manipulation and transport. These three programs have been written for highly enriched fuel and take into account U-235 only. The modification of CREMAT for LEU Cores and its combiantion with ORIGEN is in preparation. KINIK is an inverse kinetic code and widely used for absorber rod calibration at the abovementioned research reactors. It includes a special polynomial subroutine which can easily be used in other codes. (orig.) [de

  4. Studies towards the Intrinsic Function of the AVR4 and AVR9 Elicitors of the Fungal Tomato Pathogen Cladosporium fulvum

    NARCIS (Netherlands)

    Burg, van den H.A.

    2003-01-01

    Recognition of the extracellular race-specific elicitor proteins AVR4 and AVR9 produced by the pathogenic fungus Cladosporium fulvum is mediated by the tomato resistance genes Cf-4 and Cf-9 , respectively. Recognition of these elicitors triggers host defense responses

  5. Evaluation of mating type distribution and genetic diversity of three Magnaporthe oryzae avirulence genes, PWL-2, AVR-Pii and Avr-Piz-t, in Thailand rice blast isolates

    Directory of Open Access Journals (Sweden)

    Thanyaluk Sirisathaworn

    2017-02-01

    Full Text Available Rice blast disease, caused by the filamentous ascomycete fungus Magnaporthe oryzae (anamorph Pyricularia oryzae, has been ranked among the most important diseases of rice. The molecular mechanisms against this fungus follow the idea of “gene-for-gene interaction”, in which a plant resistance (R gene product recognizes a fungal avirulence (Avr effector and triggers the defense response. However, the Avr genes have been shown to be rapidly evolving resulting in high levels of genetic diversity. This study investigated genetic diversity that is influenced by sexual recombination and mutation for the adaptation of rice blast fungus to overcome the defense response. Mating type distribution and the nucleotide sequence variation of three avirulence genes were evaluated—PWL-2, Avr-Pii and Avr-Piz-t. In total, 77 rice blast isolates collected from infected rice plants in northern, northeastern and central Thailand in 2005, 2010 and 2012, were used in the analysis with mating type and avirulence gene-specific primers. The results revealed that all the tested blast isolates belonged to the mating type MAT1-2, suggesting a lack of sexual recombination within the population. The successful rates of PWL-2, Avr-Pii and Avr-Piz-t gene-specific primer amplification were 100%, 60% and 54%, respectively. Base substitution mutation was observed in coding regions of the Avr-Pii and Avr-Piz-t genes. Although these results showed a low level of genetic diversity in Thai rice blast isolates, non-synonymous mutations did occur which revealed common mechanisms of selective pressure that are prone to adaptation of Avr genes. The information on nucleotide sequence variation and the genetic diversity of Avr genes obtained from this study could be useful for planning novel strategies in the development of rice breeding programs in Thailand.

  6. A microprocessor from AVR to embedded SoC

    International Nuclear Information System (INIS)

    Jeong, Geum Seoh

    2005-01-01

    This book was divided into two parts. The first part consists of ten chapter, which are basic knowledge, code vision AVR compiler, analysis on code vision, introduction and characteristic of AVR, I/O ports, interrupt and timer/counter, LCD, serial communication, analog comparator and stepping Motor and digital control of DC Motor. In the second part, it introduces Embedded Soc including application field, its characteristic, general description, functional description, designs with Quartus II.

  7. ANKE, a new facility for medium energy hadron physics at COSY-Juelich

    International Nuclear Information System (INIS)

    Barsov, S.; Bechstedt, U.; Bothe, W.; Bongers, N.; Borchert, G.; Borgs, W.; Braeutigam, W.; Buescher, M.; Cassing, W.; Chernyshev, V.; Chiladze, B.; Dietrich, J.; Drochner, M.; Dymov, S.; Erven, W.; Esser, R.; Franzen, A.; Golubeva, Ye.; Gotta, D.; Grande, T.; Grzonka, D.; Hardt, A.; Hartmann, M.; Hejny, V.; Horn, L. van; Jarczyk, L.; Junghans, H.; Kacharava, A.; Kamys, B.; Khoukaz, A.; Kirchner, T.; Klehr, F.; Klein, W.; Koch, H.R.; Komarov, V.I.; Kondratyuk, L.; Koptev, V.; Kopyto, S.; Krause, R.; Kravtsov, P.; Kruglov, V.; Kulessa, P.; Kulikov, A.; Lang, N.; Langenhagen, N.; Lepges, A.; Ley, J.; Maier, R.; Martin, S.; Macharashvili, G.; Merzliakov, S.; Meyer, K.; Mikirtychiants, S.; Mueller, H.; Munhofen, P.; Mussgiller, A.; Nekipelov, M.; Nelyubin, V.; Nioradze, M.; Ohm, H.; Petrus, A.; Prasuhn, D.; Prietzschk, B.; Probst, H.J.; Pysz, K.; Rathmann, F.; Rimarzig, B.; Rudy, Z.; Santo, R.; Paetz Schieck, H.; Schleichert, R.; Schneider, A.; Schneider, Chr.; Schneider, H.; Schwarz, U.; Seyfarth, H.; Sibirtsev, A.; Sieling, U.; Sistemich, K.; Selikov, A.; Stechemesser, H.; Stein, H.J.; Strzalkowski, A.; Watzlawik, K.-H.; Wuestner, P.; Yashenko, S.; Zalikhanov, B.; Zhuravlev, N.; Zwoll, K.; Zychor, I.; Schult, O.W.B.; Stroeher, H.

    2001-01-01

    ANKE is a new experimental facility for the spectroscopy of products from proton-induced reactions on internal targets. It has recently been implemented in the accelerator ring of the cooler synchrotron COSY of the Forschungszentrum Juelich (FZ-Juelich), Germany. The device consists of three dipole magnets, various target installations and dedicated detection systems. It will enable a variety of hadron-physics experiments like meson production in elementary proton-nucleon processes and studies of medium modifications in proton-nucleus interactions

  8. List of scientific and technological reports of the Nuclear Research Center Juelich Jan. 1985 - March 1988. 2. ed.

    International Nuclear Information System (INIS)

    1988-01-01

    This list of scientific and technological reports at first comprises some general publications such as abbreviations used in technical reports, annual report for 1984, 1985, 1986, the list of reports 1984 and 1985, surveys and inquiries on the population to the Chernobyl accident and on the scientific personnel of Juelich to the transfer of scientific knowledge to the public. - The other reports are concerned with the subjects mathematics - computer - cybernetics, general physics, atomic- and nuclear physics, solid state physics, materials, analytical chemistry, technology, reactor techniques, measuring techniques, biology, agriculture, questions of energy, medicine, environment, plasma physics and fusion reactor technology, physical chemistry, nuclear- and radiochemistry, chemical technology, electrotechniques - electronics, geosciences. Finally a register of the authors is added. (HK) [de

  9. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    1978-09-01

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

  10. Special neutron measurement results from the spectral positions of the Juelich FKS steel irradiation capsules

    International Nuclear Information System (INIS)

    Schneider, W.; Kuepper, H.; Pott, G.; Borchardt, G.; Segelhorst, G.; Thoene, L.; Weise, L.

    1986-10-01

    For the German project 'Forschungsvorhaben Komponentensicherheit' (FKS, i.e., Structural Integrity of Components) steel specimen irradiations have been carried out in the Juelich Merlin-type reactor (FRJ-1). The neutron monitoring to these irradiations is described in a German report (Juel-2087). In this context, some special considerations and results are given here, i.e., an experimental investigation of the fast neutron spectrum variation over a thick steel plate (in a special dosimetry test experiment); a comparison of the outcome of this investigation with the results from other FKS participants; and finally, the evaluation of the neutron exposure expressed in displacements per atom (dpa) in the centre of that steel plate. (orig.)

  11. ST segment elevation in lead aVR during exercise testing is associated with LAD stenosis

    International Nuclear Information System (INIS)

    Neill, Johanne; Harbinson, Mark; Shannon, Heather J.; Morton, Amanda; Muir, Alison R.; Adgey, Jennifer A.

    2007-01-01

    To evaluate, in patients with chest pain, the diagnostic value of ST elevation (STE) in lead aVR during stress testing prior to 99m Tc-sestamibi scanning correlating ischaemic territory with angiographic findings. Consecutive patients attending for 99m Tc-sestamibi myocardial perfusion imaging (MPI) completed a treadmill protocol. Peak exercise ECGs were coded. STE ≥0.05 mV in lead aVR was considered significant. Gated perfusion images and findings at angiography were assessed. STE in lead aVR occurred in 25% (138/557) of the patients. More patients with STE in aVR had a reversible defect on imaging compared with those who had no STE in aVR (41%, 56/138 vs 27%, 114/419, p=0.003). Defects indicating a left anterior descending artery (LAD) culprit lesion were more common in the STE in aVR group (20%, 27/138 vs 9%, 39/419, p=0.001). There was a trend towards coronary artery stenosis (>70%) in a double vessel distribution involving the LAD in those patients who had STE in aVR compared with those who did not (22%, 8/37 vs 5%, 4/77, p=0.06). Logistic regression analysis demonstrated that STE in aVR (OR 1.36, p=0.233) is not an independent predictor of inducible abnormality when adjusted for STD >0.1 mV (OR 1.69, p=0.026). However, using anterior wall defect as an end-point, STE in aVR (OR 2.77, p=0.008) was a predictor even after adjustment for STD (OR 1.43, p=0.281). STE in lead aVR during exercise does not diagnose more inducible abnormalities than STD alone. However, unlike STD, which is not predictive of a territory of ischaemia, STE in aVR may indicate an anterior wall defect. (orig.)

  12. Ozone profiles at Juelich, FRG, during 1988 and 1989

    International Nuclear Information System (INIS)

    Smit, H.G.J.; Straeter, W.; Loup, H.; Kley, D.

    1989-12-01

    Ozone soundings were performed regular at Juelich, FRG (50deg 41' N, 6deg 24' E). This report, the first one of an intended series, contains information on technical aspects and presents vertical profiles obtained during 1988 and 1989. (orig.) [de

  13. Bullet Ultrasonic Obstruction Detection & Distance Measurement Using AVR Microcontroller

    Directory of Open Access Journals (Sweden)

    Satish Pandey

    2008-08-01

    Full Text Available This paper describes the practical implementation of a short range ultrasonic obstruction detection and distance measurement device. By employing an ultrasonic transducer pair for producing ultrasonic sounds and sensing the reflected sound waves, the obstructions are detected. The hardware interface uses an Atmel ATmega8 AVR microcontroller to facilitate the generation of 40 kHz signal burst which is used in the transmitter circuit, and also to process the received signal for measuring the time of flight of reflected waves and exact distance of the obstruction. The program for this device is developed in WinAVR, and the code generated is dumped into microcontroller using AVR Studio. Educational aspects of this project include the mastery of a programming language and corresponding tools, the design of a functional and intuitive embedded application, and the development of appropriate hardware to build the device.

  14. Concepts for the interim storage of spent fuel elements from research reactors in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Niephaus, D.; Bensch, D.; Quaassdorff, P.; Plaetzer, S.

    1997-01-01

    Research reactors have been operated in the Federal Republic of Germany since the late fifties. These are Material Test Reactors (MTR) and training, Research and Isotope Facilities of General Atomic (TRIGA). A total of seven research reactors, i.e. three TRIGA and four MTR facilities were still in operation at the beginning of 1996. Provisions to apply to the back-end of the fuel cycle are required for their continued operation and for already decommissioned plants. This was ensured until the end of the eighties by the reprocessing of spent fuel elements abroad. In view of impeding uncertainties in connection with waste management through reprocessing abroad, the development of a national back-end fuel cycle concept was commissioned by the Federal Minister of Education, Science, Research and Technology in early 1990. Development work was oriented along the lines of the disposal concept for irradiated light-water reactor fuel elements from nuclear power plants. Analogously, the fuel elements from research reactors are to be interim-stored on a long-term basis in adequately designed transport and storage casks and then be directly finally disposed without reprocessing after up to forty years of interim storage. As a first step in the development of a concept for interim storage, several sites with nuclear infrastructure were examined and assessed with respect to their suitability for interim storage. A reasonably feasible reference concept for storing the research reactor fuel elements in CASTOR MTR 2 transport and storage casks at the Ahaus interim storage facility (BZA) was evaluated and the hot cell facility and AVR store of Forschungszentrum Juelich (KFA) were proposed as an optional contingency concept for casks that cannot be repaired at Ahaus. Development work was continued with detailed studies on these two conceptual variants and the results are presented in this paper. (author)

  15. ST Elevation in Lead aVR and Its Association with Clinical Outcomes

    Directory of Open Access Journals (Sweden)

    Eka Ginanjar

    2018-01-01

    Full Text Available The purpose of this case repots are to evaluate the role of ST elevation in aVR lead and to make analysis between both cases. There are some atypical electrocardiogram (ECG presentations which need prompt management in patient with ischemic clinical manifestation such as ST elevation in aVR lead. In this case study, we report a 68-year old woman with chief symptoms of shortness of breath and chest discomfort. She was diagnosed with cardiogenic shock, with Killip class IV, and TIMI score of 8. The second case is a 57-year-old man with typical chest pain at rest which could not be relieved with nitrate treatment. He was diagnosed with ST elevation in inferior and aVR lead, and occlusion in left circumflex artery (LCX. Both patients underwent primary percutaneous coronary intervention (PPCI. Subsequently, both cases presented remarkable clinical improvements and improved ST elevation myocardial infarction (STEMI in aVR lead.

  16. Juelich Research Center. Annual report 1991

    International Nuclear Information System (INIS)

    1991-10-01

    The Research Centre Juelich (KFA) as one of the thirteen national research centres in the Federal Republic of Germany is probably unique in that it concentrates equally on four essentials for mankind - energy, health and environment, materials and matter as well as information. These basic requirements are reflected by the four priority programmes characterizing research at the KFA in the nineties. The research priorities are: Properties of Matter and Material Research; Basic Research on Information Technology; Health, Environment, Biotechnology; Energy Research and Technology; Nuclear Fusion; Basic Nuclear Research; Interdisciplinary Analyses and Methods. (orig./HSCH) [de

  17. IRPhE-HTR-ARCH-01, Archive of HTR Primary Documents

    International Nuclear Information System (INIS)

    2004-01-01

    Description: High Temperature Reactor Studies, including experiments in critical facilities or in prototypes have been carried out in the past. Information gathered, experience gained and experimental data produced are of value for the development of future advanced HTRs. For the purpose of knowledge, competence, information preservation and management, computer readable archives have been established. The present archive includes several relevant documents relative to the following: - Graphite Moderated Critical Facility, CESAR at Cadarache. Dragon Countries Physics Meetings (DCPM); - OTTO Pebble Bed Reactors; - Gulf - HTGR Experiments; - Zero Power MARIUS Reactor; - Pebble-bed KAHTER Critical Facility; - Helium Cooled Fast Reactor Assessment Studies; - Gas Cooled Reactor Technology Safety and Siting; - Initial Evaluation of the Gas-Turbine Modules HTGCR; - A report on Nuclear Graphite; - AVR Reactor Juelich (new in version 02); - HTR IAEA proceedings (new in version 02); - Studies at IRI Delft(new in version 02); - Studies and experiments at PSI Villigen (new in version 02); 2 - Related or auxiliary information: IRPHE-DRAGON-DPR, high Temperature Reactor Dragon Project, Primary Documents NEA-1726/01. 3 - Software requirements: Acrobat Reader, Microsoft Word, HTML Browser required

  18. Comparison of intelligent fuzzy based AGC coordinated PID controlled and PSS controlled AVR system

    Energy Technology Data Exchange (ETDEWEB)

    Mukherjee, V. [Department of Electrical Engineering, Asansol Engineering College, Asansol, West Bengal (India); Ghoshal, S.P. [Department of Electrical Engineering, National Institute of Technology, Durgapur, West Bengal (India)

    2007-11-15

    This paper attempts to investigate the performance of intelligent fuzzy based coordinated control of the Automatic Generation Control (AGC) loop and the excitation loop equipped with Proportional Integral Derivative (PID) controlled Automatic Voltage Regulator (AVR) system and Power System Stabilizer (PSS) controlled AVR system. The work establishes that PSS controlled AVR system is much more robust in dynamic performance of the system over a wide range of system operating configurations. Thus, it is revealed that PSS equipped AVR is much more superior than PID equipped AVR in damping the oscillation resulting in improved transient response. The paper utilizes a novel class of Particle Swarm Optimization (PSO) termed as Craziness based Particle Swarm Optimization (CRPSO) as optimizing tool to get optimal tuning of PSS parameters as well as the gains of PID controllers. For on-line, off-nominal operating conditions Takagi Sugeno Fuzzy Logic (TSFL) has been applied to obtain the off-nominal optimal gains of PID controllers and parameters of PSS. Implementation of TSFL helps to achieve very fast dynamic response. Fourth order model of generator with AVR and high gain thyristor excitation system is considered for PSS controlled system while normal gain exciter is considered for PID controlled system. Simulation study also reveals that with high gain exciter, PID control is not at all effective. Transient responses are achieved by using modal analysis. (author)

  19. ST Elevation in AVR: When Time May Not Mean Muscle

    Science.gov (United States)

    2017-10-31

    REPORT TYPE 1013112017 Presentation 4. TITLE AND SUBTITLE ST Elevation in AVR: When Thne May Not Mean !Vfu&cle 6. AUTHOR{S) Capt \\Villiam T...ACCF/AHA Guideline for the Management of ST -Elevation Myocardial Infarction A Report of the Arnerican College of C ardiology Foundation/ Ameri can...C7Q n ([) G) ro )::>t w :J r+ c.. < -· ro ti) ti) ro ti) OJ m ti) -ro n Take home points •Don’t ignore ST elevation in aVR •Look closely

  20. Die Energiewerke Nord GmbH. From operator of a decommissioned Russian nuclear power plant to one of Europe's leading decommissioning companies

    International Nuclear Information System (INIS)

    Philipp, Marlies

    2011-01-01

    EWN GmbH is a state-owned company with these duties: - decommissioning and demolition of the Greifswald and Rheinsberg nuclear power stations; - safe operation of the Zwischenlager Nord interim store; - development of the 'Lubminer Heide' industrial and commercial estate. Other projects for which EWN GmbH uses its know-how: - disposal of 120 decommissioned Russian nuclear submarines in Murmansk; - decommissioning and dismantling of the Juelich, NRW, AVR experimental reactor; - demolition of nuclear plants; running the Central Decontamination Operations Department at Karlsruhe, BW. Since 2008, EWN GmbH has held 25% of the shares of Deutsche Gesellschaft zum Bau- und Betrieb von Endlagern fuer Abfallstoffe mbH (DBE), a firm building and operating nuclear repositories. (orig.)

  1. AVR RISC microcontroller handbook

    CERN Document Server

    Kuhnel, Claus

    1998-01-01

    The AVR RISC Microcontroller Handbook is a comprehensive guide to designing with Atmel's new controller family, which is designed to offer high speed and low power consumption at a lower cost. The main text is divided into three sections: hardware, which covers all internal peripherals; software, which covers programming and the instruction set; and tools, which explains using Atmel's Assembler and Simulator (available on the Web) as well as IAR's C compiler.Practical guide for advanced hobbyists or design professionalsDevelopment tools and code available on the Web

  2. The Phytophthora sojae avirulence locus Avr3c encodes a multi-copy RXLR effector with sequence polymorphisms among pathogen strains.

    Directory of Open Access Journals (Sweden)

    Suomeng Dong

    Full Text Available Root and stem rot disease of soybean is caused by the oomycete Phytophthora sojae. The avirulence (Avr genes of P. sojae control race-cultivar compatibility. In this study, we identify the P. sojae Avr3c gene and show that it encodes a predicted RXLR effector protein of 220 amino acids. Sequence and transcriptional data were compared for predicted RXLR effectors occurring in the vicinity of Avr4/6, as genetic linkage of Avr3c and Avr4/6 was previously suggested. Mapping of DNA markers in a F(2 population was performed to determine whether selected RXLR effector genes co-segregate with the Avr3c phenotype. The results pointed to one RXLR candidate gene as likely to encode Avr3c. This was verified by testing selected genes by a co-bombardment assay on soybean plants with Rps3c, thus demonstrating functionality and confirming the identity of Avr3c. The Avr3c gene together with eight other predicted genes are part of a repetitive segment of 33.7 kb. Three near-identical copies of this segment occur in a tandem array. In P. sojae strain P6497, two identical copies of Avr3c occur within the repeated segments whereas the third copy of this RXLR effector has diverged in sequence. The Avr3c gene is expressed during the early stages of infection in all P. sojae strains examined. Virulent alleles of Avr3c that differ in amino acid sequence were identified in other strains of P. sojae. Gain of virulence was acquired through mutation and subsequent sequence exchanges between the two copies of Avr3c. The results illustrate the importance of segmental duplications and RXLR effector evolution in the control of race-cultivar compatibility in the P. sojae and soybean interaction.

  3. Fido, a novel AMPylation domain common to fic, doc, and AvrB.

    Directory of Open Access Journals (Sweden)

    Lisa N Kinch

    2009-06-01

    Full Text Available The Vibrio parahaemolyticus type III secreted effector VopS contains a fic domain that covalently modifies Rho GTPase threonine with AMP to inhibit downstream signaling events in host cells. The VopS fic domain includes a conserved sequence motif (HPFx[D/E]GN[G/K]R that contributes to AMPylation. Fic domains are found in a variety of species, including bacteria, a few archaea, and metazoan eukaryotes.We show that the AMPylation activity extends to a eukaryotic fic domain in Drosophila melanogaster CG9523, and use sequence and structure based computational methods to identify related domains in doc toxins and the type III effector AvrB. The conserved sequence motif that contributes to AMPylation unites fic with doc. Although AvrB lacks this motif, its structure reveals a similar topology to the fic and doc folds. AvrB binds to a peptide fragment of its host virulence target in a similar manner as fic binds peptide substrate. AvrB also orients a phosphate group from a bound ADP ligand near the peptide-binding site and in a similar position as a bound fic phosphate.The demonstrated eukaryotic fic domain AMPylation activity suggests that the VopS effector has exploited a novel host posttranslational modification. Fic domain-related structures give insight to the AMPylation active site and to the VopS fic domain interaction with its host GTPase target. These results suggest that fic, doc, and AvrB stem from a common ancestor that has evolved to AMPylate protein substrates.

  4. Requirements of, and operating experience with, gas analyses on high temperature reactors

    International Nuclear Information System (INIS)

    Nieder, R.

    1982-06-01

    Impurities in the helium coolant of the primary coolant circuit of HTGR's are mainly due to ingress of air or water, occasionally oil. Typical concentrations are given of H 2 O, H 2 , CO 2 , CO, N 2 , CH 4 and Ar in the AVR, Dragon, Peach Bottom and Fort St. Vrain reactors. A characteristic is presented of measuring devices for measuring non-active impurities in helium; measuring methods are described and a list is given of required and actual detection limits. Also given are concentrations of solid fission and activation products and tritium in the primary circuit of the AVR reactor

  5. Utility of lead aVR for identifying the culprit lesion in acute myocardial infarction

    DEFF Research Database (Denmark)

    Kühl, Jørgen Tobias; Berg, Ronan M G

    2009-01-01

    BACKGROUND: Lead aVR is a neglected, however, potentially useful tool in electrocardiography. Our aim was to evaluate its value in clinical practice, by reviewing existing literature regarding its utility for identifying the culprit lesion in acute myocardial infarction (AMI). METHODS: Based...... on a systematic search strategy, 16 studies were assessed with the intent to pool data; diagnostic test rates were calculated as key results. RESULTS: Five studies investigated if ST-segment elevation (STE) in aVR is valuable for the diagnosis of left main stem stenosis (LMS) in non-ST-segment AMI (NSTEMI......). The studies were too heterogeneous to pool, but the individual studies all showed that STE in aVR has a high negative predictive value (NPV) for LMS. Six studies evaluated if STE in aVR is valuable for distinguishing proximal from distal lesions in the left anterior descending artery (LAD) in anterior ST...

  6. A new biotype of Fusarium oxysporum f. sp. lycopersici race 2 emerged by a transposon-driven mutation of avirulence gene AVR1.

    Science.gov (United States)

    Kashiwa, Takeshi; Suzuki, Tatsuya; Sato, Akira; Akai, Kotaro; Teraoka, Tohru; Komatsu, Ken; Arie, Tsutomu

    2016-07-01

    Emergence of races in Fusarium oxysporum f. sp. lycopersici (Fol) is caused by loss or mutation of at least one avirulence (AVR) gene. The product of AVR1 is a small protein (Avr1) secreted by Fol in tomato xylem sap during infection. This protein triggers Fol race 1 specific resistance (I) in tomato, indicating that AVR1 is an AVR gene. Deletion of AVR1 in race 1 resulted in the emergence of race 2, and an additional mutation in AVR2 generated race 3. Previously, we reported a new biotype of race 3, KoChi-1, in which AVR1 was truncated by a transposon Hormin, which suggested a new route to evolution of races in Fol However, to date no race 2 isolate carrying Hormin-truncated AVR1 has been reported. In this report, we describe such isolates, represented by Chiba-5, in which Hormin insertion occurred in AVR1 at a position different from that in KoChi-1. AVR1 truncation in both isolates resulted in production of defective Avr1 proteins. Chiba-5 and KoChi-1 belong to different phylogenetic clades, A1 and A2, respectively, suggesting that insertion of Hormin in AVR1 in Chiba-5 and KoChi-1 occurred as independent evolutionary events. © FEMS 2016. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

  7. Predictions on an HTR coolant composition after operational experience with experimental reactors

    International Nuclear Information System (INIS)

    Nieder, R.

    1981-01-01

    Long-term operational experience of the HTR experimental reactors Dragon (1966 - 1975), Peach Bottom (1967 - 1974) and AVR (since 1967) has yielded a large number of common quantitative and qualitative results about the sources and behaviour of helium impurities in the primary circuits. Additional information has also been obtained from experiments made at the three reactors. The results at the AVR are particularly interesting because the gas outlet temperature can be varied from 770 0 C to 950 0 C when the reactor power is kept constant. Hence they can be studied according to the temperature dependence of all chemical reactions. It should be possible to apply the results from the operating measurements and experiments made at the reactors, in particular the interrelation of the impurity concentrations, to future reactors. The absolute values of these impurity concentrations are obtained first and foremost by the corresponding helium purification constants

  8. The development of high-temperature reactors in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Engelmann, P.; Krings, F.

    1980-01-01

    The principal features of high-temperature reactors are recalled, then the current state of technology of the line in the Federal Republic of Germany is described. Reference is made to the experience of operating the AVR reactor, the construction of the THTR-300 reactor as well as the HTT and PNP projects [fr

  9. Avirulence (AVR) Gene-Based Diagnosis Complements Existing Pathogen Surveillance Tools for Effective Deployment of Resistance (R) Genes Against Rice Blast Disease.

    Science.gov (United States)

    Selisana, S M; Yanoria, M J; Quime, B; Chaipanya, C; Lu, G; Opulencia, R; Wang, G-L; Mitchell, T; Correll, J; Talbot, N J; Leung, H; Zhou, B

    2017-06-01

    Avirulence (AVR) genes in Magnaporthe oryzae, the fungal pathogen that causes the devastating rice blast disease, have been documented to be major targets subject to mutations to avoid recognition by resistance (R) genes. In this study, an AVR-gene-based diagnosis tool for determining the virulence spectrum of a rice blast pathogen population was developed and validated. A set of 77 single-spore field isolates was subjected to pathotype analysis using differential lines, each containing a single R gene, and classified into 20 virulent pathotypes, except for 4 isolates that lost pathogenicity. In all, 10 differential lines showed low frequency (95%), inferring the effectiveness of R genes present in the respective differential lines. In addition, the haplotypes of seven AVR genes were determined by polymerase chain reaction amplification and sequencing, if applicable. The calculated frequency of different AVR genes displayed significant variations in the population. AVRPiz-t and AVR-Pii were detected in 100 and 84.9% of the isolates, respectively. Five AVR genes such as AVR-Pik-D (20.5%) and AVR-Pik-E (1.4%), AVRPiz-t (2.7%), AVR-Pita (0%), AVR-Pia (0%), and AVR1-CO39 (0%) displayed low or even zero frequency. The frequency of AVR genes correlated almost perfectly with the resistance frequency of the cognate R genes in differential lines, except for International Rice Research Institute-bred blast-resistant lines IRBLzt-T, IRBLta-K1, and IRBLkp-K60. Both genetic analysis and molecular marker validation revealed an additional R gene, most likely Pi19 or its allele, in these three differential lines. This can explain the spuriously higher resistance frequency of each target R gene based on conventional pathotyping. This study demonstrates that AVR-gene-based diagnosis provides a precise, R-gene-specific, and differential line-free assessment method that can be used for determining the virulence spectrum of a rice blast pathogen population and for predicting the

  10. A ECG Signal Gathering and Displaying System Based on AVR

    Science.gov (United States)

    Ning, Li; Ruilan, Zhang; Jian, Liu; Xiaochen, Wang; Shuying, Chen; Zhuolin, Lang

    2017-12-01

    This article introduces a kind of system which is based on the AVR to acquire the data of ECG. Such system using the A/D function of ATmega8 chip and the lattice graph LCD to design ECG heart acquisition satisfies the demands above. This design gives a composition of hardware and programming of software about the system in detail which has mainly realized the real-time gathering, the amplifier, the filter, the A/D transformation and the LCD display. Since the AVR includes A/D transformation function and support embedded C language programming, it reduces the peripheral circuit, further more it also decreases the time to design and debug this system.

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  12. Development of closed orbit diagnostics towards EDM measurements at COSY in Juelich

    Energy Technology Data Exchange (ETDEWEB)

    Hinder, Fabian [Forschungszentrum Juelich, Institut fuer Kernphysik IV (Germany); RWTH Aachen University, III. Physikalisches Institut B (Germany); Collaboration: JEDI-Collaboration

    2016-07-01

    Electric Dipole Moments (EDMs) violate parity and time reversal symmetries. Assuming the CPT-theorem, this leads to CP violation, which is needed to explain the matter over antimatter dominance in the Universe. Thus, a non-zero EDM is a hint to new physics beyond the Standard Model. The JEDI collaboration (Juelich Electric Dipole moment Investigations) has started investigations of a direct EDM measurement of protons and deuterons at a storage ring. To measure a tiny EDM signal with high precision, systematic effects have to be controlled to the same level. One major source of systematic uncertainties is a distortion of the closed orbit. To control and measure this effect, the orbit measurement system, including the readout electronics, the orbit correction system and the beam position monitor pick-ups are improved. All the mentioned developments are ongoing at the Cooler Synchrotron (COSY) at Juelich. The achievements in the mentioned fields are presented at the conference.

  13. In vitro translocation experiments with RxLR-reporter fusion proteins of Avr1b from Phytophthora sojae and AVR3a from Phytophthora infestans fail to demonstrate specific autonomous uptake in plant and animal cells.

    Science.gov (United States)

    Wawra, Stephan; Djamei, Armin; Albert, Isabell; Nürnberger, Thorsten; Kahmann, Regine; van West, Pieter

    2013-05-01

    Plant-pathogenic oomycetes have a large set of secreted effectors that can be translocated into their host cells during infection. One group of these effectors are the RxLR effectors for which it has been shown, in a few cases, that the RxLR motif is important for their translocation. It has been suggested that the RxLR-leader sequences alone are enough to translocate the respective effectors into eukaryotic cells through binding to surface-exposed phosphoinositol-3-phosphate. These conclusions were primary based on translocation experiments conducted with recombinant fusion proteins whereby the RxLR leader of RxLR effectors (i.e., Avr1b from Phytophthora sojae) were fused to the green fluorescent protein reporter-protein. However, we failed to observe specific cellular uptake for a comparable fusion protein where the RxLR leader of the P. infestans AVR3a was fused to monomeric red fluorescent protein. Therefore, we reexamined the ability of the reported P. sojae AVR1b RxLR leader to enter eukaryotic cells. Different relevant experiments were performed in three independent laboratories, using fluorescent reporter fusion constructs of AVR3a and Avr1b proteins in a side-by-side comparative study on plant tissue and human and animal cells. We report that we were unable to obtain conclusive evidence for specific RxLR-mediated translocation.

  14. Design of PID controller as an AVR in frequency-domain

    International Nuclear Information System (INIS)

    Shaikh, S.A.; Ahmed, I.

    2008-01-01

    The primary means of generator reactive-power control is the generator-excitation control, using Automatic Voltage Regulator (AVR). The role of AVR is to hold the terminal-voltage of Synchronous generator at a specified level. This paper presents the design of a proportional integral- derivative (PID) controller to work as an A VR. The PID controller has been tuned by HO-HANG-CAO method. In this method, pm parameters are computed from the gain-margin and phase-margin specifications. This method has been found much superior to the conventional Ziegler- Nichols rules. The performance of the controller has been evaluated through Simulation Studies in MATLAB environment. It has been demonstrated that the PID controller, tuned with the said method, yields highly satisfactory closed loop performance. (author)

  15. Cost accounting in Kernforschungsanlage Juelich Gesellschaft mit beschraenkter Haftung (KFA)

    International Nuclear Information System (INIS)

    Seidel, G.; Schilling, H.

    1979-01-01

    The paper gives an overview about the organization and the research program of the Kernforschungsanalage Juelich Gesellschaft mit beschraenkter Haftung (KFA). The cost accounting system is discussed in detail, cost categories, cost centers, units of production and the data flow of the cost accounting are described. The distribution of the resulting management accounting reports and all sorts of EDP-result listings of the cost accounting system is described. (A.N.)

  16. Optimization of AVR Parameters of a Multi-machine Power System ...

    African Journals Online (AJOL)

    user1

    Keywords: multi-machine power system stability, AVR system, power system stabilizer, PID controller ... The proposed controller was a fuzzy-logic-based stabilizer that has the capability to ..... Computer methods in power system analysis.

  17. Progress and problems in modelling HTR core dynamics

    International Nuclear Information System (INIS)

    Scherer, W.; Gerwin, H.

    1991-01-01

    In recent years greater effort has been made to establish theoretical models for HTR core dynamics. At KFA Juelich the TINTE (TIme dependent Neutronics and TEmperatures) code system has been developed, which is able to model the primary circuit of an HTR plant using modern numerical techniques and taking into account the mutual interference of the relevant physical variables. The HTR core is treated in 2-D R-Z geometry for both nucleonics and thermo-fluid-dynamics. 2-energy-group diffusion theory is used in the nuclear part including 6 groups of delayed neutron precursors and 14 groups of decay heat producers. Local and non-local heat sources are incorporated, thus simulating gamma ray transport. The thermo-fluid-dynamics module accounts for heterogeneity effects due to the pebble bed structure. Pipes and other components of the primary loop are modelled in 1-D geometry. Forced convection may be treated as well as natural convection in case of blower breakdown accidents. Validation of TINTE has started using the results of a comprehensive experimental program that has been performed at the Arbeitsgemeinschaft Versuchsreaktor GmbH (AVR) high temperature pebble bed reactor at Juelich. In the frame of this program power transients were initiated by varying the helium blower rotational speed or by moving the control rods. In most cases a good accordance between experiment and calculation was found. Problems in modelling the special AVR reactor geometry in two dimensions are described and suggestions for overcoming the uncertainties of experimentally determined control rod reactivities are given. The influence of different polynomial expansions of xenon cross sections to long term transients is discussed together with effects of burnup during that time. Up to now the TINTE code has proven its general applicability to operational core transients of HTR. The effects of water ingress on reactivity, fuel element corrosion and cooling gas properties are now being

  18. The Juelich compact cyclotron - a multi-purpose irradiation facility

    International Nuclear Information System (INIS)

    Hemmerich, J.; Hoelzle, R.; Kogler, W.

    1977-01-01

    A commercially available variable-energy compact cyclotron has been installed at the Kernforschungsanlage Juelich. It is equipped to accelerate protons, deuterons, 3 He- and α-particles. A +- 60 0 switching magnet allows to switch the beam to any of seven external target stations. Three separately shielded target rooms allow a flexible use of the cyclotron for a wide range of applications such as production of short-lived nuclides, activation analysis, radiation damage studies in metals and studies of biological effects of fast neutron irradiation. (orig.) [de

  19. Development of Mitsui/Juelich Incineration System and hydro-thermal ash solidification

    International Nuclear Information System (INIS)

    Suzuki, S.; Kamada, S.; Nakamori, Y.; Katakura, M.; Yamazaki, N.

    1988-01-01

    This paper summarizes the developing program for Mitsui/Juelich Incinerated System combined with Hydrothermal ash solidification. The system is an integrated one and capable for volume reduction of various kind of radioactive waste and safe disposal of residual incinerator ash. The system also has an advantage of reducing construction and operation cost. An outline of the incineration plant is also presented in this paper

  20. Incineration of wastes from nuclear installations with the Juelich incineration process

    International Nuclear Information System (INIS)

    Wilke, M.

    1979-01-01

    In the Juelich Research Center a two-stage incineration process has been developed which, due to an integral thermal treatment stage, is most suitable for the incineration of heterogeneous waste material. The major advantages of this technique are to be seen in the fact that mechanical treatment of the waste material is no longer required and that off gas treatment is considerably facilitated. (orig.) [de

  1. KFA Juelich annual report 1986/87

    International Nuclear Information System (INIS)

    1987-01-01

    Basic and applied research almost keep the balance in the program of Kernforschungsanlage Juelich GmbH (KFA), one of the FRG's national research facilities. Multidisciplinary work of KFA in various fields of the natural sciences include nuclear medicine, nuclear chemistry, biotechnology, plasma physics/nuclear fusion, nuclear physics, energy and environmental resarch, multiparticle systems research, including related solid-state, surface, and vacuum physics research, and also problems of analyzing large-scale nonlinear systems. Future work will focus on the three major programs 'Materials development', 'Environmental chemicals and ecosystems', and 'Basic research on information technologies'. The chapter 'Examples of research work' contains topical contributions. The 'Report on R and D work' is a survey of all scientific-technical activities. The chapter 'Research institutes and joint scientific and technical facilities' describes tasks and targets of the entire KFA complex. (HK) [de

  2. New experimental results on electron cooling at COSY-Juelich

    International Nuclear Information System (INIS)

    Dietrich, J.; Maier, R.; Prasuhn, D.; Stein, H.J.; Kobets, A.; Meshkov, I.; Sidorin, A.; Smirnov, A.

    2007-01-01

    Recent results of electron cooling of proton beams at COSY-Juelich are reported. Cooling at an electron energy of 70 keV has been studied for the first time. At the injection energy level of COSY, corresponding to 24.5 keV electron energy, the features of the cooled proton beam at extremely low intensities have been investigated in order to find out whether an ordering of the proton beam can be achieved. Such investigations are motivated by the results of a numerical simulation of the ordering process by the BETACOOL code. (author)

  3. Seasonal trends of NH4+ and NO3- nitrogen isotope composition in rain collected at Juelich, Germany

    International Nuclear Information System (INIS)

    Freyer, H.D.

    1978-01-01

    Data are presented on nitrogen isotope composition in ammonium and nitrate from rain-water collected over 2 years in an interior area at Juelich, Germany. The seasonal trends in these data are discussed relative to natural and anthropogenic emissions of nitrogen compounds which additionally have been measured or estimated in their isotope composition, e.g. ammonia from animal urine, fuel combustion, fertilizer use and organic soil nitrogen, and natural and anthropogenic nitric oxides from automobile exhausts as well. The 15 N content of Juelich rain ammonium is found to be different from values of Hoering (1957) and Moore (1974) and from other rain samples collected in coastal areas. (Auth.)

  4. PISCES and ALT-II: Juelich PSI papers

    International Nuclear Information System (INIS)

    Conn, R.W.; Hirooka, Y.; LaBombard, B.

    1988-08-01

    This publication comprises papers from the PISCES and ALT-II Programs at UCLA which were presented at the International Plasma Surface Interactions Meeting held in Juelich, FRG, on May 2-6, 1988. A list of publications from the PISCES and ALT-II contained in this report are: Deuterium pumping and erosion behavior of selected graphite materials under high flux plasma bombardment in PISCES; Erosion and redeposition behavior of selected NET-candidate materials under high-flux hydrogen, deuterium plasma bombardment in PISCES; Presheath profiles in simulated tokamak edge plasmas; Boundary asymmetries and plasma flow to the ALT-II toroidal belt pump limiter; ALT-II toroidal belt pump limiter performance in TEXTOR; and An in-situ spectroscopic erosion yield measurement with applications to sputtering and surface morphology alterations

  5. COSY Juelich - a cooler synchrotron for unpolarized and polarized medium-energy studies

    International Nuclear Information System (INIS)

    Seyfarth, H.

    2001-01-01

    Full text: The Forschungszentrum Juelich (Research Center Juelich) is one of the sixteen national research institutions in the 'Hermann von Helmholtz Association of German Research Centers'. It is dedicated to fundamental and applied research and development which can be summarized under five priorities: (i) structure of matter and materials research, (ii) information technology, (iii) life sciences, (iv) environment precaution research, and (v) energy technology. As one of the institutes within (i). the Institut fur Kernphysik (Institute for Nuclear Research) operates the COSY cooler synchrotron which allows to accelerate unpolarized and polarized protons and deuterons to the maximum momentum of 3450 MeV/c (2640 MeV and 2050 MeV kinetic energy for protons and deuterons, respectively). At low energy electron cooling can be used for beam preparation, whereas stochastic cooling can be applied to the accelerated beam. In the first years of operation since 1993 the experiments have been performed with the unpolarized proton beam. Since 1997 the polarized proton beam is available with increasing intensity and a typical degree of polarization of about 75 % up to the maximum beam energy. In 2000 the first unpolarized deuteron beam could be accelerated and stored at the maximum energy. Four target places exist for the internal experiments PISA. EDDA, COSY-II, and ANKE which use the circulating beam with thin solid strip or fiber targets and gas targets. The four experiments TOF, MOMO, GEM, NESSI, and JESSICA are using external beams. The programs of the experiments JESSICA (Juelich Experimental Spallation Setup in the COSY Area), NESSI (Neutron Scintillator and Silicon), and PISA (Proton Induced Spallation) aim at the measurement of data needed or the design of the target station of the planned European Spallation neutron Source (ESS). The set-up of PISA is replacing the earlier experiment COSY-13 which successfully completed its investigations on the production of

  6. The nuclear detectives. The international atomic energy agency IAEA is supposed to trace countries that produce or disseminate nuclear materials for nuclear weapons. A team from Juelich is supporting the process; Die Nukleardetektive. Die internationale Atomenergie-Organisation IAEO soll Staaten aufspueren, die heimlich nukleares Material fuer Atomwaffen herstellen oder verbreiten. Ein Juelicher Team unterstuetzt sie dabei

    Energy Technology Data Exchange (ETDEWEB)

    Frick, Frank

    2015-07-01

    The Juelich Team is part of the International Standing Advisory Group on Safeguard Implementation that is supposed to find violations of the Non-Proliferation Treaty. Using wipe cloth in nuclear facilities it is possible to find uranium or plutonium containing particles. The study of the isotopic composition of uranium particles allows the identification of the uranium enrichment in the facility. In Juelich reference uranium particles with defined isotopic composition are produced for calibration purposes. The Juelich team supporting IAEA is also developing computer programs for the automatic processing of satellite data with respect to nuclear facilities.

  7. Silencing of the PiAvr3a effector-encoding gene from Phytophthora infestans by transcriptional fusion to a short interspersed element.

    Science.gov (United States)

    Vetukuri, Ramesh R; Tian, Zhendong; Avrova, Anna O; Savenkov, Eugene I; Dixelius, Christina; Whisson, Stephen C

    2011-12-01

    Phytophthora infestans is the notorious oomycete causing late blight of potato and tomato. A large proportion of the P. infestans genome is composed of transposable elements, the activity of which may be controlled by RNA silencing. Accumulation of small RNAs is one of the hallmarks of RNA silencing. Here we demonstrate the presence of small RNAs corresponding to the sequence of a short interspersed retrotransposable element (SINE) suggesting that small RNAs might be involved in silencing of SINEs in P. infestans. This notion was exploited to develop novel tools for gene silencing in P. infestans by engineering transcriptional fusions of the PiAvr3a gene, encoding an RXLR avirulence effector, to the infSINEm retroelement. Transgenic P. infestans lines expressing either 5'-infSINEm::PiAvr3a-3' or 5'-PiAvr3a::SINEm-3' chimeric transcripts initially exhibited partial silencing of PiAvr3a. Over time, PiAvr3a either recovered wild type transcript levels in some lines, or became fully silenced in others. Introduction of an inverted repeat construct was also successful in yielding P. infestans transgenic lines silenced for PiAvr3a. In contrast, constructs expressing antisense or aberrant RNA transcripts failed to initiate silencing of PiAvr3a. Lines exhibiting the most effective silencing of PiAvr3a were either weakly or non-pathogenic on susceptible potato cv. Bintje. This study expands the repertoire of reverse genetics tools available for P. infestans research, and provides insights into a possible mode of variation in effector expression through spread of silencing from adjacent retroelements. Crown Copyright © 2011. Published by Elsevier Ltd. All rights reserved.

  8. Pathfinder irradiation of advanced fuel (Th/U mixed oxide) in a power reactor

    International Nuclear Information System (INIS)

    Brant Pinheiro, R.

    1993-01-01

    Within the joint Brazilian-German cooperative R and D Program on Thorium Utilization in Pressurized Water Reactors carried out from 1979 to 1988 by Nuclebras/CDTN, KFA-Juelich, Siemens/KWU and NUKEM, a pathfinder irradiation of Th/U mixed oxide fuel in the Angra 1 nuclear power reactor was planned. The objectives of this irradiation testing, the irradiation strategy, the work performed and the status achieved at the end of the joint Program are presented. (author)

  9. To the safety conception of the high temperature reactor with natural heat removal decay in teh case of accidents

    International Nuclear Information System (INIS)

    Petersen, K.

    1983-10-01

    On September 22, 1970, for the first time an accident simulation experiment with complete failure of the forced core cooling and the nuclear shut-down system was performed in the AVR-reactor: Due to a small heat-up of the fuel the nuclear chain-reaction was interrupted and an overheating of the core and structure was prevented due to the natural heat-convection. On the basis of the meanwhile developed computer-methods and accompanying experimental investigations it is now possible to determine exactly the behaviour of the non actively controlled core of the high temperature reactor, and to understand better the course of the AVR-experiments. On the same basis the potential and the limits of the safety conception realized in the AVR with self-stabilization in the case of accident can be determined. Such a small high temperature reactor as for example the HTR-modul of the KWU, which is characterized by a reliable and simple safety-technique with a minimum of expensive active systems, can be realized using a 2-zone-core up to a unit size of nearly 250 MW(th). (orig.) [de

  10. High-temperature reactor developments in the Netherlands

    International Nuclear Information System (INIS)

    Schram, R.P.C.; Cordfunke, E.H.P.; Heek, A.I. van.

    1996-01-01

    The high-temperature reactor development in the Netherland is embedded in the WHITE reactor program, in which several Dutch research institutes and engineering companies participate. The activities within the WHITE program are focused on the development of a small scale HTS for combined heat and power generation. In 1995, design choices for a pebble bed reactor were made at ECN. The first concept HTR will gave a closed cycle helium turbine and a power level of 40 MWth. It is intended to make the market introduction of a commercially competitive HTR feasible. The design will be an optimization of the Peu-a-Peu (PAP) concept of KFA Juelich. Computer codes necessary for the evaluation of reactor physics aspects of this reactor are developed in cooperation with international partners. An evaluation of a 20 MWth PAP concept showed that the maximum fuel termmperature after depressurization does not exceed 1300 C. (orig.)

  11. The receptor-like kinase SOBIR1 interacts with Brassica napus LepR3 and is required for Leptosphaeria maculans AvrLm1-triggered immunity

    Directory of Open Access Journals (Sweden)

    Lisong eMa

    2015-10-01

    Full Text Available AbstractThe fungus Leptosphaeria maculans (L. maculans is the causal agent of blackleg disease of canola/oilseed rape (Brassica napus worldwide. We previously reported cloning of the B. napus blackleg resistance gene, LepR3, which encodes a receptor-like protein. LepR3 triggers localised cell death upon recognition of its cognate Avr protein, AvrLm1. Here, we exploited the Nicotiana benthamiana model plant to investigate the recognition mechanism of AvrLm1 by LepR3. Co-expression of the LepR3/AvrLm1 gene pair in N. benthamiana resulted in development of a hypersensitive response (HR. However, a truncated AvrLm1 lacking its indigenous signal peptide was compromised in its ability to induce LepR3-mediated HR, indicating that AvrLm1 is perceived by LepR3 extracellularly. Structure-function analysis of the AvrLm1 protein revealed that the C-terminal region of AvrLm1 was required for LepR3-mediated HR in N. benthamiana and for resistance to L. maculans in B. napus. LepR3 was shown to be physically interacting with the B. napus receptor like kinase, SOBIR1 (BnSOBIR1. Silencing of NbSOBIR1 or NbSERK3 (BAK1 compromised LepR3-AvrLm1-dependent HR in N. benthamiana, suggesting that LepR3-mediated resistance to L. maculans in B. napus requires SOBIR1 and BAK1/SERK3. Using this model system, we determined that BnSOBIR1 and SERK3/BAK1 are essential partners in the LepR3 signalling complex and were able to define the AvrLm1 effector domain.

  12. Applications of instrumental neutron activation analysis in the Analytical Division of the Research Center Juelich (KFA)

    International Nuclear Information System (INIS)

    Erdtmann, G.

    1991-12-01

    The Radioanalytical Chemistry Section, as a part of the Central Division of Chemical Analysis of the Research Center KFA Juelich, has the task to provide and to apply nuclear methods in the analytical service for the institutes and projects of the KFA and to customers outside. A great part of this service is trace element determinations by neutron activation analysis using the research reactor FRJ-2. The procedure for the instrumental technique is described and mainly practical aspects are reported in detail. It is based on the k 0 -method developed by Simonits and DeCorte. The results are calculated from the peak areas of the γ-lines and the corresponding k 0 -factors. A new variant of this procedure is required, if the program used for the deconvolution of the γ-spectra provides absolute decay rates of the radionuclides instead of the γ-emission rates. This variant is also described. Some examples of analyses carried out in the analytical service are presented and discussed mainly with respect to accuracy of the results and detection limits. (orig.) [de

  13. Study on the construction of a combined cooler-synchroton ring at the KFA Juelich (COSY study)

    International Nuclear Information System (INIS)

    Gaul, G.; Hagedoorn, H.; Heide, J.A. van der; Hinterberger, F.; Huber, M.; Jahn, R.; Mayer-Kuckuk, T.; Paetz genannt Schieck, H.; Berg, G.; Hardt, A.; Martin, S.; Osterfeld, F.; Prasuhn, D.; Riepe, G.; Rogge, M.; Rossen, P. von; Schult, O.W.B.; Speth, J.; Turek, P.

    1984-02-01

    The project of a storage ring for the extension of the nuclear physics research facilities at the KFA Juelich is presented. Together with the construction of the ring the possibilities for physical research are described. (HSI) [de

  14. Structure-function analysis of the Fusarium oxysporum Avr2 effector allows uncoupling of its immune-suppressing activity from recognition

    NARCIS (Netherlands)

    Di, X.; Cao, L.; Hughes, R.K.; Tintor, N.; Banfield, M.J.; Takken, F.L.W.

    2017-01-01

    Plant pathogens employ effector proteins to manipulate their hosts. Fusarium oxysporum f. sp. lycopersici (Fol), the causal agent of tomato wilt disease, produces effector protein Avr2. Besides being a virulence factor, Avr2 triggers immunity in I-2 carrying tomato (Solanum lycopersicum). Fol

  15. Pembuatan Alat Ukur Kadar Alkohol Pada Minuman Menggunakan Sensor TGS822 Berbasis Mikrokontroler AVR ATMEGA8535.

    OpenAIRE

    Afniza

    2011-01-01

    Alcohol in Indonesia have been common and socially acceptable. But often in excessive consumption. For that we need a tool to test the alcohol content quickly and accurately. Alcohol Concentration Measuring Tool Using AVR Microcontroller- Based Sensor TGS822 ATMega 8535 suitable for the test. This system is done by changing the analog data from sensors TGS 822 into digital and then transmit the data to the AVR microcontroller ATMega 8535 and displayed via the LCD (Liquid Crysta...

  16. Incineration of dry burnable waste from reprocessing plants with the Juelich incineration process

    International Nuclear Information System (INIS)

    Dietrich, H.; Gomoll, H.; Lins, H.

    1987-01-01

    The Juelich incineration process is a two stage controlled air incineration process which has been developed for efficient volume reduction of dry burnable waste of various kinds arising at nuclear facilities. It has also been applied to non nuclear industrial and hospital waste incineration and has recently been selected for the new German Fuel Reprocessing Plant under construction in Wackersdorf, Bavaria, in a modified design

  17. PERANCANGAN SMARTHOME DENGAN RASBERRY BERBASIS WIRELESS MENGGUNAKAN MICROKONTOLLER AVR ATMEGA328 DAN FUZZY LOGIC

    Directory of Open Access Journals (Sweden)

    Desmira Desmira

    2016-11-01

    Full Text Available Kebutuhan Energi listrik yang diperlukan untuk kehidupan manusia sehari-hari sangat meningkat. Penggunaan energi listrik dewasa ini tidak sesuai dengan penggunaan pada umumnya yang membiarkan peralatan elektronik atau penerangan pada sebuah rumah tetap menyala pada saat tidak dipergunakan. Dengan peristiwa tersebut, terpikirlah untuk menghasilkan Smarthome yang dapat mengontrol peralatan elektronik dari jauh menggunakan Mini PC Raspberry pi dan Smartphone android. Pemanfaatan Raspberry pi adalah untuk mengontrol dan memonitoring peralatan elektronik melalui web yang diperintahkan oleh Smartphone Android, kemudian dikomunikasikan kepada mikrokontroller AVR ATMega328 untuk menyalakan relay yang tersambung ke perangkat elektronik. Berdasarkan pembahasan dan pengujian, dapat ditemukan kelebihan dan kelemahan rancangan Smarthome menggunakan Raspberry berkomunikasi Wireless berbasis Microcontroller AVR ATMega328. Dalam sistem ini, user melakukan komunikasi dari aplikasi android menuju Raspberry yang kemudian dikirimkan oleh transiver dan diterima oleh reciver lalu dibaca oleh mikrokontroler yang membuat lampu atau alat elektronik menyala. Dengan demikian dapat memberikan solusi dalam mengontrol perangkat elektronik yang ada di dalam rumah dengan cara pengontrolan terpusat pada sebuah Smartphone android disertai media internet yang meringankan kerja manusia dan mengoptimalisasikan kenyamanan dan keamanaan dari sebuah rumah. Kata kunci: rumah pintar, mini pc, raspberry pi, android smartphone, mikrokontroler AVR ATmega328.

  18. Forschungszentrum Juelich GmbH, Institut fuer Kernphysik. Annual report 1991

    International Nuclear Information System (INIS)

    1992-03-01

    During the year 1991 the work concentrated on building the Cooler Synchrotron COSY-Juelich. The experimental activity was accompanied by theoretical studies in the field of medium energy nuclear physics. The preparation of COSY experiments concerned the realization of multipurpose facilities and other experiments. Work going on in the field of theoretical nuclear physics was strongly connected with research projects at COSY and processes induced in the proton-antiproton interaction. Through nuclear spectroscopy two-phonon octupole excitation had been identified in Gd-148. The ISIS ECR source has been used for the production of ion beams for atomic physics research. (DG)

  19. Demolition to Green-Field conditions of the FRJ-1 (MERLIN) research reactor. Successes and hurdles in the demolition of a research reactor of the megawatt class; Der Rueckbau des Forschungsreaktors FRJ-1 (MERLIN) bis zur 'Gruenen Wiese'. Erfolge und Huerden beim Rueckbau eines Forschungsreaktors der Megawatt-Klasse

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, Burkhard; Printz, Rudolf; Matela, Karel; Zehbe, Carsten; Stauch, Bernhard; Zander, Iven [Forschungszentrum Juelich GmbH, Juelich (Germany)

    2010-02-15

    The Juelich-1 Research Reactor (FRJ-1), also referred to as MERLIN (Medium Energy Research Light Water Moderated Industrial Nuclear Reactor), was a light-water moderated and cooled swimming pool reactor of British design. The cornerstone in the erection of the reactor building was laid on June 11, 1958. Reactor operation was started on February 23, 1962. The plant was last run at a thermal power of 10 MW and shut down for good in 1985 after 23 years of operation. After the fuel elements had been removed and most of the experimental installations dismantled, some first steps towards demolition were taken in 1995. Demolition on a large scale began in 1996. September 8, 2008 was a special day: On the area of the former reactor hall, an oak tree was planted as a symbol of the 'green field' and of the original oak wood which had to make way for the construction of reactors in Juelich. An oak tree now stands in the place of the reactor unit. Was that all? It was not, for there were ancillary systems, operations, utility and hygiene buildings which had to be pulled down. Decontamination and clearance measurements were completed. The application for clearance was prepared and completed. Conventional demolition was started in 2009. After completion of that step, the last chapter about demolition of the FRJ-1 research reactor has been written, and the book can be closed. (orig.)

  20. Decontamination of nuclear graphite by thermal processing; Dekontamination von Nukleargraphit durch thermische Behandlung

    Energy Technology Data Exchange (ETDEWEB)

    Florjan, Monika W.

    2010-04-15

    The main problem in view of the direct disposal of the nuclear graphite is its large volume. This waste contains long-lived and short-lived radionuclides which determine the waste strategy. The irradiated graphite possess high amount of the {sup 14}C isotope. The main object of the present work was the selective separation of {sup 14}C isotope from the isotope {sup 12}C by thermal treatment (pyrolysis, partial oxidation). A successful separation could reduce the radiotoxicity and offer a different disposal strategy. Three different graphite types were investigated. The samples originate from the reflector and from the flaking of spherical fuel elements of the high-temperature reactor (AVR) Juelich. The samples from the thermal column of the research reactor (Merlin, Juelich) were also investigated. The maximum tritium releases were obtained both in inert gas atmosphere (N{sub 2}) and under water vapour-oxidizing conditions at 1280 C and 900 C. Furthermore it could be shown that 28% of {sup 14}C could be released under inert gas conditions at a 1280 C. By additive of oxidizing agent such as water vapour and oxygen the {sup 14}C release could be increased. Under water vapour-oxidizing conditions at a temperature of 1280 C up to 93% of the {sup 14}C was separated from the graphite. The matrix corrosion of 5.4% was obtained. The selective separation of the {sup 14}C is possible, because a substantial part of the radiocarbon is bound near the grain boundary surfaces. (orig.)

  1. Main research activities at the Institute of Energy Process Engineering Research Centre Juelich Germany

    International Nuclear Information System (INIS)

    Achenbach, E.

    1995-06-01

    This report summarizes four lectures been held during the author's seven-week stay at the Department of High Temperature Engineering in the period from February 2nd to March 23rd in 1995 under the JAERI foreign researcher inviting program. Though the Institute of Energy Process Engineering(IEV) in the Research Centre Juelich(KFA), has recently changed the subject of research from nuclear technology of high-temperature gas-cooled reactors(HTGRs) to fuel cell technology, there are many common items of research. In particular, the following topics presented in the lectures are of mutual interest: 1)Methane-steam reforming used at JAERI as HTGR heat utilization system and applied at KFA to internal reforming in the high temperature Solid Oxide Fuel Cell(SOFC), 2)Technology and modeling of high temperature electrolysis at JAERI as the inverse process of the SOFC developed at KFA, 3)Flow simulation of branched systems treated at JAERI for the development of high temperature heat exchangers and performed at KFA with respect to the SOFC manifold system, 4)Fundamental aspects of heat and mass transfer. The report should help to create a basis of discussing the above mentioned problems and to stimulate the research work at JAERI. (author)

  2. The Relationship of ST Segment Changes in Lead aVR with Outcomes after Myocardial Infarction; a Cross Sectional Study

    Directory of Open Access Journals (Sweden)

    Mohammad Reza Beyranvand

    2017-01-01

    Full Text Available Introduction: Among the 12 leads studied in electrocardiography (ECG, lead aVR can be considered as the most forgotten part of it since no attention is paid to it as the mirror image of other leads. Therefore, the present study has been designed with the aim of evaluating the prevalence of ST segment changes in lead aVR and its relationship with the outcome of these patients.Methods: In this retrospective cross sectional study medical profiles of patients who had presented to emergency department with the final diagnosis of myocardial infarction (MI in a 4-year period were evaluated regarding changes of ST segment in lead aVR and its relationship with in-hospital mortality, the number of vessels involved, infarct location and cardiac ejection fraction.Results: 288 patients with the mean age of 59.00 ± 13.14 (18 – 91 were evaluated (79.2% male. 168 (58.3% patients had the mentioned changes (79.2% male. There was no significant relationship between presence of ST changes in lead aVR with infarct location (p = 0.976, number of vessels involved (p = 0.269 and ejection fraction on admission (p = 0.801. However, ST elevation ≥ 1 mv in lead aVR had a significant relationship with mortality (Odds = 7.72, 95% CI: 3.07 – 19.42, p < 0.001. Sensitivity, specificity, positive and negative predictive values and positive and negative likelihood ratios of ST elevation ≥ 1 for prediction of in-hospital mortality were 41.66 (95% CI: 22.79 – 63.05, 91.53 (95% CI: 87.29 – 94.50, 31.25 (95% CI: 16.74 – 50.13, 94.44 (95% CI: 90.65 – 96.81, 0.45 (95% CI: 0.25 – 0.79, and 0.05 (95% CI: 0.03 – 0.09, respectively.Conclusion: Based on the results of the present study, the prevalence of ST segment changes in lead aVR was estimated to be 58.3%. There was no significant relationship between these changes and the number of vessels involved in angiography, infarct location and cardiac ejection fraction. However, presence of ST elevation ≥ 1 in lead aVR

  3. Intramolecular interaction influences binding of the Flax L5 and L6 resistance proteins to their AvrL567 ligands.

    Directory of Open Access Journals (Sweden)

    Michael Ravensdale

    Full Text Available L locus resistance (R proteins are nucleotide binding (NB-ARC leucine-rich repeat (LRR proteins from flax (Linum usitatissimum that provide race-specific resistance to the causal agent of flax rust disease, Melampsora lini. L5 and L6 are two alleles of the L locus that directly recognize variants of the fungal effector AvrL567. In this study, we have investigated the molecular details of this recognition by site-directed mutagenesis of AvrL567 and construction of chimeric L proteins. Single, double and triple mutations of polymorphic residues in a variety of AvrL567 variants showed additive effects on recognition strength, suggesting that multiple contact points are involved in recognition. Domain-swap experiments between L5 and L6 show that specificity differences are determined by their corresponding LRR regions. Most positively selected amino acid sites occur in the N- and C-terminal LRR units, and polymorphisms in the first seven and last four LRR units contribute to recognition specificity of L5 and L6 respectively. This further confirms that multiple, additive contact points occur between AvrL567 variants and either L5 or L6. However, we also observed that recognition of AvrL567 is affected by co-operative polymorphisms between both adjacent and distant domains of the R protein, including the TIR, ARC and LRR domains, implying that these residues are involved in intramolecular interactions to optimize detection of the pathogen and defense signal activation. We suggest a model where Avr ligand interaction directly competes with intramolecular interactions to cause activation of the R protein.

  4. Status and perspectives of the dismantling of nuclear power plants in Germany (Dismantling monitoring 2015); Stand und Perspektiven des Rueckbaus von Kernkraftwerken in Deutschland (''Rueckbau-Monitoring 2015'')

    Energy Technology Data Exchange (ETDEWEB)

    Wealer, Ben; Seidel, Jan Paul [Technische Univ. Berlin (Germany); Gerbaulet, Clemens; Hirschhausen, Christian von [Technische Univ. Berlin (Germany); Deutsches Institut fuer Wirtschaftsforschung, Berlin (Germany)

    2015-11-15

    The dismantling monitoring 2015 covers the nuclear power plants HDR Grosswelzheim, Niederaichbach (KKN), MZFR Karlsruhe, Lingen (KWL), Gundremmingen unit A (KRB-A), VAK Kahl, Muehlheim-Kaerlich (KMK), THTR-300 Hamm-Uentrop, AVR Juelich, Greifswald (KGR 1-5), KNK II Karlsruhe, Rheinsberg (KKR), Wuergassen (KWW), Stade (KKS), Obrigheim (KWO), SNR 300. The post-operational phase activities of other shut-down nuclear power plants and the active companies are summarized.

  5. International working group on gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-15

    The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

  6. Utilidad de la derivación aVR en la identificación de la arteria responsable en el infarto inferior

    Directory of Open Access Journals (Sweden)

    Yanina B. Castillo Costa

    2006-01-01

    Full Text Available El ECG es el método más simple para el diagnóstico de infarto inferior (IAMinf; sin embargo, su utilidad disminuye para determinar el vaso responsable. Objetivos 1. Analizar la utilidad del desnivel del ST en aVR en la identificación de oclusión de las arterias coronaria derecha (CD o circunfleja (CX. 2. Evaluar la utilidad agregada a los criterios clásicos. Material y métodos Se incluyeron en el estudio 65 pacientes con IAMinf a quienes se les realizó una cinecoronariografía (CCG durante la internación. Se analizaron los criterios clásicos, así como el desnivel del ST en aVR: ?ST =?1 mm, ?ST =?1 mm o nivelado (isoST a 0,08 seg del punto J y se calcularon la sensibilidad (S, la especificidad (E, el valor predictivo positivo (VPP, el valor predictivo negativo (VPN y el likelihood ratio (LR. Se incluyó la derivación aVR en un análisis multivariado para determinar su valor agregado a los criterios clásicos. Resultados La CD fue responsable de la oclusión en 47 pacientes (72% y la CX en 18 (28%. Signos clásicos : CD : la relación del ?STD3/D2 =?1 se halló en 52 pacientes (91,5%; p = 0,001 e ?STD1-aVL =?1 mm en 34 (61%; p 1 en 13 pacientes (50%; p = 0,001 y ?STV5-V6 =?1 mm en 12 (44%; p < 0,001. Derivación aVR : CD : ?ST o isoST en 46 pacientes (81%; p < 0,001; CX : ?ST =?1 mm en 19 (56%; p < 0,001. En el análisis multivariado, el ?ST o isoST en aVR identificó a la CD (OR IC 95% 4,7 (1,1-19,8; p = 0,03, mientras que el ?ST aVR identificó a la CX. Se propone un algoritmo diagnóstico para CD que refleja el valor aditivo de aVR a la presencia de los signos clásicos: ?ST o isoST en aVR (VPP 94%; LR+: 2. Conclusión La derivación aVR es una herramienta útil para la identificación de la arteria responsable en el IAMinf, ya que brinda información diagnóstica adicional a los signos clásicos.

  7. Modular high-temperature reactor launched (and wallchart)

    International Nuclear Information System (INIS)

    Steinwarz, W.

    1987-01-01

    In view of the need for a technically unsophisticated, safe and economic reactor system, the KWU group has integrated the experience gained from German light-water reactor engineering and from successful operation of the German AVR experimental high-temperature reactor into the development of the High-Temperature Reactor (HTR)-module. The main components are illustrated and explained and technical data for the HTR-module is given. Safety is also considered. This includes graphs of core heat-up temperature for pebble-bed HTR and a graph of the temperature load of the fuel elements. The operation, control and applications are considered. The latter includes use in combined heat and power generation and community heating. Feasibility studies have shown that the HTR-module is cheaper, comparatively, than coal-fired power stations. (U.K.)

  8. ST Elevation in aVR with Coexistent Multilead ST Depression

    Directory of Open Access Journals (Sweden)

    Benjamin Cooper

    2017-01-01

    Full Text Available History of present illness: An 80-year-old female with a history of Crohn’s disease presented to the emergency department with chest pain. She had two weeks of exertional chest pain that preceded an episode of chest pain immediately prior to arrival associated with diaphoresis. Her pain nearly completely resolved with sublingual nitroglycerin provided by pre-hospital personnel. She was hemodynamically stable with normal vital signs on arrival. An ECG was immediately obtained. Significant findings: The ECG shows ST-segment depressions in precordial leads V3 through V6, and limb leads I, II, and aVL, and 1 mm of ST-segment elevation in aVR. The initial troponin I was elevated at 1.37 ng/mL (upper limit of normal 0.40. Cardiology decided to delay catheterization until the next day when diffuse coronary disease was discovered (including 90% of the left circumflex stenosis, 60% proximal and 75% mid-left anterior descending stenosis, 75% third diagonal branch stenosis, and 90% posterior descending artery stenosis. The following day, the patient went to the operating room for coronary artery bypass grafting (CABG. Discussion: Traditionally, lead aVR has not received attention when interpreting acutely ischemic changes on ECG, leading some to refer to it as “the forgotten lead.”1 Current guidelines acknowledge the significance of multilead ST depression with coexistent ST elevation in aVR, and this pattern has been identified as the strongest predictor of severe left main coronary artery and/or 3-vessel disease (LM/3VD.2-3 When this ECG pattern is recognized in patients with ischemic symptoms, the emergency physician should involve cardiology early. When managing patients with suspected LM/3VD, it is important to withhold dual anti-platelet therapy as CABG is likely to be indicated,1,3 and guidelines recommend discontinuing P2Y12 inhibitors like clopidogrel or ticagrelor at least 24 hours prior to urgent CABG.2

  9. Radiation protection monitoring for #betta#-radiation at the Juelich Nuclear Research Centre

    International Nuclear Information System (INIS)

    Keller, M.; Heinzelmann, M.

    1983-01-01

    A complete system for radiation protection monitoring also includes #betta#-radiation monitoring. This requires suitable dose rate meters, personal dosemeters and last but not least detailed information about possible radiation exposure due to #betta#-radiation. Since there are at present no suitable #betta#-dosemeters available on the market yet, a large nuclear research centre such as the KFA Juelich, where radioactive substances are being handled by more than 1600 persons, has the task of developing and deploying suitable dosemeters. The centre's accomplishments in this area are described

  10. PeDaB - the personal dosimetry database at the research centre Juelich

    International Nuclear Information System (INIS)

    Geisse, C.; Hill, P.; Paschke, M.; Hille, R.; Schlaeger, M.

    1998-01-01

    In May, 1997 the mainframe based registration, processing and archiving of personal monitoring data at the research centre Juelich (FZJ) was transferred to a client server system. A complex database application was developed. The client user interface is a Windows based Microsoft ACCESS application which is connected to an ORACLE database via ODBC and TCP/IP. The conversion covered all areas of personal dosimetry including internal and external exposition as well as administrative areas. A higher degree of flexibility, data security and integrity was achieved. (orig.) [de

  11. The high-temperature reactor's attractiveness lies in passive safety

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    In the recent years the use of nuclear energy has turned from a technical and scientific issue to a political one. The high-temperature reactor (HTR) however, has always been advertised as particularly safe. The present situation and future developments of HTR-technology were the two issues that VDI-News brought up on the 27th October on an HTR-conference in an interview with the 'spiritual father' of the HTR, Prof. Dr. Rudolf Schulten of the Juelich Nuclear Research Centre. (orig.) [de

  12. Work report 1999 of the Safety and Radiation Protection Department

    International Nuclear Information System (INIS)

    Hille, R.; Frenkler, K.L.

    2000-05-01

    Research Centre Juelich is a member of the Hermann von Helmholtz Association of German Research Centres (HGF) in which Germany's 16 research institutions are joined together. The Centre's mission is future-oriented basic research and application-oriented research and development. The Centre's research and development activities are subsumed under the following five research priorities: - structure of matter and materials research, - energy technology, - information technology, - environmental precaution research, - life sciences. In order to perform its research tasks, the Research Centre also operates facilities in which radioactive substances are handled or ionizing radiation generated. The following facilities currently in operation are of particular significance in this respect due to their activity inventory or accelerator power: - DIDO research reactor (FRJ-2), - Large Hot Cells, chemistry cells, - interim store for spent AVR fuel elements, - decontamination operations with waste store and waste cells, - TEXTOR fusion experiment, - accelerator facilities such as COSY, JULIC, Compact and BABY cyclotron, - radionuclide laboratories in the fields of chemistry and medicine. The MERLIN research reactor (FRJ-1) has been shut down since 1985 and the fuel discharged, the decommissioning licensing procedure under the Atomic Energy Act is about to be completed. The high-temperature experimental reactor, AVR, has been shut down since the end of 1988 and received a decommissioning licence in 1994. Although it is not part of the Research Centre according to company law, it is supervised by the Research Centre with respect, amongst other aspects, to waste management and fuel element disposal. Core discharge was completed in 1997. Various nuclear facilities licensed pursuant to paragraph 9 of the Atomic Energy Act are being dismantled. This includes the fuel cells and several laboratories at the IFF and IWV-2 institutes. (orig.) [de

  13. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Travelli, A [Argonne National Laboratory, Argonne, IL (United States)

    1983-09-01

    Conversion of research and test reactor cores from the use of high enrichment uranium to the use of low enrichment uranium depends on the cooperation of many research organizations, reactor operators, and government agencies. At a technical level, it involves almost all aspects of the fuel cycle, including fuel development, testing, shipping and reprocessing; experiment performance; economics; and safety and licensing aspects. The reactors involved and the conversion activities are distributed among approximately 25 countries, making this a subject which is best dealt with on an international basis. To foster direct communication in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the fifth of a series which began in 1978. The previous meetings were held at Argonne (International Meeting of Research Reactor Fuel Designers, Developers, and Fabricators, Argonne National Laboratory, Argonne, U.S.A., November 910, 1978), at Saclay (IAEA Consultants' Meeting on Research Reactor Core Conversions from HEU to LEU, Centre d'etudes Nucleaires de Saclay, Saclay, France, December 12-14, 1979), at Argonne (International Meeting on Development, Fabrication and Application of Reduced Enrichment Fuels for Research and Test Reactors, Argonne National Laboratory, Argonne, U.S.A., November 12-14, 1980) and at Juelich (Seminar on Research Reactor Operation and Use, Juelich Nuclear Research Center, Juelich, F.R.G., September 48, 1981). Proceedings from the two most recent previous meetings were published as ANL/RERTR/TM-3 (CONF-801144) and IAEA-SR-77. The spirit of this meeting differs slightly from that of the previous meetings. The advances which have been made and the growing maturity of the effort have caused a gradual shift of emphasis away from those topics which dominated the floor during the first meetings, such as fuel and methods development, and towards topics which concern more

  14. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    International Nuclear Information System (INIS)

    Travelli, A.

    1983-09-01

    Conversion of research and test reactor cores from the use of high enrichment uranium to the use of low enrichment uranium depends on the cooperation of many research organizations, reactor operators, and government agencies. At a technical level, it involves almost all aspects of the fuel cycle, including fuel development, testing, shipping and reprocessing; experiment performance; economics; and safety and licensing aspects. The reactors involved and the conversion activities are distributed among approximately 25 countries, making this a subject which is best dealt with on an international basis. To foster direct communication in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the fifth of a series which began in 1978. The previous meetings were held at Argonne (International Meeting of Research Reactor Fuel Designers, Developers, and Fabricators, Argonne National Laboratory, Argonne, U.S.A., November 910, 1978), at Saclay (IAEA Consultants' Meeting on Research Reactor Core Conversions from HEU to LEU, Centre d'etudes Nucleaires de Saclay, Saclay, France, December 12-14, 1979), at Argonne (International Meeting on Development, Fabrication and Application of Reduced Enrichment Fuels for Research and Test Reactors, Argonne National Laboratory, Argonne, U.S.A., November 12-14, 1980) and at Juelich (Seminar on Research Reactor Operation and Use, Juelich Nuclear Research Center, Juelich, F.R.G., September 48, 1981). Proceedings from the two most recent previous meetings were published as ANL/RERTR/TM-3 (CONF-801144) and IAEA-SR-77. The spirit of this meeting differs slightly from that of the previous meetings. The advances which have been made and the growing maturity of the effort have caused a gradual shift of emphasis away from those topics which dominated the floor during the first meetings, such as fuel and methods development, and towards topics which concern more

  15. Demolition of the FRJ-1 research reactor (MERLIN)

    International Nuclear Information System (INIS)

    Stahn, B.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

    2003-01-01

    FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [de

  16. Comparison of measured and calculated burn-up of AVR-Fuel-Elements

    Energy Technology Data Exchange (ETDEWEB)

    Wagemann, R.

    1974-03-15

    Burn-up comparisons are made for small batches of three types of AVR fuel elements using a coupled EREBUS-MUPO neutronic analysis compared against test results from both nondestructive gamma-ray measurements of cesium-137 activity and destructive mass spectrometry measurements of the ratio of U-233 to U-235. The comparisons are relatively good for average burn-up and reasonably good for burn-up distributions.

  17. Atmel AVR Microcontroller Primer Programming and Interfacing, Second Edition

    CERN Document Server

    Barrett, Steven F

    2012-01-01

    This textbook provides practicing scientists and engineers a primer on the Atmel AVR microcontroller. In this second edition we highlight the popular ATmega164 microcontroller and other pin-for-pin controllers in the family with a complement of flash memory up to 128 kbytes. The second edition also adds a chapter on embedded system design fundamentals and provides extended examples on two different autonomous robots. Our approach is to provide the fundamental skills to quickly get up and operating with this internationally popular microcontroller. We cover the main subsystems aboard the ATmega

  18. Nuclear Research Centre Juelich (KFA). Annual report 1984/85

    International Nuclear Information System (INIS)

    1985-09-01

    The annual report from the Nuclear Research Centre in Juelich (KFA) consist of four differently coloured parts. The 'white part' - from the research - published topical contributions from authors writing popular science. Subjects were chosen from the viewpoint of a journalist rather than oriented on the research field of the KFA. The 'yellow part' - reports on research and development - is a survey of the scientific-technological work of KFA structured according to the main fields of research of KFA. The 'green part' - research institutes and scientific-technological joint ventures - describes tasks and objectives of the individual institutes/joint ventures. The 'blue part' - organisation, administration and technical infrastructure - contains information on management, organisation and administration of KFA (personnel department, PR, finances, purchase- and material disposal department, cost and planning, cooperation and use of know-how, projects, technical equipment). (orig.) [de

  19. Hypernuclear properties derived from the Juelich hyperon-nucleon interaction (in comparison with the Nijmegen interactions)

    International Nuclear Information System (INIS)

    Yamamoto, Y.; Reuber, A.; Himeno, H.; Nagata, S.; Motoba, T.

    1992-01-01

    The G-matrix interactions are derived from the Juelich YN interaction models A and B, compared with those from the Nijmegen models. The DDHF calculations for heavy Λ hypernuclei and the shell-model analysis for spin-doublet states of light hypernuclei are performed by use of the G-matrix interactions. It is demonstrated that the OBE models can be tested by the hypernuclear calculations. (author) 3 tabs., 5 figs., 23 refs

  20. Raising the four downcomers in the reactor aluminium tank of the FRJ-2 research reactor as an example of the execution of complicated work in the region of high radiation levels

    International Nuclear Information System (INIS)

    Nickel, M.; Schmitz, J.; Wolters, J.

    1975-02-01

    As a result of the planned power increase from 15 MW to 25 MW, a new emergency cooling system had to be installed in the research reactor FRJ-2 of the KFA Juelich, which called for an extension of the four standpipes in the reactor tank by 57 mm. Due to the high radiation level in the reactor tank, new techniques had to be found allowing aluminium rings of corresponding height to be welded onto the top part of the standpipes by remotecontrolled welding; moreover, the welded parts were then to be protected by a bandage made of high-quality steel. The development work was carried out in the KFA and this report gives an account of the technique applied and the results obtained. (author)

  1. Proceedings of the spring meeting of the Study Group for Electronic Instrumentation on April 3-5, 1995 in Juelich

    International Nuclear Information System (INIS)

    1995-06-01

    The 17 papers presented at the meeting give a survey of recent progress in the field of electronic instrumentation of scientific and technical measuring instruments, achieved at the research centers of KFA Juelich, TU Graz, HMI Berlin, Karlsruhe FZ, and Rossendorf FZ. (DG) [de

  2. Scientific report. Plasma-wall interaction studies related to fusion reactor materials

    International Nuclear Information System (INIS)

    Temmerman, G. De

    2006-01-01

    This scientific report summarises research done on erosion and deposition mechanisms affecting the optical reflectivity of potential materials for use in the mirrors used in fusion reactors. Work done in Juelich, Germany, at the Federal Institute of Technology in Lausanne, Switzerland, the JET laboratory in England and in Basle is discussed. Various tests made with the mirrors are described. Results obtained are presented in graphical and tabular form and commented on. The influence of various material choices on erosion and deposition mechanisms is discussed

  3. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    International Nuclear Information System (INIS)

    Shropshire, D.E.; Herring, J.S.

    2004-01-01

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

  4. Identification of distinct specificity determinants in resistance protein Cf-4 allows construction of a Cf-9 mutant that confers recognition of avirulence protein AVR4

    NARCIS (Netherlands)

    Hoorn, Van der R.A.L.; Roth, R.; Wit, De P.J.G.M.

    2001-01-01

    The tomato resistance genes Cf-4 and Cf-9 confer specific, hypersensitive response-associated recognition of Cladosporium carrying the avirulence genes Avr4 and Avr9, respectively. Cf-4 and Cf-9 encode type I transmembrane proteins with extracellular leucine-rich repeats (LRRs). Compared with Cf-9,

  5. Käyttöjärjestelmä AVR-mikro-ohjaimelle

    OpenAIRE

    Koivuranta, Janne

    2011-01-01

    Tämän opinnäytetyön tavoitteena oli suunnitella ja toteuttaa Atmel AVR mikro-ohjaimella toimiva yksinkertainen käyttöjärjestelmä. Käyttöjärjestelmän ominaisuuksiin haluttiin sisällyttää laitteistopohjaisella keskeytyksellä toimiva vuorottaja, joka jakaa suoritinaikaa ajossa oleville ohjelmille. Lisäksi käyttöjärjestelmä tarjoaa tarvittavat palvelut ajossa olevien ohjelmien väliseen synkronointiin. Kaikkien käyttöjärjestelmän palveluiden suoritukseen kuluva aika on ennakoitavissa, joten kysees...

  6. The conversion of the DIDO-type reactor FRJ-2. Studies and conclusions

    International Nuclear Information System (INIS)

    Stroemich, A.; Siebertz, Ch.; Wickert, M.

    1985-01-01

    For the FRJ-2 (23 MW) of the KFA-Juelich the conversion from HEU- to LEU-fuel was investigated. Before starting the conversion calculations our methods were qualified for the application to heavy water moderated research reactors. A combination of LEU-elements with two different U-235 loadings of 180 g and 225 g was found as suitable for conversion. With these LEU-elements a working core and a transition phase was calculated. The change of the mechanical fuel element design was taken into account. (author)

  7. Implementation of a Microcode-controlled State Machine and Simulator in AVR Microcontrollers (MICoSS

    Directory of Open Access Journals (Sweden)

    S. Korbel

    2005-01-01

    Full Text Available This paper describes the design of a microcode-controlled state machine and its software implementation in Atmel AVR microcontrollers. In particular, ATmega103 and ATmega128 microcontrollers are used. This design is closely related to the software implementation of a simulator in AVR microcontrollers. This simulator communicates with the designed state machine and presents a complete design environment for microcode development and debugging. These two devices can be interconnected by a flat cable and linked to a computer through a serial or USB interface.Both devices share the control software that allows us to create and edit microprograms and to control the whole state machine. It is possible to start, cancel or step through the execution of the microprograms. The operator can also observe the current state of the state machine. The second part of the control software enables the operator to create and compile simulating programs. The control software communicates with both devices using commands. All the results of this communication are well arranged in dialog boxes and windows. 

  8. Rapid data acquisition from the safety system of the FRJ-2 reactor

    International Nuclear Information System (INIS)

    Inhoven, H.

    1980-06-01

    The central department for research reactors (ZFR) of the Juelich Nuclear Research Centre (KFA) is operating the reactors FRJ-1 (MERLIN) and FRJ-2 (DIDO) since 1962. In 1976, a Siemens 330 computer has been put into operation especially for the processing of data from the DIDO reactor, followed by another computer of the same type for the purpose of processing data from the ZFR department in general. The present report is a result of the work investigating 'Data acquisition and data processing in the FRJ-2' and primarily discusses the complex of 'fast analog and binary signals'. The activities in this field of work have been and still are mainly concerned with general problems encountered in adapting a currently 14-year-old reactor system to a digital computer, namely problems such as data decoupling in the safety system of the reactor, data acquisition using the CAMAC system, data transfer via an 'extended branch', data acquisition software as core-resident programs, temporary storage as common data, interpreting software as peripheral - storage - resident programs. (orig./WB) [de

  9. The FRJ 1 reactor (MERLIN) at Juelich, F.R. Germany and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the FRJ 1 reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  10. Research center Juelich to install Germany's most powerful supercomputer new IBM System for science and research will achieve 5.8 trillion computations per second

    CERN Multimedia

    2002-01-01

    "The Research Center Juelich, Germany, and IBM today announced that they have signed a contract for the delivery and installation of a new IBM supercomputer at the Central Institute for Applied Mathematics" (1/2 page).

  11. Noise thermometer

    Energy Technology Data Exchange (ETDEWEB)

    Von Brixy, H. [Forschungszentrum Juelich GmbH (Germany); Kakuta, Tsunemi

    1996-03-01

    The noise thermometry (NT) is a temperature measuring method by which the absolute temperature measurement can be performed with a very high accuracy and without any influence of ambient environments and of the thermal history of its NT sensor (electric resistor). Hence it is quite suitable for application as a standard thermometry to the in-situ temperature calibration of incore thermocouples. The KFA Juelich had played a pioneering role in the development of NT and applied the results successfully to the AVR for testing its feasibility. In this report, all about the NT including its principle, sensor elements and system configurations are presented together with the experiences in the AVR and the results of investigation to apply it to high temperature measurement. The NT can be adopted as a standard method for incore temperature measurement and in situ temperature calibration in the HTTR. (author). 85 refs.

  12. Noise thermometer

    International Nuclear Information System (INIS)

    Von Brixy, H.; Kakuta, Tsunemi.

    1996-03-01

    The noise thermometry (NT) is a temperature measuring method by which the absolute temperature measurement can be performed with a very high accuracy and without any influence of ambient environments and of the thermal history of its NT sensor (electric resistor). Hence it is quite suitable for application as a standard thermometry to the in-situ temperature calibration of incore thermocouples. The KFA Juelich had played a pioneering role in the development of NT and applied the results successfully to the AVR for testing its feasibility. In this report, all about the NT including its principle, sensor elements and system configurations are presented together with the experiences in the AVR and the results of investigation to apply it to high temperature measurement. The NT can be adopted as a standard method for incore temperature measurement and in situ temperature calibration in the HTTR. (author). 85 refs

  13. Case Report: Cardiac Rehabilitation in a Patient with MVR & AVR & Tricuspid Valve Repair

    Directory of Open Access Journals (Sweden)

    Babak Gousheh

    2003-01-01

    Full Text Available Patient is a 24 year .old male with valvular heart disease, severe mitral & aortic & tricuspid valve stenosis and regurgitation. After MVR & AVR & tricuspid surgical repair, he has undergone cardiac rehabilitation for 8 weeks (24 sittings. After completion of a cardiac rehabilitation, review of cardiovascular tests showed obvious improvement in the functional capacity, blood pressure and heart rate. Physically and mentally patient feels very comfortable and hopeful of a good healthy life.

  14. Analisis Pengaruh Konsentrasi Gas LPG Menggunakan Sensor TGS 2610 Berbasis Mikrokontroler AVR ATMega8535

    OpenAIRE

    Nurhalimah

    2011-01-01

    Telah dilakukan analisis kuantitatif gas dalam LPG. Penelitian ini dilakukan untuk mengukur konsentrasi gas LPG terhadap sensor. Metoda yang digunakan untuk mengukur konsentrasi gas LPG yaitu sensor gas semikonduktor jenis TGS 2610 keluaran Figaro yang digunakan untuk mendeteksi keberadaan gas. Sementara yang menjadi pusat pengendalian dari seluruh alat yang dirancang digunakan mikrokontroler AVR ATMega8535. Selain itu sistem yang dirancang dilengkapi LCD sebagai tampilan nilai konsentrasi ga...

  15. State of development of high temperature gas-cooled reactors in foreign countries

    International Nuclear Information System (INIS)

    Sudo, Yukio

    1990-01-01

    Emphasis has been placed in the development of high temperature gas-cooled reactors on high thermal efficiency as power reactors and the reactor from which nuclear heat can be utilized. In U.K., as the international project 'Dragon Project', the experimental Dragon reactor for research use with 20 MWt output and exit coolant temperature 750 deg C was constructed, and operated till 1976. Coated fuel particles were developed. In West Germany, the experimental power reactor AVR with 46 MWt and 15 MWe output was operated till 1988. The prototype power reactor THTR-300 with 300 MWe output and 750 deg C exit temperature is in commercial operation. In USA, the experimental power reactor Peach Bottom reactor with 40 MWe output and 728 deg C exit temperature was operated till 1974. The prototype Fort Saint Vrain power reactor with 330 MWe output and 782 deg C exit temperature was operated till 1989. In USSR, the modular VGM with 200 MWh output is at the planning stage. Also in China, high temperature gas-cooled reactors are at the design stage. Switzerland has taken part in various international projects. (K.I.)

  16. The AVR2-SIX5 gene pair is required to activate I-2-mediated immunity in tomato

    NARCIS (Netherlands)

    Ma, L.; Houterman, P.M.; Gawehns, F.; Cao, L.; Sillo, F.; Richter, H.; Clavijo-Ortiz, M.J.; Schmidt, S.M.; Boeren, S.; Vervoort, J.; Cornelissen, B.J.C.; Rep, M.; Takken, F.L.W.

    2015-01-01

    Plant-invading microbes betray their presence to a plant by exposure of antigenic molecules such as small, secreted proteins called 'effectors'. In Fusarium oxysporum f. sp. lycopersici (Fol) we identified a pair of effector gene candidates, AVR2-SIX5, whose expression is controlled by a shared

  17. Natural disulfide bond-disrupted mutants of AVR4 of the tomato pathogen Cladosporium fulvum are sensitive to proteolysis, circumvent Cf-4-mediated resistance, but retain their chitin binding ability.

    NARCIS (Netherlands)

    Burg, van den H.A.; Westerink, N.; Francoijs, C.J.J.; Roth, R.; Woestenenk, E.A.; Boeren, J.A.; Wit, de P.J.G.M.; Joosten, M.H.A.J.; Vervoort, J.J.M.

    2003-01-01

    The extracellular AVR4 elicitor of the pathogenic fungus Cladosporium fulvum induces defense responses in the tomato genotype Cf-4. Here, the four disulfide bonds of AVR4 were identified as Cys-11-41, Cys-21-27, Cys-35-80, and Cys-57-72 by partial reduction with Tris-(2-carboxyethyl)-phosphine

  18. First internal and external experiments at COSY Juelich

    International Nuclear Information System (INIS)

    Prasuhn, D.; Maier, R.; Bechstedt, U.; Dietrich, J.; Hacker, U.; Martin, S.; Stockhorst, H.; Toelle, R.; Grzonka, D.; Nake, C.; Mosel, F.

    1995-01-01

    The inauguration of the cooler synchrotron COSY Juelich was celebrated on April 1st, 1993. After the first successful acceleration to proton momenta above 800 GeV/c, beamtimes for experiments were scheduled in parallel to further machine development. The first experiment was the internal target experiment EDDA, which investigated the energy dependence of the p-p interaction. It makes use of a 3x4 μm 2 thin CH 2 fiber as an internal target. The thickness of the fiber is more than adequate to achieve high luminosities, so the intensity of the stored beam has to be reduced to 10 7 p. On the other hand, it is thin enough to achieve beam lifetimes of 3 s at 1.4 GeV/c. Details of the target fabrication and the first experimental results will be discussed. Both external experimental facilities at COSY, the time-of-flight spectrometer, and the magnetic spectrometer BIG KARL use a liquid hydrogen (deuterium) target. The first experiments were carried out at proton energies between 300 MeV and 500 MeV. Also, these experimental data will be presented. Two further internal experiments are prepared for the installation into the COSY ring. The target for the first experiment is a gas-jet target, the second experiment uses ribbon targets for the interaction. The status of both experimental setups will be shown. (orig.)

  19. Burn-up measurement in the HTR-module-reactor

    International Nuclear Information System (INIS)

    Gerhards, E.

    1993-05-01

    The burn-up status of spherical HTR-fuel elements is determined by a γ-spectrometric analysis of Cs-137 activity. The γ-spectrum recorded by a semiconductor detector up to now is analyzed by complex mathematical and time-consuming methods. For the operation of the HTR-Module-Reactor, however, a fast evaluation of the burn-up status is necessary. It is shown that this can be ensured by a comparison between the measured spectra and simulation results. Using the computer-program HTROGEN and the program system SPECCALC especially developed for this problem the γ-spectra are evaluated as a function of the burn-up status. The method is applied to results available from the operation of the AVR-reactor. The burn-up status determined with different methods corresponds very well within the limits of accuracy. (orig.)

  20. Water chemistry in nuclear power stations with high-temperature reactors with particular reference to the AVR

    International Nuclear Information System (INIS)

    Nieder, R.; Resch, G.

    1976-01-01

    The water-steam cycle of a nuclear power plant with a helium-cooled high-temperature reactor differs in design data significantly and extensively from the corresponding cycles of light-water-cooled nuclear reactors and resembles to a great extent the water-steamcycle of a modern conventional power plant. The radioactive constituents of the water-steam cycle can be satisfactorily removed apart from Tritium by means of a pre-coat filter with powder-resisn, as comprehensive experiments have demonstrated. (orig.) [de

  1. CAS algorithm-based optimum design of PID controller in AVR system

    International Nuclear Information System (INIS)

    Zhu Hui; Li Lixiang; Zhao Ying; Guo Yu; Yang Yixian

    2009-01-01

    This paper presents a novel design method for determining the optimal PID controller parameters of an automatic voltage regulator (AVR) system using the chaotic ant swarm (CAS) algorithm. In the tuning process of parameters, the CAS algorithm is iterated to give the optimal parameters of the PID controller based on the fitness theory, where the position vector of each ant in the CAS algorithm corresponds to the parameter vector of the PID controller. The proposed CAS-PID controllers can ensure better control system performance with respect to the reference input in comparison with GA-PID controllers. Numerical simulations are provided to verify the effectiveness and feasibility of PID controller based on CAS algorithm.

  2. The AVR2–SIX5 gene pair is required to activate I-2-mediated immunity in tomato

    NARCIS (Netherlands)

    Ma, L.; Houterman, P.M.; Gawehns, F.; Cao, L.; Sillo, F.; Richter, H.; Clavijo-Ortiz, M.J.; Schmidt, S.M.; Boeren, J.A.; Vervoort, J.J.M.; Cornelissen, B.J.C.; Rep, M.; Takken, F.L.W.

    2015-01-01

    •Plant-invading microbes betray their presence to a plant by exposure of antigenic molecules such as small, secreted proteins called ‘effectors’. In Fusarium oxysporum f. sp. lycopersici (Fol) we identified a pair of effector gene candidates, AVR2-SIX5, whose expression is controlled by a shared

  3. Conceptual design of a rapid-cycling synchrotron for the KFA-Juelich spallation neutron source: working papers

    International Nuclear Information System (INIS)

    1983-01-01

    An accelerator group was established at ANL by the request of KFA-Juelich to carry out a conceptual design study and cost estimate for a rapid-cycling synchrotron as a possible first stage program on spallation neutron sources at KFA-Juelich. This set of notes is the individual notes which form the basis of the final report under this proposal prepared in January 1983. The topics covered include: SNQ Synchrotron Lattice-I; injection and extraction orbit; extraction from SNQ-SRA; SRA injection; capture and acceleration considerations in the SNQ-SRA; longitudinal coupling impedance; power supplies for SNQ synchrotron proposals; space charge limits in the SNQ-SRA; error analysis; SNQ-SRA ring magnets preliminary designs and cost; summary of CERN booster 4-ring arrangement; V-lattices for SNQ-SRA and extraction from the V-lattices; rf parameters for capture, acceleration and extraction; some parameters of the SNQ-SRA injector system; Keil-Schnell criterion; risetime of longitudinal resistive wall instability; beam scrapers; a design of the vacuum system; some aspects of vacuum consideration for SNQ-SRA; choice working points; ring magnet power supplies for shaped extaction of 1.1 GeV SNQ; ring magnet design and costs; tune shift due to the fringing field of the quadrupoles; coherent instability due to ions in the residual gas; transverse stabilization of bunched beams; rf acceleration system; injection into the SRA; Landau damping to get transverse stability; chromaticity and amplitude dependent tune controls in the SNQ-SRA; conversion of the SNQ-SRA to a compressor ring; comments on beam loss; summary of longitudinal stability study and transverse stability study for the SNQ-SRA; and the beam stay clear regions of the SNQ-SRA

  4. Gas-cooled reactor technology safety and siting. Report of a technical committee meeting. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-07-01

    At the invitation of the Government of the Union of Soviet Socialist Republics, the Eleventh International Conference on the HTGR and the IAEA Technical Committee Meeting on Gas-Cooled Reactor Technology, Safety and Siting were held in Dimitrovgrad, USSR, on June 21-23, 1989. The Technical Committee Meeting provided the Soviet delegates with an opportunity to display the breadth of their program on HTGRs to an international audience. Nearly one-half of the papers were presented by Soviet participants. Among the highlights of the meeting were the following: the diverse nature and large magnitude of the Soviet research and development program on high temperature gas-cooled reactors; the Government approval of the budget for the construction of the 30 MWt High Temperature Test Reactor (HTTR) in Japan (The schedule contemplates a start of construction in spring 1990 on a site at the Oarai Research Establishment and about a five year construction period.); disappointment in the announced plans to shutdown both the Fort St. Vrain (FSV) plant in the United States (US) and the Thorium High Temperature Reactor (THTR-300) in Germany (These two reactors presently represent the only operating HTGRs in the world since the AVR plant in Juelich, Germany, was also shutdown at the end of 1988.); the continuing negotiations between Germany and the USSR on the terms of the co-operation between the two countries for the construction of a HTR Module supplemented by joint research and development activities aimed at increasing coolant outlet temperatures from 750 deg. C to 950 deg. C; the continued enthusiasm displayed by both the US and German representatives for the potential of the small modular designs under development in both countries and the ability for these designs to meet the stringent requirements demanded for the future expansion of nuclear power; the combining of the HTGR technology interest of ABB-Atom and Siemens in Germany into a joint enterprise, HTR GmbH, in May 1989

  5. Gas-cooled reactor technology safety and siting. Report of a technical committee meeting. Working material

    International Nuclear Information System (INIS)

    1990-01-01

    At the invitation of the Government of the Union of Soviet Socialist Republics, the Eleventh International Conference on the HTGR and the IAEA Technical Committee Meeting on Gas-Cooled Reactor Technology, Safety and Siting were held in Dimitrovgrad, USSR, on June 21-23, 1989. The Technical Committee Meeting provided the Soviet delegates with an opportunity to display the breadth of their program on HTGRs to an international audience. Nearly one-half of the papers were presented by Soviet participants. Among the highlights of the meeting were the following: the diverse nature and large magnitude of the Soviet research and development program on high temperature gas-cooled reactors; the Government approval of the budget for the construction of the 30 MWt High Temperature Test Reactor (HTTR) in Japan (The schedule contemplates a start of construction in spring 1990 on a site at the Oarai Research Establishment and about a five year construction period.); disappointment in the announced plans to shutdown both the Fort St. Vrain (FSV) plant in the United States (US) and the Thorium High Temperature Reactor (THTR-300) in Germany (These two reactors presently represent the only operating HTGRs in the world since the AVR plant in Juelich, Germany, was also shutdown at the end of 1988.); the continuing negotiations between Germany and the USSR on the terms of the co-operation between the two countries for the construction of a HTR Module supplemented by joint research and development activities aimed at increasing coolant outlet temperatures from 750 deg. C to 950 deg. C; the continued enthusiasm displayed by both the US and German representatives for the potential of the small modular designs under development in both countries and the ability for these designs to meet the stringent requirements demanded for the future expansion of nuclear power; the combining of the HTGR technology interest of ABB-Atom and Siemens in Germany into a joint enterprise, HTR GmbH, in May 1989

  6. First internal and external experiments at COSY Juelich

    Energy Technology Data Exchange (ETDEWEB)

    Prasuhn, D [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Kernphysik; Maier, R [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Kernphysik; Bechstedt, U [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Kernphysik; Dietrich, J [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Kernphysik; Hacker, U [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Kernphysik; Martin, S [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Kernphysik; Stockhorst, H [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Kernphysik; Toelle, R [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Kernphysik; Grzonka, D [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Kernphysik; Nake, C [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Kernphysik; Mosel, F [Bonn Univ. (Germany). Inst. fuer Strahlen- und Kernphysik

    1995-08-01

    The inauguration of the cooler synchrotron COSY Juelich was celebrated on April 1st, 1993. After the first successful acceleration to proton momenta above 800 GeV/c, beamtimes for experiments were scheduled in parallel to further machine development. The first experiment was the internal target experiment EDDA, which investigated the energy dependence of the p-p interaction. It makes use of a 3x4 {mu}m{sup 2} thin CH{sub 2} fiber as an internal target. The thickness of the fiber is more than adequate to achieve high luminosities, so the intensity of the stored beam has to be reduced to 10{sup 7} p. On the other hand, it is thin enough to achieve beam lifetimes of 3 s at 1.4 GeV/c. Details of the target fabrication and the first experimental results will be discussed. Both external experimental facilities at COSY, the time-of-flight spectrometer, and the magnetic spectrometer BIG KARL use a liquid hydrogen (deuterium) target. The first experiments were carried out at proton energies between 300 MeV and 500 MeV. Also, these experimental data will be presented. Two further internal experiments are prepared for the installation into the COSY ring. The target for the first experiment is a gas-jet target, the second experiment uses ribbon targets for the interaction. The status of both experimental setups will be shown. (orig.).

  7. IRPhEP-handbook, International Handbook of Evaluated Reactor Physics Benchmark Experiments

    International Nuclear Information System (INIS)

    Sartori, Enrico; Blair Briggs, J.

    2008-01-01

    experimental series that were performed at 17 different reactor facilities. The Handbook is organized in a manner that allows easy inclusion of additional evaluations, as they become available. Additional evaluations are in progress and will be added to the handbook periodically. Content: FUND - Fundamental; GCR - Gas Cooled (Thermal) Reactor; HWR - Heavy Water Moderated Reactor; LMFR - Liquid Metal Fast Reactor; LWR - Light Water Moderated Reactor; PWR - Pressurized Water Reactor; VVER - VVER Reactor; Evaluations published as drafts 2 - Related Information: International Criticality Safety Benchmark Evaluation Project (ICSBEP); IRPHE/B and W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments; IRPHE-JAPAN, Reactor Physics Experiments carried out in Japan ; IRPHE/JOYO MK-II, JOYO MK-II core management and characteristics database ; IRPhE/RRR-SEG, Reactor Physics Experiments from Fast-Thermal Coupled Facility; IRPHE-SNEAK, KFK SNEAK Fast Reactor Experiments, Primary Documentation ; IRPhE/STEK, Reactor Physics Experiments from Fast-Thermal Coupled Facility ; IRPHE-ZEBRA, AEEW Fast Reactor Experiments, Primary Documentation ; IRPHE-DRAGON-DPR, OECD High Temperature Reactor Dragon Project, Primary Documents; IRPHE-ARCH-01, Archive of HTR Primary Documents ; IRPHE/AVR, AVR High Temperature Reactor Experience, Archival Documentation ; IRPHE-KNK-II-ARCHIVE, KNK-II fast reactor documents, power history and measured parameters; IRPhE/BERENICE, effective delayed neutron fraction measurements ; IRPhE-TAPIRO-ARCHIVE, fast neutron source reactor primary documents, reactor physics experiments. The International Handbook of Evaluated Reactor Physics Benchmark Experiments was prepared by a working party comprised of experienced reactor physics personnel from Belgium, Brazil, Canada, P.R. of China, Germany, Hungary, Japan, Republic of Korea, Russian Federation, Switzerland, United Kingdom, and the United States of America. The IRPhEP Handbook is available to authorised requesters from the

  8. High temperature reactor development in the Netherlands

    International Nuclear Information System (INIS)

    Heek, A.I. van

    1996-01-01

    This year, some clear design choices have been made in the WHITE Reactor development programme. The activities will be concentrated at the development of a small size pebble bed HTR for combined heat and power production with a closed cycle gas turbine. Objective of the development is threefold: 1. restoring social support; 2. establishing commercial viability after market introduction; and 3. making the market introduction itself feasible, i.e. limited development and first-of-a-kind costs. This design is based on the peu-a-peu design of KFA Juelich and will be optimized. The computer codes necessary for this are being prepared for this work. The dynamic neutronics code PANTHER is being coupled to the thermal hydraulics code THERMIX-DIREKT. For this reactor type, fuel temperatures are maximal in the scenario of depressurization with recriticality. Even for this scenario, fuel temperatures of the 20MWth PAP-GT do not exceed 1300 deg. C, so there should be room for upscaling for economic reasons. On the other hand, it would be convenient to fuel the reactor batchwise instead of continuously, and the use of thorium could be required. These two features may lead to a larger temperature margin. The optimal design must unite these features in the best acceptable way. To gain expertise in calculations on gas cooled graphite moderate reactors, benchmark calculations are being performed in parallel with international partners. Parallel to this, special expertise is being built up on HTR fuel and HTR reactor vessels. (author). 3 refs

  9. Pemrograman Antarmuka Modem GSM dengan Pengendali Mikro AVR Menggunakan Bahasa C

    Directory of Open Access Journals (Sweden)

    Daniel Kartawiguna

    2011-12-01

    Full Text Available Cell phones are now a cheap means of wide-ranged wireless communication. One feature widely used is the short message service (SMS. The cellular communication system is more appropriate for long-distanced applications than high speed data transfer. This study aims to develop a GSM modem interface with AVR microcontroller using C programming language. The tools controlled by the microcontroller system can gain more benefit if it is linked with GSM mobile communication system. The results of this study can be applied whether in a remote monitoring system, remote control system, or communication between the microcontroller via SMS. Many additional benefits can be obtained for the tools controlled by the microcontroller when connected to a GSM system. 

  10. Results of neutron measurements in the spectral position of the Juelich FKS steel irradiation capsules

    International Nuclear Information System (INIS)

    Schneider, W.

    1986-10-01

    This is a report on the planning and results of neutron monitoring in the capsules of the Juelich steel irradiation for the research project on component safety (FKS). The table of results and their discussion is provided specifically for the spectral positions (for characterising the neutron spectrum) in each of the types of irradiation capsules used. The results are given for the reaction rates of the neutron measurement reactions used (activation or fission reactions), for the neutron flux densities and fluxes derived from them related to the actual or at least plausible neutron spectra and finally for the radiation damage (or exposure) of the irradiated material calculated from them, expressed as the atomic displacement figure (dpa) and its percentage in sections of the neutron spectrum. (orig.) [de

  11. The role of autophagy in chloroplast degradation and chlorophagy in immune defenses during Pst DC3000 (AvrRps4 infection.

    Directory of Open Access Journals (Sweden)

    Junjian Dong

    Full Text Available BACKGROUND: Chlorosis of leaf tissue normally observed during pathogen infection may result from the degradation of chloroplasts. There is a growing evidence to suggest that the chloroplast plays a significant role during pathogen infection. Although most degradation of the organelles and cellular structures in plants is mediated by autophagy, its role in chloroplast catabolism during pathogen infection is largely unknown. RESULTS: In this study, we investigated the function of autophagy in chloroplast degradation during avirulent Pst DC3000 (AvrRps4 infection. We examined the expression of defensive marker genes and suppression of bacterial growth using the electrolyte leakage assay in normal light (N and low light (L growing environments of wild-type and atg5-1 plants during pathogen treatment. Stroma-targeted GFP proteins (CT-GFP were observed with LysoTracker Red (LTR staining of autophagosome-like structures in the vacuole. The results showed that Arabidopsis expressed a significant number of small GFP-labeled bodies when infected with avirulent Pst DC3000 (AvrRps4. While barely detectable, there were small GFP-labeled bodies in plants with the CT-GFP expressing atg5-1 mutation. The results showed that chloroplast degradation depends on autophagy and this may play an important role in inhibiting pathogen growth. CONCLUSION: Autophagy plays a role in chloroplast degradation in Arabidopsis during avirulent Pst DC3000 (AvrRps4 infection. Autophagy dependent chloroplast degradation may be the primary source of reactive oxygen species (ROS as well as the pathogen-response signaling molecules that induce the defense response.

  12. On fission product retention in the core of the low powered high temperature reactor under accident conditions

    International Nuclear Information System (INIS)

    Bastek, H.

    1984-01-01

    In the core of the high temperature reactor the fuel element and the coated particles contained herein provide the safest enclosure for fission products. The complex process of fission product transport out of the particle kernel, through the particle coating and within the fuel element graphite is described in a simplified form by the Fick's diffusion. The effective diffusion coefficient is used for calculation. Starting from the existing ideas of fission product transport five burn-up and temperature-dependent diffusion coefficients for Cesium in (Th,U)O 2 -kernels are derived in this study. The results have been gained from several fuel element radiation experiments in recent years, which showed extreme variation in regard to burn-up, temperature cycle, neutron flux and operation time. Cs-137 release measurements from single particle kernels were present from all the experiments. Furthermore, annealing tests of AVR-fuel elements were analyzed. Heat-temperatur and heating-time, the fuel element burn-up in the AVR-reactor, as well as the measured Cs-137 inventory of the fuel elements before and after annealing, are included in the investigation as essential parameters. With the aid of the derived diffusion coeffizients and already present data sets the Cs-137 release of fuel elements into a small reactor core is investigated under unrestricted core heat-up. While the released Cs-137 is derived mainly from defective particles at accident temperatures up to 1600 0 C, the main part diffuses through the particle coating at higher accident temperatures. (orig./HP) [de

  13. ST-segment depression in aVR as a predictor of culprit artery in acute inferior wall ST-segment elevation myocardial infarction

    Directory of Open Access Journals (Sweden)

    Ahmed Hafez el-neklawy

    2014-03-01

    Conclusions: ST depression in aVR is common in patients with LCX-related acute inferior myocardial infarction. The ST changes in this lead are associated with an excellent specificity and a good sensitivity in differentiating LCX from RCA as the IRA.

  14. Contributions for the third international carbon conference CARBON '80

    International Nuclear Information System (INIS)

    Delle, W.

    1980-05-01

    This report is a compilation of 8 papers prepared by KFA Juelich GmbH for the International Carbon Conference carbon 80 in Baden-Baden. The contributions deal mainly with materials problems which arise from the application of graphite and silicon carbide in High-Temperature Gas-Cooled Reactors, HTR. Most of the results described were obtained in the framework of the HTR Projects ''Hochtemperaturreaktor-Brennstoffkreislauf'' (High Temperature Reactor Fuel Cycle), HBK, that includes the partners Gesellschaft fuer Hochtemperaturreaktor-Technik mbH, Hochtemperaturreaktor-Brennelement GmbH, Hochtemperatur-Reaktorbau GmbH, Kernforschungsanlage Juelich GmbH, NUKEM GmbH and Sigri Elektrographit GmbH/Ringsdorff-Werke GmbH and ''Prototyp Nukleare Prozesswaerme'' (Prototype Nuclear Heat), PNP, for the development of procedures for the conversion of solid fossil raw materials by means of heat from High Temperature Gas-Cooled Reactors, that includes the partners Bergbau-Forschung GmbH, Gesellschaft fuer Hochtemperaturreaktor-Technik mbH, Hochtemperatur-Reaktorbau GmbH, Kernforschungsanlage Juelich GmbH and Rheinische Braunkohlenwerke AG. Both projects are financed by the Federal Ministry for Research and Technology and the State of North Rhine-Westphalia. (orig./IHOE) [de

  15. Demolition of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklearservice, Essen (Germany); Cremer, J. [SNT Siempelkamp Nukleartechnik, Heidelberg (Germany)

    2003-06-01

    FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [German] Der mit Leichtwasser gekuehlte und moderierte Schwimmbad-Forschungsreaktor FRJ-2 (MERLIN) wurde von 1958 bis 1962 fuer die damalige Kernforschungsanlage Juelich (KFA) errichtet. Von 1964 bis 1985 wurde er fuer Experimente mit zunaechst 5 MW und spaeter 10 MW thermischer Leistung bei einem maximalen thermischen Neutronenfluss von 1,1.10{sup 14} n/cm{sup 2}s genutzt. Im Jahr 1985 stellte der Reaktor seinen Betrieb ein. Die Brennelemente wurden aus der Anlage entfernt und in die USA und nach Grossbritannien verbracht. Seit 1996 erfolgen die wesentlichen Abbautaetigkeiten unter Leitung eines verantwortlichen Projektteams. Bis Ende 1998 wurde das komplette Sekundaerkuehlsystem entfernt. Dem Abbau der Kuehlkreislaeufe und Experimentiereinrichtungen folgte im Jahr 2000 der Ausbau der

  16. Iodine-123 and bromine-75 production and development program at Juelich

    International Nuclear Information System (INIS)

    Stoecklin, G.

    1985-01-01

    The iodine-123 and bromine-75 production and development program at the Nuclear Research Center in Juelich as of 1982 is described, and examples of recent 123 I- and 75 Br-analogue tracers that have been developed to the level of clinical trial are given. Iodine-123 is produced via the 127 I(d,6n) 123 Xe → 123 I process and by the 124 Te(p,2n) 123 I and 122 Te(d,n) 123 I reactions. These production methods are critically reviewed. Bromine-75-labeled benzodiazenes have been prepared for in vivo mapping of benzodiazepine receptor sites. The 7-( 75 Br)-5-(2-fluorophenyl)-1-methyl-1,3-dihydro-2H-1,4-benzodiazepine-2-one (BFB) was prepared with a specific activity of > 10 4 Ci/mmole. Finally, preparation and applications of the halogenated amino acid L-3-( 123 I)-iodo-α-methyltyrosine (IMT) and the analogous 75 Br compound (BMT) are reported. Both IMT and BMT have been successfully applied for pancreas imaging and tomography, and IMT has been used for imaging both melanotic and amelanotic malignant melanoma of the eye

  17. Design of hygrothermal detection and control intelligent system based on AVR-MCU in radon chamber

    International Nuclear Information System (INIS)

    Zheng Yongming; Fang Fang; Zhou Wei; Zheng Meiyang; Xu Jianyi

    2006-01-01

    The design of a new hygrothermal detection and control system based on AVR-MCU, which is used in minitype and medium-sized radon chamber, is introduced. The kernel of the interface among ATmega128 MCU, hygrothermal sensor, refrigeration and desiccation components is described. In addition, with the calculation of the control capability in theory, it comes to the conclusion that the design is feasible, and this control system not only can work in independence, but also can cooperate with PC by RS232 communication. (authors)

  18. Design and construction of the HESR BPM prototype wire test bench at COSY, Forschungzentrum Juelich

    Energy Technology Data Exchange (ETDEWEB)

    Srinivasan, Sudharsan; Kamerdzhiev, Vsevolod; Boehme, Christian [Forschungszentrum Juelich GmbH (Germany)

    2016-07-01

    The Institute of Nuclear Physics 4(IKP-4), of the Research Center Juelich (FZJ), is in charge of building and commissioning the High Energy Storage Ring (HESR) within the international facility, Facility for Antiproton and Ion Research (FAIR) at Darmstadt. Beam Position Monitors (BPMs) are an essential instrument for any accelerator allowing operators to accurately monitor and control the accelerated beam. The demand for a BPM test bench will be showcased which will help to assess the design's ability to meet the system requirements. The weight is on the factors considered for the development of the initial test bench, its functional components, the metrology tests for ensuring positional measurement accuracy, and the design modifications from metrology investigations leading to the conceptual development of a new test bench.

  19. Twenty-five years of Brown Boveri experience in development, design and fabrication of circulators for HTGR

    International Nuclear Information System (INIS)

    Stoelzl, D.

    1988-01-01

    The two circulators for the AVR experimental reactor in Juelich, Federal Republic of Germany, were supplied. The circulators, which are equipped with oil bearings, have been operating troublefree since the start of commissioning in 1966. As a consequence of a water ingress into the reactor resulting from a steam generator damage one bearing was replaced in 1977 after 72,000 operating hours. Up to the present date, each of the circulators has scored 115,000 hours of operation, one of them without any disassembly. In the THTR 300 in Schmehausen, Federal Republic of Germany, 6 BBC circulators are in operation. The insertable circulator units equipped with oil bearings have successfully proven their operating capability without any problems during the commissioning phase and the 100% power operation which was started recently. Currently active magnetic bearings are being developed for advanced gas-cooled reactors such as the HTR 100, the HTR 500 and the heating reactor after excellent results have been furnished by a small prototype in a test loop. This ADI circulator has since scored more than 15,000 operating hours without any trouble. A retainer bearing test stand also equipped with active magnetic bearings has been in operation for nearly 2 years. This test stand serves for developing the conditions for safe rundown of the rotors of even the largest circulators after the magnetic bearings have been deenergized unintentionally. Development work is conducted on the prototype of a safety-relevant circulator held in magnetic bearings, to be used for decay heat removal in the HTR 500. The original aim to have circulators without auxiliary medium for bearing lubrication will thus be reached. The advantages to be obtained in process and systems design are a supplementary support to the inherent safety characteristics of high-temperature reactors. Another advantage of these bearings is cost reduction. 5 refs, 7 figs

  20. Radiological consequences of the Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Hill, P.; Hille, R.

    2003-01-01

    The reactor accident at unit 4 of the Chernobyl nuclear power plant in Ukraine has deeply affected the living conditions of millions of people. Especially the health consequences have been of public concern up to the present and also been the subject of sometimes absurd claims. The current knowledge on the radiological consequences of the accident is reviewed. Though an increased hazard for some risk groups with high radiation exposure, e.g., liquidators, still cannot be totally excluded for the future, the majority of the population shows no statistically significant indication of radiation-induced illnesses. The contribution of the Research Center Juelich to the assessment of the post-accidental situation and psychological relief of the population is reported. The population groups still requiring special attention include, in particular, children growing up in highly contaminated regions and the liquidators of the years 1986 and 1987 deployed immediately after the accident. (author)

  1. Six years of operational experience with a lead acid battery in the autonomous PV-hydrogen plant phoebus Juelich

    Energy Technology Data Exchange (ETDEWEB)

    Meurer, C.; Brocke, W.A.; Emonts, B.; Heuts, G.; Mai, H.; Croe, D. [Forschungszentrum Juelich GmbH, Juelich (Germany). Inst. for Materials and Processes in Energy Systems IWV-3

    1999-07-01

    A set of 110 lead acid battery cells with a capacity of 1380 Ah was operated for six years in the PV-hydrogen plant PHOEBUS Juelich under realistic consumer and solar conditions. The plant is controlled by an energy management system that is specially designed for the use of a battery combined with a hydrogen long-term storage. The energy management system uses the state of charge SOC, which is determined by measurements of the battery current using validated models of the gassing current and the equilibrium voltage. It was found that after six years of operation there is hardly any fading of battery capacity. (orig.)

  2. Annual meeting on nuclear technology 1980. Technical meeting: Operating experiences

    International Nuclear Information System (INIS)

    1980-01-01

    In addition to general experiences, experiences in reactor operation with relation to the Phenix reactor, KNK-2 reactor, the AVR reactor, the BWR-type KKI-reactor, the Philippsburg-1 reactor and the Obrigheim reactor are described. (DG) [de

  3. Study on the profitableness of electricity generation with high temperature reactors

    International Nuclear Information System (INIS)

    Kolb, G.

    1978-08-01

    The programme group 'Systemforschung und Technologische Entwicklung' (STE) of the Nuclear Research Centre Juelich in cooperation with the internal institutions 'Projekttraegerschaft Entwicklung von Hochtemperaturreaktoren' (PTH) and 'Institut fuer Reaktorentwicklung' (IRE) on the one hand, and the external partner 'Hochtemperatur-Reactorbau GmbH' (HRB) and 'Gesellschaft fuer Hochtemperaturreactor Technik' (GHT) on the other hand have set up a study on fuel cycle costs, electricity production cost and the economical use as well as uranium resource protection by introduction of high temperature reactors (HTR) with pebble bed core to generate electricity. The pressurized-water reactor (PWR) today on the market serves as comparison. The working results obtained sofar are compiled in the present report. It was particularly noted that - the HTR can economically fully compete with the PWR for electricity generation - the necessary supply of natural uranium for the HTR in open circuit is about one third lower and in the closed circuit, almost two thirds lower than in the corresponding PWR. A further reduction is possible on a long-term basis by highly converting HTW systems. (orig.) [de

  4. Solarthermie 2000 - Partial programme 3: Solar district heat - solar campus Juelich, feasibility study of the seasonal heat storage; Solarthermie 2000, Teilprogramm 3: Solare Nahwaerme - Solar-Campus Juelich, Machbarkeitsuntersuchung des saisonalen Waermespeichers

    Energy Technology Data Exchange (ETDEWEB)

    Braxein, A. [Iwu - Ingenieurgesellschaft fuer Wasser und Umwelt mbH, Aachen (Germany); Spaete, F.; Repschlaeger, H. [Solar-Inst. Juelich (Germany); Friedel, J. [Stadtwerke Juelich (Germany)

    1998-12-31

    The heart of the solar-assisted district heating system for the solar campus Juelich is going to be a seasonal underground store with a capacity of 2,500 cubic metres, e.g. a rather small long-term storage. The aim of the project was to prove the technical and financial feasibility of the underground store by developing a building concept for the storage and to determine whether this concept was also suitable for larger storage facilities. The design of the floating lid and the suitability of various building materials were studied by means of a physical model. The resulting storage concept shows an underground storage pit whose shape is an inverted square pyramid base with a seal of double-layer PP-foil and a thermal insulation of pressure-resistant mineral wool in the upper part. The lid consists of 16 square walk-on floating containers with an edge length of 6.50 metres and a height of 1.0 metre each. Tests and calculations showed that the concept can be realised. The building costs for the small storage with a capacity of 2,500 cubic metres will amount to approx. DM 370 per cubic metre, costs for larger volumes of 25,000 cubic metres will amount to approx. DM 150 DM per cubic metre of water stored. (orig.) [Deutsch] Das Kernstueck der solargestuetzten Nahwaermeversorgung fuer den Solar-Campus Juelich soll ein 2500 m{sup 3} grosser, saisonaler Erdbeckenspeicher bilden, also ein relativ kleiner Langzeitspeicher. Ziel des Vorhabens war, die technische und finanzielle Machbarkeit des geplanten Erdbeckenspeichers durch ein baureifes Speicherkonzept nachzuweisen sowie die Uebertragbarkeit auf grosse saisonale Speicher gleicher Bauart zu untersuchen. Die konstruktive Ausbildung der schwimmenden Deckelkonstruktion sowie die Eignung verschiedener Baumaterialien wurden mit Hilfe eines physikalischen Modells untersucht. Das resultierende Speicherkonzept sieht einen Erdbeckenspeicher in Form eines umgedrehten, quadratischen Pyramidenstumpfes mit einer Abdichtung aus

  5. Postirradiation examination of HTR fuel

    International Nuclear Information System (INIS)

    Nabielek, H.; Reitsamer, G.; Kania, M.J.

    1986-01-01

    Fuel for the High Temperature Reactor (HTR) consists of 1 mm diameter coated particles uniformly distributed in a graphite matrix within a cold-molded 60 mm diameter spherical fuel element. Fuel performance demonstrations under simulated normal operation conditions are conducted in accelerated neutron environments available in Material Test Reactors and in real-time environments such as the Arbeitsgemeinschaft Versuchsreaktor (AVR) Juelich. Postirradiation examinations are then used to assess fuel element behavior and the detailed performance of the coated particles. The emphasis in postirradiation examination and accident testing is on assessment of the capability for fuel elements and individual coated particles to retain fission products and actinide fuel materials. To accomplish this task, techniques have been developed which measures fission product and fuel material distributions within or exterior to the particle: Hot Gas Chlorination - provides an accurate method to measure total fuel material concentration outside intact particles; Profile Electrolytic Deconsolidation - permits determination of fission product distribution along fuel element diameter and retrieval of fuel particles from positions within element; Gamma Spectrometry - provides nondestructive method to measure defect particle fractions based on retention of volatile metallic fission products; Particle Cracking - permits a measure of the partitioning of fission products between fuel kernel and particle coatings, and the derivation of diffusion parameters in fuel materials; Micro Gas Analysis - provides gaseous fission product and reactive gas inventory within free volume of single particles; and Mass-spectrometric Burnup Determination - utilizes isotope dilution for the measurement of heavy metal isotope abundances

  6. Neutron scattering at FRJ-2. Experimental reports 2004

    International Nuclear Information System (INIS)

    Brueckel, T.; Richter, D.; Zorn, R.

    2004-01-01

    The Research Centre FZ-Juelich is offering its neutron research facilities to a growing national and international user community for the benefit of their research using neutron beams. FZ-Juelich operates a 23 MW DIDO reactor that delivers a total neutron flux of 2.9 x 10 14 n/cm 2 s (undisturbed) for a comprehensive suite of 17 instruments installed at 5 individual thermal beam tubes and, in addition, 5 external cold neutron guides. In the year 2004 the reactor was in operation for 208 days and we are happy to announce that more than 150 individual experiments (constituting 61% of the total) were carried out by a large external user community from 85 institutions all over the world. In close collaboration with internal staff the study of soft matter systems took the largest stake. In addition, subjects of biology, magnetism and engineering were among the other main topics of the experimental programme. We gratefully acknowledge the funding programme ''Juelich Neutrons for Europe'' under the European initiative NMI3 that enables numerous external users from the EU and associated countries to come over, visit Juelich and perform their experiments. This book comprises the scientific reports of the experiments completed in 2004. We wish to thank all external users, local applicants, instrument responsibles, and technical staff for their joint efforts and contributions to the success and progress of the Juelich neutron research facility. (orig.)

  7. Core Facility of the Juelich Observatory for Cloud Evolution (JOYCE - CF)

    Science.gov (United States)

    Beer, J.; Troemel, S.

    2017-12-01

    A multiple and holistic multi-sensor monitoring of clouds and precipitation processes is a challenging but promising task in the meteorological community. Instrument synergies offer detailed views in microphysical and dynamical developments of clouds. Since 2017 The the Juelich Observatory for Cloud Evolution (JOYCE) is transformed into a Core Facility (JOYCE - CF). JOYCE - CF offers multiple long-term remote sensing observations of the atmosphere, develops an easy access to all observations and invites scientists word wide to exploit the existing data base for their research but also to complement JOYCE-CF with additional long-term or campaign instrumentation. The major instrumentation contains a twin set of two polarimetric X-band radars, a microwave profiler, two cloud radars, an infrared spectrometer, a Doppler lidar and two ceilometers. JOYCE - CF offers easy and open access to database and high quality calibrated observations of all instruments. E.g. the two polarimetric X-band radars which are located in 50 km distance are calibrated using the self-consistency method, frequently repeated vertical pointing measurements as well as instrument synergy with co-located micro-rain radar and distrometer measurements. The presentation gives insights into calibration procedures, the standardized operation procedures and recent synergistic research exploiting our radars operating at three different frequencies.

  8. Statistical investigations of the failure behaviour of components in the AVR experimental nuclear power plant. Vol. 2

    International Nuclear Information System (INIS)

    Meyna, A.; Mock, R.; Tietze, A.; Hennings, W.

    1989-08-01

    From operational records of the years 1977 to 1986, service life distributions of helium valves in gas circuits of the AVR were determined. Results are constant failure rates in the range from 3 to 6x10 -6 /h and, for some populations, indications of time dependent failure rates. Nonparametric methods showed only limited efficiency. For a Bayesian approach the necessary prior information was missing. Furthermore, the main failure causes could be determined. (orig./HP) [de

  9. Statistical investigations of the failure behaviour of components in the AVR-experimental nuclear power plant. Vol. 1

    International Nuclear Information System (INIS)

    Hennings, W.

    1989-08-01

    From operational reports of the years 1970 to 1984, failure rates of valves in gas circuits of the AVR experimental power plant were determined. Also, potential influences of environmental and operational conditions were investigated. The resulting failure rates are for manual valves app. 0,1.10 -6 /h, for pneumatic valves between 3 and 9.10 -6 /h, for solenoid valves between 1,5 and 4.10 -6 /h and for control valves between 12 and 41.10 -6 /h. (orig.) [de

  10. Pseudomonas syringae pv. Tomato DC3000 Type III secretion effector polymutants reveal an interplay between hopAD1 and AvrPtoB

    Science.gov (United States)

    The model pathogen Pseudomonas syringae pv. tomato DC3000 suppresses the two-tiered innate immune system of plants by injecting a complex repertoire of effector proteins into host cells via the type III secretion system. The model effector AvrPtoB has multiple domains and plant protein interactors i...

  11. Ultimate storage of spent fuel elements of the AVR test power station in the Asse salt mine

    International Nuclear Information System (INIS)

    Wolf, J.

    1975-02-01

    With regard to the ultimate storage of irradiated AVR pebble-bed reactor carbide fuel elements in the saline of Asse, a number of tests and calculations are presented to demonstrate that there is no credible possibility of the MCA (maximum credible accident) defined for the saline. The safety of persons is not threatened during the operation of spent fuel storage nor at any later time (extrapolation up to approx. 1,000 years after storage). 1,000 fuel elements at a time are packed up in gas-tight containers which are stacked in boreholes. The boreholes are then sealed with concrete. Lay-out and functions of the special airlock and transportation systems - from the packing of the containers in a hot cell to the final storage in the borehole - are described with special reference to aspects of the safety of the overall procedure. The possible accidents in the mine are discussed in detail. 85 Kr and T release rates are determined in laboratory tests by heating of the spherical fuel elements. Tests with fuel elements embedded in salt or stagnant brine were carried out at varies temperatures to investigate their behaviour in final storage. Kr and T release, extraction of fission products, mechanical resistance and corrosion were examined in these tests. Finally, the permeability of salt and salt concrete to radioactive gases were investigated in a special experimental arrangement. The diffusion and permeability coefficients obtained for 85 Kr, HT and HTO allow an estimation of the gas discharge of the stored fuel element. (RB/AK) [de

  12. Reactivity control in HTR power plants with respect to passive safety system. Summary

    Energy Technology Data Exchange (ETDEWEB)

    Barnert, H; Kugeler, K [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Sicherheitsforschung und Reaktortechnik

    1996-12-01

    The R and D and Demonstration of the High Temperature Reactor (HTR) is described in overview. The HTR-MODULE power plant, as the most advanced concept, is taken for the description of the reactivity control in general. The idea of the ``modularization of the core`` of the HTR has been developed as the answer on the experiences of the core melt accident at Three Miles Island. The HTR module has two shutdown systems: The ``6 rods``-system for hot shutdown at the ``18 small absorber pebbles units`` - system for cold shutdown. With respect to the definition of ``Passive Systems`` of IAEA-TECDOC-626 the total reactivity control system of the HTR-MODULE is a passive system of category D, because it is an emergency reactor shutdown system based on gravity driven rods, and devices, activated by fail-safe trip logic. But reactivity control of the HTR does not only consist of these engineered safety system but does have a self-acting stabilization by the negative temperature coefficient of the reactivity, being rather effective in reactivity control. Examples from computer calculations are presented, and, in addition, experimental results from the ``Stuck Rod Experiment`` at the AVR reactor in Juelich. On the basis of this the proposal is made that ``self-acting stabilization as a quality of the function`` should be discussed as a new category in addition to the active and passive engineered safety systems, structures and components of IAEA-TECDOC-626. The requirements for a future ``catastrophe-free`` nuclear technology are presented. In the appendix the 7th amendment of the atomic energy act of the Federal Republic of Germany, effective 28 July 94, is given. (author).

  13. Avr4 promotes Cf-4 receptor-like protein association with the BAK1/SERK3 receptor-like kinase to initiate receptor endocytosis and plant immunity.

    Science.gov (United States)

    Postma, Jelle; Liebrand, Thomas W H; Bi, Guozhi; Evrard, Alexandre; Bye, Ruby R; Mbengue, Malick; Kuhn, Hannah; Joosten, Matthieu H A J; Robatzek, Silke

    2016-04-01

    The first layer of plant immunity is activated by cell surface receptor-like kinases (RLKs) and proteins (RLPs) that detect infectious pathogens. Constitutive interaction with the SUPPRESSOR OF BIR1 (SOBIR1) RLK contributes to RLP stability and kinase activity. As RLK activation requires transphosphorylation with a second associated RLK, it remains elusive how RLPs initiate downstream signaling. We employed live-cell imaging, gene silencing and coimmunoprecipitation to investigate the requirement of associated kinases for functioning and ligand-induced subcellular trafficking of Cf RLPs that mediate immunity of tomato against Cladosporium fulvum. Our research shows that after elicitation with matching effector ligands Avr4 and Avr9, BRI1-ASSOCIATED KINASE 1/SOMATIC EMBRYOGENESIS RECEPTOR KINASE 3 (BAK1/SERK3) associates with Cf-4 and Cf-9. BAK1/SERK3 is required for the effector-triggered hypersensitive response and resistance of tomato against C. fulvum. Furthermore, Cf-4 interacts with SOBIR1 at the plasma membrane and is recruited to late endosomes upon Avr4 trigger, also depending on BAK1/SERK3. These observations indicate that RLP-mediated resistance and endocytosis require ligand-induced recruitment of BAK1/SERK3, reminiscent of BAK1/SERK3 interaction and subcellular fate of the FLAGELLIN SENSING 2 (FLS2) RLK. This reveals that diverse classes of cell surface immune receptors share common requirements for initiation of resistance and endocytosis. © 2016 The Authors. New Phytologist © 2016 New Phytologist Trust.

  14. Neutron irradiation embrittlement of reactor pressure vessel steel 20 MnMoNi55 weld

    International Nuclear Information System (INIS)

    Ghoneim, M.M.

    1987-05-01

    The effect of neutron irradiation on the mechanical and fracture properties of an 'improved' 20 MnMoNi 55 Pressure Vessel Steel (PVS) weld was investigated. In addition to very low residual element content, especially Cu (0.035 wt.%), and relatively higher Ni content (0.9 wt.%), this steel has higher strength (30% more) than the steels used currently in nuclear reactor pressure vessels. The material was irradiated to 3.5x10 19 and 7x10 19 n/cm 2 (E > 1 Mev) at 290 0 C and 2.5x10 19 n/cm 2 (E > 1 MeV) at 160 0 C in FRJ-1 and FRJ-2 research reactors at KFA, Juelich, F.R.G. Test methods used in the evaluation included instrumented impact testing of standard and precracked Charpy specimens, tensile, and fracture toughness testing. Instrumented impact testing provided load and energy vs. time (deflection) data in addition to energy absorption data. The results indicated that the investigated high strength improved steel is more resistant to irradiation induced embrittlement than conventional PVSs. (orig./IHOE)

  15. Aortic valve surgery of the 21st century: sutureless AVR versus TAVI.

    Science.gov (United States)

    Costache, Victor S; Moldovan, Horatiu; Arsenescu, Catalina; Costache, Andreea

    2018-04-01

    Surgical aortic valve replacement (sAVR) has been a safe, effective and time-proven technique and is still the standard of care all over the world for aortic valve treatment. The vast majority of centers perform this procedure by doing a median sternotomy with several disadvantages. While many others specialties went minimally invasive decades ago, in cardiovascular field transcatheter valve implantation was the first minimally invasive valvular procedure that gained rapid worldwide acceptance. Transcatheter valve replacement (TAVR) is now marketed as a procedure that should be performed under local anesthesia, by an interventional cardiologist via trans femoral route with no other healthcare professional invited to the patient selection or case planning. An increasing number of surgeons are promoting minimally invasive aortic valve replacement, which is gaining grounds, especially with the help of the new sutureless valve technology. With these two new technologies emerging, legitimate questions arise and need to be answered - which has the longest durability, lower complication rate and lower overall mortality.

  16. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    Energy Technology Data Exchange (ETDEWEB)

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next

  17. Reasons of an experimental effort for pebble bed reactors. A program of measurements in the CESAR reactor at Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    Scherer, W; Bock, H J; Krings, F; Neef, R D; Langlet, G; Dixmier, M; Laponche, B; Morier, F

    1972-06-15

    An extended experimental program on neutron physics of HTR fuel balls is being performed in the graphite moderated critical faclity CESAR at CEN Cadarache (France). The experiments are done in the frame of a cooperation between KFA Juelich and CEA Cadarache.

  18. Leptosphaeria maculans effector AvrLm4-7 affects salicylic acid (SA) and ethylene (ET) signalling and hydrogen peroxide (H2O2) accumulation in Brassica napus

    Czech Academy of Sciences Publication Activity Database

    Nováková, Miroslava; Šašek, Vladimír; Trdá, Lucie; Krutinová, Hana; Mongin, T.; Valentová, O.; Balesdent, M.H.; Rouxel, T.; Burketová, Lenka

    2016-01-01

    Roč. 17, č. 6 (2016), s. 818-831 ISSN 1464-6722 R&D Projects: GA ČR GA13-26798S Institutional support: RVO:61389030 Keywords : AvrLm4-7 * Brassica napus * effector Subject RIV: GF - Plant Pathology, Vermin, Weed, Plant Protection Impact factor: 4.697, year: 2016

  19. In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions in the LVR-15 nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, Jan [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Entler, Slavomir, E-mail: slavomir.entler@cvrez.cz [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Vsolak, Rudolf; Klabik, Tomas [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Zlamal, Ondrej [CEZ, Duhova 2/1444, 140 53 Praha 4 (Czech Republic); Bellin, Boris; Zacchia, Francesco [Fusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2015-10-15

    Highlights: • Irradiated thermal fatigue testing of the ITER primary first wall mock-ups. • Cyclic heat flux of 0.5 MW/m{sup 2} in the neutron field of the nuclear reactor core. • 17,040 thermal cycles. • Radiation damage in the range of 0.41–1.17 dpa depending on the material. - Abstract: The TW3 in-pile rig enabled the thermal fatigue testing of ITER primary first wall mock-ups in the core of the nuclear reactor. This experiment investigated the neutron irradiation influence on the design performance under high heat flux testing. A thermal flux of 0.5 MW/m{sup 2} in the neutron field of the core of the LVR-15 nuclear reactor was applied. Within the scope of the tests with simultaneous neutron irradiation, the TW3 rig reached a record of 17,040 thermal cycles with the radiation damage in the range of 0.41–1.17 dpa depending on the material. Even after a high number of thermal cycles, while being irradiated by neutrons, no damage of the tested mock-ups was visually observed. Further testing and analysis will follow in the Forschungszentrum Juelich.

  20. FPGA-based upgrade of the read-out electronics for the low energy polarimeter at COSY/Juelich

    Energy Technology Data Exchange (ETDEWEB)

    Hempelmann, Nils [Institut fuer Kernphysik, Forschungszentrum Juelich (Germany); Collaboration: JEDI-Collaboration

    2016-07-01

    The Cooler Synchrotron (COSY) is a facility for cooled polarized beams at the Forschungszentrum in Juelich. The Low Energy Polarimeter (LEP) is the polarimeter in the injection beam line of COSY. The beam polarization is measured using scattering off carbon and polyethylene (CH2) targets. The outgoing particles are detected using twelve plastic scintillators installed in groups of three to the left, to the right, above, and below the beam. The LEP is the routine tool for beam set-up, but its performance was limited by the old read-out electronics consisting of analog NIM modules. A new system using analog pulse sampling and an FPGA chip for signal processing was installed and tested. The ejectile particles were identified by relative time of flight measurement using a signal from the RF amplifier of the cyclotron used for acceleration as a reference. The new system is able to measure the time at which a particle arrives to an accuracy in the order of 50 ps. The presentation includes a review of available systems and a report about measurements in May and December 2015.

  1. Recent developments in solid oxide fuel cells at Forschungszentrum Juelich and in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Steinberger-Wilckens, Robert; Blum, L.; Buchkremer, H.P.; Haart, L.J.G. de; Malzbender, J. [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energieforschung (IEF); Pap, M.; Gross, S.M. [Forschungszentrum Juelich GmbH (DE). Zentralabteilung Technologie (ZAT)

    2010-07-01

    The SOFC group at FZJ has assembled and tested more than 350 SOFC stacks rated between 100 W and 15 kW during the last 15 years. The research topics cover the whole SOFC development area from materials over stack design, manufacturing of cells, stacks and components, mechanical and electrochemical characterisation, up to system design and demonstration. Use of improved steels, cathodes and materials processing has resulted in reduced degradation rates around 4 mV (<0.50%) per 1000 hours at 800 C and 500 mA/cm{sup 2} over tested stack lifetimes of over 15 000 hours. Other stacks operating at 700 C have already reached over 22.000 hours of lifetime. However, the target of development is directed at even further lowered degradation for commercial operation in stationary applications. All stack tests are accompanied by disassembly and post-operative examinations investigating such phenomena as cathode degradation, corrosion, and other ageing phenomena. These analyses give a deep insight into the interaction of the stack materials and supply vital data on assessing the possibilities for prolonged stack operation over some 10's of thousand hours. This paper gives an overview and summary of achievements of Juelich developments. It also discusses the European perspectives in SOFC commercialisation. European manufacturers are holding a leading edge on the planar SOFC technology with new activities developing rapidly. (orig.)

  2. Dynamics and inherent safety features of small modular high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

    1986-01-01

    Investigations were made at Oak Ridge National Laboratory to characterize the dynamics and inherent safety features of various modular high temperature gas-cooled reactor (HTGR) designs. This work was sponsored by the US Nuclear Regulatory Commission's HTGR Safety Research program. The US Department of Energy (DOE) and the Gas Cooled Reactor Associates (GCRA) have sponsored studies of several modular HTGR concepts, each having it own unique advantageous economic and inherent safety features. The DOE design team has recently choses a 350-MW(t) annular core with prismatic, graphite matrix fuel for its reference plant. The various safety features of this plant and of the pebble-bed core designs similar to those currently being developed and operated in the Federal Republic of Germany (FRG) are described. A varity of postulated accident sequences involving combinations of loss of forced circulation of the helium primary coolant, loss of primary coolant pressurization, and loss of normal and backup heat sinks were studied and are discussed. Results demonstrate that each concept can withstand an uncontrolled heatup accident without reaching excessive peak fuel temperatures. Comparisons of calculated and measured response for a loss of forced circulation test on the FRG reactor, AVR, are also presented. 10 refs

  3. Diseño del núcleo de un procesador AVR de 8 bits utilizando lógica programable de Altera

    Directory of Open Access Journals (Sweden)

    Dilaila Criado

    2010-09-01

    Full Text Available Normal 0 21 false false false ES-TRAD X-NONE X-NONE MicrosoftInternetExplorer4 /* Style Definitions */ table.MsoNormalTable {mso-style-name:"Tabla normal"; mso-tstyle-rowband-size:0; mso-tstyle-colband-size:0; mso-style-noshow:yes; mso-style-priority:99; mso-style-qformat:yes; mso-style-parent:""; mso-padding-alt:0cm 5.4pt 0cm 5.4pt; mso-para-margin:0cm; mso-para-margin-bottom:.0001pt; mso-pagination:widow-orphan; font-size:11.0pt; font-family:"Calibri","sans-serif"; mso-ascii-font-family:Calibri; mso-ascii-theme-font:minor-latin; mso-fareast-font-family:"Times New Roman"; mso-fareast-theme-font:minor-fareast; mso-hansi-font-family:Calibri; mso-hansi-theme-font:minor-latin; mso-bidi-font-family:"Times New Roman"; mso-bidi-theme-font:minor-bidi;} En este trabajo se describe, el diseño del núcleo de un microcontrolador AVR, en específico uno similar al AT90S8515. Se utiliza el lenguaje de descripción de hardware VHDL con dispositivos programables de las familias Cyclone y Cyclone II de Altera. Los objetivos que se persiguen con este trabajo son; valorar alternativas de diseño de sistemas digitales en circuitos programables, contar con el diseño de un núcleo AVR para futuras aplicaciones como modulo IP y realizar diseños de sistemas con varios procesadores (SoPC.

  4. Sana experiments for self-acting removal of the after-heat in reactors with pebble bed fuel and their interpretation

    International Nuclear Information System (INIS)

    Niessen, H.F.; Stoecker, Bernd; Amoignon, Olivier; Zuying, Gao; Jie, Liu

    1997-01-01

    For the confirmation of self-acting afterheat removal under hypothetical accident conditions from pebble bed reactors at the Research Center Juelich a test facility with an electrical heating input up to 30kW was erected and operated. A description of the test facility is given. Within the different tests the pebble diameter, the pebble material, the gas in the pebble bed, the heating-power and the arrangement of the heating were changed. Parts of the data were used within an IAEA Co-ordinated Research Program as benchmark problems for the code validation. All computer codes could simulate the test results with a sufficient good agreement, when the tests were executed with helium. For the tests with nitrogen the natural convection has to be taken into account. (author)

  5. The gas-cooled high temperature reactor: perspectives, problems and programmes

    International Nuclear Information System (INIS)

    Beckurts, K.H.; Engelmann, P.; Erb, D.E.

    1977-01-01

    For nearly 20 years, extensive research and development programs on Helium-cooled high-temperature reactors (HTR) have been carried out in several countries of the world, in particular in Germany and in the United States. This reactor system offers major potential advantages as a source of electricity or of nuclear process heat: it shows high nuclear fuel conversion efficiency, permitting a better utilization of uranium and in particular of thorium resources; it offers a high degree of inherent nuclear safety and thus a good potential for adoption to very strict safety standards; it permits high-efficiency electricity generation using either the indirect steam or the direct Helium cycle; dry air cooling can be employed without major economic penalties; it permits direct use of the nuclear heat for the production of gaseous or liquid secondary fuels from coal and other fossil fuels or - on a more extended time scale - by thermochemical water splitting. As a result of the longstanding efforts, satisfactory solutions have been found for many of the basic problems of this new reactor system, particularly in the field of high-temperature fuels and materials technology. Three small experimental plants - Peach Bottom in USA, Dragon in England, and AVR in Germany - have been operated successfully over extended periods of time. The AVR is still in operation; since 1974 it has performed satisfactorily with an average gas outlet temperature of 950 0 C. Prototype steam-cycle plants of 300 MW(e) are underway at Fort St. Vrain, USA (full-power operation scheduled for 1977), and at Schmehausen, Germany (scheduled for 1979). Major delays have occured in the construction and commissioning of these plants; they are due to various reasons and do not reveal specific problems of the HTR. Commercial market introduction of the steam-cycle electricity generating system has been attempted, but the first approach has not been successfull. Major effects by both government and industry are

  6. PENINGKATAN KINERJA PERANGKAT ELEKTRONIK BERBASIS MIKROKONTROLER AVR 8 BIT DENGAN MENGGUNAKAN RTOS (REAL TIME OPERATING SYSTEM

    Directory of Open Access Journals (Sweden)

    I Wayan Sutaya

    2015-01-01

    Full Text Available Tujuan dari penelitian ini adalah untuk mengimplementasikan RTOS (Real Time Operating System pada perangkat elektronik berbasis mikrokontroler AVR 8 bit sehingga didapatkan peningkatan kinerja pada perangkat tersebut. RTOS yang digunakan adalah freeRTOS dimana RTOS ini mendukung mikrokontroler 8 bit, berukuran kecil, dan bersifat open source. Hasil penelitian ini berguna bagi praktisi-praktisi elektronika dalam hal menekan biaya produksi pembuatan perangkat elektronik berbasis mikrokontroler 8 bit karena resource yang digunakan bisa dikurangi. Metode yang digunakan adalah metode penelitian pengembangan (Research and development dengan cara membuat studi kasus perangkat elektronik yang berbasis mikrokontroler. Perangkat elektronik yang dibuat adalah unit kendali elektronik pada sepeda motor. Selanjutnya mikrokontroler pada perangkat ini diprogram dengan menggunakan dua skenario. Skenario pertama adalah tanpa menggunakan RTOS dan skenario kedua dengan menggunakan RTOS. Dari dua skenario ini dilakukan pengujian dan analisis untuk mengetahui besar peningkatan kinerja yang didapat.

  7. Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

    International Nuclear Information System (INIS)

    Marmier, Alain

    2012-01-01

    high temperature irradiation to high burn-ups with fission gas release measurements. To this end, the HFR-EU1 fuel irradiation in the High Flux Reactor (HFR) Petten (2006-2010) explored the potential for high performance and high burn-up of existing German fuel (3 pebbles produced for the AVR reactor at the German research centre Juelich) and newly produced Chinese fuel (2 pebbles produced by INET for use in the HTR-10 test reactor in China). These five pebbles were irradiated for 445 days in separately controlled capsules, while the fission gas release was monitored by gamma spectrometry thus enabling evaluation of the characteristic release over birth fraction, indicative for the health of the fuel. In none of the pebbles, abnormally increased fission gas release was observed indicating that all of the approx. 45,000 coated particles in the pebbles had remained intact. The results presented in this thesis cover the first 332 days of irradiation. While HFR-EU1 was dedicated to a particularly high burn-up, HFR-EU1bis, performed between 2004 and 2005, investigated extremely high temperature for steady-state conditions. The comparison of both experiments confirms that temperature plays a decisive part in fuel performance and integrity. The peak fuel temperature in pebbles can be lowered with the so-called w allpaper fuel , in which the coated fuel particles are arranged in a spherical shell within a pebble. This wallpaper concept also enhances neutronic performance through improved neutron economy, resulting in reduced fissile material and/or enrichment needs or providing the potential to achieve higher burn-up. To quantify these improvements, calculations were performed using the Monte Carlo neutron transport and depletion codes MCNP/MCB (to assess conversion ratio, temperature coefficient of reactivity and neutron multiplication) and PANTHERMIX (for fuel cycle in steady state conditions and loss of coolant accident calculations). Based on PANTHERMIX steady

  8. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  9. Heat transport the cold way

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    A novel system for long-distance heat transport is being born in the 'Kernforschungsanlage Juelich' with the project being called 'Nukleare Fernenergie' (nuclear district energy). The project is also known as 'EVA/ADAM' [EVA = Einzelrohr-Versuchs-Anlage (single tube test facility); ADAM = Anlage mit Drei Adiabaten Methanisierungsreaktoren (plant provided with three adiabate methanising reactors)] and is based in principle on transport of energy in chemical bond within a closed loop. In the 60ies already this development was discussed both in the 'Kernforschungsanlage Juelich' and in the 'Rheinische Braunkohlenwerke' independent of each other. In 1975 these two organizations concluded a co-operation contract. (orig.) [de

  10. Bilateral cooperation between Germany and Brazil on fuel irradiation

    International Nuclear Information System (INIS)

    Dias, J.W.

    1977-01-01

    Within the framework of the Government Agreement on Scientific and Technical Cooperation between the Federal Republic of Germany and Brazil, the Brazilian National Atomic Commission and the Juelich Nuclear Research Center (KFA) signed on 23rd April, 1971 an Agreement on Cooperation in the field of Nuclear Research and Reactor Technology. Projects have been elaborated in fields of mutual interest to share activities between the partner institutes in both countries. A typical project is the fuel irradiation programme jointly prepared by NUCLEBRAS and KFA-Juelich. Brazil is planning to use elements of its own production in nuclear power plants to be erected within the German-Brazilian Industrial Agreement. As no material test reactor is available in Brazil it is expedient to irradiate samples of Brazilian production in Germany. Brazilian collaborators will participate in the preparation, execution and post-irradiation examination. In this way an optimum transfer of all information and results is assured. In the first phase, sample rods manufactured in Brazil are irradiated in the FRJ-2 test reactor in Juelich. These rods are assembled under clean conditions in the NUCLEBRAS research centres. The first Brazilian test rods showed excellent in-pile behaviour even under very high fuel rod capacity. In the second phase, fuel rods of original length manufactured and assembled in Brazil will be irradiated in German power plants, and, at the same time, additional irradiations of small samples will be carried out in test reactors. In the third phase, rod clusters and complete fuel elements will be manufactured in Brazil and irradiated in German power plants until target burn-up. All the necessary prerequisites have been fulfilled to meet the above requirements, i.e. mutual interest, good infrastructure maintained by both partners, qualified personnel and last but not least unbureaucratic and effective help by the coordinating offices of NUCLEBRAS and KFA

  11. The development of divertor and first wall armour parts at JAERI, Sandia N.L. and KFA Juelich

    International Nuclear Information System (INIS)

    Akiba, M.; Bolt, H.; Watson, R.; Kneringer, G.; Linke, J.

    1991-01-01

    The development of new armour materials, and fabrication and testings of the divertor and first wall mock-ups have worldwidely been carried out during the Conceptual Design Activites (CDA) of ITER. This paper is a review of the activities on the divertor and first wall armour components which has been performed by JAERI, Sandia National Laboratory, and KFA Juelich. The design requirements have instantly been reflected in material development. For instance, carbon fiber composites (CFCs) have already been developed to have a thermal conductivity as high as copper at room temperature. Further modification of CFC's has been made. Based on the extensive progress in armour materials, the fabrication and testings of mock-ups have been started. Divertor mock-ups which are able to endure a stationary heat flux of 8 to 15 MW/m 2 have already been developed. Some new high heat flux test facilities have been constructed and are able to simulate a heat load of plasma disruption. Extensive understanding on disruption erosion of the armour materials has been obtained by experiments with these facilities. Some mock-up tests and disruption simulating tests have been performed under international collaboration. (orig.)

  12. The challenge of introducing HTR plants on to the international power plant market

    International Nuclear Information System (INIS)

    Bogen, J.; Stoelzl, D.

    1987-01-01

    The international power plant market today is characterized by high increase in energy consumption for developing countries with limitations of investment capital and low increase in energy consumption for industrialized countries with limitations of additional power plant capacities. As a consequence there is a low demand for large new power stations. This leads to a tendency for small and medium sized power plant units - meeting high environmental standards - for which the total investment volume is low and full load operation of a plant can be realized earlier due to the small block capacity. - For nuclear power plants the High-Temperature-Reactor (HTR)-line with spherical fuel elements and a core structure of graphite is specially suited for this small and medium sized nuclear reactor (SMSNR) capacity. The excellent safety characteristics, the high availability, the low radiation doses for the operation personnel and the environment of the HTR line has been demonstrated by 20 years of operation of the AVR-15 MWe experimental power plant in Juelich F.R.G. and since 1985 by operation of the THTR-300 MWe prototype plant at Hamm-Uentrop F.R.G. Up-dated concepts of the HTR-line are under design for electricity generation (HTR-500), for co-generation of power and heat (HTR-100) and for district heating purposes only (GHR-10). By implementing two HTR projects the Brown Boveri Group is in the position to realize the collected experiences from design, licensing, erection, commissioning and operation for the follow-on projects. This leads to practical and sound technical solutions convenient for existing manufacturing processes, well known materials, standardized components and usual manufacturing tolerances. Specific plant characteristics can be used for advantages in the competition. (author)

  13. Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors. Proceedings of a specialists meeting held in Juelich, Germany, 6-8 July 1992

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-15

    The meeting was convened by the International Atomic Energy Agency on the recommendation of the IAEA`s International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Poland, the Russian Federation, Switzerland, the United Kingdom and the United States of America. The meeting was chaired by Prof. Dr. K. Kugeler and Prof. Dr. E. Hicken, Directors of the Institute for Safety Research Technology of the KFA Research Center, and covered the following: Design and licensing requirements for gas cooled reactors; concepts for decay heat removal in modern gas cooled reactors; analytical methods for predictions of thermal response, accuracy of predictions; experimental data for validation of predictive methods - operational experience from gas cooled reactors and experimental data from test facilities. Refs, figs and tabs.

  14. Seismic behaviour of gas cooled reactor components

    International Nuclear Information System (INIS)

    1990-08-01

    On invitation of the French Government the Specialists' Meeting on the Seismic Behaviour of Gas-Cooled Reactor Components was held at Gif-sur-Yvette, 14-16 November 1989. This was the second Specialists' Meeting on the general subject of gas-cooled reactor seismic design. There were 27 participants from France, the Federal Republic of Germany, Israel, Japan, Spain, Switzerland, the United Kingdom, the Soviet Union, the United States, the CEC and IAEA took the opportunity to present and discuss a total of 16 papers reflecting the state of the art of gained experiences in the field of their seismic qualification approach, seismic analysis methods and of the capabilities of various facilities used to qualify components and verify analytical methods. Since the first meeting, the sophistication and expanded capabilities of both the seismic analytical methods and the test facilities are apparent. The two main methods for seismic analysis, the impedance method and the finite element method, have been computer-programmed in several countries with the capability of each of the codes dependent on the computer capability. The correlations between calculation and tests are dependent on input assumptions such as boundary conditions, soil parameters and various interactions between the soil, the buildings and the contained equipment. The ability to adjust these parameters and match experimental results with calculations was displayed in several of the papers. The expanded capability of some of the new test facilities was graphically displayed by the description of the SAMSON vibration test facility at Juelich, FRG, capable of dynamically testing specimens weighing up to 25 tonnes, and the TAMARIS facility at the CEA laboratories in Gif-sur-Yvette where the largest table is capable of testing specimens weighing up to 100 tonnes. The proceedings of this meeting contain all 16 presented papers. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  15. Preliminary experiment design of graphite dust emission measurement under accident conditions for HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Wei, E-mail: pengwei@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Chen, Tao; Sun, Qi; Wang, Jie [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2017-05-15

    Highlights: • A theoretical analysis is used to predict the total graphite dust release for an AVR LOCA. • Similarity criteria must be satisfied between the experiment and the actual HTGR system. • Model experiments should be conducted to predict the graphite dust resuspension rate. - Abstract: The graphite dust movement behavior is significant for the safety analyses of high-temperature gas cooled reactor (HTGR). The graphite dust release for accident conditions is an important source term for HTGR safety analyses. Depressurization release tests are not practical in HTGR because of a radioactivity release to the environment. Thus, a theoretical analysis and similarity principles were used to design a group of modeling experiments. Modeling experiments for fan start-up and depressurization process and actual experiments of helium circulator start-up in an HTGR were used to predict the rate of graphite dust resuspension and the graphite dust concentration, which can be used to predict the graphite dust release during accidents. The modeling experiments are easy to realize and the helium circulator start-up test does not harm the reactor system or the environment, so this experiment program is easily achieved. The revised Rock’n’Roll model was then used to calculate the AVR reactor release. The calculation results indicate that the total graphite dust releases during a LOCA will be about 0.65 g in AVR.

  16. The contribution of a small TRIGA university research reactor to nuclear research on an international level

    International Nuclear Information System (INIS)

    Villa, M.; Bastuerk, M.; Boeck, H.

    2002-01-01

    The paper focuses especially on the important results in neutron- and solid state physics and the co-operation between the low power TRIGA reactor with high flux neutron sources in Europe such as the Institute Laue-Langevin (ILL) in Grenoble, the Paul Scherrer Institut (PSI) in Villigen, the Rutherford Appleton Laboratory (RAL) in Didcot and the Research Center Juelich. Experiments are set up for test purposes at the TRIGA reactor and then transferred to the powerful neutron sources. Different new perfect silicon channel-cut and interferometer crystals are prepared and then tested at the Bonse-Hart camera, which is a double crystal (or triple axis) diffractometer and at the interferometer set-up. Historically, the first verification of neutron interferometry at a perfect crystal device has been achieved at the 250 kW TRIGA-reactor in Vienna in the year 1974. Also the co-operation with the PSI and the TU Munich in the field of neutron radiography and neutron tomography and VESTA, an experiment for storing cold neutrons with a wavelength of 6.27A, installed at the pulsed neutron source ISIS at RAL are mentioned. The second topic in this paper focuses on the co-operation in the field of safeguard. Several projects have been carried out during the past years in co-operation with the IAEA such as establishing a gamma spectrum reference catalogue for CdZnTe detectors and tests of safeguard video cameras under neutron irradiation. Further an integrated safeguard surveillance network composed of a video camera, a gamma monitor and a neutron monitor is under development

  17. Aerosol size characteristics in selected working areas

    International Nuclear Information System (INIS)

    Ahmed, K.

    1984-05-01

    This report presents the work done to study the aerosol activity size distributions and their respirable fractions in some selected areas of the Juelich Nuclear Research Center. Anderson cascade impactors were used to find the aerodynamic size ranges of the airborne particles for subsequent analysis of activity associated with each size group. The aerosols were found to follow in general log-normal distributions in the hot cells with values of AMAD between 5 and 10 μm. Measurements in the AVR containment and decontamination laboratory in Uranit GmbH showed deviations from log-normal distribution. In the waste press area the distribution is sometimes log-normal and sometimes not, depending upon the origin of waste. The values of AMAD are in the range of 2 to 4 μm in these areas. The respirable fractions were calculated using ACGIH definition for respirable dust to be < 25% in hot cells and < 60% in other areas. Pulmonary depositions according to ICRP model were < 10% and < 15% respectively. (orig./HP)

  18. Investigations to the suitability of a computer-aided design system for mechanical construction in the ZAT of KFA Juelich

    International Nuclear Information System (INIS)

    Schoerner, M.; Koch, R.; Cordewiner, H.J.; Bachner, E.

    1983-06-01

    There is an extensive range of CAD systems in the marketplace which have been developed by various institutions for different types of applications and for different products, in some cases linked to specific hardware. Apart from such obvious features like prices, rates of sale and computer performance, other criteria such as documentation, ease of adaption and expansion, which are, as a rule, difficult to judge and quantify, play a decisive role. As an optimal CAD system does not exist in respect of every type of application the intended usage must be taken seriously into consideration when selecting a CAD system. On the basis of fundamental investigations of the suitability of CAD for the department for mechanical construction of ZAT at the KFA Juelich, the requirements and special needs of the construction process in the ZAT have been examined. In a short survey of the available CAD-systems a pre choice on the basis of some definitive musts has been made. The performance profile of the remaining systems has been thouroughly investigated and compared with the requirements of the ZAT. Thus the necessary decisions can be made on the basis of documented assessments. (orig.) [de

  19. Institute of Energy and Climate Research IEK-6. Nuclear waste management and reactor safety report 2009/2010. Material science for nuclear waste management

    International Nuclear Information System (INIS)

    Klinkenberg, M.; Neumeier, S.; Bosbach, D.

    2011-01-01

    Due to the use of nuclear energy about 17.000 t (27.000 m 3 ) of high level waste and about 300.000 m 3 of low and intermediated level waste will have accumulated in Germany until 2022. Research in the Institute of Energy and Climate Research (IEK-6), Nuclear Waste Management and Reactor Safety Division focuses on fundamental and applied aspects of the safe management of nuclear waste - in particular the nuclear aspects. In principle, our research in Forschungszentrum Juelich is looking at the material science/solid state aspects of nuclear waste management. It is organized in several research areas: The long-term safety of nuclear waste disposal is a key issue when it comes to the final disposal of high level nuclear waste in a deep geological formation. We are contributing to the scientific basis for the safety case of a nuclear waste repository in Germany. In Juelich we are focusing on a fundamental understanding of near field processes within a waste repository system. The main research topics are spent fuel corrosion and the retention of radionuclides by secondary phases. In addition, innovative waste management strategies are investigated to facilitate a qualified decision on the best strategy for Germany. New ceramic waste forms for disposal in a deep geological formation are studied as well as the partitioning of long-lived actinides. These research areas are supported by our structure research group, which is using experimental and computational approaches to examine actinide containing compounds. Complementary to these basic science oriented activities, IEK-6 also works on rather applied aspects. The development of non-destructive methods for the characterisation of nuclear waste packages has a long tradition in Juelich. Current activities focus on improving the segmented gamma scanning technique and the prompt gamma neutron activation analysis. Furthermore, the waste treatment group is developing concepts for the safe management of nuclear graphite

  20. Institute of Energy and Climate Research IEK-6. Nuclear waste management and reactor safety report 2009/2010. Material science for nuclear waste management

    Energy Technology Data Exchange (ETDEWEB)

    Klinkenberg, M; Neumeier, S; Bosbach, D [eds.

    2011-07-01

    Due to the use of nuclear energy about 17.000 t (27.000 m{sup 3}) of high level waste and about 300.000 m{sup 3} of low and intermediated level waste will have accumulated in Germany until 2022. Research in the Institute of Energy and Climate Research (IEK-6), Nuclear Waste Management and Reactor Safety Division focuses on fundamental and applied aspects of the safe management of nuclear waste - in particular the nuclear aspects. In principle, our research in Forschungszentrum Juelich is looking at the material science/solid state aspects of nuclear waste management. It is organized in several research areas: The long-term safety of nuclear waste disposal is a key issue when it comes to the final disposal of high level nuclear waste in a deep geological formation. We are contributing to the scientific basis for the safety case of a nuclear waste repository in Germany. In Juelich we are focusing on a fundamental understanding of near field processes within a waste repository system. The main research topics are spent fuel corrosion and the retention of radionuclides by secondary phases. In addition, innovative waste management strategies are investigated to facilitate a qualified decision on the best strategy for Germany. New ceramic waste forms for disposal in a deep geological formation are studied as well as the partitioning of long-lived actinides. These research areas are supported by our structure research group, which is using experimental and computational approaches to examine actinide containing compounds. Complementary to these basic science oriented activities, IEK-6 also works on rather applied aspects. The development of non-destructive methods for the characterisation of nuclear waste packages has a long tradition in Juelich. Current activities focus on improving the segmented gamma scanning technique and the prompt gamma neutron activation analysis. Furthermore, the waste treatment group is developing concepts for the safe management of nuclear

  1. Report on the interpretation of critical experiments in the Siemens-Argonaut-Reactor Graz to study water ingress into spherical elements. Ergebnisbericht zur Auslegung kritischer Experimente am Siemens-Argonaut-Reaktor Graz zum Studium des Wassereinbruches im Kugelhaufen

    Energy Technology Data Exchange (ETDEWEB)

    Schuerrer, F [Technische Univ., Graz (Austria). Inst. fuer Theoretische Physik und Reaktorphysik; Neef, R D [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung

    1979-04-15

    The experiments described are of interest in the study of water contamination in HTR fuel elements. The Siemens Argonaut Reactor (SAR) has been considered as a research tool for a simulation experiment. Following a brief description of the SAR, planned programs are discussed in 'dry' and 'wet' cores. Detector foil types and locations are noted. A theoretical model is developed and nuclide concentrations estimated in the various spectral zones. Reactivity calculations have been made and are summarised for various H{sub 2}O percentage concentrations. The discussion is supported by simplified core layout diagrams and graphs of core flux distributions. Neutron diffusion and spectra calculations are referenced to computer programs used by KFA-Juelich, published elsewhere, and include GAM, THERMOS, MUPO and EXTERMINATOR-2. (G.C.)

  2. Contributions for the seventeenth biennial conference on carbon

    International Nuclear Information System (INIS)

    Delle, W.

    1985-06-01

    This report is the compilation of the papers prepared by KFA Juelich GmbH for the 17th Biennial Conference on Carbon. In the contributions, results are presented which were obtained from the application of carbon in the High Temperature Gas-Cooled Reactor and for the Spallation Neutron Source planned in the Federal Republic of Germany. (orig.) [de

  3. Coal upgrading based on the high temperature reactor - engineering and economic aspects

    International Nuclear Information System (INIS)

    Barnert, H.; Neis, H.

    1987-01-01

    The development of the high temperature reactor in the FRG was opened with the aim of economically producing electricity which might be achieved with the next project works. Moreover, the HTR supplies process heat of very high temperature which may be used with metallic heat exchangers in the range up to 1000 0 C. From these reasons and regarding the world power prospects and the special power situation in our own country, the FRG began to develop methods of upgrading fossil primary energy sources using high-temperature heat from the HTR about 15 years ago. Meanwhile great progress has been achieved in research, development and test works; the modules are developed to a great extent. Economical service is expected for future utilization of selected processes. Problems to be still accomplished are first of all related to system testing as well as further exhaustion of the HTR's temperature potential. The proposed extension of the AVR to some process heat facility offers the possibility of testing the system in a low-cost and accelerated manner. (author)

  4. User's manual for ASTERIX-2: A two-dimensional modular code system for the steady state and xenon transient analysis of a pebble bed high temperature reactor

    International Nuclear Information System (INIS)

    Wu, T.; Cowan, C.L.; Lauer, A.; Schwiegk, H.J.

    1982-03-01

    The ASTERIX modular code package was developed at KFA Laboratory-Juelich for the steady state and xenon transient analysis of a pebble bed high temperature reactor. The code package was implemented on the Stanford Linear Accelerator Center Computer in August, 1980, and a user's manual for the current version of the code, identified as ASTERIX-2, was prepared as a cooperative effort by KFA Laboratory and GE-ARSD. The material in the manual includes the requirements for accessing the program, a description of the major subroutines, a listing of the input options, and a listing of the input data for a sample problem. The material is provided in sufficient detail for the user to carry out a wide range of analysis from steady state operations to the xenon induced power transients in which the local xenon, temperature, buckling and control feedback effects have been incorporated in the problem solution. (orig.)

  5. User's manual for ASTERIX-2: a two-dimensional modular-code system for the steady-state and xenon-transient analysis of a pebble-bed high-temperature reactor

    International Nuclear Information System (INIS)

    Lauer, A.; Schwiegk, H.J.; Wu, T.; Cowan, C.L.

    1982-03-01

    The ASTERIX modular code package was developed at KFA Laboratory-Juelich for the steady state and xenon transient analysis of a pebble bed high temperature reactor. The code package was implemented on the Stanford Linear Accelerator Center Computer in August, 1980, and a user's manual for the current version of the code, identified as ASTERIX-2, was prepared as a cooperative effort by KFA Laboratory and GE-ARSD. The material in the manual includes the requirements for accessing the program, a description of the major subroutines, a listing of the input options, and a listing of the input data for a sample problem. The material is provided in sufficient detail for the user to carry out a wide range of analyses from steady state operations to the xenon induced power transients in which the local xenon, temperature, buckling and control feedback effects have been incorporated in the problem solution

  6. Validation and application of 3D-methods for the design and safety analysis of high temperature reactors

    International Nuclear Information System (INIS)

    Bader, J.; Lapins, J.; Buck, M; Bernnat, W.; Laurien, E.

    2011-01-01

    Some of the concepts for future nuclear reactors are high-temperature gas-cooled reactors. Previous simulation codes for their cores were often based on one- or two-dimensional models, but today's increasing computer capabilities make an advance to 3D-codes possible now. Our thermal-hydraulic code ATTICA3D (Advanced Thermal-hydraulic Tool for In-vessel and Core Analysis in 3 Dimensions) is based on the porous media approach, including 3-D models of heat conduction and gas flow, using a coarse-grid integration method for the time-dependent conservation equations of mass, momentum and energy. Results of numerical calculations for various validation cases are presented: First, the test facility SANA is chosen, which has been used to study heat transfer phenomena inside a coolant-gas filled pebble-bed core, which was heated by embedded electrical heating elements. Calculations were carried out for different tests taken from the experimental database. Measured and calculated temperatures at different positions are compared and found in good agreement. Second, our code was used to simulate a depressurized loss of forced cooling experiment with simulated decay heat in the AVR Experimental Reactor. Due to its design with the shut-down rods located inside columnar noses, which extend into the pebble bed of the core, geometry and power distribution are genuinely three-dimensional. The power distribution was calculated by the 3D-Neutronic Diffusion Code CITATION in conjunction with the spectral code MICROX-2. The neutronics and thermal-hydraulics calculations were carried out for a 3D, 45°-degree section of the reactor. It is demonstrated, that the experimental results could be qualitatively reproduced. (author)

  7. Efficiency of inherent protection mechanisms for an improved HTR safety concept

    Energy Technology Data Exchange (ETDEWEB)

    Petersen, K.; Buescher, R.; Gerwin, H.; Schenk, W. [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung

    1981-01-15

    For a preliminary design of a 350 MWsub(th) annular core derived from AVR-reactor the efficiency of inherent protection mechanisms is discussed. After-heat removal and auto-shut down potential are demonstrated for intact and complete failure of core heat sinks.

  8. Radionuclide release from research reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Curtius, H., E-mail: h.curtius@fz-juelich.de [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany); Kaiser, G.; Mueller, E.; Bosbach, D. [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany)

    2011-09-01

    Numerous investigations with respect to LWR fuel under non oxidizing repository relevant conditions were performed. The results obtained indicate slow corrosion rates for the UO{sub 2} fuel matrix. Special fuel-types (mostly dispersed fuels, high enriched in {sup 235}U, cladded with aluminium) are used in German research reactors, whereas in German nuclear power plants, UO{sub 2}-fuel (LWR fuel, enrichment in {sup 235}U up to 5%, zircaloy as cladding) is used. Irradiated research reactor fuels contribute less than 1% to the total waste volume. In Germany, the state is responsible for fuel operation and for fuel back-end options. The institute for energy research (IEF-6) at the Research Center Juelich performs investigation with irradiated research reactor spent fuels under repository relevant conditions. In the study, the corrosion of research reactor spent fuel has been investigated in MgCl{sub 2}-rich salt brine and the radionuclide release fractions have been determined. Leaching experiments in brine with two different research reactor fuel-types were performed in a hot cell facility in order to determine the corrosion behaviour and the radionuclide release fractions. The corrosion of two dispersed research reactor fuel-types (UAl{sub x}-Al and U{sub 3}Si{sub 2}-Al) was studied in 400 mL MgCl{sub 2}-rich salt brine in the presence of Fe{sup 2+} under static and initially anoxic conditions. Within these experimental parameters, both fuel types corroded in the experimental time period of 3.5 years completely, and secondary alteration phases were formed. After complete corrosion of the used research reactor fuel samples, the inventories of Cs and Sr were quantitatively detected in solution. Solution concentrations of Am and Eu were lower than the solubility of Am(OH){sub 3}(s) and Eu(OH){sub 3}(s) solid phases respectively, and may be controlled by sorption processes. Pu concentrations may be controlled by Pu(IV) polymer species, but the presence of Pu(V) and Pu

  9. The contribution of a small triga university research reactor to nuclear research on an international level

    International Nuclear Information System (INIS)

    Villa, M.; Boeck, H.; Weber, H.W.

    2001-01-01

    The paper focuses especially on the important results in neutron- and solid state physics and the co-operation between the low power TRIGA reactor with high flux neutron sources in Europe such as the Institute Laue-Langevin (ILL) in Grenoble, the Paul Scherrer Institut (PSI) in Villigen, the Rutherford Appleton Laboratory (RAL) in Didcot and the Research Center Juelich. Experiments are set up for test purposes at the TRIGA reactor and then transferred to the powerful neutron sources. Different new perfect silicon channel-cut and interferometer crystals are prepared and then tested at the Bonse-Hart camera, which is a double crystal (or triple axis) diffractometer and at the interferometer set-up. Historically, the first verification of neutron interferometry at a perfect crystal device has been achieved at the 250 kW TRIGA-reactor in Vienna in the year 1974. Also the co-operation with the PSI and the TU Munich in the field of neutron radiography and neutron tomography and VESTA, an experiment for storing cold neutrons with a wavelength of 6.27 A, installed at the pulsed neutron source ISIS at RAL will be mentioned. The second topic treated in this paper shows the international co-operation in the field of superconductors. This research work is carried out under two European TMR-Network programs. The third topic in this paper focuses on the co-operation in the field of safeguard. Several projects have been carried out during the past years in co-operation with the IAEA such as establishing a gamma spectrum reference catalogue for CdZnTe detectors and tests of safeguard video cameras under neutron irradiation. Further an integrated safeguard surveillance network composed of a video camera, a gamma monitor and a neutron monitor is under development. (orig.)

  10. Institute of Energy and Climate Research IEK-6. Nuclear waste management and reactor safety report 2009/2010. Material science for nuclear waste management

    Energy Technology Data Exchange (ETDEWEB)

    Klinkenberg, M.; Neumeier, S.; Bosbach, D. (eds.)

    2011-07-01

    Due to the use of nuclear energy about 17.000 t (27.000 m{sup 3}) of high level waste and about 300.000 m{sup 3} of low and intermediated level waste will have accumulated in Germany until 2022. Research in the Institute of Energy and Climate Research (IEK-6), Nuclear Waste Management and Reactor Safety Division focuses on fundamental and applied aspects of the safe management of nuclear waste - in particular the nuclear aspects. In principle, our research in Forschungszentrum Juelich is looking at the material science/solid state aspects of nuclear waste management. It is organized in several research areas: The long-term safety of nuclear waste disposal is a key issue when it comes to the final disposal of high level nuclear waste in a deep geological formation. We are contributing to the scientific basis for the safety case of a nuclear waste repository in Germany. In Juelich we are focusing on a fundamental understanding of near field processes within a waste repository system. The main research topics are spent fuel corrosion and the retention of radionuclides by secondary phases. In addition, innovative waste management strategies are investigated to facilitate a qualified decision on the best strategy for Germany. New ceramic waste forms for disposal in a deep geological formation are studied as well as the partitioning of long-lived actinides. These research areas are supported by our structure research group, which is using experimental and computational approaches to examine actinide containing compounds. Complementary to these basic science oriented activities, IEK-6 also works on rather applied aspects. The development of non-destructive methods for the characterisation of nuclear waste packages has a long tradition in Juelich. Current activities focus on improving the segmented gamma scanning technique and the prompt gamma neutron activation analysis. Furthermore, the waste treatment group is developing concepts for the safe management of nuclear

  11. A PC-program for the calculation of neutron flux and element contents using the ki-method of neutron activation analysis

    International Nuclear Information System (INIS)

    Boulyga, E.G.; Boulyga, S.F.

    2000-01-01

    A computer program is described, which calculates the induced activities of isotopes after irradiation in a known neutron field, thermal and epithermal neutron fluxes from the measured induced activities and from nuclear data of 2-4 monitor nuclides as well as the element concentrations in samples irradiated together with the monitors. The program was developed for operation in Windows 3.1 (or higher). The application of the program for neutron activation analysis allows to simplify the experimental procedure and to reduce the time. The program was tested by measuring different types of standard reference materials at the FRJ-2 (Research Centre, Juelich, Germany) and Triga (University Mainz, Germany) reactors. Comparison of neutron flux parameters calculated by this program with those calculated by a VAX program developed at the Research Centre, Juelich was done. The results of testing seem to be satisfactory. (author)

  12. SWR 1000: an advanced boiling water reactor with passive safety features

    International Nuclear Information System (INIS)

    Brettschuh, W.

    1999-01-01

    The SWR 1000, an advanced BWR, is being developed by Siemens under contract from Germany's electric utilities and with the support of European partners. The project is currently in the basic design phase to be concluded in mid-1999 with the release of a site-independent safety report and costing analysis. The development goals for the project encompass competitive costs, use of passive safety systems to further reduce probabilities of occurrence of severe accidents, assured control of accidents so no emergency response actions for evacuation of the local population are needed, simplification of plant systems based on operator experience, and planning and design based on German codes, standards and specifications put forward by the Franco-German Reactor Safety Commission for future nuclear power plants equipped with PWRs, as well as IAEA specifications and the European Utility Requirements. These goals led to a plant concept with a low power density core, with large water inventories stored above the core inside the reactor pressure vessel, in the pressure suppression pool, and in other locations. All accident situations arising from power operation can be controlled by passive safety features without rise in core temperature and with a grace period of more than three days. In addition, postulated core melt is controlled by passive equipment. All new passive systems have been successfully tested for function and performance using large-scale components in experimental testing facilities at PSI in Switzerland and at the Juelich Research Centre in Germany. In addition to improvements of the safety systems, the plant's operating systems have been simplified based on operating experience. The design's safety concept, simplified operating systems and 48 months construction time yield favourable plant construction costs. The level of concept maturity required to begin offering the SWR 1000 on the power generation market is anticipated to be reached, as planned in the year

  13. R and D for back-end options for irradiated research reactor fuel in Germany

    International Nuclear Information System (INIS)

    Bruecher, H.; Curtius, H.; Fachinger, J.

    2001-01-01

    Out of 11.5 t of irradiated fuel arising from German research reactors until the end of this decade, 3.9 t are intended to be returned to the USA, and 2.3 t are expected to be recycled for reuse of uranium. The remaining 5.3 t, as well as the fuel irradiated after the year 2010, will have to follow the domestic back-end option of extended dry interim storage in Castor-type casks, followed by disposal in a deep geological repository. R and D is going on in the Research Centre Juelich to investigate the long-term behaviour of U-Al based fuel in a salt repository. First results from leaching experiments show I) a fast dissolution of the fuel with mobilization of its radionuclide inventory, and 2) the following formation of amorphous Al-Mg-hydroxide phases. Long-lived actinides from the fuel were shown to be fixed in these phases and hence immobilized. Future R and D will be to investigate the nature and stability of these phases for long-term safety assessments. Investigations will have to be extended to cover alternative disposal sites (granite clay) as well as different (e.g. silicon based) fuels. (author)

  14. Nuclear energy research in Germany 2008. Research centers and universities

    International Nuclear Information System (INIS)

    Tromm, Walter

    2009-01-01

    This summary report presents nuclear energy research at research centers and universities in Germany in 2008. Activities are explained on the basis of examples of research projects and a description of the situation of research and teaching in general. Participants are the - Karlsruhe Research Center, - Juelich Research Center (FZJ), - Dresden-Rossendorf Research Center (FZD), - Verein fuer Kernverfahrenstechnik und Analytik Rossendorf e.V. (VKTA), - Technical University of Dresden, - University of Applied Sciences, Zittau/Goerlitz, - Institute for Nuclear Energy and Energy Systems (IKE) at the University of Stuttgart, - Reactor Simulation and Reactor Safety Working Group at the Bochum Ruhr University. (orig.)

  15. Design and application of the HTR-100 industrial nuclear power plant

    International Nuclear Information System (INIS)

    Brandes, S.; Kohl, W.

    1988-01-01

    The small HTR-100 high temperature reactor combines the reactor concept of the AVR reactor, which has been proven for 20 years, with the latest component technology of the THTR power plant which has been in operation since 1985. The nuclear heat supply system is conceived so as to be applicable for the generation of electric power, district heat and process steam according to the customer's demand. The HTR-100 reactor has a thermal power of 258 MW and offers steam parameters of 190 bar/530 0 C. To cover a higher power demand HTR-100 reactors can be combined forming a larger power plant. Economic analyses have shown competitiveness with fossil power plants. (orig.)

  16. Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

    International Nuclear Information System (INIS)

    Boer, Brian

    2007-01-01

    The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

  17. High heat flux components with Be armour before and after neutron irradiation

    International Nuclear Information System (INIS)

    Lodato, A.; Derz, H.; Duwe, R.; Linke, J.; Roedig, M.

    2000-01-01

    Beryllium/copper mock-ups produced by different joining techniques have been tested in the electron beam facility JUDITH (Juelich Divertor Test Facility in Hot Cells) at Forschungszentrum Juelich. The experiments described in this paper represent the conclusive part of a test program started in 1994. The properties of non-irradiated Be/Cu joints have been characterised in a previous test campaign. Post-irradiation tests are now being carried out to investigate the neutron damage on the joints. The neutron irradiation on selected mock-ups has been carried out in the High Flux Reactor (HFR) at Petten (The Netherlands). Parametric finite element thermal analyses have been carried out to establish the allowable heat flux value to be applied during the tests. Screening tests up to power densities of ∼7 MW/m 2 and thermal fatigue tests up to 1000 cycles have been performed. None of these mock-ups showed any indication of failure. Post-mortem analyses (metallography, SEM) have also been conducted

  18. Irradiation of pressurized water reactor fuel rods in the Forschungsreaktor Juelich 2

    International Nuclear Information System (INIS)

    Gaertner, M.

    1978-10-01

    Test fuel rods have been irradiated in FRJ-2 to study the interaction between fuel and cladding as well as hydride orientation stability in the prehydrided cladding. The fuel rods achieved burn-ups of 3.500 to 10.000 MWd/tU at surface temperatures of 333 0 C and power levels up to 620 W/cm. (orig.) [de

  19. From nuclear research to multidisciplinary research

    International Nuclear Information System (INIS)

    Theenhaus, R.; Baurmann, K.W.

    1996-01-01

    Forty years ago, the North Rhine-Westphalian State Government founded the then Juelich Nuclear Research Center. After a growth period of the reactor engineering program until 1980, claiming a share of 42% of R and D resources, that share declined continuously to a present level of 8%. This development is an expression of the activities successfully completed in the past, of progress achieved in industrial reactor development, but also of the fact that the high temperature reactor, which had been run successfully for twenty years, failed as a technical scale THTR-300 version. The Center has reorientated its line of research in a process of structural reshuffle beginning some fifteen years ago and still going on. Information technology, materials research, life sciences, environmental research, and energy technology have become main activities of equal weight. Activities specific to nuclear reactors have been incorporated in this new line of work as nuclear safety research and work on safe repository storage. (orig.) [de

  20. Aortic valve replacement with or without coronary artery bypass graft surgery: the risk of surgery in patients > or =80 years old.

    Science.gov (United States)

    Maslow, Andrew; Casey, Paula; Poppas, Athena; Schwartz, Carl; Singh, Arun

    2010-02-01

    The purpose of this study was to evaluate the outcomes for elderly (> or =80 years) patients undergoing aortic valve replacement (AVR) with or without coronary artery bypass graft surgery (AVR/CABG). The authors hypothesized that the mortalities of AVR and AVR/CABG are lower than that predicted by published risk scores. A retrospective analysis of data from a single-hospital database. Single tertiary care, private practice. Consecutive patients undergoing AVR or AVR/CABG. Two hundred sixty-one elderly (> or =80 years) patients undergoing isolated AVR (145) or AVR/CABG (116) were evaluated. The majority (94.6%) underwent AVR for aortic valve stenosis. Outcomes were recorded and compared between the 2 surgical procedures with predicted mortalities based on published risk assessment scoring systems. The overall short-term mortality for the elderly group was 6.1% (AVR 5.5% and AVR/CABG 6.9%). The median long-term survival was 6.8 years. There were no significant differences in either morbidity or mortality between the AVR and AVR/CABG groups. Although predicted mortalities were similar for each surgical procedure, they overestimated observed outcome by up to 4-fold. Short- and long-term mortality was low for this group of elderly patients undergoing AVR or AVR/CABG and not significantly different between the 2 surgical groups. Predicted outcomes were worse than that observed, consistent with the hypothesis, and supportive of a more aggressive surgical treatment for aortic valve disease in the elderly patient. Copyright 2010 Elsevier Inc. All rights reserved.

  1. Gasification of coal using nuclear process heat. Chapter D

    International Nuclear Information System (INIS)

    Schilling, H.-D.; Bonn, B.; Krauss, U.

    1979-01-01

    In the light of the high price of coal and the enormous advances made recently in nuclear engineering, the possibility of using heat from high-temperature nuclear reactors for gasification processes was discussed as early as the 1960s. The advantages of this technology are summarized. A joint programme of development work is described, in which the Nuclear Research Centre at Juelich is aiming to develop a high-temperature reactor which will supply process heat at as high a temperature as possible, while other organizations are working on the hydrogasification of lignites and hard coals, and steam gasification. Experiments are at present being carried out on a semi-technical scale, and no operational data for large-scale plants are available as yet. (author)

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

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

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

  3. The ADAM and EVE project: Heat transfer at ambient temperature

    International Nuclear Information System (INIS)

    Boltendahl, U.; Harth, R.

    1980-01-01

    In the nuclear research plant at Juelich a new heating system is at present being developed as part of the Nuclear Long-distance Heating Project. Helium is heated up in a high-temperature reactor. The heat chemically converts a gas mixture in a reformer plant (EVE). The gases 'charged' with energy can be transported through tubes over any distance required at ambient temperatures. In a methanisation plant (ADAM) the gases react with one another, releasing the energy in the form of heat which can be used for heating air or water. (orig.) [de

  4. Advances in neutron scattering spectroscopy

    International Nuclear Information System (INIS)

    White, J.W.

    1977-01-01

    Some aspects of the application of neutron scattering to problems in polymer science, surface chemistry, and adsorption phenomena, as well as molecular biology, are reviewed. In all these areas, very significant work has been carried out using the medium flux reactors at Harwell, Juelich and Risoe, even without the use of advanced multidetector techniques or of a neutron cold source. A general tendency can also be distinguished in that, for each of these new fields, a distinct preference for colder neutrons rather than thermal neutron beams can be seen. (author)

  5. HET/JUPITER project assessment report

    International Nuclear Information System (INIS)

    Baxter, B.J.; Harrington, F.E.; Kaiser, G.G.; Wolf, J.

    1979-05-01

    This report is an assessment of the United States' Hot Engineering Test (HET) and the Federal Republic of Germany's Juelich Pilot Plant Thorium Element Reprocessing (JUPITER) Projects. The assessment was conducted with a view to developing mutually supportive roles in the achievement of hot engineering test objectives. Conclusions of the assessment are positive and identify several technical areas with potential for US/FRG cooperation. Recommendations presented in this report support a cost-effective US/FRG program to jointly develop high temperature gas-cooled reactor fuel recycle technology. (orig.) [de

  6. Forschungszentrum Juelich. Annual report 2012; Forschungszentrum Juelich. Jahresbericht 2012

    Energy Technology Data Exchange (ETDEWEB)

    Frick, Frank; Roegener, Wiebke

    2013-07-15

    This annual report is structured as follows: 1 Highlight Energy Research (Next-generation batteries. Innovative material for the fuel cell. Smart material for solar cells. Store from midnight - Study on electromobility. Fuels from renewable electricity, carbon dioxide and water.). 2. Knowledge management (Create knowledge, pass knowledge, share knowledge and apply knowledge), and 3. Appendix (finance, boards and committees, organizational chart). [German] Dieser Jahresbericht ist wie folgt strukturiert: 1. Highlight Energieforschung (Batterien der naechsten Generation. Innovativer Werkstoff fuer die Brennstoffzelle. Smartes Material fuer Solarzellen. Laden ab Mitternacht - Studie zur Elektromobilitaet. Kraftstoffe aus Oekostrom, Kohlendioxid und Wasser). 2. Wissensmanagement (Wissen schaffen, Wissen weitergeben, Wissen teilen und Wissen anwenden) und 3. Anhang (Finanzen, Organe und Gremien, Organigramm).

  7. Forschungszentrum Juelich. Annual report 2011; Forschungszentrum Juelich. Jahresbericht 2011

    Energy Technology Data Exchange (ETDEWEB)

    Frick, Frank; Roegener, Wiebke

    2012-07-15

    The annual report presents ten primary scientific reports selected for information of the general reader, representing the research priorities, a survey of the research and development programmes, a list of research institutes, cooperation agreements for technology transfer, facts and figures showing the organizational structure, personnel employed, financing aspects, and aspects of the service departments.

  8. Forschungszentrum Juelich. Annual report 2009; Forschungszentrum Juelich. Jahresbericht 2009

    Energy Technology Data Exchange (ETDEWEB)

    Frick, Frank; Roegener, Wiebke; Stahl-Busse, Brigitte

    2010-07-15

    The annual report presents ten primary scientific reports selected for information of the general reader, representing the research priorities, a survey of the research and development programmes, a list of research institutes, cooperation agreements for technology transfer, facts and figures showing the organizational structure, personnel employed, financing aspects, and aspects of the service departments.

  9. Forschungszentrum Juelich. Annual report 2008; Forschungszentrum Juelich. Jahresbericht 2008

    Energy Technology Data Exchange (ETDEWEB)

    Frick, Frank; Roegener, Wiebke

    2009-07-15

    The following topics are dealt with: The precise lattice QCD mass calculation of protons and neutrons by means of the JUGENE supercomputer, the early diagnosis of morbus Alzheimer, the fabrication of vertebra-column implants consisting of porus titanium, software for the improvement of the spatial resolution in electron microscopy by means of aberration corrections. (HSI)

  10. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

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

  12. Air ingress and graphite burning in HTRs: A survey on analytical examinations performed with the code React/Thermix

    International Nuclear Information System (INIS)

    Moormann, R.

    1995-05-01

    Analyses on air ingress in the pebble bed reactors PNP-500, THTR-300, HTR-Modul, AVR-II and AVR process heat plant are outlined; in addition, some results for the VHTR with block type fuel are given. Air ingress requires primary circuit depressurization and large leak(s) to reactor buildings and environment and belongs therefore to highly hypothetical events in the sense of classical safety analysis. One accident class examined is air ingress with forced flow by emergency cooling: For this case, the range of mass flow/air content in cooling gas has been evaluated, in which safe core cool down is possible resp. long term core burning occurs; for highest available emergency cooling flow, a safe cool down of the THTR-300, which has no reactor building, is possible for up to 20 vol-% of air in the cooling gas, wheras low flow allows only for about 5 vol-%. If the amount of available air is restricted to the content of a reactor building, as is examined for the PNP-500, relevant consequences have not to be expected; this remains also true for forced convection flow, if burning of CO, formed by graphite oxidation, within the building is considered. For the second accident class examined, air ingress with natural convection flow by chimney draught as studied for the HTR-Modul and some other concepts, the time span until significant fission product release begins has been determined; in case, that the bottom reflector is hot at accident start (> 600 C) and therefore consumes most of the ingressing oxygen, this time span is at least several hours and leak tightening counter measures may be possible. (orig./HP)

  13. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

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

  14. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

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

  15. Nuclear process heat at high temperature: Application, realization and development programme

    International Nuclear Information System (INIS)

    Sammeck, K.H.; Fischer, R.

    1976-01-01

    Studies in the Federal Republic of Germany (FRG), the USA and the United Kingdom have shown that high-temperature helium energy from an HTR can advantageously be utilized for coal gasification and other fossil fuel conversion processes, and that a substantial demand for substitute natural gas (SNG) can be expected in the future. These results are based on plant design studies, economic assessments and basic development efforts in the field of coal gasification with nuclear heat, which in the FRG were carried out by Arbeitsgemeinschaft Nukleare Prozesswaerme (ANP)-members, HRB and KFA Juelich. Nuclear process plants are based on different gasification processes, resulting in different concepts of the nuclear heat system. In the case of hydro-gasification it is expected that steam reformers, arranged within the primary circuit of the reactor, will be heated directly by the primary helium. In the case of steam gasification, the high-temperature energy must be transferred to the gasification process via an intermediate circuit which is coupled to a gasifier outside the containment. In both cases the design of the nuclear reactor resembles an HTR for electricity generation. The main objectives of the development of nuclear process heat are to increase the helium outlet temperature of the reactor up to 950 0 C, to develop metallic alloys for high-temperature components such as heat exchangers, to design and construct a hot-gas duct, a steam reformer and a helium-helium heat exchanger and to develop the gasification processes. The nuclear safety regulations and the interface problems between the reactor, the process plant and the electricity generating plant have to be considered thoroughly. The Arbeitsgemeinschaft Nukleare Prozesswaerme and HRB started a development programme, in close collaboration with KFA Juelich, which will lead to the construction of a prototype plant for coal gasification with nuclear heat within 5 to 5 1/2 years. A survey of the main objectives

  16. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

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

  17. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  18. Interdisciplinary study of the influence on effectiveness of catalytic hydrogen recombiners of operating conditions in the reactor containment

    International Nuclear Information System (INIS)

    Kelm, S.; Reinecke, E.A.; Schoppe, L.; Dornseiffer, J.; Leistner, F.; Juehe, S.

    2008-01-01

    At the Emsland nuclear power station, a total of 58 autocatalytic hydrogen recombiners were backfitted in 1999 as an additional measure of risk reduction in connection with major hydrogen releases after events going beyond the design basis. Annual in-service inspections after 2002 revealed that some of the catalyst sheets developed startup delays and marked evolutions of smoke and smell. Recombiners not meeting the inspection criterion were completely regenerated as a measure of precaution. A preventive study was conducted jointly with institutes of the Juelich Research Center and the Aachen Technical University to analyze the composition of the deposits, which was then compared with the chemical characteristics of potential sources in the reactor containment. At the same time, the influence on effectiveness of the catalyst sheets was examined. On the basis of a random evaluation of the in-service inspection logs of the past few years, representative samples were taken whose startup behavior and operating characteristics were studied in a test rig alongside chemical analyses so as to allow a correlation to be established between the analytical findings and the catalytic activity of the samples. The findings made allowed internal sources of the catalyst deposits to be excluded. The impurities are introduced with the outside air. As a consequence, the air ducts in the vicinity of the respective recombiners were inspected and optimization steps were taken in connection with in-service inspections and regeneration procedures. (orig.)

  19. Nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

  20. Genetic analysis of environmental strains of the plant pathogen Phytophthora capsici reveals heterogeneous repertoire of effectors and possible effector evolution via genomic island.

    Science.gov (United States)

    Iribarren, María Josefina; Pascuan, Cecilia; Soto, Gabriela; Ayub, Nicolás Daniel

    2015-11-01

    Phytophthora capsici is a virulent oomycete pathogen of many vegetable crops. Recently, it has been demonstrated that the recognition of the RXLR effector AVR3a1 of P. capsici (PcAVR3a1) triggers a hypersensitive response and plays a critical role in mediating non-host resistance. Here, we analyzed the occurrence of PcAVR3a1 in 57 isolates of P. capsici derived from globe squash, eggplant, tomato and bell pepper cocultivated in a small geographical area. The occurrence of PcAVR3a1 in environmental strains of P. capsici was confirmed by PCR in only 21 of these pathogen isolates. To understand the presence-absence pattern of PcAVR3a1 in environmental strains, the flanking region of this gene was sequenced. PcAVR3a1 was found within a genetic element that we named PcAVR3a1-GI (PcAVR3a1 genomic island). PcAVR3a1-GI was flanked by a 22-bp direct repeat, which is related to its site-specific recombination site. In addition to the PcAVR3a1 gene, PcAVR3a1-GI also encoded a phage integrase probably associated with the excision and integration of this mobile element. Exposure to plant induced the presence of an episomal circular intermediate of PcAVR3a1-GI, indicating that this mobile element is functional. Collectively, these findings provide evidence of PcAVR3a1 evolution via mobile elements in environmental strains of Phytophthora. © FEMS 2015. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

  1. Novel Mutations Detected in Avirulence Genes Overcoming Tomato Cf Resistance Genes in Isolates of a Japanese Population of Cladosporium fulvum.

    Directory of Open Access Journals (Sweden)

    Yuichiro Iida

    Full Text Available Leaf mold of tomato is caused by the biotrophic fungus Cladosporium fulvum which complies with the gene-for-gene system. The disease was first reported in Japan in the 1920s and has since been frequently observed. Initially only race 0 isolates were reported, but since the consecutive introduction of resistance genes Cf-2, Cf-4, Cf-5 and Cf-9 new races have evolved. Here we first determined the virulence spectrum of 133 C. fulvum isolates collected from 22 prefectures in Japan, and subsequently sequenced the avirulence (Avr genes Avr2, Avr4, Avr4E, Avr5 and Avr9 to determine the molecular basis of overcoming Cf genes. Twelve races of C. fulvum with a different virulence spectrum were identified, of which races 9, 2.9, 4.9, 4.5.9 and 4.9.11 occur only in Japan. The Avr genes in many of these races contain unique mutations not observed in races identified elsewhere in the world including (i frameshift mutations and (ii transposon insertions in Avr2, (iii point mutations in Avr4 and Avr4E, and (iv deletions of Avr4E, Avr5 and Avr9. New races have developed by selection pressure imposed by consecutive introductions of Cf-2, Cf-4, Cf-5 and Cf-9 genes in commercially grown tomato cultivars. Our study shows that molecular variations to adapt to different Cf genes in an isolated C. fulvum population in Japan are novel but overall follow similar patterns as those observed in populations from other parts of the world. Implications for breeding of more durable C. fulvum resistant varieties are discussed.

  2. Genetic mapping of 14 avirulence genes in an EU-B04 × 1639 progeny of Venturia inaequalis.

    Science.gov (United States)

    Broggini, Giovanni A L; Bus, Vincent G M; Parravicini, Gabriella; Kumar, Satish; Groenwold, Remmelt; Gessler, Cesare

    2011-02-01

    Durable resistance to apple scab (Venturia inaequalis (Cke) Wint; anamorph Spilocaea pomi Fries) is one of the major goals of apple (Malus) breeding programs. Since current scab resistance breeding is heavily reliant on genes with gene-for-gene relationships, a good understanding of the genetic basis of host-pathogen interactions needs to be developed for this strategy to be successful. While the genomic organization of apple scab resistance genes has been studied extensively, little is known about the avirulence genes in the pathogen. The progeny of a cross of European V. inaequalis race (1) isolate EU-B04 and race (1,2,8,9) isolate 1639 was used to generate a genetic map based on microsatellite and AFLP markers, and investigated for inheritance of avirulence traits on 20 Malus accessions representing 17 scab resistance genes. The accessions comprised scab differential hosts (0), (1), (2), (8), and (9), and hosts carrying known as well as not previously reported secondary resistance genes, including some identified in crosses that have resistant accessions 'Geneva', 'Dolgo', Malus baccata jackii, M. micromalus, or 'Antonovka' in their pedigree. The latter genes appear to be narrow spectrum genes that showed gene-for-gene relationships as a segregation ratio of Avr:avr=1:1 was observed on 12 accessions, while a ratio of 3:1 was observed on five accessions and a ratio of 7:1 on one host. All progenies were shown to be pathogenic, as all of them were able to infect hosts (0) and (1). A genetic map consisting of 15 major linkage groups (LGs) and spanning 972cM was generated with the aid of 156 markers. The map position of 12 avirulence traits was determined: eight avirulence genes mapped into two separate clusters (1: AvrVdg2, AvrVv1, AvrVu1, AvrVrjrd; and 2: AvrVu2, AvrVh3.2, AvrVs1, AvrVu4), while four avirulence genes (AvrRvi8, AvrVv2, AvrVt57 and AvrVsv) mapped to different LGs. AvrRvi2 and AvrRvi9 also are genetically linked, but showed an interaction with Avr

  3. Papers about coated particles, graphitic and metallic materials for progressive high-temperature reactors at the Reactor Meeting 1978

    International Nuclear Information System (INIS)

    Rottmann, J.

    1978-09-01

    In the contributions, questions on the development, the radiation and the high-temperature behaviour and the characterization of fuel element particles are treated. Furthermore the resistance and radiation behaviour of graphitic materials are discussed. Finally, questions on the choice of high-temperature alloys for nuclear process heat facilities are discussed and the testing-equipment of the Nuclear Research Centre as well as first results of the long-time experiments are presented. The work was performed within the frame of the projects 'HTR-Fuel Element Cycle' and 'Prototype Nuclear Process Heat', which are sponsored by the Federal Ministry of Research and Technology of the Federal Republic of Germany and of the state of North-Rhine-Westfalia. Partner firms, who participate in the two projects are Gelsenberg AG, Gesellschaft fuer Hochtemperaturreaktor-Technik mbH, Hochtemperaturreaktor-Brennelement GmbH, Hochtemperatur-Reaktorbau GmbH, Kernforschungsanlage Juelich GmbH, NUKEM GmbH, SIGRI Elektrographit GmbH/Ringsdorff-Werke GmbH, Bergbauforschung GmbH und Rheinische Braunkohlenwerke AG. (orig./UA) [de

  4. Study on the production mechanism of Co-60 in the primary loop of HTR-10

    International Nuclear Information System (INIS)

    Wang Shouang; Xie Feng; Li Hong; Cao Jianzhu; Li Fu; Wei Liqiang

    2015-01-01

    Co-60 is an activated metallic erosion product, which is very important for waste management and decommissioning work of pressurized water reactor (PWR) power plants. Recent measurement on the samples from the primary loop of HTR-10 indicates the existence of Co-60. In current paper, the preliminary experimental results in HTR-10 will be introduced, and the production mechanism of Co-60 in the pebble bed high temperature gas-cooled reactors will be summarized and compared with that in PWRs and Germany High Temperature Nuclear Reactor (AVR). The further experiments with decomposing the post-irradiation graphite spheres of HTR-10 are put forward, which will promote the further study to testify the production sources of Co-60 and be of great significance in the waste minimization and the decommissioning work of HTR-10. (author)

  5. Reoperative aortic valve replacement in the octogenarians-minimally invasive technique in the era of transcatheter valve replacement

    NARCIS (Netherlands)

    Kaneko, Tsuyoshi; Loberman, Dan; Gosev, Igor; Rassam, Fadi; McGurk, Siobhan; Leacche, Marzia; Cohn, Lawrence

    2014-01-01

    Objective: Reoperative aortic valve replacement (re-AVR) in octogenarians is considered high risk and therefore might be indicated for transcatheter AVR. The minimally invasive technique for re-AVR limits dissection and might benefit this patient population. We report the outcomes of re-AVR in

  6. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  7. Is rapid development of nuclear power purposeful

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    The questions of the development of nuclear energy are discussed with regard to the efficacy of investments. The results are given of studies carried out at the nuclear research institute in Juelich in the FRG. At the estimated 25 years' service-life of the reactor and 0.2% uranium concentration in ore the following results were obtained: Total energy consumption for the construction and operation of a light water reactor power plant makes up 4.6% of the total power production, and in high-temperature reactors it amounts to 3.5%, both with uranium enrichment by the diffusion process. In uranium enrichment by centrifugal technology, consumption drops to 1.25% for LWRms and to 0.9% for high-temperature reactors, in fast breeder reactors it makes up only 0.8% of the total power production of the nuclear power plant. The period during which a nuclear power plant produces the amount of power consumed in construction and operation is 1.2 to 2.5 months which makes it less costly and more economical than any power plant burning coal minus the negative environmental impacts of such power plants.

  8. Impact of prosthesis-patient mismatch on early and late mortality after aortic valve replacement

    NARCIS (Netherlands)

    Koene, Bart M.; Hamad, Mohamed A. Soliman; Bouma, Wobbe; Mariani, Massimo A.; Peels, Kathinka C.; van Dantzig, Jan-Melle; van Straten, Albert H.

    2013-01-01

    Background: The influence of prosthesis-patient mismatch (PPM) on survival after aortic valve replacement (AVR) remains controversial. In this study, we sought to determine the effect of PPM on early (30 days) after AVR or AVR combined with coronary artery bypass grafting (AVR with CABG). Methods:

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

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

  10. The CAREM reactor and present currents in reactor design

    International Nuclear Information System (INIS)

    Ordonez, J.P.

    1990-01-01

    INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

  11. Characterization of graphite dust produced by pneumatic lift

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Ke [Guangdong Provincial Key Laboratory of Thermal Management Engineering and Materials, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055, Guangdong (China); Peng, Wei; Liu, Bing [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Kang, Feiyu [Guangdong Provincial Key Laboratory of Thermal Management Engineering and Materials, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055, Guangdong (China); Yang, Xiaoyong; Li, Weihua [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Educations, Tsinghua University, Beijing 100084 (China)

    2016-08-15

    Highlights: • Generation of graphite dust by pneumatic lift. • Determination of morphology and particle size distribution of graphite dust. • The size of graphite dust in this study is compared to AVR and THTR-300 results. • Graphite dust originates from both filler and binder of the matrix graphite. - Abstract: Graphite dust is an important safety concern of high-temperature gas-cooled reactor (HTR). The graphite dust could adsorb fission products, and the radioactive dust is transported by the coolant gas and deposited on the surface of the primary loop. The simulation of coagulation, aggregation, deposition, and resuspension behavior of graphite dust requires parameters such as particle size distribution and particle shape, but currently very limited data on graphite dust is available. The only data we have are from AVR and THTR-300, however, the AVR result is likely to be prejudiced by the oil ingress. In pebble-bed HTR, graphite dust is generally produced by mechanical abrasion, in particular, by the abrasion of graphite pebbles in the lifting pipe of the fuel handling system. Here we demonstrate the generation and characterization of graphite dust that were produced by pneumatic lift. This graphite dust could substitute the real dust in HTR for characterization. The dust, exhibiting a lamellar morphology, showed a number-weighted average particle size of 2.38 μm and a volume-weighted average size of 14.62 μm. These two sizes were larger than the AVR and THTR results. The discrepancy is possibly due to the irradiation effect and prejudice caused by the oil ingress accident. It is also confirmed by the Raman spectrum that both the filler particle and binder contribute to the dust generation.

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

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

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

  13. Reactor science and technology: operation and control of reactors

    International Nuclear Information System (INIS)

    Qiu Junlong

    1994-01-01

    This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

  14. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

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

  15. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

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

  16. Ross Versus Non-Ross Aortic Valve Replacement in Children: A 22-Year Single Institution Comparison of Outcomes.

    Science.gov (United States)

    Brown, John W; Patel, Parth M; Ivy Lin, Jiuann-Huey; Habib, Asma S; Rodefeld, Mark D; Turrentine, Mark W

    2016-05-01

    The Ross aortic valve replacement (AVR) has been the AVR of choice for children at our center since 1993. Absence or inadequate quality of the pulmonary valve or, less commonly, family or surgeon preference caused us to select an alternative AVR prosthesis for some children. This review compares the outcomes of 42 children who received a non-Ross AVR with 115 children undergoing Ross root replacement at our institution during the most recent 22 years. A retrospective chart review of the 42 pediatric non-Ross AVRs was compared with 115 Ross AVRs. The mean age at AVR was 11.0 ± 6.5 years (range, 1 month to 18 years) for the non-Ross and 11.4 ± 5.5 years (range, 6 weeks to 18 years) for the Ross groups. Follow-up was 7.8 ± 6.2 years for the Ross group and 8.7 ± 6.5 years for the non-Ross group. The Ross AVR technique was modified in 2000; these modifications were applied to 72 patients (63%) of the total Ross AVR group. All-cause early and late mortality of the non-Ross AVR group was 17% compared with 4% for the Ross cohort (p = 0.017). The actuarial survival rate at 20 years for the non-Ross AVR group was 81% compared with 94% for the Ross group (p = 0.018). Reintervention rate was 29% (12/38) in the non-Ross group and 28% (32/115) in the entire Ross group (60% [26/43] before 2000, and 8% [6/72] after 2000). Thromboemboli, bleeding, and endocarditis occurred in 2.4%, 1.4%, and 10% in the non-Ross group compared with 0.9%, 2.6%, and 1.7% in the Ross AVR group (p = 0.46, 0.01, and 0.02), respectively. Patients in the non-Ross group had a significantly higher recurrent or persistent aortic valve gradient (>20 mm Hg) than did patients in the Ross group at most recent follow-up (p Ross AVR than with other types of AVR prostheses. The Ross AVR remains the treatment of choice for children who have an adequate pulmonary valve. Reoperation for Ross root dilatation, regurgitation, or both (only 6% since our modifications in 2000) has markedly reduced the main drawbacks

  17. Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

    International Nuclear Information System (INIS)

    Hirayama, Shoichi; Kawabe, Takaya

    1995-01-01

    The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

  18. Reactor physics aspects of CANDU reactors

    International Nuclear Information System (INIS)

    Critoph, E.

    1980-01-01

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

  19. Determination of effective cross sections and production rates for tritium in the irradiation experiment TRIDEX

    International Nuclear Information System (INIS)

    Weise, L.

    1986-04-01

    In the framework of the development of a fusion reactor blanket the irradiation experiment TRIDEX (Tritium Recovery Irradiation DIDO Experiment) takes place at the Juelich Research Reactor FRJ-2 (DIDO). For this equipment the required neutronic calculations have been performed. The aim was the determination of the neutron spectrum and several therefrom derived integral parameters for the irradiation positions in interest. From the calculated effective cross sections for the formation of Tritium resulting from irradiated Lithium samples on one hand and from measured neutron flux densities on the other hand, all needed quantities of the Tritium production could be determined. The calculation of the neutron spectrum has been performed in a two-dimensional x-y-geometry. The neutron flux densities have been gained by gamma-spectrometric measurement of the activities in irradiated activation samples. (orig.) [de

  20. Development of a manufacturing process of (Th,U)O2 sintered pellets to be used as nuclear fuel

    International Nuclear Information System (INIS)

    Neto Ferreira, R.A.; Santos, A.M. dos; Lameiras, F.S.; Cardoso, P.E.

    1989-01-01

    The R and D result of a reliable manufacturing process of sintered (Th,U)O 2 pellets meeting the operational requirements of pressurized light water nuclear reactors is presented. Available technologies were used as much as possible. The R and D effort was directed to perform the required adaptations. The gel precipitation process was adapted successfully to the specific requirements of direct pressing and sintering. This was done mainly by adjusting the composition of the feed solution. The direct pressing and sintering parameters could be kept almost unchanged in relation to the manufacturing of UO 2 pellets. The design criteria of the (Th,U)O 2 nuclear fuel for pressurized light water reactors were identified and settled in the specification for this fuel. This R and D work was made jointly with the Kernforschungsanlage - Juelich, NUKEM and SIEMENS, Group KWU [pt

  1. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

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

  2. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

  3. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

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

  4. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    Kouts, H.

    1986-01-01

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

  5. Thrombocytopenia in Moderate- to High-Risk Sutureless Aortic Valve Replacement

    Directory of Open Access Journals (Sweden)

    Puwadon Thitivaraporn

    2018-06-01

    Full Text Available Background : This study aimed to compare preliminary data on the outcomes of sutureless aortic valve replacement (SU-AVR with those of aortic valve replacement (AVR. Methods : We conducted a retrospective study of SU-AVR in moderate- to high-risk patients from 2013 to 2016. Matching was performed at a 1:1 ratio using the Society of Thoracic Surgeons predicted risk of mortality score with sex and age. The primary outcome was 30-day mortality. The secondary outcomes were operative outcomes and complications. Results : A total of 277 patients were studied. Ten patients (50% males; median age, 81.5 years underwent SU-AVR. Postoperative echocardiography showed impressive outcomes in the SU-AVR group. The 30-day mortality was 10% in both groups. In our study, the patients in the SU-AVR group developed postoperative thrombocytopenia. Platelet counts decreased from 225X103 /μL preoperatively to 94.5, 54.5, and 50.1X10 3/μL on postoperative days 1, 2, and 3, respectively, showing significant differences compared with the AVR group (p=0.04, p=0.16, and p=0.20, respectively. The median amount of platelet transfusion was higher in the AVR group (12.5 vs. 0 units, p=0.052. Conclusion : There was no difference in the 30-day mortality of moderate- to high-risk patients depending on whether they underwent SU-AVR or AVR. Although SU-AVR is associated with favorable cardiopulmonary bypass and cross-clamp times, it may be associated with postoperative thrombocytopenia.

  6. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  7. 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.

  8. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  9. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

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

  10. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  11. Research reactor standards and their impact on the TRIGA reactor community

    International Nuclear Information System (INIS)

    Richards, W.J.

    1980-01-01

    The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

  12. 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

  13. Improved nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

  14. Reactor as furnace and reactor as lamp

    International Nuclear Information System (INIS)

    Goldanskii, V.I.

    1992-01-01

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

  15. Development of Reactor Console Simulator for PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Nufarhana Ayuni Joha; Mohd Sabri Minhat

    2012-01-01

    The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)

  16. 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

  17. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  18. Control of reactor coolant flow path during reactor decay heat removal

    International Nuclear Information System (INIS)

    Hunsbedt, A.N.

    1988-01-01

    This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

  19. The research reactors their contribution to the reactors physics

    International Nuclear Information System (INIS)

    Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

    2000-01-01

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  20. Reactor noise analysis of experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

    As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

  1. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

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

  2. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  3. Reactor

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  4. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

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

  5. FBR type reactor

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  6. RB Research nuclear reactor RB reactor, Annual report for 2000

    International Nuclear Information System (INIS)

    Milosevic, M.

    2000-12-01

    Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

  7. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

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

  8. 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.

  9. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

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

  10. Co-evolutionary interactions between host resistance and pathogen avirulence genes in rice-Magnaporthe oryzae pathosystem.

    Science.gov (United States)

    Singh, Pankaj Kumar; Ray, Soham; Thakur, Shallu; Rathour, Rajeev; Sharma, Vinay; Sharma, Tilak Raj

    2018-06-01

    Rice and Magnaporthe oryzae constitutes an ideal pathosystem for studying host-pathogen interaction in cereals crops. There are two alternative hypotheses, viz. Arms race and Trench warfare, which explain the co-evolutionary dynamics of hosts and pathogens which are under continuous confrontation. Arms race proposes that both R- and Avr- genes of host and pathogen, respectively, undergo positive selection. Alternatively, trench warfare suggests that either R- or Avr- gene in the pathosystem is under balanced selection intending to stabilize the genetic advantage gained over the opposition. Here, we made an attempt to test the above-stated hypotheses in rice-M. oryzae pathosystem at loci of three R-Avr gene pairs, Piz-t-AvrPiz-t, Pi54-AvrPi54 and Pita-AvrPita using allele mining approach. Allele mining is an efficient way to capture allelic variants existing in the population and to study the selective forces imposed on the variants during evolution. Results of nucleotide diversity, neutrality statistics and phylogenetic analyses reveal that Piz-t, Pi54 and AvrPita are diversified and under positive selection at their corresponding loci, while their counterparts, AvrPiz-t, AvrPi54 and Pita are conserved and under balancing selection, in nature. These results imply that rice-M. oryzae populations are engaged in a trench warfare at least at the three R/Avr loci studied. It is a maiden attempt to study the co-evolution of three R-Avr gene pairs in this pathosystem. Knowledge gained from this study will help in understanding the evolutionary dynamics of host-pathogen interaction in a better way and will also aid in developing new durable blast resistant rice varieties in future. Copyright © 2018 Elsevier Inc. All rights reserved.

  11. Aortic Valve Replacement and the Ross Operation in Children and Young Adults.

    Science.gov (United States)

    Sharabiani, Mansour T A; Dorobantu, Dan M; Mahani, Alireza S; Turner, Mark; Peter Tometzki, Andrew J; Angelini, Gianni D; Parry, Andrew J; Caputo, Massimo; Stoica, Serban C

    2016-06-21

    There are several options available for aortic valve replacement (AVR), with few comparative reports in the literature. The optimal choice for AVR in each age group is not clear. The study sought to report and compare outcomes after AVR in the young using data from a national database. AVR procedures were compared after advanced matching, both in pairs and in a 3-way manner, using a Bayesian dynamic survival model. A total of 1,501 patients who underwent AVR in the United Kingdom between 2000 and 2012 were included. Of these, 47.8% had a Ross procedure, 37.8% a mechanical AVR, 10.9% a bioprosthesis AVR, and 3.5% a homograft AVR, with Ross patients being significantly younger when compared to the other groups. Overall survival at 12 years was 94.6%. In children, the Ross procedure had a 12.7% higher event-free probability (death or any reintervention) at 10 years when compared to mechanical AVR (p = 0.05). We also compared all procedures except the homograft in a matched population of young adults, where the bioprosthesis had the lowest event-free probability of 78.8%, followed by comparable results in mechanical AVR and Ross, with 86.3% and 89.6%, respectively. Younger age was associated with mortality and pulmonary reintervention in the Ross group and with aortic reintervention in the mechanical AVR. Of all 3 options, only the patients undergoing the Ross procedure approached the survival of the general population. AVR in the young achieves good results, with the Ross being overall better suited for this age group, especially in children. Although freedom from aortic valve reintervention is superior after the Ross procedure, the need for homograft reinterventions is an issue to take into account. All methods have advantages and limitations, with reinterventions being an issue in the long term for all, more crucially in smaller children. Copyright © 2016 American College of Cardiology Foundation. Published by Elsevier Inc. All rights reserved.

  12. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

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

  13. Association EURATOM-FZJ. Annual progress report 2011. SC-FZJ 88(12)/4.1.2

    International Nuclear Information System (INIS)

    2011-01-01

    emphasis on the topic ''plasma-wall interactions'' and the joint use of dedicated facilities in Juelich, Rijnhuizen (NL) and Mol (B) has been signed in 2010. Co-operations beyond Europe are supported by an IEA Implementing Agreement on ''Plasma-Wall Interaction in TEXTOR'' together with Japan, USA and Canada, which meanwhile also serves as a basis for the exchange of scientists to other devices than TEXTOR. In view of the limited lifetime of TEXTOR and new opportunities arising from linear plasma test facilities the parties have agreed upon a change of name and scope: ''Implementing Agreement on the Development and Research on Plasma Wall Interaction Facilities for Fusion Reactors''. This change will involve a number of existing and planned linear plasma devices. The start under the new name and scope is planned for 2013.

  14. Nuclear reactor, reactor core thereof, and device for constituting the reactor

    International Nuclear Information System (INIS)

    Takiyama, Masashi.

    1994-01-01

    A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

  15. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-08-01

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

  16. Optimal reactor strategy for commercializing fast breeder reactors

    International Nuclear Information System (INIS)

    Yamaji, Kenji; Nagano, Koji

    1988-01-01

    In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)

  17. Computerized reactor monitor and control for nuclear reactors

    International Nuclear Information System (INIS)

    Buerger, L.

    1982-01-01

    The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

  18. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

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

  19. Control of reactor coolant flow path during reactor decay heat removal

    Science.gov (United States)

    Hunsbedt, Anstein N.

    1988-01-01

    An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

  20. Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

    International Nuclear Information System (INIS)

    Hauptman, H.M.; Petro, J.N.; Jacobi, O.

    1995-04-01

    This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

  1. Reactor core for FBR type reactor

    International Nuclear Information System (INIS)

    Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

    1996-01-01

    In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

  2. Space-time reactor kinetics for heterogeneous reactor structure

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1969-11-15

    An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods.

  3. Allelic barley MLA immune receptors recognize sequence-unrelated avirulence effectors of the powdery mildew pathogen.

    Science.gov (United States)

    Lu, Xunli; Kracher, Barbara; Saur, Isabel M L; Bauer, Saskia; Ellwood, Simon R; Wise, Roger; Yaeno, Takashi; Maekawa, Takaki; Schulze-Lefert, Paul

    2016-10-18

    Disease-resistance genes encoding intracellular nucleotide-binding domain and leucine-rich repeat proteins (NLRs) are key components of the plant innate immune system and typically detect the presence of isolate-specific avirulence (AVR) effectors from pathogens. NLR genes define the fastest-evolving gene family of flowering plants and are often arranged in gene clusters containing multiple paralogs, contributing to copy number and allele-specific NLR variation within a host species. Barley mildew resistance locus a (Mla) has been subject to extensive functional diversification, resulting in allelic resistance specificities each recognizing a cognate, but largely unidentified, AVR a gene of the powdery mildew fungus, Blumeria graminis f. sp. hordei (Bgh). We applied a transcriptome-wide association study among 17 Bgh isolates containing different AVR a genes and identified AVR a1 and AVR a13 , encoding candidate-secreted effectors recognized by Mla1 and Mla13 alleles, respectively. Transient expression of the effector genes in barley leaves or protoplasts was sufficient to trigger Mla1 or Mla13 allele-specific cell death, a hallmark of NLR receptor-mediated immunity. AVR a1 and AVR a13 are phylogenetically unrelated, demonstrating that certain allelic MLA receptors evolved to recognize sequence-unrelated effectors. They are ancient effectors because corresponding loci are present in wheat powdery mildew. AVR A1 recognition by barley MLA1 is retained in transgenic Arabidopsis, indicating that AVR A1 directly binds MLA1 or that its recognition involves an evolutionarily conserved host target of AVR A1 Furthermore, analysis of transcriptome-wide sequence variation among the Bgh isolates provides evidence for Bgh population structure that is partially linked to geographic isolation.

  4. Nuclear Reactor RA Safety Report, Vol. 4, Reactor

    International Nuclear Information System (INIS)

    1986-11-01

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

  5. Strain Specific Factors Control Effector Gene Silencing in Phytophthora sojae.

    Directory of Open Access Journals (Sweden)

    Sirjana Devi Shrestha

    Full Text Available The Phytophthora sojae avirulence gene Avr3a encodes an effector that is capable of triggering immunity on soybean plants carrying the resistance gene Rps3a. P. sojae strains that express Avr3a are avirulent to Rps3a plants, while strains that do not are virulent. To study the inheritance of Avr3a expression and virulence towards Rps3a, genetic crosses and self-fertilizations were performed. A cross between P. sojae strains ACR10 X P7076 causes transgenerational gene silencing of Avr3a allele, and this effect is meiotically stable up to the F5 generation. However, test-crosses of F1 progeny (ACR10 X P7076 with strain P6497 result in the release of silencing of Avr3a. Expression of Avr3a in the progeny is variable and correlates with the phenotypic penetrance of the avirulence trait. The F1 progeny from a direct cross of P6497 X ACR10 segregate for inheritance for Avr3a expression, a result that could not be explained by parental imprinting or heterozygosity. Analysis of small RNA arising from the Avr3a gene sequence in the parental strains and hybrid progeny suggests that the presence of small RNA is necessary but not sufficient for gene silencing. Overall, we conclude that inheritance of the Avr3a gene silenced phenotype relies on factors that are variable among P. sojae strains.

  6. A bacterial E3 ubiquitin ligase targets a host protein kinase to disrupt plant immunity.

    Science.gov (United States)

    Rosebrock, Tracy R; Zeng, Lirong; Brady, Jennifer J; Abramovitch, Robert B; Xiao, Fangming; Martin, Gregory B

    2007-07-19

    Many bacterial pathogens of plants and animals use a type III secretion system to deliver diverse virulence-associated 'effector' proteins into the host cell. The mechanisms by which these effectors act are mostly unknown; however, they often promote disease by suppressing host immunity. One type III effector, AvrPtoB, expressed by the plant pathogen Pseudomonas syringae pv. tomato, has a carboxy-terminal domain that is an E3 ubiquitin ligase. Deletion of this domain allows an amino-terminal region of AvrPtoB (AvrPtoB(1-387)) to be detected by certain tomato varieties leading to immunity-associated programmed cell death. Here we show that a host kinase, Fen, physically interacts with AvrPtoB(1-387 )and is responsible for activating the plant immune response. The AvrPtoB E3 ligase specifically ubiquitinates Fen and promotes its degradation in a proteasome-dependent manner. This degradation leads to disease susceptibility in Fen-expressing tomato lines. Various wild species of tomato were found to exhibit immunity in response to AvrPtoB(1-387 )and not to full-length AvrPtoB. Thus, by acquiring an E3 ligase domain, AvrPtoB has thwarted a highly conserved host resistance mechanism.

  7. Feasible reactor power cutback logic development for an integral reactor

    International Nuclear Information System (INIS)

    Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

    2013-01-01

    Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

  8. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

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

  9. 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...

  10. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

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

  11. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

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

  12. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

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

  13. Strategic planning for research reactors. Guidance for reactor managers

    International Nuclear Information System (INIS)

    2001-04-01

    The purpose of this publication is to provide guidance on how to develop a strategic plan for a research reactor. The IAEA is convinced of the need for research reactors to have strategic plans and is issuing a series of publications to help owners and operators in this regard. One of these covers the applications of research reactors. That report brings together all of the current uses of research reactors and enables a reactor owner or operator to evaluate which applications might be possible with a particular facility. An analysis of research reactor capabilities is an early phase in the strategic planning process. The current document provides the rationale for a strategic plan, outlines the methodology of developing such a plan and then gives a model that may be followed. While there are many purposes for research reactor strategic plans, this report emphasizes the use of strategic planning in order to increase utilization. A number of examples are given in order to clearly illustrate this function

  14. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1991-04-01

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

  15. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

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

  16. Multiregion reactors

    International Nuclear Information System (INIS)

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

    1979-08-01

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

  17. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

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

  18. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

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

  19. Reactor physics challenges in GEN-IV reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, Michael K.; Hejzlar, Pavel [Massachusetts Institute of Technology, MA (United States)

    2005-02-15

    An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

  20. Reactor physics challenges in GEN-IV reactor design

    International Nuclear Information System (INIS)

    Driscoll, Michael K.; Hejzlar, Pavel

    2005-01-01

    An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources

  1. 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.

  2. The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1983-01-01

    A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

  3. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

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

  4. Tokamak reactor studies

    International Nuclear Information System (INIS)

    Baker, C.C.

    1981-01-01

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

  5. Computerized reactor monitor and control for research reactors

    International Nuclear Information System (INIS)

    Buerger, L.; Vegh, E.

    1981-09-01

    The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

  6. Survival after aortic valve replacement for severe aortic stenosis with low transvalvular gradients and severe left ventricular dysfunction

    Science.gov (United States)

    Pereira, Jeremy J.; Lauer, Michael S.; Bashir, Mohammad; Afridi, Imran; Blackstone, Eugene H.; Stewart, William J.; McCarthy, Patrick M.; Thomas, James D.; Asher, Craig R.

    2002-01-01

    OBJECTIVE: We sought to assess whether aortic valve replacement (AVR) among patients with severe aortic stenosis (AS), severe left ventricular (LV) dysfunction and a low transvalvular gradient (TVG) is associated with improved survival. BACKGROUND: The optimal management of patients with severe AS with severe LV dysfunction and a low TVG remains controversial. METHODS: Between 1990 and 1998, we evaluated 68 patients who underwent AVR at our institution (AVR group) and 89 patients who did not undergo AVR (control group), with an aortic valve area < or = 0.75 cm(2), LV ejection fraction < or = 35% and mean gradient < or = 30 mm Hg. Using propensity analysis, survival was compared between a cohort of 39 patients in the AVR group and 56 patients in the control group. RESULTS: Despite well-matched baseline characteristics among propensity-matched patients, the one- and four-year survival rates were markedly improved in patients in the AVR group (82% and 78%), as compared with patients in the control group (41% and 15%; p < 0.0001). By multivariable analysis, the main predictor of improved survival was AVR (adjusted risk ratio 0.19, 95% confidence interval 0.09 to 0.39; p < 0.0001). The only other predictors of mortality were age and the serum creatinine level. CONCLUSIONS: Among select patients with severe AS, severe LV dysfunction and a low TVG, AVR was associated with significantly improved survival.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  8. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  9. Reactor. Mind picture of the future Jules-Horowitz Reactor (RHJ)

    International Nuclear Information System (INIS)

    Eustache, S.

    1999-01-01

    This paper gives information about the future research reactor, named Reactor Jules-Horowitz (RJH). This irradiation reactor will be placed at industrialists disposal, for research concerning the competitiveness and the safety french electro-nuclear park. Principles and innovations are detailed. This reactor will respect the ALARA principle (as low as reasonably achievable). (A.L.B.)

  10. The analysis for inventory of experimental reactor high temperature gas reactor type

    International Nuclear Information System (INIS)

    Sri Kuntjoro; Pande Made Udiyani

    2016-01-01

    Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

  11. Impact of Valvuloarterial Impedance on Concentric Remodeling in Aortic Stenosis and Its Regression after Valve Replacement.

    Science.gov (United States)

    Jang, Jeong Yoon; Seo, Jeong-Sook; Sun, Byung Joo; Kim, Dae-Hee; Song, Jong-Min; Kang, Duk-Hyun; Song, Jae-Kwan

    2016-09-01

    Left ventricle (LV) in patients with aortic stenosis (AS) faces a double hemodynamic load incorporating both valvular stenosis and reduced systemic arterial compliance (SAC). This study aimed to evaluate the impact of global LV afterload on LV hypertrophy (LVH) before and after aortic valve replacement (AVR). The study cohort included 453 patients (247 males; mean age, 64 ± 11 years) who underwent AVR. Pre- and post-AVR echocardiographic examinations were retrospectively analyzed including an index of valvuloarterial impedance (Z VA ) and LV mass index/LV end-diastolic volume index (LVMI/LVEDVI) as a parameter of LVH. Pre-AVR LVMI/LVEDVI was 2.7 ± 0.9 g/mL with an aortic valve area (AVA) of 0.6 ± 0.2 cm 2 . Z VA was 5.9 ± 1.9 mm Hg/mL/m 2 and showed a stronger correlation (β = 0.601, p regression in 322 patients with follow-up duration >1 year after AVR. Z VA is a major determinant of concentric remodeling in AS before AVR and LVH regression after AVR, which should be incorporated in routine evaluation of AS.

  12. FBR type reactors

    International Nuclear Information System (INIS)

    Suzuoki, Akira; Yamakawa, Masanori.

    1985-01-01

    Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

  13. FBR type reactor

    International Nuclear Information System (INIS)

    Hayase, Tamotsu.

    1991-01-01

    The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

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

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

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

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

  18. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-01

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

  20. Nuclear reactor instrumentation at research reactor renewal

    International Nuclear Information System (INIS)

    Baers, B.; Pellionisz, P.

    1981-10-01

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

  1. Association EURATOM-FZJ. Annual progress report 2012. SC-FZJ 90(13)/4.1.2

    International Nuclear Information System (INIS)

    2012-01-01

    to the neighbouring EURATOM associations in Belgium and The Netherlands. In 1996 they together have founded the Trilateral Euregio Cluster (TEC) which provides a clustering of resources in order to perform a co-ordinated R and D programme, to operate or construct large facilities (TEXTOR, MAGNUM-PSI), to combine different kinds of expertise and to allow for the forming of a strong partner-ship as a consortium within the ITER construction phase. An updated TEC agreement with a strong emphasis on the topic ''plasma-wall interactions'' and the joint use of dedicated facilities in Juelich, Rijnhuizen (NL) and Mol (B) has been signed in 2010. Cooperations beyond Europe are supported by an IEA Implementing Agreement on ''Plasma-Wall Interaction in TEXTOR'' together with Japan, USA and Canada. In view of the fact that TEXTOR will be closed at the end of 2013 and that new opportunities are arising from new linear plasma test facilities, the parties of the Implementing Agreement have agreed upon a change of name and scope: ''Implementing Agreement on the Development and Research on Plasma Wall Interaction Facilities for Fusion Reactors''. This change will involve a number of existing and planned linear plasma devices in Europe, Japan and USA. The start under the new name and scope is planned in the course of 2013.

  2. Reactor System Design

    International Nuclear Information System (INIS)

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

    2006-08-01

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

  3. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1983-01-01

    Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

  4. Prevention device for rapid reactor core shutdown in BWR type reactors

    International Nuclear Information System (INIS)

    Koshi, Yuji; Karatsu, Hiroyuki.

    1986-01-01

    Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

  5. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  6. High-temperature and breeder reactors - economic nuclear reactors of the future

    International Nuclear Information System (INIS)

    Djalilzadeh, A.M.

    1977-01-01

    The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

  7. Heterogeneous reactors

    International Nuclear Information System (INIS)

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

    1979-08-01

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

  8. Nuclear reactor neutron shielding

    Science.gov (United States)

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

    2017-09-12

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

  9. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

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

  10. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

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

  11. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    Powell, J.R.

    1977-01-01

    The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

  12. Project study of a small-angle neutron scattering apparatus

    International Nuclear Information System (INIS)

    Schedler, E.; Pollet, J.L.

    1979-03-01

    This design study deals with the set up of a low angle scattering apparatus in the HMI reactor hall in Berlin. The experiences of other institutes with facilities of a similar type, - especially with D11 and D17 of the ILL in Grenoble, the set up the KFA in Juelich and of the PTB in Braunschweig -, are included to a large extend. The aim of this paper is - to define the necessary boundary conditions for the construction (including: installation of a cold source, the beam line, the neutron guide pipe and an extention of the reactor hall), -to determine the properties of the planned apparatus, especially in comparison with D11, probably the most versatile instrument, - to make desitions for the design of the components, - to work out the detailed drawings for construction - to estimate the costs and the time necessary for construction, if industrial manufacturers set up the project. (orig.) [de

  13. Fusion-fission hybrid reactors

    International Nuclear Information System (INIS)

    Greenspan, E.

    1984-01-01

    This chapter discusses the range of characteristics attainable from hybrid reactor blankets; blanket design considerations; hybrid reactor designs; alternative fuel hybrid reactors; multi-purpose hybrid reactors; and hybrid reactors and the energy economy. Hybrid reactors are driven by a fusion neutron source and include fertile and/or fissile material. The fusion component provides a copious source of fusion neutrons which interact with a subcritical fission component located adjacent to the plasma or pellet chamber. Fissile fuel and/or energy are the main products of hybrid reactors. Topics include high F/M blankets, the fissile (and tritium) breeding ratio, effects of composition on blanket properties, geometrical considerations, power density and first wall loading, variations of blanket properties with irradiation, thermal-hydraulic and mechanical design considerations, safety considerations, tokamak hybrid reactors, tandem-mirror hybrid reactors, inertial confinement hybrid reactors, fusion neutron sources, fissile-fuel and energy production ability, simultaneous production of combustible and fissile fuels, fusion reactors for waste transmutation and fissile breeding, nuclear pumped laser hybrid reactors, Hybrid Fuel Factories (HFFs), and scenarios for hybrid contribution. The appendix offers hybrid reactor fundamentals. Numerous references are provided

  14. Reactor protection systems for the Replacement Research Reactor, ANSTO

    International Nuclear Information System (INIS)

    Morris, C.R.

    2003-01-01

    The 20-MW Replacement Research Reactor Project which is currently under construction at ANSTO will have a combination of a state of the art triplicated computer based reactor protection system, and a fully independent, and diverse, triplicated analogue reactor protection system, that has been in use in the nuclear industry, for many decades. The First Reactor Protection System (FRPS) consists of a Triconex triplicated modular redundant system that has recently been approved by the USNRC for use in the USA?s power reactor program. The Second Reactor Protection System is a hardwired analogue system supplied by Foxboro, the Spec 200 system, which is also Class1E qualified. The FRPS is used to drop the control rods when its safety parameter setpoints have been reached. The SRPS is used to drain the reflector tank and since this operation would result in a reactor poison out due to the time it would take to refill the tank the FRPS trip setpoints are more limiting. The FRPS and SRPS have limited hardwired indications on the control panels in the main control room (MCR) and emergency control centre (ECC), however all FRPS and SRPS parameters are capable of being displayed on the reactor control and monitoring system (RCMS) video display units. The RCMS is a Foxboro Series I/A control system which is used for plant control and monitoring and as a protection system for the cold neutron source. This paper will provide technical information on both systems, their trip logics, their interconnections with each other, and their integration into the reactor control and monitoring system and control panels. (author)

  15. The future of HTR development and market chances

    International Nuclear Information System (INIS)

    Baust, E.; Weisbrodt, I.

    1989-01-01

    In more than thirty years of development, the pebble bed high-temperature reactor has been brought to the threshold of commercial maturity. On the basis of the experience accumulated with the 15 MW AVR reactor and the THTR-300, unit sizes tailored to demand (HTR-500, modular HTR, GHR-10) will be developed for the electricity and heat markets of the future. The high-temperature reactor is a meaningful supplement to the proven line of light-water reactors and is particularly suitable for being exported to developing countries and industrial threshold countries because of its special technical and inherent safeguards properties. There is broad worldwide interest in the HTR, as is evidenced by several existing agreements on cooperation. It is for this reason that market chances are believed to exist for the HTR after the expected revival of the nuclear power market. ABB and Siemens therefore have decided to develop and market the HTR jointly in the future as a matter of long term strategy by working through a joint subsidiary, HTR-GmbH. (orig.) [de

  16. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

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

  17. LV reverse remodeling imparted by aortic valve replacement for severe aortic stenosis; is it durable? A cardiovascular MRI study sponsored by the American Heart Association

    Directory of Open Access Journals (Sweden)

    Caruppannan Ketheswaram

    2011-04-01

    Full Text Available Abstract Background In patients with severe aortic stenosis (AS, long-term data tracking surgically induced effects of afterload reduction on reverse LV remodeling are not available. Echocardiographic data is available short term, but in limited fashion beyond one year. Cardiovascular MRI (CMR offers the ability to serially track changes in LV metrics with small numbers due to its inherent high spatial resolution and low variability. Hypothesis We hypothesize that changes in LV structure and function following aortic valve replacement (AVR are detectable by CMR and once triggered by AVR, continue for an extended period. Methods Tweny-four patients of which ten (67 ± 12 years, 6 female with severe, but compensated AS underwent CMR pre-AVR, 6 months, 1 year and up to 4 years post-AVR. 3D LV mass index, volumetrics, LV geometry, and EF were measured. Results All patients survived AVR and underwent CMR 4 serial CMR's. LVMI markedly decreased by 6 months (157 ± 42 to 134 ± 32 g/m2, p 2. Similarly, EF increased pre to post-AVR (55 ± 22 to 65 ± 11%,(p 2. LV stroke volume increased rapidly from pre to post-AVR (40 ± 11 to 44 ± 7 ml, p Conclusion After initial beneficial effects imparted by AVR in severe AS patients, there are, as expected, marked improvements in LV reverse remodeling. Via CMR, surgically induced benefits to LV structure and function are durable and, unexpectedly express continued, albeit markedly incomplete improvement through 4 years post-AVR concordant with sustained improved clinical status. This supports down-regulation of both mRNA and MMP activity acutely with robust suppression long term.

  18. A multi-purpose reactor

    International Nuclear Information System (INIS)

    Changwen Ma

    2000-01-01

    An integrated natural circulation self pressurized reactor can be used for sea water desalination, electrogeneration, ship propulsion and district or process heating. The reactor can be used for ship propulsion because it has following advantages: it is a integrated reactor. Whole primary loop is included in a size limited pressure vessel. For a 200 MW reactor the diameter of the pressure vessel is about 5 m. It is convenient to arranged on a ship. Hydraulic driving facility of control rods is used on the reactor. It notably decreases the height of the reactor. For ship propulsion, smaller diameter and smaller height are important. Besides these, the operation reliability of the reactor is high enough, because there is no rotational machine (for example, circulating pump) in safety systems. Reactor systems are simple. There are no emergency water injection system and boron concentration regulating system. These features for ship propulsion reactor are valuable. Design of the reactor is based on existing demonstration district heating reactor design. The mechanic design principles are the same. But boiling is introduced in the reactor core. Several variants to use the reactor as a movable seawater desalination plant are presented in the paper. When the sea water desalination plant is working to produce fresh water, the reactor can supply electricity at the same time to the local electricity network. Some analyses for comprehensive application of the reactor have been done. Main features and parameters of the small (Thermopower 200 MW) reactor are given in the paper. (author)

  19. On Recording the Unipolar ECG Limb Leads via the Wilson's vs the Goldberger's Terminals: aVR, aVL, and aVF Revisited

    Directory of Open Access Journals (Sweden)

    John E. Madias

    2008-11-01

    Full Text Available The augmented unipolar limb leads aVR, aVL, and aVF, introduced by Goldberger in 1942, are an integral part of the 12-lead ECG.1,2 Leads I, II, and III have 2 dedicated electrodes, but the other 9 leads have a single dedicated electrode, and another one constructed from the averaged inputs of multiple electrodes. This Viewpoint discusses whether an indifferent pole for the recording of unipolar limb leads is best provided by the Wilson's central terminal (WCT, or by inputs from 2 limb electrodes (Goldberger's central terminal (GCT, as done currently, and whether the latter have any advantages over the former. The term "unipolar", popularized by Wilson, is a misnomer, since no leads can be truly "unipolar", all requiring positive and negative poles. Thus the term unipolar is used herein in the quasi-unipolar sense, as when first introduced by Wilson and Goldberger, who also realized that such leads were not truly unipolar. The popularity of the unipolar leads reflected the quest of recording the ECG from various vantage points of the body, considering the limitations of the 3 bipolar leads, introduced by Einthoven,3 which register the difference of 2 ECG curves recorded at the 2 poles of these leads, and no variation in potential at each of these poles.4 In contrast the unipolar leads were thought to register such variation of absolute potential, something not really true. Initially the WCT was used to record the unipolar limb leads,5 but the amplitude was low, and the inscribed ECGs, then, and for many decades later,6 were thick-lined (≥2 mm (Figure 1.

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  1. Supercritical Water Reactors

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  2. Tank type reactor

    International Nuclear Information System (INIS)

    Otsuka, Fumio.

    1989-01-01

    The present invention concerns a tank type reactor capable of securing reactor core integrity by preventing incorporation of gases to an intermediate heat exchanger, thgereby improving the reliability. In a conventional tank type reactor, since vortex flows are easily caused near the inlet of an intermediate heat exchanger, there is a fear that cover gases are involved into the coolant main streams to induce fetal accidents. In the present invention, a reactor core is suspended by way of a suspending body to the inside of a reactor vessel and an intermediate heat exchanger and a pump are disposed between the suspending body and the reactor vessel, in which a vortex current preventive plate is attached at the outside near the coolant inlet on the primary circuit of the intermediate heat exchanger. In this way vortex or turbulence near the inlet of the intermediate heata exchanger or near the surface of coolants can be prevented. Accordingly, the cover gases are no more involved, to insure the reactor core integrity and obtain a tank type nuclear reactor of high reliability. (I.S.)

  3. New about research reactors

    International Nuclear Information System (INIS)

    Egorenkov, P.M.

    2001-01-01

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

  4. Evaluation of Pressure Changes in HANARO Reactor Hall after a Reactor Shutdown

    International Nuclear Information System (INIS)

    Han, Geeyang; Han, Jaesam; Ahn, Gukhoon; Jung, Hoansung

    2013-01-01

    The major objective of this work is intended to evaluate the characteristics of the thermal behavior regarding how the decay heat will be affected by the reactor hall pressure change and the increase of pool water temperature induced in the primary coolant after a reactor shutdown. The particular reactor pool water temperature at the surface where it is evaporated owing to the decay heat resulting in the local heat transfer rate is related to the pressure change response in the reactor hall associated with the primary cooling system because of the reduction of the heat exchanger to remove the heat. The increase in the pool water temperature is proportional to the heat transfer rate in the reactor pool. Consequently, any limit on the reactor pool water temperature imposes a corresponding limit on the reactor hall pressure. At HANARO, the decay heat after a reactor shutdown is mainly removed by the natural circulation cooling in the reactor pool. This paper is written for the safety feature of the pressure change related leakage rate from the reactor hall. The calculation results show that the increase of pressure in the reactor hall will not cause any serious problems to the safety limits although the reactor hall pressure is slightly increased. Therefore, it was concluded that the pool water temperature increase is not so rapid as to cause the pressure to vary significantly in the reactor hall. Furthermore, the mathematical model developed in this work can be a useful analytical tool for scoping and parametric studies in the area of thermal transient analysis, with its proper representation of the interaction between the temperature and pressure in the reactor hall

  5. Design guide for Category III reactors: pool type reactors

    International Nuclear Information System (INIS)

    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

  6. The computerized reactor period measurement system for China fast burst reactor-II

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1996-01-01

    The article simply introduces the hardware, principle, and software of the computerized reactor period measurement system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between fission yield and pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computerized measurement system makes the reactor period measurement into automatical and intelligent and also improves the speed and precision of period data on-line process

  7. Computer measurement system of reactor period for China fast burst reactor-II

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1997-01-01

    The author simply introduces the hardware, principle, and software of the reactor period computer measure system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between Fission yield and Pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computer measure system makes the reactor period measurement into automation and intellectualization and also improves the speed and precision of period data process on-line

  8. Reactor safety method

    International Nuclear Information System (INIS)

    Vachon, L.J.

    1980-01-01

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

  9. Small propulsion reactor design based on particle bed reactor concept

    International Nuclear Information System (INIS)

    Ludewig, H.; Lazareth, O.; Mughabghab, S.; Perkins, K.; Powell, J.R.

    1989-01-01

    In this paper Particle Bed Reactor (PBR) designs are discussed which use 233 U and /sup 242m/Am as fissile materials. A constant total power of 100MW is assumed for all reactors in this study. Three broad aspects of these reactors is discussed. First, possible reactor designs are developed, second physics calculations are outlined and discussed and third mass estimates of the various candidates reactors are made. It is concluded that reactors with a specific mass of 1 kg/MW can be envisioned of 233 U is used and approximately a quarter of this value can be achieved if /sup 242m/Am is used. If this power level is increased by increasing the power density lower specific mass values are achievable. The limit will be determined by uncertainties in the thermal-hydraulic analysis. 5 refs., 5 figs., 6 tabs

  10. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    Baeten, Peter

    2006-01-01

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

  11. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    2010-03-01

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

  12. Guide to power reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-15

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

  13. Analysis of dynamic stability and safety of reactor system by reactor simulator

    International Nuclear Information System (INIS)

    Raisic, N.

    1963-11-01

    In order to enable qualitative analysis of dynamic properties of reactors RA and RB, mathematical models of these reactors were formulated and adapted for solution on analog computer. This report contains basic assessments for creating the model and complete equations for each reactor. Model was used to analyse three possible accidents at the RA reactor and possible hypothetical accidents at the RB reactor

  14. 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

  15. Association EURATOM-FZJ. Annual progress report 2013. SC-FZJ-92-(14)-4.1.3

    International Nuclear Information System (INIS)

    2013-01-01

    or construct large facilities (TEXTOR, MAGNUM-PSI), to combine different kinds of expertise and to allow for the forming of a strong partner-ship as a consortium within the ITER construction phase. An updated TEC agreement with a strong emphasis on the topic ''plasma-wall interactions'' and the joint use of dedicated facilities in Juelich, Rijnhuizen (NL) and Mol (B) has been signed in 2010. Co-operations beyond Europe are supported by an IEA Implementing Agreement on ''Plasma-Wall Interaction in TEXTOR'' together with Japan, USA and Canada. In view of the fact that TEXTOR has been closed at the end of 2013 and that new opportunities are arising from new linear plasma test facilities, the parties of the Implementing Agreement have agreed upon a change of name and scope: ''Implementing Agreement on the Development and Research on Plasma Wall Interaction Facilities for Fusion Reactors''. This change will involve a number of existing and planned linear plasma devices in Europe, Japan and USA. With the ending of the Association EURATOM-FZJ also the TEXTOR experiment closed operations in December 2013.

  16. Association EURATOM-FZJ. Annual progress report 2013. SC-FZJ-92-(14)-4.1.3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-07-01

    or construct large facilities (TEXTOR, MAGNUM-PSI), to combine different kinds of expertise and to allow for the forming of a strong partner-ship as a consortium within the ITER construction phase. An updated TEC agreement with a strong emphasis on the topic ''plasma-wall interactions'' and the joint use of dedicated facilities in Juelich, Rijnhuizen (NL) and Mol (B) has been signed in 2010. Co-operations beyond Europe are supported by an IEA Implementing Agreement on ''Plasma-Wall Interaction in TEXTOR'' together with Japan, USA and Canada. In view of the fact that TEXTOR has been closed at the end of 2013 and that new opportunities are arising from new linear plasma test facilities, the parties of the Implementing Agreement have agreed upon a change of name and scope: ''Implementing Agreement on the Development and Research on Plasma Wall Interaction Facilities for Fusion Reactors''. This change will involve a number of existing and planned linear plasma devices in Europe, Japan and USA. With the ending of the Association EURATOM-FZJ also the TEXTOR experiment closed operations in December 2013.

  17. Sodium-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hammers, H.W.

    1982-01-01

    The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

  18. Atomic reactor thermal engineering

    International Nuclear Information System (INIS)

    Kim, Gwang Ryong

    1983-02-01

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

  19. Argentinean integrated small reactor design and scale economy analysis of integrated reactor

    International Nuclear Information System (INIS)

    Florido, P. C.; Bergallo, J. E.; Ishida, M. V.

    2000-01-01

    This paper describes the design of CAREM, which is Argentinean integrated small reactor project and the scale economy analysis results of integrated reactor. CAREM project consists on the development, design and construction of a small nuclear power plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors. The CAREM is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a highly level of safety: integrated primary cooling system, self pressurized, primary cooling by natural circulation and safety system relying on passive features. For a fully doupled economic evaluation of integrated reactors done by IREP (Integrated Reactor Evaluation Program) code transferred to IAEA, CAREM have been used as a reference point. The results shows that integrated reactors become competitive with power larger than 200MWe with Argentinean cheapest electricity option. Due to reactor pressure vessel construction limit, low pressure drop steam generator are used to reach power output of 200MWe for natural circulation. For forced circulation, 300MWe can be achieved. (author)

  20. HTR core physics and transient analyses by the Panthermix code system

    International Nuclear Information System (INIS)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.

    2005-01-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes

  1. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

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

  2. Reactor power measuring device

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  3. Reactor container

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  4. Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

    2014-01-01

    The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

  5. Accident Testing of High Temperature Reactor Fuel Elements with the KueFA Device

    International Nuclear Information System (INIS)

    Seeger, O.; Laurie, M.; Bottomley, P.D.W.; Ferreira-Teixeira, A.E.; Van Winckel, S.; Rondinella, V.V.; Allelein, H.J.

    2013-06-01

    The High Temperature Reactor (HTR) is characterised by an advanced design with passive safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with Tri-Isotropic (TRISO) coating, designed to provide high fission product retention. During a loss of coolant accident scenario in a HTR the maximum temperature is foreseen to be in the range of 1600-1650 deg. C, remaining well below the melting point of the fuel. The Cold Finger Apparatus (KueFA) is used to observe the combined effects of Depressurization and Loss of Forced Circulation (DLOFC) accident scenarios on HTR fuel. Originally designed at the Forschungszentrum Juelich (FZJ), an adapted KueFA operates on irradiated fuel in hot cell at JRC-ITU. A fuel pebble is heated in He atmosphere for several hundred hours, mimicking accident temperatures up to 1800 deg. C and realistic temperature transients. Non-gaseous volatile fission products released from the fuel condense on a water cooled stainless steel plate dubbed 'Cold Finger'. Exchanging plates frequently during the experiment and analysing plate deposits by means of HPGe gamma spectroscopy allows a reconstruction of the fission product release as a function of time and temperature. In order to achieve a good quantification of the release, a careful calibration of the setup is mandatory. An especially tailored collimator was designed to perform plate scanning with high spatial resolution, thus yielding information about the fission product distribution on the condensation plates. The analysis of condensation plates from recent KueFA tests shows that fission product release quantification is possible at high and low activity levels. Chemical dissolution has been performed for some condensation plates in order to assess beta nuclides of interest such as 90 Sr and possibly 129 I using an Inductively Coupled Plasma - Mass Spectrometer (ICP-MS) and to cross check the HPGe gamma spectroscopy measurements

  6. Nuclear power reactors: reactor safety and military and civil defence

    International Nuclear Information System (INIS)

    Hvinden, T.

    1976-01-01

    The formation of fission products and plutonium in reactors is briefly described, followed by a short general discussion of reactor safety. The interaction of reactor safety and radioactive release considerations with military and civil defence is thereafter discussed. Reactors and other nuclear plants are factors which must be taken into account in the defence of the district around the site, and as potential targets of both conventional and guerilla attacks and sabotage, requiring special defence. The radiological hazards arising from serious damage to a power reactor by conventional weapons are briefly discussed, and the benefits of underground siting evaluated. Finally the author discusses the significance of the IAEA safeguards work as a preventive factor. (JIW)

  7. A randomized multicenter trial of minimally invasive rapid deployment versus conventional full sternotomy aortic valve replacement.

    Science.gov (United States)

    Borger, Michael A; Moustafine, Vadim; Conradi, Lenard; Knosalla, Christoph; Richter, Markus; Merk, Denis R; Doenst, Torsten; Hammerschmidt, Robert; Treede, Hendrik; Dohmen, Pascal; Strauch, Justus T

    2015-01-01

    Minimally invasive surgical procedures (MIS) may offer several advantages over conventional full sternotomy (FS) aortic valve replacement (AVR). A novel class of aortic valve prostheses has been developed for rapid-deployment AVR (RDAVR). We report a randomized, multicenter trial comparing the outcomes for MIS-RDAVR with those of conventional FS-AVR. A total of 100 patients with aortic stenosis were enrolled in a prospective, multicenter, randomized comparison trial (CADENCE-MIS). Exclusion criteria included ejection fraction below 25%, AVR requiring concomitant procedures, and recent myocardial infarction or stroke. Patients were randomized to undergo MIS-RDAVR through an upper hemisternotomy (n = 51) or AVR by FS with a conventional stented bioprosthesis (n = 49). Three patients were excluded before the procedure, and 3 more patients who were randomized to undergo RDAVR were excluded because of their anatomy. Procedural, early clinical outcomes, and functional outcomes were assessed for the remaining 94 patients. Hemodynamic performance was assessed by an echocardiography core laboratory. Implanted valve sizes were similar between groups (22.9 ± 2.1 vs 23.0 ± 2.1 mm, p = 0.9). MIS-RDAVR was associated with significantly reduced aortic cross-clamp times compared with FS-AVR (41.3 ± 20.3 vs 54.0 ± 20.3 minutes, p quality of life measures. The RDAVR patients had a significantly lower mean transvalvular gradient (8.5 vs 10.3 mm Hg, p = 0.044) and a lower prevalence of patient-prosthesis mismatch (0% vs 15.0%, p = 0.013) 3 months postoperatively compared with the FS-AVR patients. RDAVR by the MIS approach is associated with significantly reduced myocardial ischemic time and better valvular hemodynamic function than FS-AVR with a conventional stented bioprosthesis. Rapid deployment valves may facilitate the performance of MIS-AVR. Copyright © 2015 The Society of Thoracic Surgeons. Published by Elsevier Inc. All rights reserved.

  8. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Carlos Ruben Calabrese

    1999-01-01

    Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

  9. To the analysis of reactor noise in boiling water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1972-01-01

    The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de

  10. The CEA research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1993-01-01

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

  11. 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.

  12. Virtual nuclear reactor for education of nuclear reactor physics

    International Nuclear Information System (INIS)

    Tsuji, Masashi; Narabayashi, Takashi; Shimazu, Youichiro

    2008-01-01

    As one of projects that were programmed in the cultivation program for human resources in nuclear engineering sponsored by the Ministry of Economy, Trade and Industry, the development of a virtual reactor for education of nuclear reactor physics started in 2007. The purpose of the virtual nuclear reactor is to make nuclear reactor physics easily understood with aid of visualization. In the first year of this project, the neutron slowing down process was visualized. The data needed for visualization are provided by Monte Carlo calculations; The flights of the respective neutrons generated by nuclear fissions are traced through a reactor core until they disappear by neutron absorption or slow down to a thermal energy. With this visualization and an attached supplement textbook, it is expected that the learners can learn more clearly the physical implication of neutron slowing process that is mathematically described by the Boltzmann neutron transport equation. (author)

  13. Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

  14. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

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

  15. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  16. Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

    International Nuclear Information System (INIS)

    Reuscher, J.A.; Pickard, P.S.

    1976-01-01

    The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

  17. Identification of nuclear reactor characteristics by the reactor noise analysis

    International Nuclear Information System (INIS)

    Yashima, Hideyuki

    1980-01-01

    Reactor noise analysis method was applied to TRIGA II Research Reactor (Atomic Research Laboratory, Musashi Institute of Technology) and computed power spectral density (PSD) from the CIC current record. PSD has provided many valuable informations regarding to the reactor kinetics, including the effect of control rods vibration. Another information of neutron physics parameters were obtained and this result was compared with the parameter which was formerly measured by the Feynman-α experiment. Through these experiments we could find overall frequency characteristics of TRIGA II Reactor. (author)

  18. Cost-effectiveness of transcatheter aortic valve replacement compared with surgical aortic valve replacement in high-risk patients with severe aortic stenosis: results of the PARTNER (Placement of Aortic Transcatheter Valves) trial (Cohort A).

    Science.gov (United States)

    Reynolds, Matthew R; Magnuson, Elizabeth A; Lei, Yang; Wang, Kaijun; Vilain, Katherine; Li, Haiyan; Walczak, Joshua; Pinto, Duane S; Thourani, Vinod H; Svensson, Lars G; Mack, Michael J; Miller, D Craig; Satler, Lowell E; Bavaria, Joseph; Smith, Craig R; Leon, Martin B; Cohen, David J

    2012-12-25

    The aim of this study was to evaluate the cost-effectiveness of transcatheter aortic valve replacement (TAVR) compared with surgical aortic valve replacement (AVR) for patients with severe aortic stenosis and high surgical risk. TAVR is an alternative to AVR for patients with severe aortic stenosis and high surgical risk. We performed a formal economic analysis based on cost, quality of life, and survival data collected in the PARTNER A (Placement of Aortic Transcatheter Valves) trial in which patients with severe aortic stenosis and high surgical risk were randomized to TAVR or AVR. Cumulative 12-month costs (assessed from a U.S. societal perspective) and quality-adjusted life-years (QALYs) were compared separately for the transfemoral (TF) and transapical (TA) cohorts. Although 12-month costs and QALYs were similar for TAVR and AVR in the overall population, there were important differences when results were stratified by access site. In the TF cohort, total 12-month costs were slightly lower with TAVR and QALYs were slightly higher such that TF-TAVR was economically dominant compared with AVR in the base case and economically attractive (incremental cost-effectiveness ratio economically dominated by AVR in the base case and economically attractive in only 7.1% of replicates. In the PARTNER trial, TAVR was an economically attractive strategy compared with AVR for patients suitable for TF access. Future studies are necessary to determine whether improved experience and outcomes with TA-TAVR can improve its cost-effectiveness relative to AVR. Copyright © 2012 American College of Cardiology Foundation. Published by Elsevier Inc. All rights reserved.

  19. Long-Term Outcomes of the Ross Procedure Versus Mechanical Aortic Valve Replacement: Propensity-Matched Cohort Study.

    Science.gov (United States)

    Mazine, Amine; David, Tirone E; Rao, Vivek; Hickey, Edward J; Christie, Shakira; Manlhiot, Cedric; Ouzounian, Maral

    2016-08-23

    The ideal aortic valve substitute in young and middle-aged adults remains unknown. We sought to compare the long-term outcomes of patients undergoing the Ross procedure and those receiving a mechanical aortic valve replacement (AVR). From 1990 to 2014, 258 patients underwent a Ross procedure and 1444 had a mechanical AVR at a single institution. Patients were matched into 208 pairs through the use of a propensity score. Mean age was 37.2±10.2 years, and 63% were male. Mean follow-up was 14.2±6.5 years. Overall survival was equivalent (Ross versus AVR: hazard ratio, 0.91, 95% confidence interval, 0.38-2.16; P=0.83), although freedom from cardiac- and valve-related mortality was improved in the Ross group (Ross versus AVR: hazard ratio, 0.22; 95% confidence interval, 0.034-0.86; P=0.03). Freedom from reintervention was equivalent after both procedures (Ross versus AVR: hazard ratio, 1.86; 95% confidence interval, 0.76-4.94; P=0.18). Long-term freedom from stroke or major bleeding was superior after the Ross procedure (Ross versus AVR: hazard ratio, 0.09; 95% confidence interval, 0.02-0.31; PRoss procedure and mechanical AVR. However, the Ross procedure was associated with improved freedom from cardiac- and valve-related mortality and a significant reduction in the incidence of stroke and major bleeding. In specialized centers, the Ross procedure represents an excellent option and should be considered for young and middle-aged adults undergoing AVR. © 2016 American Heart Association, Inc.

  20. Research on the reactor physics and reactor safety of VVER reactors. AER Symposium 2016

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, S.

    2017-09-15

    The selected paperscan be attributed to the following main subjects: Reactor start-up tests and use of corresponding data for code validation, code development and application, approaches for safety analyses, closure of nuclear fuel cycle, prospective reactor concepts.

  1. The fast breeder reactor

    International Nuclear Information System (INIS)

    Collier, J.

    1990-01-01

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

  2. Reactor feedwater system

    International Nuclear Information System (INIS)

    Kagaya, Hiroyuki; Tominaga, Kenji.

    1993-01-01

    In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

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

    International Nuclear Information System (INIS)

    Bussac, J.; Reuss, P.

    1985-01-01

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

  4. Ulysse, mentor reactor

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  5. Siting of research reactors

    International Nuclear Information System (INIS)

    1987-01-01

    The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

  6. Experience in using a research reactor for the training of power reactor operators

    International Nuclear Information System (INIS)

    Blotcky, A.J.; Arsenaut, L.J.

    1972-01-01

    A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

  7. Reactor transients tests for SNR fuel elements in HFR reactor

    International Nuclear Information System (INIS)

    Plitz, H.

    1989-01-01

    In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

  8. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

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

  9. Training reactor deployment. Advanced experimental course on designing new reactor cores

    International Nuclear Information System (INIS)

    Skoda, Radek

    2009-01-01

    Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

  10. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

  11. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    International Nuclear Information System (INIS)

    Kyrki-Rajamaeki, R.

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

  12. Reactor performances and microbial communities of biogas reactors: effects of inoculum sources.

    Science.gov (United States)

    Han, Sheng; Liu, Yafeng; Zhang, Shicheng; Luo, Gang

    2016-01-01

    Anaerobic digestion is a very complex process that is mediated by various microorganisms, and the understanding of the microbial community assembly and its corresponding function is critical in order to better control the anaerobic process. The present study investigated the effect of different inocula on the microbial community assembly in biogas reactors treating cellulose with various inocula, and three parallel biogas reactors with the same inoculum were also operated in order to reveal the reproducibility of both microbial communities and functions of the biogas reactors. The results showed that the biogas production, volatile fatty acid (VFA) concentrations, and pH were different for the biogas reactors with different inocula, and different steady-state microbial community patterns were also obtained in different biogas reactors as reflected by Bray-Curtis similarity matrices and taxonomic classification. It indicated that inoculum played an important role in shaping the microbial communities of biogas reactor in the present study, and the microbial community assembly in biogas reactor did not follow the niche-based ecology theory. Furthermore, it was found that the microbial communities and reactor performances of parallel biogas reactors with the same inoculum were different, which could be explained by the neutral-based ecology theory and stochastic factors should played important roles in the microbial community assembly in the biogas reactors. The Bray-Curtis similarity matrices analysis suggested that inoculum affected more on the microbial community assembly compared to stochastic factors, since the samples with different inocula had lower similarity (10-20 %) compared to the samples from the parallel biogas reactors (30 %).

  13. The fast reactor

    International Nuclear Information System (INIS)

    1980-02-01

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

  14. PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Masood, Z.

    2016-01-01

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

  15. 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...

  16. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1985-08-01

    Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

  17. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

    1982-09-01

    Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

  18. Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

    International Nuclear Information System (INIS)

    D. Kokkinos

    2005-01-01

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

  19. Multi-purpose reactor

    International Nuclear Information System (INIS)

    1991-05-01

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

  20. Mirror reactor studies

    International Nuclear Information System (INIS)

    Moir, R.W.; Barr, W.L.; Bender, D.J.

    1977-01-01

    Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

  1. Reactor Simulations for Safeguards with the MCNP Utility for Reactor Evolution Code

    International Nuclear Information System (INIS)

    Shiba, T.; Fallot, M.

    2015-01-01

    To tackle nuclear material proliferation, we conducted several proliferation scenarios using the MURE (MCNP Utility for Reactor Evolution) code. The MURE code, developed by CNRS laboratories, is a precision, open-source code written in C++ that automates the preparation and computation of successive MCNP (Monte Carlo N-Particle) calculations and solves the Bateman equations in between, for burnup or thermal-hydraulics purposes. In addition, MURE has been completed recently with a module for the CHaracterization of Radioactive Sources, called CHARS, which computes the emitted gamma, beta and alpha rays associated to any fuel composition. Reactor simulations could allow knowing how plutonium or other material generation evolves inside reactors in terms of time and amount. The MURE code is appropriate for this purpose and can also provide knowledge on associated particle emissions. Using MURE, we have both developed a cell simulation of a typical CANDU reactor and a detailed model of light water PWR core, which could be used to analyze the composition of fuel assemblies as a function of time or burnup. MURE is also able to provide, thanks to its extension MURE-CHARTS, the emitted gamma rays from fuel assemblies unloaded from the core at any burnup. Diversion cases of Generation IV reactors have been also developed; a design of Very High Temperature Reactor (a Pebble Bed Reactor (PBR), loaded with UOx, PuOx and ThUOx fuels), and a Na-cooled Fast Breeder Reactor (FBR) (with depleted Uranium or Minor Actinides in the blanket). The loading of Protected Plutonium Production (P3) in the FBR was simulated. The simulations of various reactor designs taking into account reactor physics constraints may bring valuable information to inspectors. At this symposium, we propose to show the results of these reactor simulations as examples of the potentiality of reactor simulations for safeguards. (author)

  2. Reactor Sharing Program

    International Nuclear Information System (INIS)

    Tehan, Terry

    2002-01-01

    Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

  3. 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.

  4. 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.

  5. Reactor control device

    International Nuclear Information System (INIS)

    Fukami, Haruo; Morimoto, Yoshinori.

    1981-01-01

    Purpose: To operate a reactor always with safety operation while eliminating the danger of tripping. Constitution: In a reactor control device adapted to detect the process variants of a reactor, control a control rod drive controlling system based on the detected signal to thereby control the driving the control rods, control the reactor power and control the electric power generated from an electric generator by the output from the reactor, detection means is provided for the detection of the electric power from said electric generator, and a compensation device is provided for outputting control rod driving compensation signals to the control rod driving controlling system in accordance with the amount of variation in the detected value. (Seki, T.)

  6. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    Veeraraghaven, N.

    1990-01-01

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  7. Fully integrated analysis of reactor kinetics, thermalhydraulics and the reactor control system in the MAPLE-X10 research reactor

    International Nuclear Information System (INIS)

    Shim, S.Y.; Carlson, P.A.; Baxter, D.K.

    1992-01-01

    A prototype research reactor, designated MAPLE-X10 (Multipurpose Applied Physics Lattice Experimental - X 10MW), is currently being built at AECL's Chalk River Laboratories. The CATHENA (Canadian Algorithm for Thermalhydraulic Network Analysis) two-fluid code was used in the safety analysis of the reactor to determine the adequacy of core cooling during postulated reactivity and loss-of-forced-flow transients. The system responses to a postulated transient are predicted including the feedback between reactor kinetics, thermalhydrauilcs and the reactor control systems. This paper describes the MAPLE-X10 reactor and the modelling methodology used. Sample simulations of postulated loss-of-heat-sink and loss-of-regulation transients are presented. (author)

  8. Power Trip Set-points of Reactor Protection System for New Research Reactor

    International Nuclear Information System (INIS)

    Lee, Byeonghee; Yang, Soohyung

    2013-01-01

    This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

  9. Volcanism and hydrothermalism on a hotspot-influenced ridge: Comparing Reykjanes Peninsula and Reykjanes Ridge, Iceland

    Science.gov (United States)

    Pałgan, Dominik; Devey, Colin W.; Yeo, Isobel A.

    2017-12-01

    Current estimates indicate that the number of high-temperature vents (one of the primary pathways for the heat extraction from the Earth's mantle) - at least 1 per 100 km of axial length - scales with spreading rate and should scale with crustal thickness. But up to present, shallow ridge axes underlain by thick crust show anomalously low incidences of high-temperature activity. Here we compare the Reykjanes Ridge, an abnormally shallow ridge with thick crust and only one high-temperature vent known over 900 km axial length, to the adjacent subaerial Reykjanes Peninsula (RP), which is characterized by high-temperature geothermal sites confined to four volcanic systems transected by fissure swarms with young (Holocene) volcanic activity, multiple faults, cracks and fissures, and continuous seismic activity. New high-resolution bathymetry (gridded at 60 m) of the Reykjanes Ridge between 62°30‧N and 63°30‧N shows seven Axial Volcanic Ridges (AVR) that, based on their morphology, geometry and tectonic regime, are analogues for the volcanic systems and fissure swarms on land. We investigate in detail the volcano-tectonic features of all mapped AVRs and show that they do not fit with the previously suggested 4-stage evolution model for AVR construction. Instead, we suggest that AVR morphology reflects the robust or weak melt supply to the system and two (or more) eruption mechanisms may co-exist on one AVR (in contrast to 4-stage evolution model). Our interpretations indicate that, unlike on the Reykjanes Peninsula, faults on and around AVRs do not cluster in orientation domains but all are subparallel to the overall strike of AVRs (orthogonal to spreading direction). High abundance of seamounts shows that the region centered at 62°47‧N and 25°04‧W (between AVR-5 and -6) is volcanically robust while the highest fault density implies that AVR-1 and southern part of AVR-6 rather undergo period of melt starvation. Based on our observations and interpretations we

  10. Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Khang, Ngo Phu [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    The Dalat nuclear research reactor was reconstructed from TRIGA MARK II reactor, built in 1963 with nominal power of 250 KW, and reached its planned nominal power of 500 kW for the first time in Feb. 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc. (author). 10 refs., 10 figs., 2 tabs.

  11. Reactor safety device

    International Nuclear Information System (INIS)

    Okada, Yasumasa.

    1987-01-01

    Purpose: To scram control rods by processing signals from a plurality of temperature detectors and generating abnormal temperature warning upon occurrence of abnormal temperature in a nuclear reactor. Constitution: A temperature sensor comprising a plurality of reactors each having a magnetic body as the magnetic core having a curie point different from each other and corresponding to the abnormal temperature against which reactor core fuels have to be protected is disposed in an identical instrumentation well near the reactor core fuel outlet/inlet of a reactor. A temperature detection device actuated upon detection of an abnormal temperature by the abrupt reduction of the reactance of each of the reactors is disposed. An OR circuit and an AND circuit for conducting OR and AND operations for each of the abnormal temperature detection signals from the temperature detection device are disposed. The output from the OR circuit is used as the abnormal temperature warning signal, while the output from the AND circuit is utilized as a signal for actuating the scram operation of control rod drive mechanisms. Accordingly, it is possible to improve the reliability of the reactor scram system, particularly, improve the reliability under a high temperature atmosphere. (Kamimura, M.)

  12. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1980-09-01

    Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

  13. Reactor scram device for FBR type reactor

    International Nuclear Information System (INIS)

    Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

    1991-01-01

    In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

  14. Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

    International Nuclear Information System (INIS)

    Glasstone, S.; Sesonske, A.

    1994-01-01

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future

  15. Refuelling nuclear reactors

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  16. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

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

  17. Multipurpose research reactors

    International Nuclear Information System (INIS)

    1988-01-01

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

  18. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1978-10-01

    Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

  19. An internally illuminated monolith reactor: Pros and cons relative to a slurry reactor

    NARCIS (Netherlands)

    Carneiro, Joana T.; Carneiro, J.T.; Berger, Rob; Moulijn, Jacob A.; Mul, Guido

    2009-01-01

    In the present study, kinetic models for the photo-oxidation of cyclohexane in two different photoreactor systems are discussed: a top illumination reactor (TIR) representative of a slurry reactor, and the so-called internally illuminated monolith reactor (IIMR) representing a reactor containing

  20. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  1. Advances in reactor physics education: Visualization of reactor parameters

    International Nuclear Information System (INIS)

    Snoj, L.; Kromar, M.; Zerovnik, G.

    2012-01-01

    Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

  2. Reactor feedwater system

    International Nuclear Information System (INIS)

    Hikabe, Katsumi.

    1978-01-01

    Purpose: In order to prevent thermal stresses of a core of PWR type reactor, described has been a method for feeding heated recirculating water to the core in the case of the reactor start-up or shut-down. Constitution: A recirculating water is degassed, cleaned up and heated in the steam condensers, and then feeds the water to the reactor, characterized in that heaters are provided in the bypasses of the turbine, so that heated water is constantly supplied to the reactor. (Nakamura, S.)

  3. New reactor concepts

    International Nuclear Information System (INIS)

    Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A.

    1998-11-01

    The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

  4. 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.

  5. Fission reactors and materials

    International Nuclear Information System (INIS)

    Frost, B.R.T.

    1981-12-01

    The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

  6. Spectral shift reactor control method

    International Nuclear Information System (INIS)

    Impink, A.J. Jr.

    1981-01-01

    A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core wherein is established a reator coolant temperature set point at which it is desired to operate said reactor and first reactor coolant temperature band limits are provided within which said set point is located and it is desired to operate said reactor charactrized in that said reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in said core as said reactor coolant temperature approaches said first band limits thereby to maintain said reactor coolant temperature near said set point and within said first band limits

  7. Method of operating a reactor

    International Nuclear Information System (INIS)

    Oosumi, Katsumi; Yamamoto, Michiyoshi.

    1980-01-01

    Purpose: To prevent stress corrosion cracking in the structural material of a reactor pressure vessel. Method: Prior to the starting of a reactor, the reactor pressure vessel is evacuated to carry out degassing of reactor water, and, at the same time, reactor water is heated. After reactor water is heated to a predetermined temperature, control rods are extracted to start nuclear heating. While the temperature of the reactor water is in a temperature range where elution of a metal which is a structural material of the reactor pressure vessel becomes vigorous and the sensitivity to the stress corrosion cracks increases, the reactor is operated at the maximum permissible temperature raising speed or maximum permissible cooling speed. (Aizawa, K.)

  8. Reactor container

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  9. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1993-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  10. Reactor engineering department annual report

    International Nuclear Information System (INIS)

    1990-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  11. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Hirota, Jitsuya; Asaoka, Takumi; Suzuki, Tomoo; Mitani, Hiroshi; Akino, Fujiyoshi

    1977-09-01

    Research activities in the Division of Reactor Engineering in fiscal 1976 are described. Works of the division concern mainly the development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and the development of Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and activities of the Committee on Reactor Physics. (auth.)

  12. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1984-08-01

    Research and development activities in the Department of Reactor Engineering in fiscal 1983 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  13. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1976-09-01

    Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

  14. Calculation of the emergency condenser of the indirect cycle reactor CAREM

    International Nuclear Information System (INIS)

    Walter, D.; Schaffrath, A.

    2002-08-01

    CAREM is an Argentine project to achieve the development, design and construction of an innovative, simple and small Nuclear Power Plant (NPP). This NPP has an indirect cycle reactor with some distinctive and characteristic features (e.g. integrated primary cooling system by natural circulation, self-pressurized primary system and passive safety systems, etc.). In the frame of an IAEA forwarded HUMAN RESOURCES DEVELOPMENT - 'Advice on CAREM passive emergency condenser' at the Centro Atomico in San Carlos de Bariloche (Argentina) German experimental as well as analytical investigations of passive safety systems - mainly performed at the research centers Juelich and Rossendorf - were presented. During this visit it was decided that FZR should determine the emergency condenser capacity by using the ATHLET code with and without the condensation module extension KONWAR (condensation inside horizontal tube). These results should be compared and assessed with existing RELAP calculations. The ATHLET and ATHLET/KONWAR calculations of the emergency condenser were used for minimizing the user influence performed with exactly the same nodalization scheme and the same initial and boundary conditions than the RELAP calculations. The comparison of all three computational results shows that the required condenser capacities were achieved. The magnitudes were in order of up to 12%. An additionally performed evaluation of local flow parameters show large deviations especially for the heat transfer coefficients at the inner and outer tube wall. Due to the good validation progress of KONWAR by an extensive module validation against NOKO and HORUS experiments, the focus of the further investigation were concentrated on the boiling heat transfer. A comparison of the ATHLET boiling model (according to Forster and Zuber in combination with the equation of Thom for the determination of temperature difference between wall and saturated fluid) with other models from literature (Borishansky

  15. Department of reactor technology

    International Nuclear Information System (INIS)

    1980-01-01

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

  16. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-03-15

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

  17. Competitiveness and range of applications of nuclear power, as seen in the light of recent developments in the field of energy economic and energy policy

    International Nuclear Information System (INIS)

    Michaelis, H.

    1975-01-01

    At the reactor conferences in Karlsruhe in 1973 and in Berlin in 1974 the author gave his views on the competitive position of nuclear energy in the German Federal Republic and described how the determining factors in cost development, both for conventional energy generation and for nuclear power generation, have developed since October 1971. Basic data were provided by the paper by B. Bergmann and H. Kraemer (KFA Juelich) 'Technical and economic state in October 1971 as well as prospects for nuclear energy in power economics in the German Federal Republic' (Juel-827-HT - February 1972). The author now shows to what extent the determining factors for the competitive capacity of nuclear energy in the Federal Republic of Germany have changed until April 1975. (orig.) [de

  18. Reactive flow simulations in complex geometries with high-performance supercomputing

    International Nuclear Information System (INIS)

    Rehm, W.; Gerndt, M.; Jahn, W.; Vogelsang, R.; Binninger, B.; Herrmann, M.; Olivier, H.; Weber, M.

    2000-01-01

    In this paper, we report on a modern field code cluster consisting of state-of-the-art reactive Navier-Stokes- and reactive Euler solvers that has been developed on vector- and parallel supercomputers at the research center Juelich. This field code cluster is used for hydrogen safety analyses of technical systems, for example, in the field of nuclear reactor safety and conventional hydrogen demonstration plants with fuel cells. Emphasis is put on the assessment of combustion loads, which could result from slow, fast or rapid flames, including transition from deflagration to detonation. As a sample of proof tests, the special tools have been tested for specific tasks, based on the comparison of experimental and numerical results, which are in reasonable agreement. (author)

  19. Reactor Safety Commission Code of Practice for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

  20. Remote Reactor Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-10-21

    The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

  1. The program of reactors and nuclear power plants; Programa de reactores y centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Centro Atomico Constituyentes

    2001-07-01

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined.

  2. Basal longitudinal strain predicts future aortic valve replacement in asymptomatic patients with aortic stenosis

    DEFF Research Database (Denmark)

    Carstensen, Helle Gervig; Larsen, Linnea Hornbech; Hassager, Christian

    2016-01-01

    analysis and coronary angiography by MDCT. The combined endpoint was indication for aortic valve replacement (AVR) and sudden cardiac death. During a median follow-up of 2.3 years (interquartile range 1.7-3.6) 43 patients (41%) met the endpoint of indication for AVR. The basal (13.4 ± 3.1% vs. 15.7 ± 3.......1%) and mid-ventricular segments (14.9 ± 2.7% vs. 16.2 ± 2.9%) were significantly reduced, but with sparing of the apical segments, in patients who later underwent AVR. In various multivariable Cox regression models, including only BLS, but not GLS, remained an independent predictor of AVR. CONCLUSION...

  3. Canada-India Reactor (CIR)

    Energy Technology Data Exchange (ETDEWEB)

    None

    1960-12-15

    Design information on the Canada-India Reactor is presented. Data are given on reactor physics, the core, fuel elements, core heat transfer, control, reactor vessel, fluid flow, reflector and shielding, containment, cost estimates, and research facilities. Drawings of vertical and horizontal sections of the reactor and fluid flow are included. (M.C.G.)

  4. REACTOR: an expert system for diagnosis and treatment of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Nelson, W.R.

    1982-01-01

    REACTOR is an expert system under development at EG and G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system technology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future plans for validation and evaluation of REACTOR are presented. The concept of using both event-oriented and function-oriented strategies for accident diagnosis is discussed. The response tree concept for representing expert knowledge is also introduced

  5. One piece reactor removal

    International Nuclear Information System (INIS)

    Chia, Wei-Min; Wang, Song-Feng

    1993-01-01

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

  6. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

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

  7. Reactor decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1984-01-01

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

  8. Status of advanced technology and design for water cooled reactors: Heavy water reactors

    International Nuclear Information System (INIS)

    1989-07-01

    In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

  9. Compact torsatron reactors

    International Nuclear Information System (INIS)

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

    1988-05-01

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

  10. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    2000-01-01

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

  11. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    1998-01-01

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

  12. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    1996-01-01

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

  13. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki

    1982-07-01

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

  14. FBR type reactor core

    International Nuclear Information System (INIS)

    Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

    1995-01-01

    Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

  15. An Immunity-Triggering Effector from the Barley Smut Fungus Ustilago hordei Resides in an Ustilaginaceae-Specific Cluster Bearing Signs of Transposable Element-Assisted Evolution

    KAUST Repository

    Ali, Shawkat

    2014-07-03

    The basidiomycete smut fungus Ustilago hordei was previously shown to comprise isolates that are avirulent on various barley host cultivars. Through genetic crosses we had revealed that a dominant avirulence locus UhAvr1 which triggers immunity in barley cultivar Hannchen harboring resistance gene Ruh1, resided within an 80-kb region. DNA sequence analysis of this genetically delimited region uncovered the presence of 7 candidate secreted effector proteins. Sequence comparison of their coding sequences among virulent and avirulent parental and field isolates could not distinguish UhAvr1 candidates. Systematic deletion and complementation analyses revealed that UhAvr1 is UHOR_10022 which codes for a small effector protein of 171 amino acids with a predicted 19 amino acid signal peptide. Virulence in the parental isolate is caused by the insertion of a fragment of 5.5 kb with similarity to a common U. hordei transposable element (TE), interrupting the promoter of UhAvr1 and thereby changing expression and hence recognition of UhAVR1p. This rearrangement is likely caused by activities of TEs and variation is seen among isolates. Using GFP-chimeric constructs we show that UhAvr1 is induced only in mated dikaryotic hyphae upon sensing and infecting barley coleoptile cells. When infecting Hannchen, UhAVR1p causes local callose deposition and the production of reactive oxygen species and necrosis indicative of the immune response. UhAvr1 does not contribute significantly to overall virulence. UhAvr1 is located in a cluster of ten effectors with several paralogs and over 50% of TEs. This cluster is syntenous with clusters in closely-related U. maydis and Sporisorium reilianum. In these corn-infecting species, these clusters harbor however more and further diversified homologous effector families but very few TEs. This increased variability may have resulted from past selection pressure by resistance genes since U. maydis is not known to trigger immunity in its corn host.

  16. An immunity-triggering effector from the Barley smut fungus Ustilago hordei resides in an Ustilaginaceae-specific cluster bearing signs of transposable element-assisted evolution.

    Directory of Open Access Journals (Sweden)

    Shawkat Ali

    2014-07-01

    Full Text Available The basidiomycete smut fungus Ustilago hordei was previously shown to comprise isolates that are avirulent on various barley host cultivars. Through genetic crosses we had revealed that a dominant avirulence locus UhAvr1 which triggers immunity in barley cultivar Hannchen harboring resistance gene Ruh1, resided within an 80-kb region. DNA sequence analysis of this genetically delimited region uncovered the presence of 7 candidate secreted effector proteins. Sequence comparison of their coding sequences among virulent and avirulent parental and field isolates could not distinguish UhAvr1 candidates. Systematic deletion and complementation analyses revealed that UhAvr1 is UHOR_10022 which codes for a small effector protein of 171 amino acids with a predicted 19 amino acid signal peptide. Virulence in the parental isolate is caused by the insertion of a fragment of 5.5 kb with similarity to a common U. hordei transposable element (TE, interrupting the promoter of UhAvr1 and thereby changing expression and hence recognition of UhAVR1p. This rearrangement is likely caused by activities of TEs and variation is seen among isolates. Using GFP-chimeric constructs we show that UhAvr1 is induced only in mated dikaryotic hyphae upon sensing and infecting barley coleoptile cells. When infecting Hannchen, UhAVR1p causes local callose deposition and the production of reactive oxygen species and necrosis indicative of the immune response. UhAvr1 does not contribute significantly to overall virulence. UhAvr1 is located in a cluster of ten effectors with several paralogs and over 50% of TEs. This cluster is syntenous with clusters in closely-related U. maydis and Sporisorium reilianum. In these corn-infecting species, these clusters harbor however more and further diversified homologous effector families but very few TEs. This increased variability may have resulted from past selection pressure by resistance genes since U. maydis is not known to trigger immunity

  17. An Immunity-Triggering Effector from the Barley Smut Fungus Ustilago hordei Resides in an Ustilaginaceae-Specific Cluster Bearing Signs of Transposable Element-Assisted Evolution

    KAUST Repository

    Ali, Shawkat; Laurie, John D.; Linning, Rob; Cervantes-Chá vez, José Antonio; Gaudet, Denis; Bakkeren, Guus

    2014-01-01

    The basidiomycete smut fungus Ustilago hordei was previously shown to comprise isolates that are avirulent on various barley host cultivars. Through genetic crosses we had revealed that a dominant avirulence locus UhAvr1 which triggers immunity in barley cultivar Hannchen harboring resistance gene Ruh1, resided within an 80-kb region. DNA sequence analysis of this genetically delimited region uncovered the presence of 7 candidate secreted effector proteins. Sequence comparison of their coding sequences among virulent and avirulent parental and field isolates could not distinguish UhAvr1 candidates. Systematic deletion and complementation analyses revealed that UhAvr1 is UHOR_10022 which codes for a small effector protein of 171 amino acids with a predicted 19 amino acid signal peptide. Virulence in the parental isolate is caused by the insertion of a fragment of 5.5 kb with similarity to a common U. hordei transposable element (TE), interrupting the promoter of UhAvr1 and thereby changing expression and hence recognition of UhAVR1p. This rearrangement is likely caused by activities of TEs and variation is seen among isolates. Using GFP-chimeric constructs we show that UhAvr1 is induced only in mated dikaryotic hyphae upon sensing and infecting barley coleoptile cells. When infecting Hannchen, UhAVR1p causes local callose deposition and the production of reactive oxygen species and necrosis indicative of the immune response. UhAvr1 does not contribute significantly to overall virulence. UhAvr1 is located in a cluster of ten effectors with several paralogs and over 50% of TEs. This cluster is syntenous with clusters in closely-related U. maydis and Sporisorium reilianum. In these corn-infecting species, these clusters harbor however more and further diversified homologous effector families but very few TEs. This increased variability may have resulted from past selection pressure by resistance genes since U. maydis is not known to trigger immunity in its corn host.

  18. Role of Global Longitudinal Strain in the Prediction of Outcome in Patients With Severe Aortic Valve Stenosis.

    Science.gov (United States)

    Fries, Bastian; Liu, Dan; Gaudron, Philipp; Hu, Kai; Nordbeck, Peter; Ertl, Georg; Weidemann, Frank; Herrmann, Sebastian

    2017-08-15

    In the present study, we assessed the role of Global Longitudinal Strain (GLS) as a predictor of all-cause mortality in patients with severe aortic valve stenosis (AS), irrespective of their type of treatment. Data of 807 patients with AS receiving complete echocardiographic and clinical examination were retrospectively analyzed. Valve area <1 cm 2 and sufficient image quality were inclusion criteria; patients with severe concomitant valvulopathy were excluded. Patients were grouped into treatment (aortic valve replacement [AVR]) and conservative (non-AVR) groups. Multivariable Cox analysis was used to assess predictors of all-cause mortality. Five hundred fourteen patients were included and 53.3% were of male gender. Mean age at inclusion was 76.4 ± 9.8 years; 326 received AVR. Death from any cause occurred in 72.9% of non-AVR group and 17.8% of AVR group (p <0.001). GLS (expressed as |%|) was found to be an independent predictor of all-cause mortality in non-AVR group (hazard ratio [HR] 0.933, 95% CI 0.854 to 0.987, p = 0.038). In patients receiving AVR, GLS and history of coronary artery bypass graft were found to be independent predictors of all-cause mortality (HR for GLS 0.912, 95% CI 0.730 to 0.999, p = 0.048; HR for coronary artery bypass graft 2.977, 95% CI 1.014 to 6.273, p = 0.013). In non-AVR patients, GLS <9.7% showed a higher 1- and 5-year mortality (log rank p values of 0.002 and 0.010, respectively). In conclusion, GLS is an independent predictor of all-cause mortality in severe AS, irrespective of their type of treatment. GLS <9.7% indicates a significantly higher 1- and 5-year mortality in non-AVR patients. Therefore, GLS should be regularly assessed for enhanced risk stratification and clinical decision-making. Copyright © 2017 Elsevier Inc. All rights reserved.

  19. Reactor operation method

    International Nuclear Information System (INIS)

    Osumi, Katsumi; Miki, Minoru.

    1979-01-01

    Purpose: To prevent stress corrosion cracks by decreasing the dissolved oxygen and hydrogen peroxide concentrations in the coolants within a reactor container upon transient operation such as at the start-up or shutdown of bwr type reactors. Method: After a condensate has been evacuated, deaeration operation is conducted while opening a main steam drain line, as well as a main steam separation valve and a by-pass valve in a turbine by-pass line connecting the main steam line and the condenser without by way of a turbine, and the reactor is started-up by the extraction of control rods after the concentration of dissolved oxygen in the cooling water within a pressure vessel has been decreased below a predetermined value. Nuclear heating is started after the reactor water has been increased to about 150 0 C by pump heating after the end of the deaeration operation for preventing the concentration of hydrogen peroxide and oxygen in the reactor water from temporarily increasing immediately after the start-up. The corrosive atmosphere in the reactor vessel can thus be moderated. (Horiuchi, T.)

  20. Nuclear reactor buildings

    International Nuclear Information System (INIS)

    Nagashima, Shoji; Kato, Ryoichi.

    1985-01-01

    Purpose: To reduce the cost of reactor buildings and satisfy the severe seismic demands in tank type FBR type reactors. Constitution: In usual nuclear reactor buildings of a flat bottom embedding structure, the flat bottom is entirely embedded into the rock below the soils down to the deck level of the nuclear reactor. As a result, although the weight of the seismic structure can be decreased, the amount of excavating the cavity is significantly increased to inevitably increase the plant construction cost. Cross-like intersecting foundation mats are embedded to the building rock into a thickness capable withstanding to earthquakes while maintaining the arrangement of equipments around the reactor core in the nuclear buildings required by the system design, such as vertical relationship between the equipments, fuel exchange systems and sponteneous drainings. Since the rock is hard and less deformable, the rigidity of the walls and the support structures of the reactor buildings can be increased by the embedding into the rock substrate and floor responsivity can be reduced. This enables to reduce the cost and increasing the seismic proofness. (Kamimura, M.)

  1. Advanced reactor development

    International Nuclear Information System (INIS)

    Till, C.E.

    1989-01-01

    Consideration is given to what the aims of advanced reactor development have to be, if a new generation of nuclear power is really to play an important role in man's energy generation activities in a fragile environment. The background given briefly covers present atmospheric evidence, the current situation in nuclear power, how reactors work and what can go wrong with them, and the present magnitudes of world energy generation. The central part of the paper describes what is currently being done in advanced reactor development and what can be expected from various systems and various elements of it. A vigorous case is made that three elements must be present in any advanced reactor development: (1) breeding; (2) passive safety; and (3) shorter-live nuclear waste. All three are possible. In the right advanced reactor systems the ways of achieving them are known. But R and D is necessary. That is the central argument made in the paper. Not advanced reactor prototype construction at this point, but R and D itself. (author)

  2. The reactor Cabri

    International Nuclear Information System (INIS)

    Ailloud, J.; Millot, J.P.

    1964-01-01

    It has become necessary to construct in France a reactor which would permit the investigation of the conditions of functioning of future installations, the choice, the testing and the development of safety devices to be adopted. A water reactor of a type corresponding to the latest CEA constructions in the field of laboratory or university reactors was decided upon: it appeared important to be able to evaluate the risks entailed and to study the possibilities of increasing the power, always demanded by the users; on the other hand, it is particularly interesting to clarify the phenomena of power oscillation and the risks of burn out. The work programme for CABRI will be associated with the work carried out on the American Sperts of the same type, during its construction, very useful contacts were made with the American specialists who designed the se reactors. A brief description of the reactor is given in the communication as well as the work programme for the first years with respect to the objectives up to now envisaged. Rough description of the reactor. CABRI is an open core swimming-pool reactor without any lateral protection, housed in a reinforced building with controlled leakage, in the Centre d'Etudes Nucleaires de Cadarache. It lies alone in the middle of an area whose radius is 300 meters long. Control and measurements equipment stand out on the edge of that zone. It consumes MTR fuel elements. The control-safety rods are propelled by compressed air. The maximum flow rate of cooling circuit is 1500 m 3 /h. Transient measurements are recorded in a RW330 unit. Aims and work programme. CABRI is meant for: - studies on the safety of water reactors - for the definition of the safety margins under working conditions: research of maximum power at which a swimming-pool reactor may operate with respect to a cooling accident, of local boiling effect on the nuclear behaviour of the reactor, performances of the control and safety instruments under exceptional

  3. Spectral shift reactor control method

    International Nuclear Information System (INIS)

    Impink, A.J.

    1982-01-01

    A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core where there is established a reactor coolant temperature set point at which it is desired to operate the reactor and first reactor coolant temperature band limits within which the set point is characterized. The reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in the core as the reactor coolant temperature approaches the first band limits to maintain the reactor coolant temperature near the set point and within the first band limits. The reactivity charges associated with movement of respective coolant displacer element clusters is calculated and compared with a calculated derived reactivity charge in order to select the cluster to be moved. (author)

  4. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

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

  5. Innovative hybrid biological reactors using membranes; Reactores biologico hibrido innovadores utilizando membranas

    Energy Technology Data Exchange (ETDEWEB)

    Diez, R.; Esteban-Garcia, A. L.; Florio, L. de; Rodriguez-Hernandez, L.; Tejero, I.

    2011-07-01

    In this paper we present two lines of research on hybrid reactors including the use of membranes, although with different functions: RBPM, biofilm reactors and membranes filtration RBSOM, supported biofilm reactors and oxygen membranes. (Author) 14 refs.

  6. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Joo, Young Sang; Lee, Jae Han

    2007-02-01

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification

  7. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Lee, Jae Han

    2007-02-15

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

  8. The Maple reactor project

    International Nuclear Information System (INIS)

    Malkoske, G.R.; Labrie, J.-P.

    2003-01-01

    MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

  9. Repairing liner of the reactor; Reparacion del liner del reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-15

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

  10. Methanogenesis in Thermophilic Biogas Reactors

    DEFF Research Database (Denmark)

    Ahring, Birgitte Kiær

    1995-01-01

    Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process as ....... Experiments using biogas reactors fed with cow manure showed that the same biogas yield found at 550 C could be obtained at 610 C after a long adaptation period. However, propionate degradation was inhibited by increasing the temperature.......Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process...... as indicated by a lower concentration of volatile fatty acids in the effluent from the reactors. The specific methanogenic activity in a thermophilic pilot-plant biogas reactor fed with a mixture of cow and pig manure reflected the stability of the reactor. The numbers of methanogens counted by the most...

  11. A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

  12. Future view of total energy system and reactor engineering and reactor physics

    International Nuclear Information System (INIS)

    Ozawa, T.

    1974-01-01

    This paper outlines the present status of fission reactors and fusion reactors. The conversion ratio of light water reactors is 0.5, and the efficiency is 32% because of relatively low temperature. Both pressurized water reactors and boiling water reactors are technically well developed, their performances are well known, and the fuel cycle is well developed, so that both reactors have monopolized power reactor market. But the reprocessing of spent fuel and the treatment of their hazards are inevitable, and the construction and enlargement of reprocessing facilities are indispensable. In LMFBR's tight sealing is easy because they are non-pressurized, and the efficiency is 41%. But liquid sodium is strongly activated and recirculated, so that chemical obstruction due to the breakage of recirculating pumps, pipings, and heat exchangers may occur, and the hazard of plutonium is large. Regarding controlled thermo-nuclear fusion reactors, because Lawson criterion must be satisfied, two methods of plasma confinement are now experimented. One is the plasma confinement by strong magnetic field of 50 KG to 100 KG, and the other is the confinement by the implosion method with high-power laser beam. The latter has much more uncertainties than the former, but recently both methods have made much progress. (Tai, I)

  13. Sutureless versus Conventional Aortic Valve Replacement: Outcomes in 70 High-Risk Patients Undergoing Concomitant Cardiac Procedures.

    Science.gov (United States)

    Hanedan, Muhammet Onur; Yuruk, Mehmet Ali; Parlar, Ali Ihsan; Ziyrek, Ugur; Arslan, Ali Kemal; Sayar, Ufuk; Mataraci, Ilker

    2018-02-01

    In elderly, high-risk surgical patients, sutureless aortic valve replacement (AVR) can often be an alternative to conventional AVR; shorter aortic cross-clamp and cardiopulmonary bypass times are the chief advantages. We compared the outcomes of sutureless AVR with those of conventional AVR in 70 elderly patients who underwent concomitant cardiac surgical procedures. We retrospectively analyzed the cases of 42 men and 28 women (mean age, 70.4 ± 10.3 yr; range, 34-93 yr) who underwent cardiac operations plus AVR with either a sutureless valve (group 1, n=38) or a conventional bioprosthetic or mechanical valve (group 2, n=32). Baseline patient characteristics were similar except for worse New York Heart Association functional status and the prevalence of diabetes mellitus in group 1. In group 1, the operative, cross-clamp, and cardiopulmonary bypass times were shorter (all P =0.001), postoperative drainage amounts were lower ( P =0.009), hospital stays were shorter ( P =0.004), and less red blood cell transfusion was needed ( P =0.037). Echocardiograms before patients' discharge from the hospital showed lower peak and mean aortic gradients in group 1 (mean transvalvular gradient, 8.4 ± 2.8 vs 12.2 ± 5.2 mmHg; P =0.012). We found that elderly, high-risk patients who underwent multiple cardiac surgical procedures and sutureless AVR had better hemodynamic outcomes and shorter ischemic times than did patients who underwent conventional AVR.

  14. Applications: fission, nuclear reactors. Fission: the various ways for reactors and cycles

    International Nuclear Information System (INIS)

    Bacher, P.

    1997-01-01

    A historical review is presented concerning the various nuclear reactor systems developed in France by the CEA: the UNGG (graphite-gas) system with higher CO 2 pressures, bigger fuel assemblies and powers higher than 500 MW e, allowed by studies on reactor physics, cladding material developments and reactor optimization; the fast neutron reactor system, following the graphite-gas development, led to the Superphenix reactor and important progress in simulation based on experiment and return of experience; and the PWR system, based on the american license, which has been successfully accommodated to the french industry and generates up to 75% of the electric power in France

  15. 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

  16. Research reactor support

    International Nuclear Information System (INIS)

    2005-01-01

    Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

  17. The prototype fast reactor

    International Nuclear Information System (INIS)

    Broomfield, A.M.

    1985-01-01

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

  18. NCSU Reactor Sharing Program

    International Nuclear Information System (INIS)

    Perez, P.B.

    1993-01-01

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

  19. Reactor safety research program. A description of current and planned reactor safety research sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research

    International Nuclear Information System (INIS)

    1975-06-01

    The reactor safety research program, sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, is described in terms of its program objectives, current status, and future plans. Elements of safety research work applicable to water reactors, fast reactors, and gas cooled reactors are presented together with brief descriptions of current and planned test facilities. (U.S.)

  20. 2012 review of French research reactors

    International Nuclear Information System (INIS)

    Estrade, Jerome

    2013-01-01

    Proposed by the French Reactor Operators' Club (CER), the meeting and discussion forum for operators of French research reactors, this report first gives a brief presentation of these reactors and of their scope of application, and a summary of highlights in 2012 for each of them. Then, it proposes more detailed presentations and reviews of characteristics, activities, highlights, objectives and results for the different types of reactors: neutron beam reactors (Orphee, High flux reactor-Laue-Langevin Institute or HFR-ILL), technological irradiation reactors (Osiris and Phenix), training reactors (Isis and Azur), reactors for safety research purposes (Cabri and Phebus), reactors for neutronic studies (Caliban, Prospero, Eole, Minerve and Masurca), and new research reactors (the RES facility and the Jules Horowitz reactor or JHR)